ML14258A179
| ML14258A179 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 09/11/2014 |
| From: | Dent J A Entergy Nuclear Operations |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Morgan N S | |
| References | |
| 2.14.065 | |
| Download: ML14258A179 (14) | |
Text
S En t rgy Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power Station600 Rocky Hill RoadPlymouth, MA 02360LETTER NUMBER: 2.14.065 John A. Dent, Jr.Site Vice President September 11,2014U.S. Nuclear Regulatory Commission Attn: Document Control DeskWashington, DC 20555-0001
SUBJECT:
Supplement to Proposed License Amendment Request to ModifyTechnical Specification 4.3.4, "Heavy Loads" to Facilitate Dry StorageHandling Operations Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power StationDocket No. 50-293License No. DPR-35
REFERENCES:
- 1. Entergy Letter to NRC "Proposed License Amendment Request toModify Technical Specification 4.3.4, "Heavy Loads" to Facilitate DryStorage 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 theattached supplement to the license amendment for Pilgrim Nuclear Power Station(PNPS) requested in Reference
- 1. The proposed supplemental amendment revisesTechnical Specification (TS) 4.3.4, "Heavy Loads" to reflect the removal of the energyabsorbing pad from the spent fuel pool and installation of a leveling platform.
Theproposed 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 couldbe potentially affected by a cask during cask handling operations.
The proposed TSchange would replace the energy absorbing pad point of reference with a levelingplatform point of reference.
In addition, the requirement is being clarified to apply onlywhen cask handling operations are in progress in the spent fuel pool or when a cask isin the spent fuel pool. A single-failure-proof handling system will be used for the caskhandling operations.
The single-failure-proof handling system complies with the NRCguidance 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 2.14.065Page 2 of 2The 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 changeinvolves no significant hazards.
The bases for this determination are included in theattached submittal.
Entergy requests approval of the proposed amendment by October 27, 2014, in supportof 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 caskstorage 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 2 2014.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 ManagerDivision of Operating Reactor Licensing Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission One White Flint North O-8C2-A11555 Rockville PikeRockville, MD 20852John Giarrusso, Jr.Planning and Preparedness Section ChiefMass Emergency Management Agency400 Worcester RoadFramingham, MA 01702Regional Administrator, Region 1U.S. Nuclear Regulatory Commission 2100 Renaissance Blvd, Suite 100King of Prussia, PA 19406-2713 John Priest, Director, Massachusetts Department of Public HealthRadiation Control ProgramCommonwealth of Massachusetts 529 Main Street, Suite 1M2ACharlestown, MA 02129-1121 NRC Resident Inspector Pilgrim Nuclear Power Station Letter 2.14.065Attachment 1Page 1 of 10ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGETABLE OF CONTENTS1.0 DESCRIPTION 2.0 PROPOSED CHANGES3.0 BACKGROUND 4.0 TECHNCIAL ANALYSIS4.1 Description of the Crane Upgrade to Single-Failure-Proof 4.2 Description of the Energy Absorbing Pad and the Leveling Platform4.3 Analysis5.0 REGULATORY SAFETY ANALYSIS5.1 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 2.14.065Attachment 1Page 2 of 10ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGE1.0 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 aleveling platform.
In addition, the requirement is being clarified to apply only when caskhandling operations are in progress in the spent fuel pool using a single-failure-proof handling system.2.0 PROPOSED CHANGEThe current TS Section 4.0 Design Features specifies in Section 4.3.4.b, Heavy Loads,as follows:4.3.4.b Heavy LoadsNo fuel which has decayed for less than 200 days shall be stored in racks within anarc described by the height of the cask around the periphery of the energyabsorbing pad.Entergy proposes the following changes:4.3.4.b Heavy LoadsNo fuel which has decayed for less than 200 days shall be stored in racks within anarc described by the height of the cask around the periphery of the levelingplatform during cask handling operations in the spent fuel pool or when a cask is inthe spent fuel pool.
Letter 2.14.065Attachment 1Page 3 of 103.0 BACKGROUND Entergy has determined that based on current inventory of spent fuel in the pool andadditional 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 spentnuclear fuel assemblies until the end of the plant life, in 2032. Therefore, additional spent fuel storage space is required.
