ML14295A685

Issuance of Amendment Regarding Heavy Loads to Facilitate Dry Storage Handling Operations

ML14295A685
Person / Time
Site:	Pilgrim
Issue date:	10/31/2014
From:	Nadiyah Morgan Plant Licensing Branch 1
To:	Dent J Entergy Nuclear Operations
Morgan N
References
TAC MF3237
Download: ML14295A685 (27)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 31, 2014 Mr. John A. Dent, Jr.

Site Vice President Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360

SUBJECT:

PILGRIM NUCLEAR POWER STATION -ISSUANCE OF AMENDMENT REGARDING HEAVY LOADS TO FACILITATE DRY STORAGE HANDLING OPERATIONS (TAC NO. MF3237)

Dear Mr. Dent:

The Commission has issued the enclosed Amendment No. 240 to Renewed Facility Operating License No. DPR-35 for the Pilgrim Nuclear Power Station. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 26, 2013, as supplemented by letters dated July 11, September 11, October 3, and October 16, 2014.

This amendment revises TS 4.3.4, "Heavy Loads," by modifying the limit imposed on the maximum weight that could travel over the irradiated fuel in the spent fuel pool. The amendment also revises TS 4.3.4 to reflect the removal of the energy absorbing pad from the spent fuel pool and installation of a leveling platform.

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

N diyah S. Morgan, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-293

Enclosures:

1. Amendment No. 240 to License No. DPR-35

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

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

PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 240 License No. DPR-35

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

A. The application for amendment filed by Entergy Nuclear Operations, Inc. (the licensee) dated November 26, 2013, as supplemented by letters dated July 11, September 11, October 3, and October 16, 2014, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations; 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. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 3.B. of Renewed Facility Operating License No. DPR-35 is hereby amended to read as follows:

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 240, are hereby incorporated in the renewed operating license. The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of issuance and shall be implemented prior to the start of the dry cask storage operations.

FOR THE NUCLEAR REGULATORY COMMISSION Benjamin G. Beasley, Chief Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: October 31 , 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 240 RENEWED FACILITY OPERATING LICENSE NO. DPR-35 DOCKET NO. 50-293 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert 3 3 Replace the following page of the Appendix A Technical Specifications with the attached revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change.

Remove Insert 4.0-2 4.0-2

provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

A. Maximum Power Level ENO is authorized to operate the facility at steady state power levels not to exceed 2028 megawatts thermal.

B. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 240, are hereby incorporated in the renewed operating license.

The licensee shall operate the facility in accordance with the Technical Specifications.

C. Records ENO shall keep facility operating records in accordance with the requirements of the Technical Specifications.

D. Equalizer Valve Restriction - DELETED E. Recirculation Loop Inoperable- DELETED F. Fire Protection ENO shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SER dated December 21, 1978 as supplemented subject to the following provision:

ENO may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

G. Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Pilgrim Nuclear Power Station Physical Security, Training and Qualification, and Safeguards Contingency Plan, Revision 0" submitted by letter dated October 13, 2004, as supplemented by letter dated May 15, 2006.

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The licensee's CSP was approved by License Amendment No. 236, as supplemented by a change approved by Amendment No. 238.

Amendment 240 Renewed License No. DPR-35

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) 4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Ke11 5;0.95 if fully flooded with water, which includes an allowance for uncertainties as described in Section 10.2.5 of the FSAR;

b. Ke11 5;0.90 when dry, which includes an allowance for uncertainties as described in Section 10.2.5 of the FSAR; and

c. A nominal 6.60 inch center to center distance between fuel assemblies placed in storage racks.

4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 115 ft.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 3859 fuel assemblies.

4.3.4 Heavy Loads

a. Loads in excess of 2000 lb. shall be prohibited from travel over fuel assemblies in the spent fuel storage pool with the exception that heavy load handling over irradiated fuel in the Multi-Purpose Canister is permitted using a single-failure-proof handling system.

b. No fuel which has decayed for less than 200 days shall be stored in racks within an arc described by the height of the cask around the periphery of the leveling platform during cask handling operations in the spent fuel pool or when a cask is in the spent fuel pool.

PNPSTS 4.0-2 Amendment No. 4+7, 240

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• • • • "'" SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 240 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-35 ENTERGY NUCLEAR GENERATION COMPANY ENTERGY NUCLEAR OPERATIONS. INC.

PILGRIM NUCLEAR POWER STATION DOCKET NO. 50-293

1.0 INTRODUCTION

By application dated November 26, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13346A026), as supplemented by letters dated July 11 (ADAMS Accession No. ML14237A328), September 11 (ADAMS Accession No.

ML14258A179), October 3 (ADAMS Accession No. ML14280A230), and October 16, 2014 (ADAMS Accession No. ML14293A063), Entergy Nuclear Operations, Inc., the licensee, submitted a request to the Nuclear Regulatory Commission (NRC) for changes to the Pilgrim Nuclear Power Station (Pilgrim) Technical Specifications (TSs). Portions of the letter dated October 3, 2014, contain sensitive unclassified non-safeguards information (proprietary) and, accordingly, have been withheld from public disclosure.

The supplement dated July 11, 2014, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on July 22, 2014 (79 FR 42545).

The supplement dated September 11, 2014, expanded the scope of the application as originally noticed and, therefore, the September 11, 2014, supplement was noticed in the Federal Register on September 22, 2014 (79 FR 56608). The supplements dated October 3 and October 16, 2014, provided additional information that clarified the September 11, 2014, supplement, did not expand the scope as noticed, and did not change the NRC staffs proposed no significant hazards consideration determination as published in the Federal Register on September 22, 2014 (79 FR 56608).

