we Entergy Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power Station600 Rocky Hill RoadPlymouth, MA 02360

November 26, 2013

U.S. Nuclear Regulatory CommissionAttn: Document Control DeskWashington, DC 20555-0001

John A. Dent, Jr.Site Vice President

SUBJECT: Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power StationDocket No. 50-293License No. DPR-35

Proposed License Amendment Request to Modify Technical Specification4.3.4, "Heavy Loads" to Facilitate Dry Storage Handling Operations

REFERENCES:

1. Final Safety Analysis Report for the Holtec International Storage and TransferOperation Reinforced Module Cask System (HI-STORM 100 Cask System),HOLTEC Report H12002444, Docket 72-1014, Rev. 9, February 13, 2010.

2. NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, U.S.Nuclear Regulatory Commission, July 1980. (ML070250180)

3. NUREG-0800 Section 9.1.5 Rev. 1, Standard Review Plan for OverheadHeavy Load Handling Systems, March 2007. (ML062260190)

4. ANSI N14.6, Radioactive Materials - Special Lifting Devices for ShippingContainers Weighing 10,000 Pounds (4500 kg) or More, American NationalStandards Institute, January 1993.

5. ASME B30.9, Slings, American Society of Mechanical Engineers, 2003.

6. NRC Regulatory Issue Summary 2005-25: Clarification of NRC Guidelines forControl of Heavy Loads, October 31, 2005. (ML052340485)

7. NRC Regulatory Issue Summary 2005-25, Supplement 1, Clarification ofNRC Guidelines for Control of Heavy Loads, May 29, 2007. (ML071210434)

8. Waterford 3 Steam Electric Station License Amendment 227, ModifyTechnical Specification 3/4.9.7, "Crane Travel-Fuel Handling Building" (TACNO. ME2221), dated September 13, 2010

9. NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, May1979

LETTER NUMBER: 2.13.042

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the followingamendment for Pilgrim Nuclear Power Station (PNPS). The proposed amendment revisesTechnical Specification (TS) 4.3.4, "Heavy Loads" limitation imposed on maximum weight thatcould travel over the irradiated fuel in the spent fuel pool. The proposed revision is associatedwith the Independent Spent Fuel Storage Installation (ISFSI) activity of loading of spent fuelassemblies into a Multi-Purpose Canister (MPC) in the spent fuel pool.

Entergy Letter No. 2.13.042, Page 2 of 3

Current wording of TS 4.3.4 prohibits travel of heavy loads in excess of 2,000 lbs over fuelassemblies in the spent fuel pool. Dry storage cask operations involve loading irradiated fuelassemblies into an MPC in the spent fuel pool, and then lifting the canister lid over those fuelassemblies to permit installation onto the canister. The MPC lid weighs approximately 10,000lbs. The proposed TS change would permit travel of loads in excess of 2,000 lbs over theloaded MPC containing irradiated fuel assemblies to place or remove the MPC lid. A single-failure-proof handling system will be used for the canister operation, while continuing to prohibittravel of heavy loads in excess of 2,000 lbs over irradiated fuel assemblies in other areas of thespent fuel pool.

The single-failure-proof handling system would comply with the NRC guidance included inReference 3.

Attachment 1 provides an analysis of the proposed Technical Specification change. Attachment2 provides a mark-up of the proposed changed page.

The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteriaspecified in 10 CFR 50.92(c), and it has been determined that this change involves nosignificant hazards. The bases for this determination are included in the attached submittal.

Entergy requests approval of the proposed amendment by September 30, 2014, in support ofthe dry cask storage operations necessary to store spent fuel at an onsite ISFSI. Onceapproved, the amendment shall be implemented prior to the start of the dry cask storageoperations.

The proposed License Amendment is similar to the Waterford 3 License Amendment No. 227,for changes to the heavy loads limitations to permit MPC operations in the spent fuel pool(Reference 8).

This application for License Amendment does not contain any new regulatory commitments.

If you have any questions regarding the subject matter, please contact Joseph R. Lynch at (508)830-8403.

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

Executed on the AX 7/1 day of , 2013.