Accordingly, Entergy has commenced plans tobuild an onsite Independent Spent Fuel Storage Installation (ISFSI) for dry cask storageat PNPS using a General License issued in accordance with 10 CFR 72.210. TheISFSI is designed for storage of 40 casks, with each unit holding 68 spent fuelassemblies.
The ISFSI capacity provides space in the spent fuel pool for one full coreoff-load and to store projected discharged spent fuel assemblies until the end of theplant life, in 2032.Entergy has selected Holtec International's (HOLTEC)
HI-STORM 100 dry cask storagesystem with a Multi-Purpose Canister (MPC) for the Pilgrim ISFSI. For spent fueloperations, the MPC will be housed in the HI-TRAC transfer cask located in the spentfuel 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 MPClid 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 inaccordance with the appropriate requirements of ASME NOG-l (Reference 8.14). Thesingle-failure-proof upgrade of the existing Reactor Building crane was made under theprovisions of 10 CFR 50.59. The 100 ton capacity of the crane main hoist was notchanged.The transfer cask lift yoke and lift yoke extension lifting devices are designed per ANSIN14.6 (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 thelifting device are constructed of metallic wire rope and comply with the requirements ofASME B30.9 (Reference 8.12) and NUREG-0612 (Reference 8.13).To accommodate the Holtec system, the energy absorbing pad was removed andreplaced with the leveling platform.
Since a single-failure-proof handling system will beused for the cask handling operations, the energy absorbing pad is not required.
Thesingle-failure-proof handling system complies with the NRC guidance included inReference 8.13.
Letter 2.14.065Attachment 1Page 4 of 104.0 TECHNICAL ANALYSIS4.1 Description of the Crane Upgrade to Single-Failure-Proof:
Entergy modified the Reactor Building crane under the provisions of 10 CFR 50.59. Thecrane and its operation are described in Sections 10.3, 12.2 and 12.4 of the PNPSUFSAR (Reference 8.1). The Reactor Building crane operates over the entire area ofthe Refuel Floor, including the spent fuel pool, and is designed to handle heavy loadsup 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 inaccordance with the appropriate requirements of ASME NOG-l (Reference 8.14). Theupgraded crane and trolley can safely handle the HI-TRAC 100D transfer cask asspecified in the HI-STORM 100 FSAR to support the dry cask storage operations.
Thecrane control system was also replaced so that the operator has finer control of themain hoist, bridge and trolley movements in order to be more precise with the caskmovements.
4.2 Description of the Energy Absorbing Pad and the Leveling PlatformThe 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 caskloading area using the original non-single-failure-proof Reactor Building crane. The useof the single-failure-proof hoist on the upgraded Reactor Building crane as part of asingle-failure-proof handling system to handle heavy loads in the cask loading area ofthe SFP precludes the need to postulate a transfer cask load drop.The leveling platform provides a stable level platform for placement of the HI-TRAC100D transfer cask to support the dry cask fuel storage operations.
The transfer caskplacement on the leveling platform was evaluated to confirm it's stability during thedesign basis seismic event.4.3 AnalysisCurrent 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 couldbe potentially affected by a cask during cask handling operations.
The proposed TSchange would replace the energy absorbing pad point of reference with a levelingplatform point of reference.
The requirement to limit placement of spent fuel that hasdecayed for less than 200 days in racks within an arc described by the height of thecask around the periphery of the point of reference is being maintained.
In addition, therequirement is being clarified to apply only when cask handling operations are in Letter 2.14.065Attachment 1Page 5 of 10progress 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 NRCguidance included in Reference 8.13.5.0 REGULATORY SAFETY ANALYSIS5.1 Applicable Regulatory Requirements/Criteria General Design Criterion (GDC) 4, Environmental and Dynamic Effects Design Bases ofAppendix A to 10 CFR Part 50 specifies, in part, that structures,
- systems, andcomponents important to safety shall be appropriately protected against dynamiceffects, including the effects of missiles, that may result from equipment failures.
GDC2, Design Bases for Protection Against Natural Phenomena, specifies, in part, thatstructures,
- 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 ofNUREG-0612 for implementation of these criteria in the design of overhead heavy loadhandling 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 safehandling 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 orpermit continued decay heat removal.