The proposed changes would revise the TSs associated with the limitation imposed on the maximum weight that could travel over the irradiated fuel in the spent fuel pool (SFP); and the removal of the energy absorbing pad from the SFP and installation of a leveling platform.

Enclosure 2

2.0 BACKGROUND

The licensee determined that additional spent fuel storage space is required for the remaining licensed life of Pilgrim. Therefore, the licensee "has commenced plans to build an onsite Independent Spent Fuel Storage Installation (ISFSI) for dry cask storage at Pilgrim using a General License issued in accordance with [Title 10 of the Code of Federal Regulations (10 CFR) Section] 72.210 ... The ISFSI capacity [would] provide space in the SFP for one full core off-load and to store projected discharged spent fuel assemblies until the end of

[operations of] the plant life, in 2032."

3.0 REGULATORY EVALUATION

The construction permit for Pilgrim was issued by the Atomic Energy Commission (AEC) on August 26, 1968, a low-power license was issued on June 8, 1972, and a full-power license was issued on September 15, 1972. The plant design approval for the construction phase was based on the proposed General Design Criteria (GDC) published by the AEC in the Federal Register (32 FR 10213) on July 11, 1967 (hereinafter referred to as "draft GDC"). The AEC published the final rule that added Appendix A to 10 CFR Part 50, "General Design*criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971 (hereinafter referred to as "final GDC").

Differences between the draft GDC and final GDC included a consolidation from 70 to 64 criteria. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J.

M. Taylor, "SECY-92-223- Resolution of Deviations Identified during the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, this includes Pilgrim. The Pilgrim Updated Final Safety Analysis Report (UFSAR), Appendix F, provides an evaluation of the design bases of Pilgrim against the draft GDC.

Although the original approval basis for Pilgrim was the draft GDC, the licensees for Pilgrim have made changes to the facility over the life of the plant that may have invoked some of the final GDC. The extent to which the final GDC have been invoked can be found in specific sections of the UFSAR and in other Pilgrim design and licensing basis documentation. For convenience, the licensee and the NRC staff usually refer to the final GDC rather than the draft GDC when discussing licensing actions.

3.1 Description of System The ISFSI will be designed for storage of 40 casks, with each unit holding 68 spent fuel assemblies. The licensee selected the Holtec International (Holtec) HI-STORM 1OOS Version B Multi-Purpose Canister (MPC)-68 dry cask storage system for the Pilgrim ISFSI. In its letter dated November 26, 2013, the licensee stated, in part, that:

The [HI-STORM dry cask storage] system is comprised of three primary components: MPC-68, HI-TRAC 100D, and HI-STORM 100S. The MPC-68 is a leak-tight metal canister that has a storage capacity of 68 BWR [boiling water

reactor] spent fuel assemblies. The HI-TRAG 1000 [transfer cask] is a metal transfer cask that provides a means to lift and handle the canister as well as providing radiological shielding of the spent fuel assemblies. The HI-STORM 1OOS Version B storage overpack is a steel-encased concrete storage cask that provides physical protection and radiological shielding for the canister when in storage.

In its letter dated September 11, 2014, the licensee stated, in part, that:

To accommodate the Holtec system, the energy absorbing pad was removed

[from the SFP] and replaced with the leveling platform.

3.2 Proposed TS Changes The current TS 4.3.4, "Heavy Loads," states:

4.3.4 Heavy Loads

a. Loads in excess of 2,000 lbs shall be prohibited from travel over fuel assemblies in the spent fuel storage pool.

b. No fuel which has decayed for less than 200 days shall be stored in racks within an arc described by the height of the cask around the periphery of the energy absorbing pad.

The licensee proposed to modify the specification as follows:

4.3.4 Heavy Loads

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

b. No fuel which has decayed for less than 200 days shall be stored in racks within an arc described by the height of the cask around the periphery of the leveling platform during cask handling operations in the spent fuel pool or when a cask is in the spent fuel pool.

3.3 Regulatory Requirements and Guidance Section 182a of the Atomic Energy Act (the "Act") requires applicants for nuclear power plant operating licenses to include TSs as part of the license application. The Commission's regulatory requirements related to the content of the TS are contained in Section 50.36(c) of Title 10 of the Code of Federal Regulations (1 0 CFR). That regulation requires that the TSs include, among other things, items in the following categories: (1) safety limits, limiting safety

systems settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls.

The regulation at 10 CFR 50.36(c)(4) requires the TSs to include facility design features such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in other categories described in 10 CFR 50.36(c).

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

The GDC 4, "Environmental and dynamic effects design bases," of Appendix A to 10 CFR Part 50 specifies, in part, that SSCs important to safety shall be appropriately protected against dynamic effects, including the effects of missiles, that may result from equipment failures.

The NRC requested licensees to address control of heavy load movements in 1980. The NRC staff provided regulatory guidelines to support this action in NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic Technical Activity A-36," dated July 1980 (ADAMS Accession No. ML070250180). Implementation of these guidelines assures safe handling of heavy loads in areas where a load drop could impact on stored spent fuel, fuel in the reactor core, or equipment that may be required to achieve safe shutdown or permit continued decay heat removal. Section 5.1.1 of NUREG-0612 provides guidelines for reducing the likelihood of dropping heavy loads and provides 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 analyses of the impact of heavy load drops.

