ML102150466

Issuance of Amendment No. 227, Modify Technical Specification 3/4.9.7, Crane Travel - Fuel Handling Building to Permit Certain Operations Needed for Dry Cask Storage of Spent Nuclear Fuel

ML102150466
Person / Time
Site:	Waterford
Issue date:	09/13/2010
From:	Kalyanam N Plant Licensing Branch IV
To:	Entergy Operations
Kalyanam N, NRR/DORL/LP4, 415-1480
References
TAC ME2221
Download: ML102150466 (16)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 13, 2010 Vice President, Operations Entergy Operations, Inc.

Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093 SUB~IECT: WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENT RE: MODIFY TECHNICAL SPECIFICATION 3/4.9.7, "CRANE TRAVEL - FUEL HANDLING BUILDING" (TAC NO. ME2221)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 227 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 September 9, 2009, as supplemented by letters dated June 8 and July 22, 2010.

The amendment revises TS 3/4.9.7, "Crane Travel - Fuel Handling Building," to permit certain operations needed for dry cask storage of spent nuclear fuel. The current wording of TS 3/4.9.7 prohibits travel of the lid for the spent fuel storage canister over irradiated fuel in the canister during canister operations. The change to this TS, while continuing to prohibit travel of a heavy load over irradiated fuel assemblies in the spent fuel pool, permits travel of loads in excess of 2,000 pounds over a transfer cask containing irradiated fuel assemblies, provided a single failure-proof handling system is used.

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. 227 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. 227 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 September 9, 2009, as supplemented by letters dated June 8 and July 22, 2010, 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

-2

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. 227, 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 prior to the start of the dry cask storage operations.

FOR THE NUCLEAR REGULATORY COMMISSION 1fl4~1?VCIMt/6w Michael T. 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: September 13, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 227 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 pages are identified by amendment number and contain marginal lines indicating the areas of change.

Facility Operating License REMOVE INSERT

-4 Technical Specifications REMOVE INSERT 3/49-7 3/49-7

- 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 staffs 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. 227, and the Environmental Protection Plan contained in Appendix 8, are hereby incorporated in the license. EOI shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

AMENDMENT NO. 227

REFUELING OPERATIONS 3/4.9.7 CRANE TRAVEL - FUEL HANDLING BUILDING LIMITING CONDITION FOR OPERATION 3.9.7 Cranes in the fuel handling building shall be restricted as follows:

a. The spent fuel handling machine shall be used* for the movement of fuel assemblies (with or without CEAs) and shall be OPERABLE with:

1. A minimum hoist capacity of 1800 pounds, and

2. An overload cutoff limit of less than or equal to 1900 pounds, and,

b. Loads in excess of 2000 pounds shall be prohibited from travel over irradiated fuel fuel assemblies in the Fuel Handling Building, except over assemblies in a transfer cask using a single-failure-proof handling system.

APPLICABILITY: During movement of irradiated fuel assemblies in the fuel handling building, or with irradiated fuel assemblies in the Fuel Handling Building.

ACTION:

a. With the spent fuel handling machine inoperable, suspend the use of the spent fuel handling machine for movement of fuel assemblies and place the crane load in a safe position.

b. With loads in excess of 2000 pounds over irradiated fuel assemblies in the Fuel Handling Building, except over assemblies in a transfer cask using a single failure-proof handling system, place the crane load in a safe position.

c. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.7.1 The spent fuel handling machine shall be demonstrated OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to the start of fuel assembly movement and at least once per 7 days thereafter by performing a load test of at least 1800 pounds and demonstrating the automatic load cutoff when the hoist load exceeds 1900 pounds.

4.9.7.2 The electrical interlock system which prevents crane main hook travel over irradiated fuel assemblies in the Fuel Handling Building, except over assemblies in a transfer cask using a single-failure-proof handling system, shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.

