NL-16-118, Proposed License Amendment Regarding the Inter-Unit Transfer of Spent Fuel

Proposed License Amendment Regarding the Inter-Unit Transfer of Spent Fuel

ML16355A067
Person / Time
Site:	Indian Point
Issue date:	12/14/2016
From:	Vitale A Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16355A066	List: ML16355A068 NL-16-118, Indian Point Nuclear Power Plant, Units 2 & 3, Proposed License Amendment Regarding the Inter-Unit Transfer of Spent Fuel NL-16-118, Proposed License Amendment Regarding the Inter-Unit Transfer of Spent Fuel
References
NL-16-118
Download: ML16355A067 (54)

Text

Entergy Nuclear Northeast

• Entergx

• ===*

Indian Point Energy Center 450 Broadway, GSB P.O. Box249 Buchanan, NY 10511-0249 Tel 914 254 6700 Anthony Vitale Site Vice President NL-16-118 December 14, 2016 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station O-P1-17 Washington, DC 20555-0001

Subject:

Indian Point Nuclear Power Plant Units 2 and 3 Proposed License Amendment Regarding the Inter-Unit Transfer of Spent Fuel Indian Point Units 2 & 3 Docket Nos. 50-247 an~ 50-286 License Nos. DPR-26 and DPR-64

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (Entergy) hereby requests amendments to the Technical Specifications (TS) for Indian Point Unit 2 (IP2) and Indian Point Unit 3 (IP3). This request proposes to:

• Revise Unit 2 TS LCO 3.7.13 Spent Fuel Pit Storage;

• Revise Unit 2 Appendix C TS LCO 3.1.2 Shielded Transfer Canister (STC) Loading; and

• Revise Unit 3 Appendix C TS LCO 3.1.2 Shielded Transfer Canister (STC) Loading The proposed changes will expand the population of assemblies that can be transferred frbm IP3 to IP2.

Entergy has evaluated the proposed changes in accordance with 10 CFR 50.91 (a)(1) using the criteria of 10 CFR 50.92(c) and has determined that these proposed changes involve no .,

significant hazards considerations, as described in Attachment 1. The marked up pages showing the proposed IP2 TS and TS Bases are provided in Attachments 2 and 3, respectively and in Attachments 4 and 5 for IP3. The TS Bases markups are provided for information only.

The evaluation of the proposed changes is supported by Holtec Report Hl-2094289 Revision 8.

Proprietary and non-proprietary versions of this report are included in Enclosures 1 and 2 respectively. Enclosure 3 contains Holtec supporting reports anc;I evaluations.

Enclosures 1 and 3 contain information that Westinghouse and/or Holtec consider to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390. In accordance with 10 CFR 2.390 and in support of this request for withholding, affidavits executed

"-by the respective parties are provided in Enclosure 4.

In accordance with 10 CFR 50.91, a copy of this application, with attachments and enclosures is being provided to the designated New York State official. ;t DD l N~~

(

NL-16-118 Docket Nos. 50-247 and 50-286 Page 2 of 2 Entergy requests approval of the proposed license amendments in one calendar year in order to support inter-unit fuel transfer operations in the second quarter of 2018. Once approved, 7 Entergy requests 60 days for implementation. There are no new regulatory commitments being made in this submittal. If you have any question$ or require additional information, please contact Mr. Robert Walpole, Licensing Manager at 914-734-6710.

I declare under penalty of perjury that the foregoing is true and correct to the best of my knowledge. Executed on :l)ic,. f _4 1 ~ D \ 6 * . /

Sincerely, AV/rw Attachments and

Enclosures:

Attachments: 1. Analysis of proposed technical specification changes

2. Proposed IP2 technical specification changes (Marked-up)

3. IP2 technical specification bases (Marked-up) (for information only)

4. Proposed IP3 technical specification changes (Marked-up)

5. IP3 technical specification bases (Marked-up) (for information only)

Enclosures:

1. Holtec International Licensing Report Hl-2094289, Revision 8 (Holtec and Westinghouse Proprietary)

2. Holtec International Licensing Report Hl-2094289, Revision 8 (Holtec and Westinghouse Non-Proprietary)

3. Holtec International Supporting-Reports and Evaluations (Holtec Proprietary)

-Hl-2084109R12, "Shielding Design Calculation of Transfer Bell for IP3"

-Hl-:-2084146R8, "Thermal Hydraulic Analysis of IP3 STC"

-Hl-2084118R6, "STC Structural Calculation Package"

-Hl-210470~R3, "Non-Mechanistic Tipover of HI-TRAC 100D for IP3"

4. Affidavits in Support of Request to Withhold Information cc: NRC Resident Inspector's Office Mr. DougJas Pickett, Senior Project Manager, NRC NRR DORL Mr. Daniel H. Dorman, Regional Administrator, NRC Region 1 Mr. John B. Rhodes, President and CEO, NYSERDA (w/o proprietary information)

Ms. Bridget Frymire, New York State Dept. of Public Service (w/o proprietary information)

ATTACHMENT 1 TO NL-16-118 ANALYSIS OF PROPOSED TECHNICAL SPECIFICATION CHANGES Entergy Nuclear Operations, Inc.

Indian Point Units 2 and 3 Docket Nos. 50-247 and 50-286

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 1 of 16

1.0 DESCRIPTION

This letter is a request to amend Operating License Nos. DPR-26 for Indian Point Unit 2 (IP2) and DPR-64 for Indian Point Unit 3 (IP3). The proposed changes revise the Appendix C Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.1.2 for IP2 and IP3 and Appendix A TS LCO 3.7.13 for IP2. These LCOs ensure that the fuel to be loaded into the Shielded Transfer Canister (STC) meets the design basis for the STC and has an acceptable rack location in the IP2 spent fuel pit before the STC is loaded with fuel. The proposed changes to these LCOs will increase the population of IP3 fuel eligible for transfer to the IP2 spent fuel pit and maintain full core offload capability for IP3.

2.0 PROPOSED CHANGE

S Attachments 2 and 4 to this submittal contain the marked-up pages showing the specific proposed changes to the IP2 and IP3 Technical Specifications respectively. The changes are summarized below.