The guidelines in Sections 5.1.2 through 5.1.5 of NUREG-0612 address alternatives to either further reduce the probability of a load-handling accident or mitigate the consequences of heavy load drops in specific areas of the facility. These alternatives include using a single-failure-proof crane for increased handling system reliability, employing electrical interlocks and mechanical stops to restrict crane travel to safe areas, or performing load drop consequence analyses to assess the effect of dropped loads on plant safety and operations. Section 5.1.6 of NUREG-0612 specifically addresses measures to further reduce the probability of a load handling accident through installation and operation of highly reliable load handling system.

These measures include use of a single-failure-proof crane to improve reliability through increased factors of safety and through redundancy or duality in certain active components.

Criteria for design of single-failure-proof cranes are included in NUREG-0554, "Single-Failure-Proof Cranes for Nuclear Power Plants," dated May 1979 (ADAMS Accession No.

ML110450636). Appendix C to NUREG-0612 provided alternative criteria for certain design features to satisfy the intent of NUREG-0554 guidance when upgrading existing cranes to single-failure-proof criteria.

In Revision 1 to Section 9.1.5, "Overhead Heavy Load Handling System," of NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants:

LWR Edition" (SRP), the NRC staff provided review guidelines for control of heavy load handling activities. For heavy load handling activities over irradiated fuel, Section 9.1.5 of the SRP

specified the use of a single-failure-proof handling system in conjunction with general NRC guidelines for design, operation, testing, maintenance, and inspection of heavy load handling systems. The guidelines specify that a single-failure-proof handling system should consist of the following elements:

• The crane should be designed to the criteria of NUREG-0554. Cranes designed to the criteria for a Type I crane specified in ASME [American Society of Mechanical Engineers] NOG-1-2004, "Rules for Construction of Overhead and Gantry Cranes," are acceptable under the guidelines of NUREG-0554 for construction of a single-failure-proof crane.

• Special lifting devices that satisfy the criteria of American National Standards Institute (ANSI) N14.6-1993, "Radioactive Materials- Special Lifting Devices for Shipping Containers Weighing 10 000 Pounds (4500 kg) or More," 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.

• Slings should satisfy the criteria of ASME B30.9-2003, "Slings," and be constructed of metal (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 Regulatory Issue Summary 2005-25, Supplement 1, "Clarification of NRC Guidelines for Control of Heavy Loads," dated October 31, 2005 (ADAMS Accession No. ML071210434), the NRC staff described operating experience that indicated single operational errors resulted in synthetic round sling failures. The NRC staff considered this condition incompatible with the intent of single-failure-proof handling systems. In addition, operating experience suggested that metallic slings resist similar load handling errors. Therefore, the NRC staff modified the guidelines in Section 9.1.5 of the SRP to specify use of metal slings for single-failure-proof handling systems.

4.0 TECHNICAL EVALUATION

4.1 MPC Loading Process In its letter dated November 26, 2013, the licensee stated, in part, that:

Loading the MPC[-68] metal canisters with spent fuel assemblies takes place underwater in the spent fuel pool Cask Loading Area. The MPC contained inside the transfer cask is loaded with spent fuel assemblies utilizing the Refuel Bridge fuel handling crane. Once the MPC is 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, the transfer cask lift yoke (and the lift yoke extension, if used) and metallic wire rope sling and

shackle rigging arrangement. Immediately following MPC lid placement, the lift yoke is used to engage the HI-TRAC transfer cask lifting trunnions and move the transfer cask containing the loaded MPC to the Cask Decontamination Area, where the canister is welded shut, drained, dried, and backfilled with helium.

Further, the licensee stated, in part, that, The MPC lid weighs approximately 10,000 lbs. In addition, it will be necessary to use the HI-TRAC transfer cask lift yoke (and lift yoke extension, if required) lifting devices over spent fuel assemblies in the MPC during dry cask loading operations.

4.2 Load Handling System Design and Operation The Pilgrim heavy load handling program is described in Section 12.2.3.7 of the Pilgrim UFSAR.

This program established administrative controls addressing safe load paths, safe load handling procedures, crane operator training, standards for lifting devices and cranes, and special requirements when handling heavy loads in areas where fuel or safe shutdown equipment could be damaged. In a safety evaluation dated March 6, 1985 (ADAMS Legacy Accession No. 8503220131 ), the NRC staff found the program acceptable with respect to satisfying the general guidance in Section 5.1.1 of NUREG-0612.

The licensee stated that the existing reactor building overhead crane would be modified under the provisions of 10 CFR 50.59, which allows changes to the facility under certain conditions without specific NRC review and approval. The 100-ton rated reactor building crane operates over the entire area of the refueling operating floor, including the SFP. The modifications would retain the 100-ton load rating while upgrading the crane to meet the single-failure-proof guidance of NUREG-0554, as modified by guidance applicable to upgrading of existing cranes to single-failure-proof provided in Appendix C to NUREG-0612. The modifications also include a replacement single-failure-proof main hoist and trolley designed and qualified to the appropriate criteria specified in ASME NOG-1-2004. A crane properly constructed to the above criteria would satisfy NRC guidelines for a single-failure-proof crane.