4.9.7.3 Administrative controls which prevent crane auxiliary hook travel with loads in excess of 2000 pounds over the irradiated fuel assemblies in the Fuel Handling Building, including over assemblies in a transfer cask, shall be enforced during crane operations.

• Not required for movement of new fuel assemblies outside the spent fuel pool and Cask Storage Pit.

WATERFORD - UNIT 3 3/4 9-7 AMENDMENT NO. 6, 144,227

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 227 TO FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS, INC.

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

1.0 INTRODUCTION

By application dated September 9, 2009 (Reference 1), as supplemented by letters dated June 8 and July 22, 2010 (References 2 and 3), Entergy Operations, Inc. (Entergy, the licensee), requested changes to the Technical Specifications (TSs) for Waterford Steam Electric Station, Unit 3 (Waterford 3). The supplemental letters dated June 8 and July 22, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and 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 November 17, 2009 (74 FR 59261).

The amendment revises TS 3/4.9.7, "Crane Travel - Fuel Handling Building," to permit certain operations needed for dry cask storage of spent nuclear fuel. The current wording of TS 3/4.9.7 prohibits travel of the lid for the spent fuel storage canister over irradiated fuel in the canister during canister operations. The change to this TS, while continuing to prohibit travel of a heavy load over irradiated fuel assemblies in the spent fuel pool, permits travel of loads in excess of 2,000 pounds over a transfer cask containing irradiated fuel assemblies, provided a single failure-proof handling system is used.

Specifically, the following changes to TS 3/4.9.7, "Crane Travel - Fuel Handling Building," are requested:

Existing TS Limiting Condition for Operation (LCO) 3.9.7.b 3.9.7 Cranes in the fuel handling building shall be restricted as follows:

b. Loads in excess of 2000 pounds shall be prohibited from travel over irradiated fuel assemblies in the Fuel Handling Building.

Enclosure 2

-2 Changed to 3.9.7 Cranes in the fuel handling bUilding shall be restricted as follows:

b. Loads in excess of 2000 pounds shall be prohibited from travel over irradiated fuel assemblies in the Fuel Handling Building, except over assemblies in a transfer cask using a single-failure proof handling system.

Existing 3.9.7 Action Statement b

b. With loads in excess of 2000 pounds over irradiated fuel assemblies in the Fuel Handling Building, place the crane load in a safe position.

Changed to

b. With loads in excess of 2000 pounds over irradiated fuel assemblies in the Fuel Handling Building, except over assemblies in a transfer cask using a single-failure-proof handling system, place the crane load in a safe position.

Existing TS Surveillance Requirement (SR) 4.9.7.2 and 4.9.7.3 4.9.7.2 The electrical interlock system which prevents crane main hook travel over irradiated fuel assemblies in the Fuel Handling Building shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.#

4.9.7.3 Administrative controls which prevent crane auxiliary hook travel with loads in excess of 2000 pounds over the irradiated fuel assemblies in the Fuel Handling Building shall be enforced during crane operations.

Changed to 4.9.7.2 The electrical interlock system which prevents crane main hook travel over irradiated fuel assemblies in the Fuel Handling Building, except over assemblies in a transfer cask using a single-failure-proof handling system, shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.

4.9.7.3 Administrative controls which prevent crane auxiliary hook travel with loads in excess of 2000 pounds over the irradiated fuel assemblies in the Fuel Handling Building, including over assemblies in a transfer cask, shall be enforced during crane operations.

-3 In addition, the licensee proposed deleting a footnote to SR 4.9.7.2, that permitted bypass of the electrical interlock system under administrative control during implementation of the spent fuel pool rack replacement prior to refueling outage 9 because it is no longer applicable.

2.0 REGULATORY EVALUATION

Section 182a of the Atomic Energy Act of 1954 (the Act), as amended, requires applicants for nuclear power plant operating licenses to include the TSs as part of the license. The Commission's regulatory requirements related to the content of TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR) Section 50.36, "Technical specifications." The regulations require that the TSs include items in specific categories, including: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in the TSs.

