ML13346A026

Proposed License Amendment Request to Modify Technical Specification 4.3.4, Heavy Loads to Facilitate Dry Storage Handling Operations

ML13346A026
Person / Time
Site:	Pilgrim
Issue date:	11/26/2013
From:	Dent J Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME2221
Download: ML13346A026 (15)

Text

we Entergy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 November 26, 2013 John A. Dent, Jr.

Site Vice President U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35 Proposed License Amendment Request to Modify Technical Specification 4.3.4, "Heavy Loads" to Facilitate Dry Storage Handling Operations

REFERENCES:

1. Final Safety Analysis Report for the Holtec International Storage and Transfer Operation 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 Overhead Heavy Load Handling Systems, March 2007. (ML062260190)

4. ANSI N14.6, Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4500 kg) or More, American National Standards 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 for Control of Heavy Loads, October 31, 2005. (ML052340485)

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

8. Waterford 3 Steam Electric Station License Amendment 227, Modify Technical Specification 3/4.9.7, "Crane Travel-Fuel Handling Building" (TAC NO. ME2221), dated September 13, 2010

9. NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, May 1979 LETTER NUMBER: 2.13.042

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests the following amendment for Pilgrim Nuclear Power Station (PNPS). The proposed amendment revises Technical Specification (TS) 4.3.4, "Heavy Loads" limitation imposed on maximum weight that could travel over the irradiated fuel in the spent fuel pool. The proposed revision is associated with the Independent Spent Fuel Storage Installation (ISFSI) activity of loading of spent fuel assemblies 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 fuel assemblies in the spent fuel pool. Dry storage cask operations involve loading irradiated fuel assemblies into an MPC in the spent fuel pool, and then lifting the canister lid over those fuel assemblies to permit installation onto the canister. The MPC lid weighs approximately 10,000 lbs. The proposed TS change would permit travel of loads in excess of 2,000 lbs over the loaded 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 prohibit travel of heavy loads in excess of 2,000 lbs over irradiated fuel assemblies in other areas of the spent fuel pool.

The single-failure-proof handling system would comply with the NRC guidance included in Reference 3. provides an analysis of the proposed Technical Specification change. Attachment 2 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 criteria specified in 10 CFR 50.92(c), and it has been determined that this change involves no significant 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 of the dry cask storage operations necessary to store spent fuel at an onsite ISFSI. Once approved, the amendment shall be implemented prior to the start of the dry cask storage operations.

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

cc: Ms. Nadiyah Morgan, Project Manager Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North O-8C2-A 11555 Rockville Pike Rockville, MD 20852

Enterav Letter No. 2.13.042. Paae 3 of 3 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100 King of Prussia, PA 19406-2713 NRC Resident Inspector Pilgrim Nuclear Power Station John Giarrusso, Jr.

Planning and Preparedness Section Chief Mass Emergency Management Agency (MEMA) 400 Worcester Road Framingham, MA 01702 Beverly Anderson, Acting Director, Massachusetts Department of Public Health (MDPH)

Radiation Control Program Commonwealth of Massachusetts 529 Main Street, Suite 1M2A Charlestown, MA 02129-1121

ATTACHMENT 1 To Entergy Letter No. 2.13.042 Proposed License Amendment Request to Modify Technical Specification 4.3.4, "Heavy Loads" to Facilitate Dry Storagqe Handling Operations Analysis of Proposed Technical Specification Change 9 Pages

ATTACHMENT 1 To Enterqv Letter No. 2.13.042 ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGE

2.0 PROPOSED CHANGE

S

3.0 BACKGROUND

4.0 TECHNCIAL ANALYSIS 4.1 HI-STORM 100 Dry Cask Storage and Heavy Load Associated with Spent 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 1 Entergy 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, to permit certain operations needed for dry cask storage of spent nuclear fuel. Current wording of TS 4.3.4 would prohibit travel of the lid for the spent fuel storage canister over irradiated fuel in the canister. The proposed change would continue to prohibit travel of loads in excess of 2,000 lbs over irradiated fuel assemblies in the spent fuel pool, but would permit heavy load handling over irradiated fuel in the Multi-Purpose Canister (MPC) using a single-failure-proof handling system. Section 3.0, Background, provides information related to the design features of the single-failure-proof handling system.

2.0 PROPOSED CHANGE

S 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 fuel assemblies 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 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.