Sincerely,

John A. nt, Jr.Site Vice President

Attachment 1: Evaluation of Proposed TS Changes (9 pages).Attachment 2: Marked-up Pages of the Current TS (1 page).

cc: Ms. Nadiyah Morgan, Project ManagerDivision of Operating Reactor LicensingOffice of Nuclear Reactor RegulationU. S. Nuclear Regulatory CommissionOne White Flint North O-8C2-A11555 Rockville PikeRockville, MD 20852

Enterav Letter No. 2.13.042. Paae 3 of 3

Regional Administrator, Region 1U.S. Nuclear Regulatory Commission2100 Renaissance Blvd, Suite 100King of Prussia, PA 19406-2713

NRC Resident InspectorPilgrim Nuclear Power Station

John Giarrusso, Jr.Planning and Preparedness Section ChiefMass Emergency Management Agency (MEMA)400 Worcester RoadFramingham, MA 01702

Beverly Anderson, Acting Director,Massachusetts Department of Public Health (MDPH)Radiation Control ProgramCommonwealth of Massachusetts529 Main Street, Suite 1M2ACharlestown, MA 02129-1121

ATTACHMENT 1To Entergy Letter No. 2.13.042

Proposed License Amendment Request to Modify Technical Specification 4.3.4, "HeavyLoads" to Facilitate Dry Storagqe Handling Operations

Analysis of Proposed Technical Specification Change

9 Pages

ATTACHMENT 1

To Enterqv Letter No. 2.13.042

1.0

2.0

3.0

4.0

ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGE

DESCRIPTION

PROPOSED CHANGES

BACKGROUND

TECHNCIAL ANALYSIS

4.1 HI-STORM 100 Dry Cask Storage and Heavy Load Associated withSpent Fuel Handling

4.2 Description of the Crane Upgrade to Single-Failure-Proof

4.3 Analysis

5.0 REGULATORY SAFETY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria

5.2 No Significant Hazards Consideration Determination

5.3 Environmental Consideration

6.0 PRECEDENCE

7.0 COORDINATION WITH PENDING PROPOSED LICENSE AMANEDMENTS

8.0 REFERENCES

ATTACHMENT 1Entergy Letter No. 2.13.042

Page 1 of 9

ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGE

1.0 DESCRIPTION

The proposed amendment revises Pilgrim Technical Specification (TS) 4.3.4, Heavy Loads, topermit certain operations needed for dry cask storage of spent nuclear fuel. Current wording ofTS 4.3.4 would prohibit travel of the lid for the spent fuel storage canister over irradiated fuel inthe canister. The proposed change would continue to prohibit travel of loads in excess of 2,000lbs over irradiated fuel assemblies in the spent fuel pool, but would permit heavy load handlingover irradiated fuel in the Multi-Purpose Canister (MPC) using a single-failure-proof handlingsystem. Section 3.0, Background, provides information related to the design features of thesingle-failure-proof handling system.

2.0 PROPOSED CHANGES

The current TS Section 4.0 Design Features specifies in Section 4.3.4, Heavy Loads, as follows:

4.3.4 Heavy Loads

a. Loads in excess of 2000 lb. shall be prohibited from travel over fuelassemblies in the spent fuel storage pool.

Entergy proposes the following changes:

4.3.4 Heavy Loads

a. Loads in excess of 2,000 lbs shall be prohibited from travel over fuelassemblies in the spent fuel storage pool with the exception thatheavy load handling over irradiated fuel in the Multi-Purpose Canisteris permitted using a single-failure-proof handling system.

3.0 BACKGROUND

Entergy has determined that based on current inventory of spent fuel in the pool and additionalspent fuel projected to be discharged into the pool during the remaining licensed life of the plant,the spent fuel pool capacity is not adequate to store all spent nuclear fuel assemblies until theend of the plant life, in 2032. Therefore, additional spent fuel storage space is required.Accordingly, Entergy has commenced plans to build an onsite Independent Spent Fuel StorageInstallation (ISFSI) for dry cask storage at PNPS using a General License issued in accordancewith 10 CFR 72.210. The ISFSI will be designed for storage of 40 casks, with each unit holding68 spent fuel assemblies. The proposed ISFSI capacity would provide space in the spent fuelpool for one full core off-load and to store projected discharged spent fuel assemblies until theend of the plant life, in 2032.