The four criteria defined by 10 CFR 50.36(c)(2)(ii) for determining whether particular items are required to be included in the TS LCOs, are as follows:

(A) Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary.

(8) Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier; (C) Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

(D) Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety.

The regulations in 10 CFR 50.36(c)(2) specify that, when an LCO of a nuclear reactor plant is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the TSs until the condition is met.

The regulations in 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 2, "Design bases for protection against natural phenomena," specify, in part, that structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena, such as earthquakes.

GDC 4, "Environmental and dynamic effects design bases," of Appendix A to 10 CFR Part 50 specifies, in part, that structures, systems, and components important to safety shall be

- 4 appropriately protected against dynamic effects, including the effects of missiles, that may result from equipment failures.

NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," (SRP), Section 9.1.5, "Overhead Heavy Load Handling Systems,"

references the guidelines of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants,"

for implementation of these GDCs in the design of overhead heavy-load handling systems.

NUREG-0612 includes the NRC staffs recommendations and guidelines including definition of safe load paths, use of load-handling procedures, training of crane operators, guidelines on slings and special lifting devices, periodic inspection and maintenance for the crane, as well as various alternatives that include: use of a single-failure-proof handling system, use of mechanical stops or electrical interlocks to keep heavy loads away from fuel or safe shutdown equipment.

NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power Plants," contains the criteria for design of single-failure-proof cranes to the appropriate standards.

3.0 TECHNICAL EVALUATION

3.1 Background The licensee has determined that the spent fuel pool capacity is not adequate to store all spent fuel until end of plant life. Therefore, it plans to use dry cask storage with an onsite Independent Spent Fuel Storage Installation. The dry cask storage system will use the multi-purpose canister (MPC)-32 canister to hold spent fuel that is transferred from the spent fuel pool to the MPC-32 in the cask storage area.

The original Fuel Handling Building (FHB) crane was designed, fabricated, and installed by Harnischfeger Corporation, also known as P&H cranes, during plant construction as a non single-failure-proof crane with a design-rated load of 125 tons.

For spent fuel operations, the MPC-32 canister will be housed in a transfer cask in the cask storage area. Spent fuel assemblies will be moved via the spent fuel handling machine from the spent fuel pool to the MPC-32 canister. After the MPC-32 canister has been loaded with spent fuel assemblies, the licensee needs to install a lid, weighing approximately 9,650 pounds, on the MPC-32 canister. In addition, the transfer cask lift yoke and lift yoke extension lifting devices each weigh in excess of 2,000 pounds. TS 3/4.9.7 currently prohibits loads in excess of 2,000 pounds from traveling over irradiated fuel assemblies in the FHB. Since the MPC-32 canister and transfer cask are in the FHB, installing the lid would not be permissible under the current TS requirements. The proposed change to the TS would permit travel of loads in excess of 2000 pounds over the transfer cask containing irradiated fuel assemblies conditional upon use a single-failure-proof handling system in accordance with SRP Section 9.1.5, paragraph III.4.C. The revised TS would continue to prohibit travel of loads in excess of 2,000 pounds over irradiated fuel assemblies in the remainder of the FHB, including the spent fuel pool, even if the load is carried by a single-failure-proof handling system.