3.0 BACKGROUND

Entergy has determined that based on current inventory of spent fuel in the pool and additional spent 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 the end 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 Storage Installation (ISFSI) for dry cask storage at PNPS using a General License issued in accordance with 10 CFR 72.210. The ISFSI will be designed for storage of 40 casks, with each unit holding 68 spent fuel assemblies. The proposed ISFSI capacity would provide space in the spent fuel pool for one full core off-load and to store projected discharged spent fuel assemblies until the end of the plant life, in 2032.

Entergy has selected Holtec International's (HOLTEC) HI-STORM 100 dry cask storage system with 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

ATTACHMENT 1 Entergy Letter No. 2.13.042 Page 2 of 9 be moved using a single-failure-proof handling system. After the MPC has been loaded with spent fuel assemblies, the MPC lid will be placed to close the MPC. 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. 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 would be 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 the spent fuel pool. The proposed change to this Technical Specification would permit travel of loads in excess of 2,000 lbs over a transfer cask containing irradiated fuel assemblies to facilitate 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 as consisting of a crane designed to the criteria of NUREG-0554 (Reference 8.10), with lifting devices selected to satisfy the requirements of ANSI N14.6 (Reference 8.11) or ASME B30.9 metallic slings (Reference 8.12). The single failure-proof fuel handing system would comply with these requirements. The revised Technical Specification would continue to prohibit travel of loads in excess of 2,000 lbs over irradiated fuel assemblies in the remainder of the spent fuel pool, 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-proof guidance of NUREG 0554 (Reference 8.10) and the NUREG 0612 (Reference 8.13) guidance applicable for the modification of an existing non-single-failure-proof crane. The replacement single-failure-proof main hoist and trolley are designed and qualified in accordance with the appropriate requirements of ASME NOG-I (Reference 8.14). All upgrade modifications will be made prior to commencing dry cask storage operations. The single-failure-proof upgrade of the existing Reactor Building crane will be made under the provisions of 10 CFR 50.59. The 100 ton capacity 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 lid is connected to the lift yoke, and lift yoke extension if used, the slings that connect the lid to the lifting device shall be constructed of metallic wire rope and comply with the requirements of ASME 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 100S Version B MPC-68 dry cask storage system developed by Holtec. This is a canister-based storage system licensed by the NRC (Reference 8.4-8.8) for storage of spent nuclear fuel at an ISFSI using the General License in accordance with 10 CFR 72.210. The 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 spent fuel assemblies. The HI-TRAC 1OOD (hereafter "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

ATTACHMENT 1 Entergy Letter No. 2.13.042 Page 3 of 9 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 the spent fuel pool Cask Loading Area. The MPC contained inside the transfer cask, would be 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.

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. This is done in the spent fuel pool following canister spent fuel loading operations, immediately prior to engaging the lift yoke onto the upper trunnions of the transfer cask for lifting to the location where canister sealing operations will be performed. When the MPC lid is connected to the lift yoke during travel 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 (Reference 8.12) 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 the guidance 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 drop accident (i.e., drop of the lift yoke, lift yoke extension, and/or MPC lid) is not considered to be a credible 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. The crane 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-proof guidance of NUREG 0554 (Reference 8.10) and the NUREG 0612 (Reference 8.13) guidance applicable for the modification of an existing non single-failure-proof crane. The replacement single-failure-proof main hoist and trolley are designed and qualified in accordance with the

ATTACHMENT 1 Entergy Letter No. 2.13.042 Page 4 of 9 appropriate requirements of ASME NOG-I (Reference 8.14). The upgraded crane and trolley can safely handle the HI-TRAC 1OOD transfer cask as specified in the HI-STORM 100 FSAR to support the dry cask storage operations. The crane control system is also being replaced so that the operator has finer control of the main hoist, bridge and trolley movements in order to be more precise with the cask movements.

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

5.0 REGULATORY SAFETY ANALYSIS 5.1 Applicable Regulatory Requirements/Criteria General Design Criterion (GDC) 4, Environmental and Dynamic Effects Design Bases of Appendix A to 10 CFR Part 50 specifies, in part, that structures, systems, and components important to safety shall be appropriately protected against dynamic effects, including the effects of missiles, that may result from equipment failures. GDC 2, Design Bases for Protection Against Natural Phenomena, specifies, in part, that structures, systems, and components important to safety shall be designed to withstand the effects of natural phenomena, such as earthquakes. Section 9.1.5 of NUREG-0800 (Reference 8.9), Overhead Heavy Load Handling Systems, refers to the guidelines of NUREG-0612 for implementation of these criteria in the design of overhead heavy load handling systems.