Section 5.1.1 of NUREG-0612 providesguidelines for reducing the likelihood of dropping heavy loads and provides criteria forestablishing safe load paths; procedures for load handling operations; training of craneoperators; design, testing, inspection, and maintenance of cranes and lifting devices;and analyses of the impact of heavy load drops. The guidelines in Sections 5.1.2through 5.1.6 address alternatives to either further reduce the probability of a load-handling accident or mitigate the consequences of heavy load drops. Thesealternatives include using a single-failure-proof crane for increased handling systemreliability, employing electrical interlocks and mechanical stops for restricting cranetravel to safe areas, or performing load drop consequence analyses for assessing theimpact 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 Cto NUREG-0612 provides alternative guidance for upgrading the reliability of existingcranes 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 cranesspecified in ASME NOG-1 2004 (Reference 8.14) as acceptable under the guidelines ofNUREG-0554 for construction of a single-failure-proof crane. Paragraph 1.4.C ofSection 9.1.5 of NUREG-0800 states the following:
Letter 2.14.065Attachment 1Page 6 of 10"The probability for a load drop is minimized by an overhead handling systemdesigned to comply with the guidelines of NUREG-0554 and lifting devices thatcomply with American National Standards Institute (ANSI) N14.6 or an alternative based on American Society of Mechanical Engineers (ASME) B30.9. An overheadhandling system that complies with ASME NOG-I criteria for Type 1 cranes is anacceptable 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 handlingsystem. A single failure-proof handling system consists of the following twoelements:
- i. The crane should be designed to the criteria of NUREG-0554.
Cranes designed tothe criteria of ASME NOG-I 2004 for a Type 1 crane are acceptable under theguidelines of NUREG-0554 for construction of a single-failure-proof crane.Consistent with Paragraph 10 of NUREG-0554, a quality assurance program shouldcover the procurement, design, fabrication, installation, inspection,
- testing, andoperation of the crane. The program should include at least the following elements:
(1) design and procurement document control; (2) instructions, procedures, anddrawings; (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 loadmovements in critical areas (reactor head lifting, reactor vessel internals, spentfuel casks). The lifting device should have either dual, independent load paths ora single load path with twice the design safety factor specified by ANSI N14.6 forthe load.(2) Slings should satisfy the criteria of ASME B30.9 and be constructed of metallicmaterial (chain or wire rope). The slings should be either (a) configured to providedual or redundant load paths or (b) selected to support a load twice the weight ofthe 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 supportcommencement of dry cask storage operations.
The single-failure-proof handlingsystem will permit the heavy load lifts required to perform dry cask storage operations tobe performed with design margins sufficient to preclude the necessity of postulating load drop accidents and evaluating consequences.
Letter 2.14.065Attachment 1Page 7 of 10NUREG-0612 evaluation of offsite release potential due to load drop accidents showsthat a decay time of 44 days for Boiling Water Reactors that exhaust through charcoalfilters (I.e. the Standby Gas Treatment System) will assure that offsite releases, due todropping of postulated heavy loads on fuel which has been subcritical for the requireddecay time, will not cause doses that approach 1 OCFR1 00 limits.5.2 No Significant Hazards Consideration Determination Entergy has evaluated whether or not a significant hazards consideration is involvedwith the proposed amendments by focusing on the three standards set forth in 10 CFR50.92, "Issuance of amendment,"
as discussed below:1. Does the proposed change involve a significant increase in the probability orconsequences 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 existingnon 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 energyabsorbing pad point of reference with a leveling platform point of reference.
Inaddition, the requirement is being clarified to apply only when cask handlingoperations 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 200days 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 newload drop accidents are postulated and no changes to the probabilities orconsequences 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 accidentfrom any accident previously evaluated?
Response:
No.Section 10.3 of the PNPS UFSAR evaluates spent fuel storage and cask handlingoperations.
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 aleveling platform point of reference.
Under these circumstances, no new or different Letter 2.14.065Attachment 1Page 8 of 10load drop accidents are postulated to occur and there are no changes in any of theload 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 anymargin of safety. The proposed change replaces the energy absorbing pad point ofreference with a leveling platform point of reference.
In addition, the requirement isbeing clarified to apply only when cask handling operations are in progress in thespent 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 beingmaintained.