-5 3.2 Upgrading to Single-Failure Proof The licensee's single-failure-proof handling system will consist of a single-failure-proof FHB cask crane and special lifting devices that satisfy American National Standards Institute (ANSI) standard N14.6, "Special Lifting Devices for Shipping Containers Weighing 10000 Pounds (4500 kg) or More," and slings that are chain or wire rope and satisfy the American Society of Mechanical Engineers (ASME) standard B30.9, "Slings." The current FHB cask crane is not single-failure proof, but the licensee plans to upgrade the FHB crane to single-failure proof. In Reference 1, the licensee provided a description of the FHB crane upgrade modifications to single-failure proof. The subject of this amendment request was a change to TS 3/4.9.7 while the description of the upgrade was provided for information only. The licensee stated that the single-failure-proof upgrade of the existing FHB cask crane will be made under the provisions of 10 CFR Section 50.59, "Changes, tests and experiments." Therefore, the process of upgrading the FHB crane is not addressed here. However, the licensee has committed to a scheduled completion date for the upgrade of "prior to the first dry cask storage loading campaign." The NRC staff concludes that, with the commitment that travel of loads in excess of 2,000 pounds over a transfer cask containing irradiated fuel assemblies will be permitted only provided a single-failure-proof handling system is used, the proposed changes to the TSs are acceptable.

3.3 Lifting Devices In Reference 1, Section 5.0, Regulatory Safety Analysis, the licensee described the regulatory guidelines of the special lifting devices and slings. Lifting devices in single-failure handling systems should be selected to satisfy either of the following criteria:

(1) A special lifting device that satisfies ANSI N14.6 should be used for recurrent load movements in critical areas (reactor head lifting, reactor vessel internals, spent fuel casks). The lifting device should have either dual, independent load paths or a single load path with twice the design safety factor specified by ANSI N14.6 for the load.

(2) Slings should satisfy the criteria of ASME B30.9 and be constructed of metallic material (chain or wire rope). The slings should be either (a) configured to provide dual or redundant load paths or (b) selected to support a load twice the weight of the handled load.

In Reference 2, the licensee confirmed that the special lifting devices will have either (a) dual and independent load paths, or (b) a single load path with twice the design safety factor specified by ANSI N14.6, and that slings, in addition of being constructed of metallic wire rope, will be either (a) configured to provide dual or redundant load paths, or (b) selected to support a load twice the weight of the handled load.

This is in conformance with the requirements of ANSI N14.6 and ASME B30.9 and, therefore, the NRC staff concludes the licensee's response is acceptable.

Additionally, in Reference 2, the licensee stated that (1) the lifting devices below the hook on the FHB overhead crane that will be used during travel over irradiated fuel in the MPC canister

-6 during canister operations will be designed to meet either ANSI N14.6 or ASME 830.9; (2) the special lifting devices that are used during canister operations will be designed to meet ANSI N14.6; (3) the special lifting devices referred to above will have either dual, independent load paths or a single load path with twice the design safety factor specified by ANSI N14.6; and (4) the ASME 830.9 standard, which applies to slings, will be used for any required slings used during canister operations that are not a special lifting device and the slings will be either (a) configured to provide dual or redundant load paths, or (b) selected to support a load twice the weight of the handled load.

The NRC staff reviewed the licensee's response and concluded it was acceptable, since the lift yoke and the lift yoke extension met the ANSI N14.6 requirements and the slings meet the requirements of ASME 830.9 and guidance of NUREG-0612.

3.4 Control of Heavy Loads NUREG-0612 provides regulatory guidelines for the control of heavy loads to assure the safe handling of heavy loads in areas where a load drop could impact stored spent fuel, fuel in the reactor core, or equipment that may be required to achieve safe shutdown or permit continued decay heat removal. NRC Generic Letter (GL)80-113, "Control of Heavy Loads;' dated December 22,1980, as supplemented by GL 81-07, "Control of Heavy Loads;'dated February 3, 1981, requested the licensee to report implementation of NUREG-0612 in Phases I and II.

Phase I guidelines addressed measures for reducing the likelihood of dropping heavy loads and provided criteria for establishing safe load paths; procedures for load-handling operations; training of crane operators; design, testing, inspection, and maintenance of cranes and lifting devices; and selection and use of slings. Phase II guidelines addressed alternatives to reduce further the probability of a load-handling accident or mitigate the consequences of heavy-load drops. These alternatives include using a single-failure-proof crane for increased handling system reliability, employing electrical interlocks and mechanical stops for restricting crane travel to safe areas, or performing load drop and consequence analyses for assessing the impact of dropped loads on plant safety and operations. Later, in GL 85-11, "Completion of Phase II of 'Control of Heavy Loads at Nuclear Power Plants' NUREG-0612;' dated June 28, 1985, the NRC staff concluded that a detailed review of Phase II responses was not necessary.