In NUREG-0612 (Reference 8.13), "Control of Heavy Loads at Nuclear Power Plants", the NRC staff provided regulatory guidelines for control of heavy load lifts to assure 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 5.1.6 address alternatives to either further reduce the probability of a load-handling accident or mitigate the consequences of heavy load drops. These alternatives include using a single-failure-proof crane for increased handling system reliability, employing electrical interlocks and mechanical stops for restricting crane travel to safe areas, or performing load drop consequence analyses for assessing the impact of dropped loads on plant safety and

ATTACHMENT 1 Entergy Letter No. 2.13.042 Page 5 of 9 operations.

Guidelines for design of single-failure-proof cranes are included in NUREG-0554 (Reference 8.10), "Single-Failure Proof Cranes for Nuclear Power Plants." Appendix C to NUREG-0612 provides 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 recognizes cranes designed to the criteria for Type I cranes specified in ASME NOG-1 2004 (Reference 8.14) 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 to comply with the guidelines of NUREG-0554 and lifting devices that comply with American National Standards Institute (ANSI) N14.6 or an alternative based on American Society of Mechanical Engineers (ASME) B30.9. An overhead handling system that complies with ASME NOG-l criteria for Type 1 cranes is an acceptable method for compliance with the NUREG-0554 guidelines."

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 the criteria of ASME NOG-I 2004 for a Type 1 crane are acceptable under the guidelines of NUREG-0554 for construction of a single-failure-proof crane. Consistent with Paragraph 10 of NUREG-0554, a quality assurance program should cover the procurement, design, fabrication, installation, inspection, testing, and operation of the crane. The program should include at least the following elements: (1) design and procurement document control; (2) instructions, procedures, and drawings; (3) control of purchased material, equipment, and services; (4) inspection; (5) testing and test control; (6) non-conforming items; (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 load movements in critical areas (reactor head lifting, reactor vessel internals, spent fuel casks). The lifting device should have either dual, independent load paths or a single load path with twice the design safety factor specified by ANSI N14.6 for the load.

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

As discussed above, Entergy is in the process of upgrading the Reactor Building crane in conformance with the single-failure-proof guidelines of NUREG-0612 and NUREG-0554 to support commencement of dry cask storage operations. The single-failure-proof handling system will permit the heavy load lifts required to perform dry cask storage operations to be performed with design margins sufficient to preclude the necessity of postulating load drop accidents and evaluating consequences.

ATTACHMENT 1 Entergy Letter No. 2.13.042 Page 6 of 9 5.2 No Significant Hazards Consideration Determination Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of 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-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 shall continue to be prohibited from travel over irradiated fuel assemblies in the spent fuel pool by the PNPS 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 ll1.4.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. 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.

ATTACHMENT 1 Entergy Letter No. 2.13.042 Page 7 of 9 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.

5.3 Environmental Consideration Entergy review has determined that the proposed amendment would permit dry cask storage operations by making provisions for loads in excess of 2,000 lbs to travel over irradiated fuel assemblies in a transfer cask using a single-failure-proof handling system. The proposed changes do not involve (i) significant hazards consideration, (ii) any changes in the types or any increase in the amounts of any effluent that may be released offsite, or (iii) significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment 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 or environmental 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). Pilgrim proposed license amendment follows the Waterford 3 License Amendment application and NRC 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 NRC License 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 coordination with 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 1 Entergy Letter No. 2.13.042 Page 8 of 9 Reinforced 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 Holtec International HI-STORM 100 Cask System, December 28, 2009. (ML093620052)

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

6. NRC Amendment No. 7 Final Safety Evaluation Report Docket No. 72-1014 Holtec 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 Technical Specifications for the HI-STORM 100 Cask System, December 28, 2009.

(ML093620062)

8. NRC Amendment No. 7 Certificate of Compliance No. 1014 Appendix B Approved 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 Heavy Load 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 Shipping Containers Weighing 10,000 Pounds (4500 kg) or More, American National Standards 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. Nuclear Regulatory Commission, July 1980. (ML070250180)

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

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

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

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

ATTACHMENT 1 Entergy Letter No. 2.13.042 Page 9 of 9

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

19. Acceptance Review Result for Waterford 3 LAR - "Modify TS 3/4.9.7, Crane Travel- Fuel Handling 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 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.