Entergy has selected Holtec International's (HOLTEC) HI-STORM 100 dry cask storage systemwith MPC for the Pilgrim ISFSI. For spent fuel operations, the MPC will be housed in the HI-TRAC transfer cask located in the spent fuel pool Cask Loading Area. Spent fuel assemblies will

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be moved using a single-failure-proof handling system. After the MPC has been loaded withspent fuel assemblies, the MPC lid will be placed to close the MPC. The MPC lid weighsapproximately 10,000 lbs. In addition, it will be necessary to use the HI-TRAC transfer cask liftyoke (and lift yoke extension, if required) lifting devices over spent fuel assemblies in the MPCduring dry cask loading operations. The weight of each of these items is in excess of 2,000 lbs.The gross weight that could travel over the spent fuel assemblies contained in the MPC wouldbe approximately 15,000 lbs while placing the lid to close the canister.

TS 4.3.4 currently prohibits loads in excess of 2,000 lbs traveling over fuel assemblies in thespent fuel pool. The proposed change to this Technical Specification would permit travel ofloads in excess of 2,000 lbs over a transfer cask containing irradiated fuel assemblies tofacilitate dry cask storage operations, using a single-failure-proof handling system. NUREG-0800 Section 9.1.5, Rev. 1 (Reference 8.9) describes a single-failure-proof handling system asconsisting of a crane designed to the criteria of NUREG-0554 (Reference 8.10), with liftingdevices selected to satisfy the requirements of ANSI N14.6 (Reference 8.11) or ASME B30.9metallic slings (Reference 8.12). The single failure-proof fuel handing system would comply withthese requirements. The revised Technical Specification would continue to prohibit travel ofloads in excess of 2,000 lbs over irradiated fuel assemblies in the remainder of the spent fuelpool, even if the load is carried by a single-failure-proof handling system.

Entergy is upgrading the existing Reactor Building crane to meet the single-failure-proofguidance of NUREG 0554 (Reference 8.10) and the NUREG 0612 (Reference 8.13) guidanceapplicable for the modification of an existing non-single-failure-proof crane. The replacementsingle-failure-proof main hoist and trolley are designed and qualified in accordance with theappropriate requirements of ASME NOG-I (Reference 8.14). All upgrade modifications will bemade prior to commencing dry cask storage operations. The single-failure-proof upgrade of theexisting Reactor Building crane will be made under the provisions of 10 CFR 50.59. The 100 toncapacity of the crane main hoist is not being changed.

The transfer cask lift yoke and lift yoke extension lifting devices are designed per ANSI N14.6(Reference 8.11) as prescribed in the HI-STORM 100 FSAR (Reference 8.3). When the MPC lidis connected to the lift yoke, and lift yoke extension if used, the slings that connect the lid to thelifting device shall be constructed of metallic wire rope and comply with the requirements ofASME B30.9 (Reference 8.12) and NUREG-0612 (Reference 8.13).

4.0 TECHNICAL ANALYSIS

4.1 HI-STORM 100 Dry Cask Storage and Heavy Load Associated with Spent Fuel Loading:

The spent fuel dry storage system selected for use for the PNPS ISFSI is the HI-STORM 100SVersion B MPC-68 dry cask storage system developed by Holtec. This is a canister-basedstorage system licensed by the NRC (Reference 8.4-8.8) for storage of spent nuclear fuel at anISFSI using the General License in accordance with 10 CFR 72.210. The system is comprisedof three primary components: MPC-68, HI-TRAC 100D, and HI-STORM 100S. The MPC-68 is aleak-tight metal canister that has a storage capacity of 68 BWR spent fuel assemblies. The HI-TRAC 1OOD (hereafter "transfer cask") is a metal transfer cask that provides a means to lift andhandle the canister as well as providing radiological shielding of the spent fuel assemblies. TheHI-STORM 1OOS Version B storage overpack is a steel-encased concrete storage cask that

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provides physical protection and radiological shielding for the metal canister when in storage.The storage cask is vented for natural convection cooling to dissipate the spent fuel decay heat.The casks are stored in a vertical position outdoors onsite on a storage pad at the ISFSI.

Loading the MPC metal canisters with spent fuel assemblies takes place underwater in thespent fuel pool Cask Loading Area. The MPC contained inside the transfer cask, would beloaded with spent fuel assemblies utilizing the Refuel Bridge fuel handling crane. Once the MPCis loaded with spent fuel assemblies, the MPC lid is placed on the canister using a single-failure-proof handling system. That system consists of the upgraded Reactor Building crane, thetransfer cask lift yoke (and the lift yoke extension if used) and metallic wire rope sling andshackle rigging arrangement. Immediately following MPC lid placement, the lift yoke is used toengage the HI-TRAC transfer cask lifting trunnions and move the transfer cask containing theloaded MPC to the Cask Decontamination Area, where the canister is welded shut, drained,dried, and backfilled with helium.