The Waterford 3 Updated Final Safety Analysis Report (UFSAR), Section 9.1.4.2.2.16, "Control of Heavy Loads Requirements;' states that the licensee has implemented NUREG-0612 Phase I guidelines. Accordingly, the NRC staff concludes that the issue of heavy-load handling is consistent with the requirements of NUREG-0612 and GL 85-11, that a detailed review of Phase II responses were not necessary and, therefore, the issue of heavy-load handling was resolved.

In Reference 3, the licensee provided its commitment for the upgrade of the existing (FH8) cask crane main hoist to meet the single-failure-proof criteria of NUREG-0554 and NUREG-0612, as is applicable, for the modification of the existing non-single-failure-proof crane. This is further discussed in this safety evaluation in Section 3.2, Upgrading to Single-Failure Proof, and Section 3.7, Regulatory Commitment.

-7 3.5 Compliance with 10 CFR 50.36 Criteria 1, 3 and 4 of 10 CFR 50.36( c)(2)(ii) are not applicable to this license amendment request (LAR). Criterion 2 applies to an operating restriction that is an initial condition of a design-basis accident or transient analysis described in the UFSAR that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.

The prohibition against handling loads greater than 2,000 pounds over irradiated fuel assemblies in the FHB is to prevent operation in a condition that could potentially lead to an unanalyzed load-drop accident. With the handling system upgraded to satisfy single-failure proof criterion, the licensee has met the regulations and NRC guidelines for handling heavy loads over the transfer cask and the proposed action would not lead to an unanalyzed event.

Therefore, the NRC staff concludes that the LAR TS change continues to be in compliance with 10 CFR 50.36(c)(2)(ii).

3.6 Seismic Qualification The ASME NOG-1-2004, "Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder)," June 2004, requires that the new trolley, the existing bridge, the bridge with the newly installed trolley, and any additional weight resulting from the single-failure-proof upgrade, will be seismically qualified.

In Reference 2, the licensee stated that the documentation related to the change to the single failure-proof upgrade for the FHB overhead crane contains a detailed step-by-step comparison of the ASME NOG-1-2004 requirements and the results of the design of the new trolley. Since the newly upgraded FHB overhead crane meets the seismic requirements of ASME NOG-1 2004, the NRC staff concludes the licensee's response is acceptable.

In a letter dated March 31, 2010 (Reference 4), the NRC staff requested the licensee to confirm that the runway crane supporting structure will be seismically qualified in accordance with the current licensing basis criteria to support the crane with the new trolley and additional weight resulting from the single-failure-proof upgrade. In Reference 2, the licensee stated that calculations have been performed to confirm that the seismic qualification has been performed and the crane rails and the supporting structure of the FHB meet seismic and structural requirements for the upgraded overhead crane. The staff concludes that the seismic qualification adequately meets the Code requirements and, therefore, is acceptable.

3.7 Regulatory Commitment In Reference 3, Entergy has made the following regulatory commitment:

Entergy will upgrade the existing Fuel Handling Building (FHB) cask crane main hoist to meet the single-failure-proof criteria of NUREG 0554 and NUREG 0612 as is applicable for the modification of the existing non single failure proof crane.

-8 The licensee's scheduled completion date is provided as "prior to the first dry cask storage loading campaign."

The NRC staff concludes that the proposed commitment satisfies the need for continuing compliance and is, therefore, acceptable.

4.0 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.

5.0 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 November 17,2009 (74 FR 59261). 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.

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.