In its letter dated November 26, 2013, the licensee provided the following statements from the Holtec HI-STORM 100 FSAR, summarizing the design and operation of the special lifting devices related to placement of the MPC lid:

The HI-STORM 100 FSAR Section 2.0.3, HI-TRAC Transfer Cask Design Criteria, states: "The lifting trunnions and associated attachments are designed in accordance with the requirements of NUREG-0612 and ANSI N14.6 for non-redundant lifting devices". The HI-STORM 100 FSAR Section 2.2.1.2, Handling, states: "Lifting attachments and special lifting devices shall meet the requirements of ANSI N14.6". The HI-STORM 100 FSAR Table 8.1.6, HI-STORM 100 System Ancillary 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 with ANSI N14.6." Section 8

of the HI-STORM 100 FSAR, "Operating Procedures" describes the procedure for placement and removal of the MPC lid on the loaded MPC ....

By e-mail dated May 30, 2014 (ADAMS Accession No. ML14234A006), the NRC staff issued a request for additional information (RAI) to support its review. In RAI-1, the NRC staff asked the licensee to verify that the lifting devices will 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. In its July 11, 2014, response letter, the licensee stated that the lift yoke used in the single-failure-proof handling system is designed with twice the normal stress design safety factor, as specified by Section 6.2.1 of ANSI N14.6 for the bounding condition of a loaded HI-TRAC cask. The NRC staff finds that the licensee's response is acceptable.

In RAI-3 and RAI-4, the NRC staff asked the licensee to confirm that the runway crane supporting structure, and bridge with the new trolley, including qualification of the existing bridge, will be seismically qualified to the requirements of ASME NOG-1-2004, and to confirm that the new trolley and the existing bridge will be adequate to support the Maximum Critical Load rating during a seismic event. In its response, the licensee stated that the existing crane bridge was upgraded by evaluation to NUREG-0554, supplemented by NUREG-0612, Appendix C, and is capable of sustaining a 100-ton load on the Main Hoist hook during design basis seismic event. The licensee also stated that analytic confirmation is contained in calculations prepared by Holtec International's subcontractor American Crane & Equipment Corporation. The licensee provided a sketch of the layout showing the arrangements for the Safe Load Path, per NUREG-0612. The NRC staff finds that the licensee's response is acceptable.

In RAI-5, the NRC staff asked the licensee to provide the weights of the old trolley, the new trolley, the lift yoke, the lift yoke extension, and the slings. In its response dated July 11, 2014, the licensee provided reference to the crane drawings, which listed the weight of the old trolley as 66,000 pounds (lbs) and the weight of the new trolley as 90,000 lbs. The licensee stated, in part, that, The total weight of the Lift Yoke Ancillary indicated on Holtec Drawing 7941 R2 is 3,644 lb. inclusive of the lift yoke, main pin shackles and slings. A lift yoke extension device will not be used at Pilgrim Station.

The NRC staff finds that the licensee's response is acceptable.

Therefore, the design and operation of the special lifting devices would satisfy NRC guidance and be acceptable for use as part of a single-failure-proof handling system.

The licensee also described the design and operation of the lifting device used for the MPC lid.

The travel path of the MPC lid would be controlled over the transfer cask and MPC, such that, the MPC lid would not extend over the remainder of the SFP. The rigging connecting the MPC lid to the lift yoke, during travel of the MPC lid over the loaded MPC canister (i.e., shackles and metallic slings), would be selected to meet the guidance of ASME 830.9 and NUREG-0612.

In RAI-1, the NRC staff also asked the licensee to verify that the slings are 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 its response, the licensee stated, in part, that:

[T]he ASME B30.9 slings that connect the MPC Lid to the Lift Yoke use wire rope and other metallic rigging components and meet the same criteria [Section 6.2.1 of ANSI N14.6] with four load paths, collectively designed to support a load twice the weight of the handled load."

The NRC staff finds that the licensee's response is acceptable.

Therefore, the use of slings, in the placement of the MPC lid over the loaded MPC, would be consistent with NRC guidance for single-failure-proof handling systems.

4.3 Cask Leveling Platform In its supplement dated September 11, 2014, the licensee described the purpose of the energy absorbing pad and the justification for its removal as follows:

The energy absorbing pad was installed to protect the liner of the spent fuel pool (SFP) from damage due to the drop of a spent fuel cask or other heavy loads in the SFP cask loading area using the original non-single-failure-proof Reactor Building crane. The use of the single-failure-proof hoist on the upgraded Reactor Building crane as part of a single-failure-proof handling system to handle heavy loads in the cask loading area of the SFP precludes the need to postulate a transfer cask load drop.

The licensee stated that the leveling platform provided a stable surface for support of the HI-TRAC 1000 transfer cask to support the dry cask fuel storage operations. Also, the licensee stated that the cask placement on the leveling platform had been evaluated to ensure its stability during the design-basis seismic event.

By letter dated September 26, 2014 (ADAMS Accession No. ML14269A063), the NRC staff issued another RAI. In RAI 1, 2, 4, and 5, the NRC staff asked the licensee to (1) provide a discussion of the details of the leveling platform such as the design, materials of construction, layout, and operational characteristics; (2) describe the technical evaluation performed to confirm the stability of the transfer cask placement on the leveling platform during a design-basis seismic event; (3) provide a discussion of the protections in-place for the SFP liner given the removal of the energy absorbing pad; and (4) to describe any evaluations completed to demonstrate that liner integrity would be maintained during cask loading operations.