2.1 IP2 Technical Specification Changes The following changes are proposed to the IP2 TS:

Appendix A LCO 3. 7 .13 Spent Fuel Pit Storage From:

IP3 fuel assemblies shall be stored in Region 1-2 of the Spent Fuel Pit.

Only assemblies with initial enrichment=::: 3.2 and s 4.4 w/o U235 and discharged prior to IP3 Cycle 12 shall be stored in the Spent Fuel Pit.

To:

IP3 fuel assemblies shall be stored in Region 1-2 of the Spent Fuel Pit.

The Appendix C Inter-Unit Fuel Transfer Technical Specifications applies to both IP2 and IP3 and controls inter-unit fuel transfer operations via LCOs, Design Features, and Programs.

Appendix C LCO 3.1.2 Shielded Transfer Canister (STC) Loading LCO 3.1.2 a.2 From:

Decay heat including NON FUEL HARDWARE s 650 Watts (cells 5 through 12);

To:

Decay heat including NON FUEL HARDWARE s 1.2 kW (any cell);

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 2of16 LCO 3.1.2 a.3 From:

Decay heat including NON FUEL HARDWARE~ 1105 Watts (cell 1, 2, 3, or 4);

To:

Total STC decay heat from all cell locations including NON FUEL HARDWARE~ 9.621 kW; LCO 3.1.2 b.2 From:

Decay heat including NON FUEL HARDWARE ~ 650 Watts; To:

Decay heat including NON FUEL HARDWARE ~ 1.2 kW; LCO 3.1.2 c Delete:

c. Only INTACT FUEL ASSEMBLIES with initial enrichment<:: 3.2 and~ 4.4 w/o U-235 and discharged prior to IP3 Cycle 12 shall be placed in the STC basket.

Table 3.1.2-3 Replace existing TS Table 3.1.2-3 with the proposed TS Table 3.1.2-3 as shown in Attachment 2.

2.2 IP3 Technical Specification Changes The following changes are proposed to the IP3 TS:

Appendix C LCO 3.1.2 Shielded Transfer Canister (STCJ Loading LCO 3.1.2 a.2 From:

Decay heat including NON FUEL HARDWARE~ 650 Watts (cells 5 through 12);

To:

Decay heat including NON FUEL HARDWARE ~ 1.2 kW (any cell);

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 3 of 16 LCO 3.1.2 a.3 From:

Decay heat including NON FUEL HARDWARE s 1105 Watts (cell 1, 2, 3, or 4);

To:

Total STC decay heat from all cell locations including NON FUEL HARDWARE s 9.621 kW; LCO 3.1.2 b.2 From:

Decay heat including NON FUEL HARDWARE s 650 Watts; To:

Decay heat including NON FUEL HARDWARE s 1.2 kW; LCO 3.1.2 c Delete:

c. Only INTACT FUEL ASSEMBLIES with initial enrichment~ 3.2 ands 4.4 w/o U-235 and discharged prior to IP3 Cycle 12 shall be placed in the STC basket.

Table 3.1.2-3 Replace existing TS Table 3.1.2-3 with the proposed TS Table 3.1.2-3 as shown in Attachment 4.

Appropriate Bases changes (for information only in Attachments 3 and 5) would also be made consistent with the TS changes discussed above.

2.3 Need for Proposed Changes In July of 2012, the NRC issued License Amendment Numbers 268 and 246 for IP2 and IP3, respectively (Ref. 1). These license amendments provided the necessary controls for the inter-unit transfer of spent fuel from IP3 to IP2. In accordance with these amendments fuel assemblies to be transferred are selected at IP3 based on restrictions for loading in the STC and for storage in the IP2 SFP. These restrictions on discharge cycle, initial enrichment, burnup and cooling time, provide a significant limitation on the population of assemblies that can be transferred to IP2, and subsequently be loaded into the HI-STORM dry storage systems. The proposed changes will expand the population of assemblies that can be transferred to IP2.

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 4of16

3.0 BACKGROUND

As required by plant operations IP3 spent fuel is transferred to the IP2 spent fuel pit in order to maintain adequate fuel storage capacity in the IP3 spent fuel pit. IP3 spent fuel moved to the IP2 spent fuel pit is subsequently transferred to dry cask storage at the IPEC on-site Independent Spent Fuel Storage Installation (ISFSI) as part of spent fuel inventory management in the IP2 spent fuel pit.

Fuel assemblies to be transferred are selected at IP3 based on the requirements for loading in the STC. The STC fuel loading requirements are such that the fuel selected for transfer to IP2 is suitable for storage in the designated IP2 spent fuel pit storage rack as controlled by IP2 Appendix A TS LCO 3.7.13 Spent Fuel Pit Storage. Fuel move sheets govern the transfer of the spent fuel from IP3 to IP2.

Appendix C TS Table 3.1.2-1 Minimum Bumup Requirements at Varying Initial Enrichments is used to classify each fuel assembly as either Type 1 or Type 2 based on initial U-235 enrichment and average assembly burnup. This classification is used to determine if, and where, the fuel assembly can be placed in the STC fuel basket. Type 1 and Type 2 fuel assemblies are subject to additional loading restrictions that include post-irradiation cooling time, initial enrichment, allowable average burnup and decay heat of fuel and non fuel hardware. These requirements are specified in Appendix C TS Table 3.1.2-2 Non Fuel Hardware Post Irradiation Cooling Times and Allowable Average Bumup and Appendix C TS Table 3.1.2-3 Allowable STC Loading Configurations.

Together, the limits on Type 1 and Type 2 fuel assemblies and associated non fuel

. hardware ensure the criticality, shielding, structural and thermal analyses performed for the STC remain bounding.

The proposed amendment seeks to relax the limits imposed on Type 1 and Type 2 fuel assemblies by the STC criticality analyses, the IP2 SFP criticality analyses and the thermal, shielding and dose analyses. The acceptability of the proposed revisions is provided below.

4.0 TECHNICAL ANALYSIS

The Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center (Ref. 2 and Enclosure 1) provides full details of the system design, design calculations and evaluations, material selection, operation, and maintenance requirements.