The HI-STORM 100 FSAR Section 2.0.3, HI-TRAC Transfer Cask Design Criteria, states: "Thelifting trunnions and associated attachments are designed in accordance with the requirementsof NUREG-0612 and ANSI N14.6 for non-redundant lifting devices". The HI-STORM 100 FSARSection 2.2.1,2, Handling, states: "Lifting attachments and special lifting devices shall meet therequirements of ANSI N14.6". The HI-STORM 100 FSAR Table 8.1.6, HI-STORM 100 SystemAncillary Equipment Operational Description, under "HI-TRAC Lift Yoke/Lifting Links", states:"Lift yoke and lifting devices for loaded HI-TRAC handling shall be provided in accordance withANSI N14.6." Section 8 of the HI-STORM 100 FSAR, "Operating Procedures" describes theprocedure for placement and removal of the MPC lid on the loaded MPC. This is done in thespent fuel pool following canister spent fuel loading operations, immediately prior to engagingthe lift yoke onto the upper trunnions of the transfer cask for lifting to the location where canistersealing operations will be performed. When the MPC lid is connected to the lift yoke duringtravel of the MPC lid over the loaded MPC canister, the rigging (shackles and metallic slings)connecting the lid to the lifting device shall meet the requirements of ASME B30.9 (Reference8.12) and NUREG-0612 (Reference 8.13). This, in combination with the upgraded crane,provides for a single-failure-proof handling system as identified in Section 9.1.5 of NUREG-0800(Reference 8.9), "Overhead Heavy Load Handling Systems", as discussed in Section 5.0,Regulatory Safety Analysis. This single-failure-proof handling system is also responsive to theguidance in Regulatory Issue Summary (RIS) 2005-25 (Reference 8.16), including RIS 2005-25,Supplement 1 (Reference 8.17). Due to the reliability of this handling system, a load dropaccident (i.e., drop of the lift yoke, lift yoke extension, and/or MPC lid) is not considered to be acredible event and is not evaluated.

4.2 Description of the Crane Upgrade to Single-Failure-Proof:

Entergy is modifying the Reactor Building crane under the provisions of 10 CFR 50.59. Thecrane and its operation are described in Sections 10.3, 12.2 and 12.4 of the PNPS UFSAR(Reference 8.1). The Reactor Building crane operates over the entire area of the Refuel Floor,including the spent fuel pool, and is designed to handle heavy loads up to its rating of 100 tons.

Entergy is upgrading the existing Reactor Building crane to meet the single-failure-proofguidance of NUREG 0554 (Reference 8.10) and the NUREG 0612 (Reference 8.13) guidanceapplicable for the modification of an existing non single-failure-proof crane. The replacementsingle-failure-proof main hoist and trolley are designed and qualified in accordance with the

ATTACHMENT 1Entergy Letter No. 2.13.042

Page 4 of 9

appropriate requirements of ASME NOG-I (Reference 8.14). The upgraded crane and trolleycan safely handle the HI-TRAC 1OOD transfer cask as specified in the HI-STORM 100 FSAR tosupport the dry cask storage operations. The crane control system is also being replaced sothat the operator has finer control of the main hoist, bridge and trolley movements in order to bemore precise with the cask movements.

4.3 Analysis

Technical Specification 4.3.4 currently prohibits loads in excess of 2,000 lbs from traveling overirradiated fuel assemblies in the spent fuel pool. The proposed change to this TechnicalSpecification would permit travel of loads in excess of 2,000 lbs over a transfer cask containingirradiated fuel assemblies to facilitate dry cask storage operations using a single-failure-proofcrane and lifting devices that comply with the applicable requirements of ANSI N14.6 and/orASME B30.9. The travel path is limited and controlled over the transfer cask and the travel pathdoes not extend over the remaining portion of the spent fuel pool. Usage of the single-failureproof main lifting hoist of the crane and lifting devices that comply with the applicablerequirements of ANSI N14.6 and/or ASME B30.9, ensure that the lifting system is sufficientlyreliable as to preclude drop of the MPC lid, transfer cask lift yoke or lift yoke extension, and theconsequences of such a drop need not be analyzed. No other dry cask storage lifts in excessof 2,000 lbs will be conducted over the remainder of the spent fuel pool over the spent fuelassemblies.

