ML14258A179

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

ML14258A179
Person / Time
Site:	Pilgrim
Issue date:	09/11/2014
From:	Dent J Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Morgan N
References
2.14.065
Download: ML14258A179 (14)

Text

S En t rgy Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 LETTER NUMBER: 2.14.065 John A.Dent, Jr.

Site Vice President September 11,2014 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

SUBJECT:

Supplement to Proposed License Amendment Request to Modify Technical Specification 4.3.4, "Heavy Loads" to Facilitate Dry Storage Handling Operations Entergy Nuclear Operations, Inc.

Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35

REFERENCES:

1. Entergy Letter to NRC "Proposed License Amendment Request to Modify Technical Specification 4.3.4, "Heavy Loads" to Facilitate Dry Storage Handling Operations, dated November 26, 2013 (2.13.042)

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

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby submits the attached supplement to the license amendment for Pilgrim Nuclear Power Station (PNPS) requested in Reference 1. The proposed supplemental amendment revises Technical Specification (TS) 4.3.4, "Heavy Loads" to reflect the removal of the energy absorbing pad from the spent fuel pool and installation of a leveling platform. 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.

Current wording of TS 4.3.4.b utilizes the energy absorbing pad as a point of reference to establish a limitation on fuel placement within an area in the spent fuel pool that could be potentially affected by a cask during cask handling operations. The proposed TS change would replace 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 when a cask is in the spent fuel pool. A single-failure-proof handling system will be used for the cask handling operations. The single-failure-proof handling system complies with the NRC guidance included in Reference 2.

Attachment 1 provides an analysis of the proposed Technical Specification change.

Attachment 2 provides a mark-up of the proposed change. ) ,

Letter 2.14.065 Page 2 of 2 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 October 27, 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.

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

If you have any questions regarding the subject matter, please contact Everett P.

Perkins at (508) 830-8323.

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

Executed on the _____/ dday of 22014.

Sincerely, John A. Dent, J ---

Site Vice President Attachment 1: Analysis of Proposed Technical Specification Change (10 pages).

Attachment 2: Marked-up Page of the Current Technical Specifications (1 page).

cc: Ms. Nadiyah Morgan, Project Manager Regional Administrator, Region 1 Division of Operating Reactor Licensing U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation 2100 Renaissance Blvd, Suite 100 U. S. Nuclear Regulatory Commission King of Prussia, PA 19406-2713 One White Flint North O-8C2-A 11555 Rockville Pike Rockville, MD 20852 John Giarrusso, Jr. John Priest, Director, Planning and Preparedness Section Chief Massachusetts Department of Public Health Mass Emergency Management Agency Radiation Control Program 400 Worcester Road Commonwealth of Massachusetts Framingham, MA 01702 529 Main Street, Suite 1M2A Charlestown, MA 02129-1121 NRC Resident Inspector Pilgrim Nuclear Power Station

Letter 2.14.065 Attachment 1 Page 1 of 10 ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGE TABLE OF CONTENTS

2.0 PROPOSED CHANGE

S

3.0 BACKGROUND

4.0 TECHNCIAL ANALYSIS 4.1 Description of the Crane Upgrade to Single-Failure-Proof 4.2 Description of the Energy Absorbing Pad and the Leveling Platform 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 AMENDMENTS

8.0 REFERENCES

Letter 2.14.065 Attachment 1 Page 2 of 10 ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGE

1.0 DESCRIPTION

The proposed supplemental amendment revises Technical Specification (TS) 4.3.4, "Heavy Loads" to reflect the removal of the energy absorbing pad and installation of a leveling platform. In addition, the requirement is being clarified to apply only when cask handling operations are in progress in the spent fuel pool using a single-failure-proof handling system.

2.0 PROPOSED CHANGE

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

4.3.4.b Heavy Loads 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.

Entergy proposes the following changes:

4.3.4.b Heavy Loads 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.

Letter 2.14.065 Attachment 1 Page 3 of 10

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 is designed for storage of 40 casks, with each unit holding 68 spent fuel assemblies. The ISFSI capacity provides 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 a Multi-Purpose Canister (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. The cask will 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.

Entergy upgraded 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-l (Reference 8.14). The single-failure-proof upgrade of the existing Reactor Building crane was made under the provisions of 10 CFR 50.59. The 100 ton capacity of the crane main hoist was not 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, the slings that connect the lid to the lifting device are constructed of metallic wire rope and comply with the requirements of ASME B30.9 (Reference 8.12) and NUREG-0612 (Reference 8.13).

To accommodate the Holtec system, the energy absorbing pad was removed and replaced with the leveling platform. Since a single-failure-proof handling system will be used for the cask handling operations, the energy absorbing pad is not required. The single-failure-proof handling system complies with the NRC guidance included in Reference 8.13.

Letter 2.14.065 Attachment 1 Page 4 of 10

4.0 TECHNICAL ANALYSIS

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

Entergy modified 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 modified 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-l (Reference 8.14). The upgraded crane and trolley can safely handle the HI-TRAC 100D transfer cask as specified in the HI-STORM 100 FSAR to support the dry cask storage operations. The crane control system was also 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.

4.2 Description of the Energy Absorbing Pad and the Leveling Platform 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 leveling platform provides a stable level platform for placement of the HI-TRAC 100D transfer cask to support the dry cask fuel storage operations. The transfer cask placement on the leveling platform was evaluated to confirm it's stability during the design basis seismic event.