The proposed technical specification changes are supported by revisions to the following Licensing Report analyses and evaluations:

• Chapter 2 Fuel Acceptance Criteria and Engineered Measures of Safety,

• Chapter 4 Criticality Evaluation,

• Chapter 5 Thermal-Hydraulic Evaluation

• Chapter 6 Structural Evaluation if Normal and Accident Conditions

• Chapter 7 Shielding Design and ALARA Considerations

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 5of16 A summary of each of these revisions to the Licensing Report is provided below.

4.1 Chapter 2 Fuel Acceptance Criteria and Engineered Measures of Safety The proposed TS relax the burnup and minimum cooling time fuel acceptance criteria.

The acceptability of these changes are discussed in Sections 4.2, 4.3 and 4.4 below.

4.2 Chapter 4 Criticality Evaluation STC Criticality Analysis The proposed TS remove the restrictions that IP3 fuel assemblies must have initial enrichments~ 3.2 ands 4.4 w/o U235 , and the fuel assembly discharge be Cycle~ 1 and s 11.

The NRC Safety Evaluation Report (SER) (Ref. 1) provided approval for the transfer of spent fuel from the IP3 SFP to the IP2 SFP. The SER was based, in part, on Rev. 6 of Ref. 2. The STC criticality analysis in Rev. 6 did not take credit for these enrichment and cycle restrictions. However, the SER did credit these restrictions. This credit allowed the NRC to conclude that a more precise determination of the STC mechanical tolerances was not required and that it was acceptable to place IP3 fuel in Region 1-2 of the IP2 SFP provided these restrictions were in place.

Section 4.7.5.1 Basket Tolerances of Ref. 2 shows that a more precise determination of the STC mechanical tolerances is not required even in the absence of the subject enrichment and cycle restrictions.

Therefore, from an STC criticality perspective, it is acceptable to remove the restrictions that IP3 fuel assemblies must have initial enrichments~ 3.2 ands 4.4 w/o U235 , and the fuel assembly discharge be Cycle ~ 1 and s 11.

Acceptability of Storing IP3 Fuel in the IP2 Pool The proposed TS remove the restrictions that IP3 fuel assemblies must have initial 235 enrichments~ 3.2 ands 4.4 w/o U , and the fuel assembly discharge be Cycle~ 1 and s 11.

The proposed TS continue to restrict IP3 fuel assemblies to Region 1-2 of the IP2 SFP.

Region 1-2 of the IP2 SFP is qualified for fresh fuel. Section 4.8 of Ref. 2 shows that IP2 and IP3 fuel assemblies are identical as regards those parameters that are utilized in the design basis criticality analysis (OBA) to qualify fresh fuel in Region 1-2. Therefore, all the IP3 spent fuel assemblies are directly covered by the OBA for storage in Region 1-2, independent of any differences in the in-core operation of the fuel in the two units.

Therefore, from an IP2 SFP criticality perspective, it is acceptable to remove the restrictions that IP3 fuel assemblies must have initial enrichments~ 3.2 ands 4.4 w/o U235 , and the fuel assembly discharge be Cycle~ 1 ands 11.

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 6 of 16 4.3 Chapter 5 Thermal-Hydraulic Evaluation The proposed TS relax the restrictions that IP3 fuel assemblies including NFH must have decay heats :::; 650 Watts in the STC peripheral cells (Types 1 and 2) and :::; 1105 Watts in the inner cells (Type 2) to :::; 1.2 kW in any cell. The proposed TS impose a new restriction that the total STC decay heat from all cell locations including NFH be:::; 9.621 kW. This new restriction, as evaluated in Ref. 2 Chapter 5, preserves the maximum STC decay heat assumed in the current STC thermal analyses i.e. 4 inner cells at 1105 Watts and 8 peripheral cells 650 Watts for a total of 9.62 kW.

The representative heat load pattern of 1105.2 Wand 650 W decay heat in four interior cells and eight peripheral cells respectively continues to be used for the thermal evaluation of the normal onsite transfer condition. Since the proposed maximum allowable per assembly decay heat can be higher than that used in the representative evaluations, an evaluation has been performed to demonstrate that the predicted temperatures and cask cavity pressures remain essentially unaffected for the same total heat load when the STC remains in a vertical configuration during normal, off normal and accident conditions.

The evaluation considered two different scenarios:

• Scenario 1: Four inner locations at the maximum per assembly decay heat of 1.2 kW with the decay heat in the remaining locations adjusted to comply with the maximum total heat load of 9.621 kW.

• Scenario 2: Eight outer locations at the maximum per assembly decay heat of 1.2 kW with the remaining four inner locations empty.

The steady state temperatures for both scenarios are essentially the same as the representative heat load pattern. The STC and annulus cavity pressures are also essentially the same. This study demonstrates that the proposed TS are sufficient to ensure that adequate thermal margins are maintained under all conditions in which the STC remains vertical. Therefore, it is reasonable to use the representative heat load distribution for all licensing basis evaluations for normal, off-normal and accident conditions in which the STC is in a vertical configuration.

As described in Ref. 2 Chapter 5 the thermal analyses were performed under the following normal and accident scenarios:

• Normal on-site transfer of IP3 fuel

• Loss of HI-TRAC water jacket

• External fire (rupture of transporter fuel tank)

• Simultaneous Loss of Water from the HI-TRAC Water Jacket and HI-TRAC annulus

• Fuel misleading

• Non-mechanistic tipover accident

• Crane malfunction

)

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 7of16 The only scenario that involves a non vertical STC configuration is the non-mechanistic tipover accident. In support of the proposed TS changes this accident has been re-evaluated in Section 5.4.5 of Rev. 1. The evaluation assumed that the maximum permissible assembly decay heat of 1.2 kW is placed in the two basket storage locations that are partially submerged in water. Additionally, it is conservatively assumed that the basket storage locations close to the water surface also have the highest decay heat of 1.2 kW. This is assumed to ensure the water surface temperature is maximized to predict bounding STC temperatures.

As a result of this evaluation the accident pressure has been increased from 90 psig to 165 psig. This change is documented in Ref. 2 Table 3.2.1 and evaluated in Ref. 2 Chapter 6.

The results of this evaluation show that the cladding temperature remains below the ISG-11 temperature limit, and that the STC and HI-TRAC component temperatures remain below design limits and the co-incident STC and HI-TRAC pressures are bounded by the accident pressure limits.