7.0 REFERENCES

1. Kowalewski, J. A., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "License Amendment Request to Modify Technical Specification 3/4.9.7, Crane Travel-Fuel Handling Building Waterford Steam Electric Station, Unit 3 (Waterford 3)," dated September 9,2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092540575).

2. Kowalewski, J.A., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Response to Request for Additional Information Associated with Technical Specification 3/4.9.7, Crane Travel-Fuel Handling Building Waterford Steam Electric Station, Unit 3," dated June 8,2010 (ADAMS Accession No. ML101620130).

-9

3. Kowalewski, J. A., Entergy Operations, Inc., letter to U.S. Nuclear Regulatory Commission, "Technical Specification 3/4.9.7, Crane Travel- Fuel Handling Building Commitment Response Waterford Steam Electric Station, Unit 3," dated July 22,2010 (ADAMS Accession No. ML102070092).

4. Kalyanam, N., U.S. Nuclear Regulatory Commission, letter to Entergy Operations, Inc.,

"Waterford Steam Electric Station, Unit 3 - Request for Additional Information Re:

License Amendment Request to Modify Technical Specification 3/4.9.7, "Crane Travel Fuel Handling Building" (TAC No. ME2221 )," dated March 31, 2010 (ADAMS Accession No. ML100680144).

Principal Contributors: G. Purciarello, DSS/SBPB D. Hoang, DE/EMCB Date: September 13, 2010

September 13, 2010 Vice President, Operations Entergy Operations, Inc.

Waterford Steam Electric Station, Unit 3 17265 River Road Killona, LA 70057-3093 SUB~IECT: WATERFORD STEAM ELECTRIC STATION, UNIT 3 - ISSUANCE OF AMENDMENT RE: MODIFY TECHNICAL SPECIFICATION 3/4.9.7, "CRANE TRAVEL - FUEL HANDLING BUILDING" (TAC NO. ME2221)

Dear Sir or Madam:

The Commission has issued the enclosed Amendment No. 227 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 September 9, 2009, as supplemented by letters dated June 8 and July 22, 2010.

The amendment revises TS 3/4.9.7, "Crane Travel - Fuel Handling Building," to permit certain operations needed for dry cask storage of spent nuclear fuel. The current wording of TS 3/4.9.7 prohibits travel of the lid for the spent fuel storage canister over irradiated fuel in the canister during canister operations. The change to this TS, while continuing to prohibit travel of a heavy load over irradiated fuel assemblies in the spent fuel pool, permits travel of loads in excess of 2,000 pounds over a transfer cask containing irradiated fuel assemblies, provided a single failure-proof handling system is used.

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. 227 to NPF-38

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

PUBLIC RidsNrrDorlDpr Resource DHoang, NRR/DE/EMCB LPLIV r/f RidsNrrDssSbpb Resource MHartzman, NRR/DE/EMCB RidsAcrsAcnw_MailCTR Resource RidsNrrPMWaterford Resource RGrover, NRRlDIRSIISTB RidsNrrDeEmcb Resource RidsNrrLAJBurkhardt Resource GPurciarello, NRRlDSS/SBPB RidsNrrDirsltsb Resource RidsOgcRp Resource RidsNrrDorlLpl4 Resource RidsRgn4MailCenter Resource ADAMS Accession No ML102150466 *SE memo dated OFFICE NRRlLPL4/PM NRRlLPL4/LA NRRlDIRSIITSB/BC NRRlDE/EMCB/BC (A)

NAME NKalyanam JBurkhardt RElliott KManoly

• DATE 8/16/10 8/6/10 9/1/10 7/12/10 OFFICE NRRlDSS/SBPB/BC OGC NRRlLPL4/BC NRRlLPL4/PM NAME GCasto* AJones MMarkley MThadani for NKalyanam DATE 7/22/10 9/9/10 9/10/10 9/13/10 OFFICIAL RECORD COpy