In its October 3, 2014, response letter, the licensee stated that the leveling platform protects the SFP liner plate in the cask storage area. The structure is comprised of a 2-inch stainless steel plate, supported by six adjustable stainless steel pedestals with 2-inch thick bearing pads, which rest on the 1-inch thick SFP liner plate in the cask storage area. The bearing pads distribute and transfer cask loadings (dead and seismic) to the SFP concrete structure. The licensee

provided an analysis demonstrating a large factor of safety for punching shear failure from the peak vertical cask loading during a Safe Shutdown Earthquake.

The licensee also provided plant drawings showing that three vertical monitoring trenches had been formed in each of the four concrete walls of the SFP behind the liner plate. The vertical trenches communicated with a horizontal monitoring trench running along the perimeter of the floor at the intersection with the wall. The drawing for the SFP floor liner did not indicate the presence of other monitoring trenches in the central portion of the pool that could be affected by the leveling platform bearing pads. Therefore, the NRC staff has determined that the cask leveling platform satisfied the design function to protect the liner from cask handling operations.

5.0 FINAL NO SIGNIFICANT HAZARDS CONSIDERATION

DETERMINATION The regulations at 10 CFR 50.92 state that the Commission may make a final determination that a license amendment involves no significant hazards consideration, if operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

The Commission may issue the license amendment before the expiration of the 60-day hearing period provided that its final determination is that the amendment involves no significant hazards consideration. This amendment is being issued prior to the expiration of the 60-day period. Therefore, a final finding of no significant hazards consideration follows.

The Commission has made a final determination that the license amendment involves no significant hazards consideration. Under the Commission's regulations in 10 CFR 50.92, this means that operation of the facility, in accordance with the proposed amendment does not:

(1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety. As required by 10 CFR 50.91 (a), in its initial application dated November 26, 2013, the licensee provided its analysis of the issue of no significant hazards consideration, which is presented 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-proof criteria 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 [pounds] shall continue to be prohibited from travel over irradiated fuel assemblies in the spent fuel pool by the PNPS [Pilgrim Nuclear Power Station] Technical Specifications, an MPC lid will be permitted to travel over irradiated fuel assemblies in a transfer cask, using a single-failure-proof handling system as described in NUREG-0800

Section 9.1.5 Paragraph 111.4.C, to enable the conduct of dry cask storage loading and unloading operations. Specifically, this will enable the Multi-Purpose Canister (MPC) lid and 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 is reduced as a result of the reliability of the single-failure-proof handling system.

The proposed change does not affect the consequences of any accidents previously evaluated in the PNPS UFSAR [Updated Final Safety Analysis Report]. The change involves the travel of heavy loads over irradiated 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 changes to 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 handling accidents involving drop of an irradiated assembly resulting in multiple fuel rod failures and consequent release of radioactivity. The change involves the travel of heavy loads over irradiated fuel assemblies in a transfer cask using a single-failure-proof handling system. Under these circumstances, no new or different load drop accidents are postulated to occur and there are no changes in any of the load drop accidents previously evaluated.

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

Response: No.

The revised Technical Specification changes do not involve a reduction in any margin of safety. Technical Specification 4.3.4 currently prohibits travel of heavy loads in excess of 2,000 lbs over irradiated fuel assemblies in the spent fuel pool. The proposed change will continue to restrict travel of heavy loads in excess of 2,000 lbs over irradiated fuel assemblies in the spent fuel pool, with the exception of the MPC lid over irradiated fuel assemblies in the canister to enable dry cask storage operations. This exception is only permitted 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 of NUREG-0800 Section 9.1.5 for a single-failure-proof handling system, a load drop accident is not considered a credible event. Under

these circumstances, no new load drop accidents are postulated and no reductions in margins of safety are involved.

By supplemental letter dated September 11, 2014, the licensee expanded the scope of its original application, and provided its analysis for the expanded scope of the issue of no significant hazards consideration, which is presented 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-proof criteria of NUREG-0554 and NUREG-0612 for the modification of the existing non single-failure-proof crane.

The proposed change does not affect the consequences of any accidents previously evaluated in the PNPS [Pilgrim Nuclear Power Station] UFSAR

[Updated Final Safety Analysis Report]. The proposed change replaces the energy absorbing pad point of reference with a leveling platform point of reference. In addition, the requirement is being clarified to apply only when cask handling operations are in progress in the spent fuel pool or a cask is in the spent fuel pool. The requirement to limit placement of spent fuel that has decayed for less than 200 days in racks within an arc described by the height of the cask around the periphery of the point of reference is being maintained. Under these circumstances, no new load drop accidents are postulated and no changes to the probabilities or consequences of accidents previously evaluated are involved.

The single-failure proof handling system used for handling operations precludes the need to postulate a transfer cask load drop.

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 cask handling operations. Consequences of a dropped fuel cask are described in Section 10.3.6. The proposed change replaces the energy absorbing pad point of reference with a leveling platform point of reference. Under these circumstances, no new or different load drop accidents are postulated to occur and there are no changes in any of the load drop accidents previously evaluated.

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

Response: No.

The revised Technical Specification change does not involve a reduction in any margin of safety. The proposed change replaces the energy absorbing pad point of reference with a leveling platform point of reference. In addition, the requirement is being clarified to apply only when cask handling operations are in progress in the spent fuel pool.

The requirement to limit placement of spent fuel that has decayed for less than 200 days in racks within an arc described by the height of the cask around the periphery of the point of reference is being maintained. Due to the reliability of the upgraded handling system that complies with the guidance of NUREG-0800 Section 9.1.5 for a single-failure-proof handling system, a load drop accident with a transfer cask is not considered a credible event. Under these circumstances, no new load drop accidents are postulated and no reductions in margins of safety are involved.