4.3 Analysis Current wording of TS 4.3.4.b utilizes the energy absorbing pad as a point of reference to establish a limitation on fuel placement within an area in the spent fuel pool that could be potentially affected by a cask during cask handling operations. The proposed TS change would replace the energy absorbing pad point of reference with a leveling platform point of reference. 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. In addition, the requirement is being clarified to apply only when cask handling operations are in

Letter 2.14.065 Attachment 1 Page 5 of 10 progress in the spent fuel pool or there is a cask in the pool. Since a single-failure-proof handling system will be used for the cask handling operations, the energy absorbing pad is not required. The single-failure-proof handling system complies with the NRC guidance included in Reference 8.13.

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 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:

Letter 2.14.065 Attachment 1 Page 6 of 10 "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-I 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 has upgraded 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.

Letter 2.14.065 Attachment 1 Page 7 of 10 NUREG-0612 evaluation of offsite release potential due to load drop accidents shows that a decay time of 44 days for Boiling Water Reactors that exhaust through charcoal filters (I.e. the Standby Gas Treatment System) will assure that offsite releases, due to dropping of postulated heavy loads on fuel which has been subcritical for the required decay time, will not cause doses that approach 10CFR1 00 limits.

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 has been 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 UFSAR. 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 cask 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 spent 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

Letter 2.14.065 Attachment 1 Page 8 of 10 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 proposed 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 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. 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.

5.3 Environmental Consideration NUREG-0612 evaluation of offsite release potential due to load. drop accidents shows that a decay time of 44 days for Boiling Water Reactors that exhaust through charcoal filters (I.e. the Standby Gas Treatment System) will assure that offsite releases, due to dropping of postulated heavy loads on fuel which has been subcritical for the required decay time, will not cause doses that approach 10CFRI1O limits. Since the limitation on placement of spent fuel that has decayed for less than 200 days is being maintained, sufficient margin to the IOCFR1OO limits is demonstrated. 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). The Pilgrim proposed license amendment'follows the Waterford 3 License Amendment application and NRC acceptance letter (Reference 8.19).

Letter 2.14.065 Attachment 1 Page 9 of 10 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 AMENDMENTS This proposed amendment request supplements the change to TS 4.3.4.a submitted in Reference 8.20. At this time, there are no other pending proposed license amendment requests requiring coordination with this proposed Technical Specification Change.

8.0 REFERENCES

8.1 Pilgrim Nuclear Power Station, Updated Final Safety Analysis Report, Rev 29.

8.2 Pilgrim Nuclear Power Plant, Technical Specification, Section 4.3.4 8.3 Final Safety Analysis Report for the Holtec International Storage and Transfer Operation Reinforced Module Cask System (HI-STORM 100 Cask System),

Holtec Report HI-2002444, Docket 72-1014, Revision 9, February 13, 2010.)

8.4 NRC Letter Amendment No. 7 to Certificate of Compliance No. 1014 for the Holtec International HI-STORM 100 Cask System, December 28, 2009.

(ML093620052) 8.5 NRC Amendment No. 7 Certificate of Compliance No. 1014 for the Holtec International HI-STORM 100 Cask System, December 28, 2009.

(ML093620057) 8.6 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) 8.7 NRC Amendment No. 7 Certificate of Compliance No. 1014 Appendix A Technical Specifications for the HI-STORM 100 Cask System, December 28, 2009. (ML093620062) 8.8 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) 8.9 NUREG-0800 Section 9.1.5 Rev. 1, Standard Review Plan for Overhead Heavy Load Handling Systems, March 2007. (ML062260190) 8.10 NUREG-0554, Single-Failure-Proof Cranes for Nuclear Power Plants, U.S.

Nuclear Regulatory Commission, May 1979.

Letter 2.14.065 Attachment 1 Page 10 of 10 8.11 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.

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

8.13 NUREG-0612, Control of Heavy Loads at Nuclear Power Plants, U.S. Nuclear Regulatory Commission, July 1980. (ML070250180) 8.14 ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (Top Running Bridge, Multiple Girder), American Society of Mechanical Engineers, 2004.

8.15 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) 8.16 NRC Regulatory Issue Summary 2005-25: Clarification of NRC Guidelines for Control of Heavy Loads, October 31, 2005. (ML052340485) 8.17 NRC Regulatory Issue Summary 2005-25, Supplement 1, Clarification of NRC Guidelines for Control of Heavy Loads, May 29, 2007. (ML071210434) 8.18 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 8.19 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 8.20 Entergy Letter to NRC "Proposed License Amendment Request to Modify Technical Specification 4.3.4, "Heavy Loads" to Facilitate Dry Storage Handling Operations, dated November 26, 2013 (2.13.042)

ATTACHMENT 2 To Entergy Letter No. 2.14.065 Marked-up Page of the Current Technical Specifications (1 Page)

Letter 2.14.065 Attachment 2 Page 1 of 1 Marked-up Page of the Current Technical Specifications MARKED UP OF CURRENT TS (Deleted words are shown by strikethrough. Added words are underlined) 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 usinq 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 energy abSorbing pad leveling platform during cask handling operations in the spent fuel pool or when a cask is in the spent fuel pool.

Note: The proposed revision to TS 4.3.4.a was previously submitted by Entergy Letter to NRC 2.13.042, dated November 26, 2013