4.4 Chapter 7 Shielding Design and ALARA Considerations The shielding and dose evaluations have been performed under the following normal and accident scenarios:

Shielding and Dose

• Normal on-site transfer

• Crane Hang Up

• Loss of water from the water jacket (e.g. due to fire)

• Simultaneous loss of water from the water jacket and HI-TRAC annulus from tip over

• STC off-center within the HI-TRAC due to tip over resulting in crushing of the centering assembly accompanying with the simultaneous loss of water from HI-TRAC jacket and HI-TRAC annulus and inside STC The proposed TS add six new allowable STC loading configurations. These additional configurations typically allow the transfer of IP3 fuel with higher burnups and lower cooling times. These additional configurations have been developed using realistic, but still conservative, source term assumptions regarding cobalt-60 and non-fuel hardware.

Specifically, these additional configurations took credit for:

• reduced Cobalt-59 content of stainless steel components

• BPRA decay time, and

• reduced RCCA activation based on IP3 operating experience

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 8of16 Revision of the source term assumptions is supported by IP3 STC operational dose rate measurements that have been shown to be significantly lower than the predicted values for the same fuel and NFH loading [L.G].

The current TS (Appendix C TS 5.4) include dose rate limits for the STC lid and the side of the HI-TRAC. These dose rate limits are based on the bounding dose rates determined in the shielding and dose evaluations and remain unchanged by this amendment request. Therefore, the occupational doses remain ALARA and the dose limits for members of the public will not be exceeded.

Therefore, the wet spent fuel transfer operations will continue to meet the shielding and radiation requirements of 10 CFR Parts 20 and 50, and the intent of 10 CFR Part 72.

5.0 REGULATORY ANALYSIS

5.1 IP2 No Significant Hazards Considerations Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment for IP2 by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment would modify the IP2 and IP3 Technical Specifications (TS) to incorporate the results of revised criticality, thermal, and shielding and dose

. analyses and evaluations.

The proposed amendment was evaluated for impact on the following previously evaluated events and accidents: STC Criticality Accidents, SFP Criticality Accidents, Boron Dilution Accidents, Fuel Handling Accidents, Loss of Spent Fuel Pool Cooling, and Natural Events.

STC Criticality Accidents The STC criticality accident cons.idered were: abnormal temperature, dropped, mislocated, and misloaded fuel assemblies, and misalignment between the active fuel region and the neutron absorber.

The probability of an STC criticality accident will not increase significantly due to the proposed changes because the individual fuel assemblies will be loaded into the STC in the sa,me manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used.

The consequences of an STC criticality accident are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for these accidents.

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 9 of 16 SFP Criticality Accidents The SFP criticality accident of record considered the following accidents 1.) a dropped fuel assembly or an assembly placed alongside a rack, 2.) a misloaded fuel assembly, and 3.) abnormal heat loads. Because the IP2 and IP3 fuel assemblies are identical as regards those parameters that are utilized in the design basis criticality analysis (OBA) to qualify fresh fuel these accidents are bounding for IP3 fuel.

The probability of an SFP criticality accident will not increase significantly due to the proposed changes because the individual fuel assemblies will be loaded into the SFP in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used.

The consequences of an SFP criticality accident are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for this accident.

STC Thermal Accidents The thermal analyses demonstrate that the postulated accidents (rupture of the HI-TRAC water jacket, 50-gallon transported fuel tank rupture and fire, simultaneous loss-of water from the water jacket and HI-TRAC annulus, fuel mislead, hypothetical tipover, and crane malfunction) continue to meet their acceptance criteria.

The probability of an STC thermal accident will not increase significantly because the individual fuel assemblies will be loaded into the SFP in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used.

The consequences of an STC thermal accident will not increase significantly because the thermal analysis demonstrates that the same thermal acceptance criteria and requirements continue to be met for this accident.

Boron Dilution Accident The probability of a boron dilution event remains the same because the proposed change does not alter the manner in which the IP2 spent fuel cooling system or any other plant system is operated, or otherwise increase the likelihood of adding significant quantities of unborated water into the spent fuel pit.

The consequences of the boron dilution event remains the same. The reactivity of the STC filled with the most reactive combination of approved fuel assemblies in unborated water results in a ketr less than 0.95. Thus, even in the unlikely event of a complete dilution of the spent fuel pit water, the STC will remain safely subcritical.

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 10of16 Fuel Handling Accident The probability of an FHA will not increase significantly due to the proposed changes because the individual fuel assemblies will be moved between the STC and the spent fuel pit racks and the STC and HI-TRAC will be moved in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used.

The consequences of the existing fuel handling accident remain bounding because the IP3 fuel assembly design is essentially the same as the IP2 design and the IP3 fuel assemblies to be transferred to IP2 will be cooled a minimum of 6 years. This compares with a cooling time of 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> used in the existing FHA radiological analysis. The 6-year cooling time results in a significant reduction in the radioactive source term available for release from a damaged fuel assembly compared to the source term considered in the design basis FHA radiological analysis. The consequences of the previously analyzed fuel assembly drop accident, therefore, continue to provide a bounding estimate of offsite dose for this accident.

Loss of Spent Fuel Pool Cooling The probability of a loss of spent fuel pit cooling remains the same because the proposed change does not alter the manner in which the IP2 spent fuel cooling loop is operated, designed or maintained.

The consequences of a loss of spent fuel pit cooling remains the same because the thermal design basis for the spent fuel pit cooling loop provides for all fuel pit rack locations to be filled at the end of a full core discharge and therefore the design basis heat load effectively includes any heat load associated with the assemblies within the STC.

Natural Events The natural events considered include the following accidents 1.) a seismic event, 2.)

high winds, tornado and tornado missiles, 3.) flooding and 4) a lightning strike.

The probability of natural event will not increase due to the proposed changes because there are no elements of the proposed changes that influence the occurrence of any natural event.

The consequences of a natural event will not increase due to the proposed changes because the structural analyses design limits continue to be met. A lightning strike may cause ignition of the VCT fuel but this event is addressed under STC thermal accidents.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 11of16

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. No new modes of operation are introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident, from any accident previously evaluated.

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

Response: No.