Therefore, the proposed changes do not involve a significant reduction in a margin of safety.

Based on the above evaluations, the NRC staff has concluded that the three standards of 10 CFR 50.92(c) are satisfied. Therefore, the NRC staff has made a final determination that no significant hazards consideration is involved for the proposed amendment and that the amendment should be issued as allowed by the criteria contained in 10 CFR 50.91.

6.0 PUBLIC COMMENTS On July 22, 2014, as supplemented on September 22, 2014, the NRC staff published a notice of consideration of issuance of an amendment to the Pilgrim license in the Federal Register (79 FR 42545) and (79 FR 56608). In accordance with the requirements in 10 CFR 50.91, the notice provided a 30-day period for public comment on the proposed NSHC determination and the opportunity to request a hearing and petition for leave to intervene. Public comments were received. A summary of the comments and the NRC staff responses are provided below.

6.1 Public Participation Public Comment The NRC's plan is to "issue the license amendment before expiration of the 60-day notice period," "prior to the expiration of the 30-day comment period," and certainly in time to insure that Entergy can move spent fuel assemblies without violating its license any more than it already has.

NRC Response The Commission may issue a license amendment before the expiration of the 60-day hearing period provided that its final determination is that the amendment involves no significant hazards consideration, as discussed above in Section 5.0 of the safety evaluation (SE).

Processing a license amendment under exigent circumstances allows a reduced period for public comment. The subject license amendment was not processed under exigent circumstances, and therefore, the normal 30-day period (September 26- October 26, 2014) for public comment was permitted.

6.2 Proprietary Information Public Comment Entergy's response to NRC's September 26, 2014 RAI withholds necessary information by hiding behind a claim of "Proprietary Information." ... This affidavit does not comply with 10 CFR 2.390 in that it fails to identify the official position of the person making the affidavit. ... Because Holtec/Entergy has NOT complied with regulatory requirements for disclosures, [public commenters] assume that NRC will make ALL the information publicly available or REJECT the bogus affidavit.

NRC Response Although there was a minor omission in the September 26, 2014, Holtec affidavit, the NRC staff would not release documents with information that could potentially contain trade secrets and commercial or financial information. By letter dated October 16, 2014, the licensee submitted a revised affidavit. As stated in its October 28, 2014 (ADAMS Accession No. ML14297A529),

letter, the NRC staff reviewed the document and determined that the information contained proprietary commercial information and should be withheld from public disclosure.

6.3 Seismic Event Analysis Public Comment

[The Public Commenters] believe that because the required post-Fukushima reassessment of Pilgrim's vulnerability to earthquakes will not be completed until 2017, it is premature to grant approval.

NRC Response The Commission's Near-Term Task Force (NTTF) found that U.S. nuclear power plants are safe and will remain safe under even difficult circumstances brought on by natural disasters.

However, the Japan Lessons Learned Directorate plans, develops, and implements the actions found necessary to enhance the safety of power plants, such as evaluations of beyond-design-basis events.

The licensee's current TSs change request was evaluated against its current licensed design-basis seismic event.

6.4 Single-Failure-Proof Upgrades Public Comment

• Was the existing crane bridge evaluated by NRC to single failure proof standards? If so how? Was there, for example, cold proof testing? What were the results of NRC's crane bridge evaluations?

• What critical welds were tested (including those in bridge girders and truck structure)? How were they tested and what were the results?

• What assumptions were made in the analysis?

• Was the proper updated seismic risk used in the analysis?

NRC Response As discussed in Attachment 1 to Reference 3, the licensee stated that the reactor building crane had been modified to a single-failure-proof design under the provisions of 10 CFR 50.59. This change control process allows licensees to make changes to the facility without prior NRC approval, provided certain conditions specified in the regulation are satisfied. The NRC verifies correct implementation of this regulation and the associated change through the inspection program, which uses a sampling approach. Section 40A5.2 of NRC Integrated Inspection Report 05000293/2014003 and ISFSI Report 07201044/2014002 (ADAMS Accession Number ML14224A067), issued August 11, 2014, included the above issues within the inspection scope and stated that no findings were identified.

6.5 Leveling Platform Public Comment There is no information regarding the thickness of the pad that was removed and the space available under the newly installed platform. Some Mark I BWR reactors have transferred assemblies to casks for some time. In those reactors, how many have removed the pads and how many have pads; what is the thickness of those pads; and how many also have platforms?

NRC Response Section 10.3.6 of the Pilgrim FSAR, Revision 27, described the energy absorbing pad as a 3-foot thick aluminum honeycomb core enclosed in a watertight stainless steel box, which was mounted on a high strength steel load distribution plate. The energy absorber was designed to limit the damage to the SFP. The NRC guidelines allow either a load drop analysis that demonstrates limited consequences of a load drop or the use of a single-failure-proof handling system as acceptable methods to demonstrate reasonable assurance of safety. Pilgrim, like all

other operators of ISFSis at BWR sites, chose to upgrade the reactor building cask handling system (the crane and associated lifting devices) to satisfy single-failure-proof criteria rather than perform a load drop analysis for the planned movement of a fuel storage canister and associated transfer overpack. The use of a single-failure-proof handling system reduces the frequency of a load drop through the use of redundancy, independent safety devices, and enhanced design margins, such that a drop is no longer considered a credible event. However, a leveling platform is typically used to protect the pool liner from potential damage due to normal operational loads.