5.0 REGULATORY SAFETY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria

General Design Criterion (GDC) 4, Environmental and Dynamic Effects Design Bases ofAppendix A to 10 CFR Part 50 specifies, in part, that structures, systems, and componentsimportant to safety shall be appropriately protected against dynamic effects, including theeffects of missiles, that may result from equipment failures. GDC 2, Design Bases for ProtectionAgainst Natural Phenomena, specifies, in part, that structures, systems, and componentsimportant to safety shall be designed to withstand the effects of natural phenomena, such asearthquakes. Section 9.1.5 of NUREG-0800 (Reference 8.9), Overhead Heavy Load HandlingSystems, refers to the guidelines of NUREG-0612 for implementation of these criteria in thedesign of overhead heavy load handling systems.

In NUREG-0612 (Reference 8.13), "Control of Heavy Loads at Nuclear Power Plants", the NRCstaff provided regulatory guidelines for control of heavy load lifts to assure safe handling ofheavy loads in areas where a load drop could impact on stored spent fuel, fuel in the reactorcore, or equipment that may be required to achieve safe shutdown or permit continued decayheat removal. Section 5.1.1 of NUREG-0612 provides guidelines for reducing the likelihood ofdropping heavy loads and provides criteria for establishing safe load paths; procedures for loadhandling operations; training of crane operators; design, testing, inspection, and maintenance ofcranes and lifting devices; and analyses of the impact of heavy load drops. The guidelines inSections 5.1.2 through 5.1.6 address alternatives to either further reduce the probability of aload-handling accident or mitigate the consequences of heavy load drops. These alternativesinclude using a single-failure-proof crane for increased handling system reliability, employingelectrical interlocks and mechanical stops for restricting crane travel to safe areas, or performingload drop consequence analyses for assessing the impact of dropped loads on plant safety and

ATTACHMENT 1Entergy Letter No. 2.13.042

Page 5 of 9operations.

Guidelines for design of single-failure-proof cranes are included in NUREG-0554 (Reference8.10), "Single-Failure Proof Cranes for Nuclear Power Plants." Appendix C to NUREG-0612provides alternative guidance for upgrading the reliability of existing cranes to single-failure--proof standards. In Section 9.1.5 of NUREG-0800 (Reference 8.9), the NRC staff recognizescranes designed to the criteria for Type I cranes specified in ASME NOG-1 2004 (Reference8.14) as acceptable under the guidelines of NUREG-0554 for construction of a single-failure--proof crane. Paragraph 1.4.C of Section 9.1.5 of NUREG-0800 states the following:

"The probability for a load drop is minimized by an overhead handling system designed tocomply with the guidelines of NUREG-0554 and lifting devices that comply with AmericanNational Standards Institute (ANSI) N14.6 or an alternative based on American Society ofMechanical Engineers (ASME) B30.9. An overhead handling system that complies with ASMENOG-l criteria for Type 1 cranes is an acceptable method for compliance with the NUREG-0554guidelines."

Paragraph 111.4.C of Section 9.1.5 of NUREG-0800 states the following:

"The likelihood of failure is extremely low due to a single-failure-proof handling system.A single failure-proof handling system consists of the following two elements:

i. The crane should be designed to the criteria of NUREG-0554. Cranes designed to thecriteria of ASME NOG-I 2004 for a Type 1 crane are acceptable under the guidelines ofNUREG-0554 for construction of a single-failure-proof crane. Consistent with Paragraph10 of NUREG-0554, a quality assurance program should cover the procurement, design,fabrication, installation, inspection, testing, and operation of the crane. The programshould include at least the following elements: (1) design and procurement documentcontrol; (2) instructions, procedures, and drawings; (3) control of purchased material,equipment, and services; (4) inspection; (5) testing and test control; (6) non-conformingitems; (7) corrective action; and (8) records.

ii. The lifting devices 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 loadmovements in critical areas (reactor head lifting, reactor vessel internals, spent fuelcasks). The lifting device should have either dual, independent load paths or a singleload 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 metallicmaterial (chain or wire rope). The slings should be either (a) configured to provide dualor redundant load paths or (b) selected to support a load twice the weight of the handledload."