The proposed amendment would modify the TS to incorporate the results of revised criticality, thermal and shield and dose analyses. The margin of safety required by 10 CFR 50.58(b)(4) remains unchanged. New criticality evaluations for both the STC and the IP2 SFP confirm that operation in accordance with the proposed amendment continues to meet the required subcriticality margins. The thermal analyses demonstrate that the postulated accidents (rupture of the HI-TRAC water jacket, 50-gallon transported fuel tank rupture and fire, simultaneous loss of water from the water jacket and HI-TRAC annulus, fuel misload, hypothetical tipover, and crane malfunction) continue to meet their acceptance criteria without a significant loss of safety margin. The shielding and dose analyses demonstrate that the shielding and radiation protection requirements continue to be met without a significant loss of safety margin.

Therefore, the proposed change does not. involve a significant reduction in a margin of safety.

• Based on the above, Entergy concludes that the proposed amendment to the IP2 Operating License and Technical Specifications presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.2 IP3 No Significant Hazards Considerations Entergy has evaluated whether or not a significant hazards consideration is involved with the proposed amendment for IP3 by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 12of16 The proposed amendment would modify the IP2 and IP3 Technical Specifications (TS) to incorporate the results of revised criticality, thermal, and shielding and dose analyses and evaluations.

The proposed amendment was evaluated for impact on the following previously evaluated events and accidents: STC Criticality Accidents, SFP Criticality Accidents, Boron Dilution Accidents, Fuel Handling Accidents, Loss of Spent Fuel Pool Cooling, and Natural Events.

STC Criticality Accidents The STC criticality accident considered were: abnormal temperature, dropped, mislocated, and misloaded fuel assemblies, and misalignment between the active fuel region and the neutron absorber.

The probability of an STC criticality accident will not increase significantly due to the proposed changes because the individual fuel assemblies will be loaded into the STC in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used.

The consequences of an STC criticality accident are not changed because the reactivity analysis demonstrates that the same subcriticality criteria and requirements continue to be met for these accidents.

STC Thermal Accidents The thermal analyses demonstrate that the postulated accidents (rupture of the HI-TRAC water jacket, 50-gallon transported fuel tank rupture and fire, simultaneous loss of water from the water jacket and HI-TRAC annulus, fuel mislead, hypothetical tipover, and crane malfunction) continue to meet their acceptance criteria.

The probability of an STC thermal accident will not increase significantly because the individual fuel assemblies will be loaded into the SFP in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used.

The consequences of an STC thermal accident will not increase significantly because the thermal analysis demonstrates that the same thermal acceptance criteria and requirements continue to be met for this accident.

Boron Dilution Accident The probability of a boron dilution event remains the same because the proposed change does not alter the manner in which the IP3 spent fuel cooling system or any other plant system is operated, or otherwise increase the likelihood of adding significant quantities of unborated water into the spent fuel pit. ,

_)

The consequences of the boron dilution event remains the same. The reactivity of the STC filled with the most reactive combination of approved fuel assemblies in

NL-16-118 Docket Nos. 50-24 7 and 50-286 Attachment 1 Page 13 of 16 unborated water results in a keff less than 0.95. Thus, even in the unlikely event of a complete dilution of the spent fuel pit water, the STC will remain safely subcritical.

Fuel Handling Accident The probability of an FHA will not increase significantly due to the proposed changes because the individual fuel assemblies will be moved between the STC and the spent fuel pit racks and the STC and HI-TRAC will be moved in the same manner, using the same equipment, procedures, and other administrative controls (i.e. fuel move sheets) that are currently used.

The consequences of the existing fuel handling accident remain bounding because only IP3 fuel is moved in the IP3 spent fuel pit. The IP3 fuel assemblies to be transferred to IP2 will be cooled a minimum of 6 years. This compares with a cooling time of 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> used in the existing FHA radiological analysis. The 6-year cooling time results in a significant reduction in the radioactive source term available for release from a damaged fuel assembly compared to the source term considered in the design basis FHA radiological analysis. The consequences of the previously analyzed fuel assembly drop accident, therefore, continue to provide a bounding estimate of offsite dose for this accident.

Loss of Spent Fuel Pool Cooling The probability of a loss of spent fuel pit cooling remains the same because the proposed change does not alter the manner in which the IP3 spent fuel cooling loop is operated, designed or maintained.

The consequences of a loss of spent fuel pit cooling remains the same because the thermal design basis for the spent fuel pit cooling loop provides for all fuel pit rack locations to be filled at the end of a full core discharge and therefore the design basis heat load effectively includes any heat load associated with the assemblies within the

• STC.

Natural Events The natural events considered include the following accidents 1.) a seismic event, 2.)

high winds, tornado and tornado missiles, 3.) flooding and 4) a lightning strike.

The probability of natural event will not increase due to the proposed changes because there are no elements of the proposed changes that influence the occurrence of any natural event.

The consequences of a natural event will not increase due to the proposed changes because the structural analyses design limits continue to be met. A lightning strike may cause ignition of the VCT fuel but this event is addressed under STC thermal accidents.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 14of16

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated. No new modes of operation are

~) introduced by the proposed changes. The proposed changes will not create any failure mode not bounded by previously evaluated accidents.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident, from any accident previously evaluated.

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

Response: No.

The proposed amendment would modify the TS to incorporate the results of revised criticality, thermal and shield and dose analyses. The margin of safety required by 10 CFR 50.58(b)(4) remains unchanged. New criticality evaluations for both the STC confirm that operation in accordance with the proposed amendment continues to meet the required subcriticality margins. The thermal analyses demonstrate thatthe postulated accidents (rupture of the HI-TRAC water jacket, 50-gallon transported fuel tank rupture and fire, simultaneous loss of water from the water jacket and HI-TRAC annulus, fuel mislead, hypothetical tipover, and crane malfunction) continue to meet their acceptance criteria without a significant loss of safety margin. The shielding and dose analyses demonstrate that the shielding and radiation protection requirements continue to be met_ without a significant loss of safety margin.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, Entergy concludes that the proposed amendment to the IP3 Operating License and Technical Specifications presents no significant hazards consideration under the standards set forth in 10 CFR 50.92( c), and, accordingly, a finding of "no significant hazards consideration" is justified.