6.6 Factor of Safety Public Comment What does a "large factor of safety" mean and what assumptions went into that determination?

NRC Response The factor of safety refers to the punching shear stress acting on the reinforced concrete slab of the SFP assuming the maximum vertical seismic load on the leveling platform is equally divided across two of the six bearing pads (i.e., the leveling platform was treated as rigid and rocking on two of six feet).

6.7 Acts of Malice Public Commit Part of the single-failure proof crane requirements is periodic inspections. Unless security is provided from the most recent inspection until the next, there is reduced assurance that someone tampering with the crane/hoist/rigging will not be detected.

NRC Response Nuclear power plants are protected by physical barriers, armed guards, intrusion detection systems, area surveillance systems, access controls, and access authorization requirements for employees working inside of the plants. Physical Protection, Safeguards Contingency, and Training and Qualification Plans are submitted to NRC for approval with license applications, as required by NRC's regulations in 10 CFR Part 73. The NRC's routine inspections of power reactor security include security evaluations of the licensee's ability to protect the plant from the design basis threats of radiological sabotage, theft, and diversion.

6.8 Human Error Public Comment There also is no information on how, or if, human error in operations and manufacturing is accounted for in Entergy's analysis of the single failure proof crane upgrades or of the removal of the energy absorbing pad and installation of a leveling platform.

NRC Response As stated in this SE, Section 5.1.1 of NUREG-0612 provides guidelines for reducing the likelihood of dropping heavy loads and provides 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 analyses of the impact of heavy load drops.

The Pilgrim heavy load handling program is described in Section 12.2.3.7 of the Pilgrim UFSAR.

In a SEdated March 6, 1985 (ADAMS Legacy Accession No. 8503220131), the NRC staff found the program acceptable with respect to satisfying the general guidance in Section 5.1.1 of NUREG-0612.

6. 9 Load Drop Accident Public Comment

a. If the load drop resulted in a spent fuel pool fire, is it not clear how there can be charcoal filtering?

b. Were release consequences of a load drop accident that does not involve exhaust through charcoal filters, analyzed, what were the results; and if not analyzed, what conceivably could be the rationale?

c. Were load drop accidents analyzed that, as RIS 2005-25 explains, penetrate the floor and "could simultaneously initiate an accident and disable equipment necessary to mitigate the accident?" What were the results; and if not analyzed, why not?

d. Entergy quoted from NUREG-0612; but we believe that is a 1980 document. Please explain how its 34 year old conclusions remain applicable to Pilgrim's spent fuel pool inventory in 2014.

e. The attachments to the license amendment say "The probability of dropping this load onto an irradiated assembly is reduced as a result of the single-failure-proof failure system." (ML13346A025, pg., 6) It does not explain how much it is reduced and the assumptions that went into the calculation.

NRC Response The NRC staff established guidelines for the control of heavy loads in NUREG-0612 in a manner that provides defense-in-depth. The guidelines accomplish that through a combination of administrative controls, design, and analyses that minimize the potential for a load handling system failure to occur; minimize the potential consequences of a load handling system failure, should one occur; and verify through analyses that the consequences of a handling system failure are acceptable. The measures that minimize the potential for a handling system failure include design of the crane and lifting devices, training of operators, operating procedures, and maintenance, inspection, and testing. Measures that minimize the consequences of a handling system failure include establishment of safe load paths and lift heights in operating procedures. The consequence analyses of handling system failures are either analyses that demonstrate a single-failure-proof handling system can withstand design-basis events and single handling system failures while maintaining control of the lifted load or analyses of a dropped load to ensure radiological consequences would be small, fuel would remain safely subcritical, damage to the reactor vessel or SFP would not cause leakage that would uncover the fuel, and damage to other equipment would not cause a loss of an essential safe shutdown function. While NRC guidelines have been updated, the NUREG-0612 guidelines remain safe, effective, and are part of the Pilgrim operations described in the FSAR.

The Pilgrim FSAR included analysis of a transportation cask drop in the SFP to support transfer of radioactive material from the SFP using the previous non-single-failure-proof reactor building crane. The consequence analyses for this event included credit for the energy absorbing pad to limit damage to the SFP liner such that water inventory would not be lost, evaluation of impact of a tipping cask on fuel to ensure the radiological consequences would be small, and verification that the fuel would remain safely subcritical. The evaluation of the tipping cask resulted in the Pilgrim TS restriction on fuel age near the cask handling area. Since a separate analysis would ensure the pool retains water, the analysis of radiological consequences may credit filtration by safety-related ventilation systems if the system is required to be operable during cask handling operations. Pilgrim has revised the analyses to credit a single-failure-proof handling system to protect the fuel, the SFP, and essential equipment. Nevertheless, the licensee has proposed to retain the TS restriction on the age of fuel near the cask handling area during cask movements.

6.10 Shutdown during transfer operations Public Comment If the license amendment is granted [the Public Commenters] request that NRC require the reactor to be shutdown during transfer operations.

NRC Response The licensee is updating the existing crane to a single-failure-proof crane. Crane upgrades and plant procedures assure that fuel and cask handling operations will be performed safely. There are no new accidents created nor are there increases in consequences of postulated accidents.

Thus, there is no basis to require that the reactor be shutdown, nor would there be a substantial

increase in safety if it were shutdown. In the past, the NRC has not required any licensee to shutdown during dry storage cask operations and it is not written within the NRC regulatory requirements.