As discussed above, Entergy is in the process of upgrading the Reactor Building crane inconformance with the single-failure-proof guidelines of NUREG-0612 and NUREG-0554 tosupport commencement of dry cask storage operations. The single-failure-proof handlingsystem will permit the heavy load lifts required to perform dry cask storage operations to beperformed with design margins sufficient to preclude the necessity of postulating load dropaccidents and evaluating consequences.

ATTACHMENT 1Entergy Letter No. 2.13.042

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5.2 No Significant Hazards Consideration Determination

Entergy has evaluated whether or not a significant hazards consideration is involved with theproposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuanceof amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences

of an accident previously evaluated?

Response: No.

The Reactor Building crane is being upgraded to meet the applicable single-failure-proofcriteria of NUREG 0554 and NUREG 0612 for the modification of the existing non single-failure-proof crane. While loads in excess of 2,000 lbs shall continue to be prohibitedfrom travel over irradiated fuel assemblies in the spent fuel pool by the PNPS TechnicalSpecifications, an MPC lid will be permitted to travel over irradiated fuel assemblies in atransfer cask, using a single-failure-proof handling system as described in NUREG-0800Section 9.1.5 Paragraph ll1.4.C, to enable the conduct of dry cask storage loading andunloading operations. Specifically, this will enable the Multi-Purpose Canister (MPC) lidand its associated lifting apparatus to travel over irradiated fuel assemblies in a MPC.The probability of dropping this load onto an irradiated fuel assembly in the canister isreduced as a result of the reliability of the single-failure-proof handling system.

The proposed change does not affect the consequences of any accidents previouslyevaluated in the PNPS UFSAR. The change involves the travel of heavy loads overirradiated fuel assemblies in a transfer cask using a single-failure-proof handling system.Under these circumstances, no new load drop accidents are postulated and no changesto the probabilities or consequences of accidents previously evaluated are involved.

2. Does the proposed change create the possibility of a new or different kind of accident from

any accident previously evaluated?

Response: No.

Section 10.3 of the PNPS UFSAR evaluates fuel storage and handling operations.Section 14 of the PNPS UFSAR discusses the analysis of design basis fuel handlingaccidents involving drop of an irradiated assembly resulting in multiple fuel rod failuresand consequent release of radioactivity. The change involves the travel of heavy loadsover irradiated fuel assemblies in a transfer cask using a single-failure-proof handlingsystem. Under these circumstances, no new or different load drop accidents arepostulated to occur and there are no changes in any of the load drop accidentspreviously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

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The revised Technical Specification changes do not involve a reduction in any margin ofsafety. Technical Specification 4.3.4 currently prohibits travel of heavy loads in excess of2,000 lbs over irradiated fuel assemblies in the spent fuel pool. The proposed changewill continue to restrict travel of heavy loads in excess of 2,000 lbs over irradiated fuelassemblies in the spent fuel pool, with the exception of the MPC lid over irradiated fuelassemblies in the canister to enable dry cask storage operations. This exception is onlypermitted when the heavy load is handled using a single-failure-proof handling system.Due to the reliability of this upgraded handling system that complies with the guidance ofNUREG-0800 Section 9.1.5 for a single-failure-proof handling system, a load dropaccident is not considered a credible event. Under these circumstances, no new loaddrop accidents are postulated and no reductions in margins of safety are involved.

5.3 Environmental Consideration

Entergy review has determined that the proposed amendment would permit dry cask storageoperations by making provisions for loads in excess of 2,000 lbs to travel over irradiated fuelassemblies in a transfer cask using a single-failure-proof handling system. The proposedchanges do not involve (i) significant hazards consideration, (ii) any changes in the types or anyincrease in the amounts of any effluent that may be released offsite, or (iii) significant increasein individual or cumulative occupational radiation exposure. Accordingly, the proposedamendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement orenvironmental assessment needs to be prepared in connection with the proposed amendment.

6.0 PRECEDENCE

The NRC approved Waterford 3 License Amendment No. 227 (Reference 8.18). Pilgrimproposed license amendment follows the Waterford 3 License Amendment application andNRC acceptance letter (Reference 8.19).