5.3 Applicable Regulatory Requirements and Criteria The following 10 CFR 50, Appendix A, General Design Criteria (GDC) have been identified as applicable to this request and the manner in which Entergy continues to comply with each GDC is described in each case. Appropriate modifications to the IP2 and IP3 updated Final Safety Analysis Reports will be made after issuance of the amendment.

GDC 61, Fuel Storage and Handling and Radioactivity Control specifies, in part, that the fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under

I NL-16-118 Docket Nos. 50-247 and 50-286 Attachment 1 Page 15 of 16 normal and postulated accident conditions. These systems shall be designed (1) with a capability to permit appropriate periodic inspection and testing of components important to safety, (2) with suitable shielding for radiation protection, (3) with appropriate containment, confinement, and filtering systems, (4) With a residual heat removal capability having reliability and testability that reflects the importance to safety of decay heat and other residual heat removal, and (5) to prevent significant reduction in fuel storage coolant inventory under accident conditions.

The proposed revision to the limitations of TS 3.1.2 continues to ensure that an acceptable inventory of fuel and non-fuel hardware would be loaded into the STC to maintain heat generation rates within analyzed values. The proposed limitations are that the maximum allowable decay heat per storage location is 1.2 kW and that the maximum allowable STC decay heat is 9.621 kW (Ref. 2)

The analyses of the various accident conditions determined that heat generation rates consistent with the prescribed limits would ensure that STC design pressure will remain below design pressure for all analyzed conditions. Therefore, the proposed TS limits on fuel loading are acceptable with respect to heat generation rates. (Ref. 2)

The dose rates originally determined in Ref 2 Rev. 6 have been shown to be bounding for the STC by itself, as well as for the STC inside a HI-TRAC. In addition, occupational exposures as originally determined are bounding for transferring the STC into and out of the HI-TRAC, moving the HI-TRAC from IP3 to IP2, and unloading the STC in IP2. Further, the original site boundary dose rates from transferring the HI-TRAC containing the STC between IP3 and IP2 for normal and accident conditions are bounding.

GDC 62, Prevention of Criticality in Fuel Storage and Handling requires that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

The STC is a device that is licensed under 10CFR50, and its criticality safety performance and acceptance criteria are therefore governed by 10CFR50.68.

When evaluating the STC criticality analysis the NRC SER took credit for the IP3 cycle and initial enrichment restrictions being removed by this proposed amendment. The analyses performed in support of this license amendment request have shown that 10CFR50.68 continues to be met when these restrictions are removed. (Ref. 2)

The criticality analysis that covers IP3 spent fuel in the IP2 SFP is addressed in Section 4.8 of the Ref. 2. The evaluations performed in support of this license amendment request have shown that 10CFR50.68 continues to be met when the IP3 cycle and initial enrichment restrictions are removed by this proposed amendment. (Ref. 2) 10 CFR Part 20, 10 CFR Part 100, and the intent of 10 CFR 72.104 and 10 CFR 72.106 apply to the shielding and confinement features of the STC and HI-TRAC. 10 CFR 20 applies to the methods for controlling and limiting occupational exposures ALARA. The

NL-16-118 Docket Nos. 50-24 7 and 50-286 Attachment 1 Page 16of16 shielding, radiation protection, and confinement design features and design criteria for the STC and HI-TRAC are not impacted by the proposed amendment. In addition, the methods for controlling and limiting occupational exposures within the dose and ALARA requirements are not impacted by the proposed amendment. Therefore, the wet spent fuel transfer operations will continue to meet the shielding and radiation requirements of 10 CFR Parts 20 and 50, and the intent of 10 CFR Part 72.

5.4 Environmental Considerations Entergy has evaluated the proposed changes and determined that the changes do not involve (1) a significant hazards consideration (2) a significant change in the types or significant increase in the amounts of any effluents that may be released off-site, or (3) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criteria 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 need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. NRC, "Safety Evaluation Report by the Office of NRR Related to Amendment No.

268 to Facility Operating License No. DPR-16 and Amendment No. 246 to Facility Operating License No. DPR-64", July 13, 2012 (ADAMS Accession No. ML121230011)

2. Hl-2094289 Revision 8, "Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center".

ATTACHMENT 2 TO NL-16-118 Indian Point Unit 2 Proposed Technical Specification Changes (Marked-up)

Affected Appendix A Tech Spec Page: 3.7.13-1 Affected Appendix C Tech Spec Pages: 3.1.2-1 3.1.2-6 3.1.2-7 (new)

Entergy Nuclear Operations, Inc.

Indian Point Unit 2 Docket No. 50-247

Spent Fuel Pit Storage 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Spent Fuel Pit Storage LCO 3.7.13 IP2 fuel assemblies stored in the Spent Fuel Pit shall be classified in accordance with Figure 3.7.13-1, Figure 3.7.13-2, Figure 3.7.13-3, and Figure 3.7.13-4, based on initial enrichment, burnup, cooling time and number of Integral Fuel Burnable Absorbers (IFBA) rods; and, Fuel assembly storage location within the Spent Fuel Pit shall be restricted to Regions identified in Figure 3.7.13-5 as follows:

a. Fuel assemblies that satisfy requirements of Figure 3.7.13-1 maybe stored in any location in Region 2-1, Region 2-2, Region 1-2 or Region 1-1;

b. Fuel assemblies that satisfy requirements of Figure 3.7.13-2 may be stored in any location in Region 2-2, Region 1-2 or Region 1-1;

c. Fuel assemblies that satisfy requirements of Figure 3.7.13-3 may be stored in any location in Region 1-2, Region 1-1, or in locations designated as "peripheral" cells in Region 2-2; and

d. Fuel assemblies that satisfy requirements of Figure 3.7.13-4 may be stored:

1) In any location in Region 1-2, or

2) In a checkerboard loading configuration (1 out of every two cells with every other cell vacant) in Region 1-1 ; or

3) In locations designated as "peripheral" cells in Region 2-2.

IP3 fuel assemblies shall be stored in Region 1-2 of the Spent Fuel Pit.

Only assemblies with initial enrichment > 3.2 and < 4 .4 w/o U2aa--aoo discharged prior to IP3 Cycle 12 shall be stored in the Spent Fuel Pit.

APPLICABILITY: Whenever any fuel assembly is stored in the Spent Fuel Pit.