Due to the reliability of the upgraded handling system that complies withthe guidance of NUREG-0800 Section 9.1.5 for a single-failure-proof handlingsystem, 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 noreductions in margins of safety are involved.
5.3 Environmental Consideration NUREG-0612 evaluation of offsite release potential due to load. drop accidents showsthat a decay time of 44 days for Boiling Water Reactors that exhaust through charcoalfilters (I.e. the Standby Gas Treatment System) will assure that offsite releases, due todropping of postulated heavy loads on fuel which has been subcritical for the requireddecay 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 donot involve (i) significant hazards consideration, (ii) any changes in the types or anyincrease in the amounts of any effluent that may be released.
- offsite, or (iii) significant increase in individual or cumulative occupational radiation exposure.
Accordingly, theproposed amendment meets the eligibility criterion for categorical exclusion set forth in10 CFR 51 .22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b),
no environmental impactstatement or environmental assessment needs to be prepared in connection with theproposed amendment.
6.0 PRECEDENCE
The NRC approved Waterford 3 License Amendment No. 227 (Reference 8.18). ThePilgrim proposed license amendment'follows the Waterford 3 License Amendment application and NRC acceptance letter (Reference 8.19).
Letter 2.14.065Attachment 1Page 9 of 10The NRC approved a similar change to Technical Specifications for Kewaunee PowerStation, 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 inReference 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 4.3.48.3 Final Safety Analysis Report for the Holtec International Storage and TransferOperation 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 theHoltec International HI-STORM 100 Cask System, December 28, 2009.(ML093620052) 8.5 NRC Amendment No. 7 Certificate of Compliance No. 1014 for the HoltecInternational HI-STORM 100 Cask System, December 28, 2009.(ML093620057) 8.6 NRC Amendment No. 7 Final Safety Evaluation Report Docket No. 72-1014Holtec International HI-STORM 100 Cask System Certificate of Compliance No.1014, December 28, 2009. (ML093620075) 8.7 NRC Amendment No. 7 Certificate of Compliance No. 1014 Appendix ATechnical Specifications for the HI-STORM 100 Cask System, December 28,2009. (ML093620062) 8.8 NRC Amendment No. 7 Certificate of Compliance No. 1014 Appendix BApproved Contents and Design Features for the HI-STORM 100 Cask System,December 28, 2009. (ML093620068) 8.9 NUREG-0800 Section 9.1.5 Rev. 1, Standard Review Plan for Overhead HeavyLoad 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 2.14.065Attachment 1Page 10 of 108.11 ANSI N14.6, Radioactive Materials
-Special Lifting Devices for ShippingContainers Weighing 10,000 Pounds (4500 kg) or More, American NationalStandards 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. NuclearRegulatory Commission, July 1980. (ML070250180) 8.14 ASME NOG-1, Rules for Construction of Overhead and Gantry Cranes (TopRunning Bridge, Multiple Girder),
American Society of Mechanical Engineers, 2004.8.15 NRC Amendment Kewaunee Power Station -Issuance of Amendment toRelocate 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 forControl of Heavy Loads, October 31, 2005. (ML052340485) 8.17 NRC Regulatory Issue Summary 2005-25, Supplement 1, Clarification of NRCGuidelines 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, 20108.19 Acceptance Review Result for Waterford 3 LAR -"Modify TS 3/4.9.7, CraneTravel- Fuel Handling Building,"
dated October 15, 20098.20 Entergy Letter to NRC "Proposed License Amendment Request to ModifyTechnical Specification 4.3.4, "Heavy Loads" to Facilitate Dry Storage HandlingOperations, dated November 26, 2013 (2.13.042)
ATTACHMENT 2To Entergy Letter No. 2.14.065Marked-up Page of the Current Technical Specifications (1 Page)
Letter 2.14.065Attachment 2Page 1 of 1Marked-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 Loadsa. Loads in excess of 2000 lb. shall be prohibited from travel over fuelassemblies in the spent fuel storage pool with the exception that heavy loadhandling over irradiated fuel in the Multi-Purpose Canister is permitted usinqa single-failure-proof handling system.b. No fuel which has decayed for less than 200 days shall be stored in rackswithin an arc described by the height of the cask around the periphery of theenergy abSorbing pad leveling platform during cask handling operations inthe 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 Letterto NRC 2.13.042, dated November 26, 2013