The NRC approved a similar change to Technical Specifications for Kewaunee Power Station,who upgraded their 125 ton Auxiliary Building crane to a single-failure-proof design, in an NRCLicense Amendment and associated Safety Evaluation Report (SER) dated November 20,2008 (Reference 8.15).

7.0 COORDINATION WITH PENDING PROPOSED LICENSE AMANEDMENTS

At this time, there are no pending proposed license amendment requests requiring coordinationwith this proposed Technical Specification Change.

8.0 REFERENCES

1. Pilgrim Nuclear Power Plant, Updated Final Safety Analysis Report, Revision 28,October 2011.

2. Pilgrim Nuclear Power Plant, Technical Specification, Section 4.3.4

3. Final Safety Analysis Report for the Holtec International Storage and Transfer Operation

ATTACHMENT 1Entergy Letter No. 2.13.042

Page 8 of 9Reinforced Module Cask System (HI-STORM 100 Cask System), Holtec Report HI-2002444, Docket 72-1014, Revision 9, February 13, 2010.)

4. NRC Letter Amendment No. 7 to Certificate of Compliance No. 1014 for the HoltecInternational HI-STORM 100 Cask System, December 28, 2009. (ML093620052)

5. NRC Amendment No. 7 Certificate of Compliance No. 1014 for the Holtec InternationalHI-STORM 100 Cask System, December 28, 2009. (ML093620057)

6. NRC Amendment No. 7 Final Safety Evaluation Report Docket No. 72-1014Holtec International HI-STORM 100 Cask System Certificate of Compliance No.1014, December 28, 2009. (ML093620075)

7. NRC Amendment No. 7 Certificate of Compliance No. 1014 Appendix A TechnicalSpecifications for the HI-STORM 100 Cask System, December 28, 2009.(ML093620062)

8. NRC Amendment No. 7 Certificate of Compliance No. 1014 Appendix BApproved Contents and Design Features for the HI-STORM 100 Cask System,December 28, 2009. (ML093620068)

9. NUREG-0800 Section 9.1.5 Rev. 1, Standard Review Plan for Overhead HeavyLoad Handling Systems, March 2007. (ML062260190)

10. NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, U.S.Nuclear Regulatory Commission, May 1979.

11. ANSI N14.6, Radioactive Materials - Special Lifting Devices for ShippingContainers Weighing 10,000 Pounds (4500 kg) or More, American NationalStandards Institute, January 1993.

12. ASME B30.9, Slings, American Society of Mechanical Engineers, 2003.

13. NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, U.S. NuclearRegulatory Commission, July 1980. (ML070250180)

14. ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (TopRunning Bridge, Multiple Girder), American Society of Mechanical Engineers, 2004.

15. NRC Amendment Kewaunee Power Station - Issuance of Amendment to RelocateSpent Fuel Pool Crane Requirements from the Technical Specifications to theTechnical Requirements Manual, November 20, 2008. (ML082971079)

16. NRC Regulatory Issue Summary 2005-25: Clarification of NRC Guidelines for Controlof Heavy Loads, October 31, 2005. (ML052340485)

17. NRC Regulatory Issue Summary 2005-25, Supplement 1, Clarification of NRCGuidelines for Control of Heavy Loads, May 29, 2007. (ML071210434)

ATTACHMENT 1Entergy Letter No. 2.13.042

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18. Waterford 3 Steam Electric Station License Amendment 227, Modify TechnicalSpecification 3/4.9.7, "Crane Travel-Fuel Handling Building" (TAC NO. ME2221), datedSeptember 13, 2010

19. Acceptance Review Result for Waterford 3 LAR - "Modify TS 3/4.9.7, Crane Travel- FuelHandling Building," (TAC No. ME2221), dated October 15, 2009

ATTACHMENT 2

To Entergy Letter No. 2.13.042

Marked-Up TS Page

MARKED-UP CURRENT TS

4.3.4 Heavy Loads

a. Loads in ex-ess of 2000 lb. shall be prohibited from-, travol •ovor fuel

assemblies in the spent fuol Storago pool.

PROPSED TS CHANGE

4.3.4 Heavy Loads

a. Loads in excess of 2,000 lbs shall be prohibited from travel over fuelassemblies in the spent fuel storage pool with the exception thatheavy load handling over irradiated fuel in the Multi-Purpose Canisteris permitted using a single-failure-proof handling system.