INDIAN POINT 2 3.7.13-1 Amendment No. TBD

STC Loading 3.1.2 3.1 INTER-UNIT FUEL TRANSFER 3.1.2 Shielded Transfer Canister (STC) Loading LCO 3.1.2 INTACT FUEL ASSEMBLIES placed into the Shielded Transfer Canister (STC) shall be classified in accordance with Table 3.1.2-1 based on initial enrichment and burnup and shall be restricted based on the following:

a. INTACT FUEL ASSEMBLIES classified as Type 2 may be placed in the STC basket (see Figure 3.1.2-1) with the following restrictions:

1. Post-irradiation cooling time, initial enrichment, and allowable average burnup shall be within the limits for the cell locations as specified in Table 3.1.2-3;

2. Decay heat including NON FUEL HARDWARE::;; 650 Watts (cells 5 through 12)1_1.2 k"'! (any cell];

3.

FUEL HARDWARE::;; 9.621 k ;

4. Post-irradiation cooling time and the maximum average burnup of NON FUEL HARDWARE shall be within the cell locations and limits specified in Table 3.1.2-2. In accordance with Table 3.1.2-2 RCCAs and Hafnium Flux Suppressors cannot be placed in locations 5, 6, 7, 8, 9, 1O, 11, 12 of the STC basket.

- NOTE -

If one or more Type 1 fuel assemblies are in the STC, cells 1, 2, 3, AND 4 must be empty, with a cell blocker installed that prevents inserting fuel assemblies and/or NON-FUEL HARDWARE.

b. INTACT FUEL ASSEMBLIES classified as Type 1 or Type 2 may be placed in locations 5, 6, 7, 8, 9, 10, 11, 12 of the STC basket (see Figure 3.1.2-1) with the following restrictions:

1. Post-irradiation cooling time, initial enrichment, and allowable average burnup shall be within the limits for the cell locations as specified i'n Table 3.1.2-3;

2. Decay heat including NON FUEL HARDWARE::;; 650 VVattsl1.2 kWt

3. Post-irradiation cooling time and the maximum average burnup of NON FUEL HARDWARE shall be within the cell locations and limits specified in Table 3.1.2-2. In accordance with Table 3.1.2-2 RCCAs and Hafnium Flux Suppressors cannot be placed in locations 5, 6, 7, 8, 9, 1O, 11, 12 of the STC basket.

c. Only INTACT FUEL ASSEMBLIES Y.'ith initial average enriohment > 3.2 and< 4.4 vJt% U 235 and disoharged prior to IP3 Cyole 12 shall be plaoed in the STC basl<:et.

INDIAN POINT 2 3.1.2-1 Amendment No. TBD

STC Loading 3.1.2

. Table 3.1.2-3l(Sheet 1 of 2) I Allowable STC Loading Configurations Configuration(c) Cells 1 2 3 4(a)(b) Cells 5, 6, 7, 8, 9, 10, 11, 12(a)(b)

Burnup ::;; 55,000 MWD/MTU Burnup ::;; 40,000 MWD/MTU 1 Cooling time <:: 1O years Cooling time <:: 25 years Initial Enrichment<:: 3.4 wt% U-235 Initial Enrichment<:: 2.3 wt% U-235 Burnup ::;; 45,000 MWD/MTU Burnup ::;; 45,000 MWD/MTU 2 Cooling time <:: 1O years Cooling time <:: 20 years Initial Enrichment<:: 3.2 wt% U-235 Initial Enrichment<:: 3.2 wt% U-235 Burnup ::;; 55,000 MWD/MTU Burnup ::;; 45,000 MWD/MTU 3 Cooling time <:: 1O years Cooling time <:: 20 years Initial Enrichment<:: 3.4 wt% U-235 Initial Enrichment<:: 3.2 wt% U-235 Burnup ::;; 45,000 MWD/MTU Burnup ::;; 40,000 MWD/MTU 4 Cooling time <:: 1O years Cooling time <:: 12 years Initial Enrichment<:: 3.6 wt% U-235 Initial Enrichment<:: 3.2 wt% U-235 Burnup ::;; 45,000 MWD/MTU Burnup ::;; 40,000 MWD/MTU 5 Cooling time <:: 14 years Cooling time <:: 12 years Initial Enrichment<:: 3.4 wt% U-235 Initial Enrichment<:: 3.2 wt% U-235 Burnup ::;; 45,000 MWD/MTU Burnup ::;; 40,000 MWD/MTU 6 Cooling time <:: 20 years Cooling time <:: 20 years Initial Enrichment<:: 3.2 wt% U-235 Initial Enrichment<:: 2.3 wt% U-235 INDIAN POINT 2 . 3.1.2-6 Amendment No. TBD

STC Loading 3.1.2 Table 3.1.2-3 (Sheet 2 of 2)

Allowable STC Loading Configurations Configuration(c) Cells 1, 2, 3, 4(a)(b) Cells 5, 6, 7, 8, 9, 10, 11, 12(a)(b)

Burnup s 45,000 MWD/MTU Burnup s 45,000 MWD/MTU 7 Cooling time ~ 10 years Cooling time ~ 12 years Initial Enrichment~ 3.2 wt% U-235 Initial Enrichment~ 3.2 wt% U-235 Burnup ::;;; 55,000 MWD/MTU Burnup ::;;; 55,000 MWD/MTU 8 Cooling time ~ 1O years Cooling time ~ 15 years Initial Enrichment~ 3.4 wt% U-235 Initial Enrichment~ 3.4 wt% U-235 Burnup ::;;; 55,000 MWD/MTU Burnup ::;;; 45,000 MWD/MTU 9 Cooling time ~ 11 years Cooling time ~ 12 years Initial Enrichment~ 3.4 wt% U-235 Initial Enrichment~ 3.2 wt% U-235 Burnup ::;;; 45,000 MWD/MTU . Burnup ::;;; 55,000 MWD/MTU 10 Cooling time ~ 1O years Cooling time ~ 15 years Initial Enrichment ~ 3.2 wt% U-235 Initial Enrichment~ 3.4 wt% U-235 Burnup ::;;; 45,000 MWD/MTU Burnup s 45,000 MWD/MTU 11 Cooling time ~ 6 years Cooling time ~ 14 years Initial Enrichment ~.3.2 wt% U-235 Initial Enrichment~ 3.2 wt% U-235

' Burnup ::;;; 60,000 MWD/MTU Burnup s 50,000 MWD/MTU 12 Cooling time ~ 9 years Cooling time ~ 14 years Initial Enrichment~ 4.2 wt% U-235 Initial Enrichment~ 3.6 wt% U-235 (a) Initial enrichment is the assembly average enrichment. Natural or enriched uranium blankets are not considered in determining the fuel assembly average enrichment for comparison to the minimum allowed initial average enrichment.