7.0 STATE CONSULTATION

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

8.0 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 made a final finding that the amendment involves no significant hazards consideration. 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.

9.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) the amendments do not (a) involve a significant increase in the probability or consequences of an accident previously evaluated; or (b) create the possibility of a new or different kind of accident from any accident previously evaluated; or (c) involve a significant reduction in a margin of safety; (2) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (3) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (4) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

10.0 REFERENCES

1. Dent, J. A., Jr., Entergy Nuclear Operations, letter to U.S. Nuclear Regulatory Commission, "Proposed License Amendment Request to Modify Technical Specification 4.3.4, 'Heavy Loads' to Facilitate Dry Storage Handling Operations," dated November 26, 2013 (ADAMS Accession No. ML13346A026).

2. Dent, J. A., Jr., Entergy Nuclear Operations, letter to U.S. Nuclear Regulatory Commission, "Entergy Response to NRC Request for Additional Information (RAI),

Pilgrim Proposed License Amendment Request to Modify Technical Specification 4.3.4,

'Heavy Loads' to Facilitate Dry Storage Handling Operations," dated July 11, 2014 (ADAMS Accession No. ML14237A328).

3. Dent, J. A., Jr., Entergy Nuclear Operations, letter to U.S. Nuclear Regulatory Commission, "Supplement to Proposed License Amendment Request to Modify Technical Specification 4.3.4, 'Heavy Loads' to Facilitate Dry Storage Handling Operations," dated September 11, 2014 (ADAMS Accession No. ML14258A179).

4. Dent, J. A., Jr., Entergy Nuclear Operations, letter to U.S. Nuclear Regulatory Commission, "Entergy Response to NRC Request for Additional Information (RAI),

Regarding the Heavy Loads License Amendment Request (TAC NO. MF3237)," dated October 3, 2014 (ADAMS Accession No. ML14280A230).

5. Final Safety Analysis Report for the Holtec International Storage and Transfer Operation Reinforced Module Cask System (HI-STORM 100 Cask System), Holtec Report Hl-2002444, Docket 72-1014, Revision 9, February 13, 2010.

6. U.S. Nuclear Regulatory Commission, NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, July 1980 (ADAMS Accession No. ML070250180).

7. U.S. Nuclear Regulatory Commission, NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Section 9.1.5, Rev. 1, Standard Review Plan for Overhead Heavy Load Handling Systems, March 2007 (ADAMS Accession No. ML062260190).

8. American National Standards Institute, ANSI N14.6, Radioactive Materials- Special Lifting devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More, January 1993.

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

10. U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2005-25, "Clarification of NRC Guidelines for Control of Heavy Loads," October 31, 2005 (ADAMS Accession No. ML052340485).

11. U.S. Nuclear Regulatory Commission, Regulatory Issue Summary 2005-25, Supplement 1, "Clarification of NRC Guidelines for Control of Heavy Loads," May 29, 2007 (ADAMS Accession No. ML071210434).

12. U.S. Nuclear Regulatory Commission, NUREG-0554, Single-Failure Proof Cranes for Nuclear Power Plants, May 1979. (ADAMS Accession No. ML110450636).

13. Vassallo, D. B., U.S. Nuclear Regulatory Commission, letter and safety evaluation to W. D. Harrington, Boston Edison Company related to Review of Control of Heavy Loads (Phase 1), dated March 6, 1985 (ADAMS Legacy Accession Nos. 8503220127 and 8503220131 ).

Principal Contributors: S. Jones R. Pettis Date: October 31 , 2014

October 31, 2014 Mr. John A. Dent, Jr.

Site Vice President Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360

SUBJECT:

PILGRIM NUCLEAR POWER STATION- ISSUANCE OF AMENDMENT REGARDING HEAVY LOADS TO FACILITATE DRY STORAGE HANDLING OPERATIONS (TAC NO. MF3237)

Dear Mr. Dent:

The Commission has issued the enclosed Amendment No. 240 to Renewed Facility Operating License No. DPR-35 for the Pilgrim Nuclear Power Station. This amendment consists of changes to the Technical Specifications (TSs) in response to your application dated November 26, 2013, as supplemented by letters dated July 11, September 11, October 3, and October 16, 2014.

This amendment revises TS 4.3.4, "Heavy Loads," by modifying the limit imposed on the maximum weight that could travel over the irradiated fuel in the spent fuel pool. The amendment also revises TS 4.3.4 to reflect the removal of the energy absorbing pad from the spent fuel pool and installation of a leveling platform.

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

Sincerely, IRA/

Nadiyah S. Morgan, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-293

Enclosures:

1. Amendment No. 240 to License No. DPR-35

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

See next page Amendment No.: ML14295A685 *See dated memo OFFICE NRR/DORLILPL 1-1/PM NRR/DORL/LPL 1-1/LA NRR/DSS/SBPB/BC NRR/DE/EMCB/BC NAME NMorgan KGoldstein JBurkhardt for GCasto Tlupold DATE 10/24/2014 10/24/2014 10/17/2014 10/22/2014 OFFICE NRR/DSS/STSB/BC OGC NRR/DORLILPL 1-1/BC NRR/DORLILPL 1-1/PM NAME REIIiott EWilliamson BBeasley NMorgan DATE 10/23/2014 10/30/2014 10/31/2014 10/31/2014 OFFICIAL RECORD COPY