(b) Rounding to one decimal place to determine initial enrichment is permitted.

(c) Fuel with five middle lnconel spacers are limited to cells 1, 2, 3, and 4 for all loading configurations except loading configuration 6 which allows fuel with lnconel spacers in all cells.

INDIAN POINT 2 3.1.2-7 Amendment No. TBD

ATTACHMENT 3 TO NL-16-118 Indian Point Unit 2 Technical Specification Bases Changes (Marked-up)

(for information only)

Affected Appendix A Tech Spec Bases Pages: B 3.7.13-5 B 3.7.13-7 B 3.7.13-11 Affected Appendix C Tech Spec Bases Pages: B3.1.2-1 B 3.1.2-2 B 3.1.2-5 Entergy Nuclear Operations, Inc.

Indian Point Unit 2 Docket No. 50-247

Spent Fuel Pit Storage B 3.7.13 BASES BACKGROUND (continued)

Peripheral" Cells, consisting of six select cells along the SFP west wall in Region 2-2, are shown in Figure 3.7.13-5. These six "peripheral" cells may be used to store fuel that meets the requirements for storage in any other location in the SFP. Cells between and adjacent to the "peripheral" cells may be filled with fuel assemblies that meet the requirements of Figure 3.7.13-2 (i.e., meet the requirements for storage in Region 2-2). The two prematurely discharged fuel assemblies meet the requirements of Figure 3.7.13-4 and qualify for storage in the "periphr"-'er~a~l"~c~e~lls~*---------.

Amendments 268 and TBD IP3 Fuel Assemblies The SFP is also used to stor spent fuel transferred from the IP3 SFP. The IP3 fuel assembly storage I cation is also restricted in accordance with LCO 3.7.13 that limits IP3 fuel ssemblies to Region 1-2 of the IP2 SFP. The NRC Refs. 8 and 12 has issued Amendment 268 for the inter-unit transfer of spent nuclear fuel

( . ). The Amendment is based on evaluations conducted for each aspect of the inter-unit transf r of fuel as documented in Reference 9.

APPLICABLE IP2 Fuel Assemblies These Amendments are SAFETY ANALYSES As required by 10 CFR 50.68, "Criticality Accident Requirements" (Ref. 1), if the Spent Fuel Pit takes credit for soluble boron, then "the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water."

NET-173-01, "Criticality Analysis for Soluble Boron and Burnup Credit in the Con Edison Indian Point Unit No. 2 Spent Fuel Storage Racks" (Ref. 4) and NET-173-02, "Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis," (Ref. 5) determined that 10 CFR 50.68(b)(4) will be met during normal SFP operation and all credible accident scenarios (including the

• affects of boraflex degradation) if: a) Spent Fuel Pit boron concentration is maintained within the limits of LCO 3.7.12, "Spent Fuel Pit Boron Concentration," and, b) fuel assembly storage location within the Spent Fuel Pit is restricted based on the fuel assembly's initial enrichment, burnup, decay of Pu 241 (i.e., cooling time) and number of Integral Fuel Burnable Absorbers (IFBA) rods.

INDIAN POINT 2 B3.7.13-5 Revision

Spent Fuel Pit Storage B 3.7.13 BASES APPLICABLE SAFETY ANALYSIS (continued) 2000 ppm to 786 ppm is not a credible event because of the low frequency of postulated initiating events and because the event would be readily detected and mitigated by plant personnel through alarms, flooding, and operator rounds through the SFP area.

Reference 4 and 5 are based on conservative projections of amount of Boraflex absorber panel degradation assumed in each sub-region. These projections are valid through the end of the year 2006. These compensatory measures for boraflex degradation in the SFP were evaluated by the NRC in Reference 6. Based upon BADGER testing in calendar years 2003, 2006

• and 201 O and RACKLIFE code projections, the validity of the critically and boron dilution analysis documented in References 4, 5, 7 and 10 can be extended through the end of the current license (September 28, 2013).

Based on Reference 11, BADGER testing was performed in 2013, to confirm the progression of localized Boraflex dissolution. The continued validity of the criticality and boron dilution analysis will be verified based on the boron monitoring program as defined in the License Renewal Application.

IP3 Fuel Assemblies An analysis, documented in.Reference 9, evaluated the effeot of modeling IP3 integral* and disorete burnable absorbers on reaotivity in the IP2 spent fuel pool using ourrent methodologies. A reaotivity bias was determined. In order to offset this bias, and maintain the validity of the IP2 SFP oritioality analysis, it *.vas determined that IP3 fuel assemblies can be stored in the IP2 SFP with the following restrictions: *

a. IP3 fuel assemblies shall be stored in Region 1-2 of the IP2 Spent Fuel Pit, and

b. The fuel assembly initial enriohment ~ 3.2 and ,s. 4 .4 Nl-o 1

U235,and

o. The fuel assembly disoharge Cyole > 1 and ,s. 11 .

The configuration of fuel assemblies in the Spent Fuel Pit satisfies Criterion 2 of 10 CFR 50.36{c)(2)(ii). .

INDIAN POINT 2 B 3.7.13 - 7 Revision

Spent Fuel Pit Storage B3.7.13 BASES REFERENCES (continued)

7. NETCO Letter to M. R. Hansler from E. Lindquist, Northeast Technology Corp. dated 12/19/06, Subject - Reference 4 and 5 extension.

8. Safety Evaluation by the- Office of Nuclear Reactor Regulation Related to Amendment No. 268 to Facility Operating License No.

DPR-26, July 13, 2012.

9. Holtec Report Hl-2094289, Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at Indian Point Energy Center, Revision 6.

10. NETCO Letter to Floyd Gumble from Matt Harris dated 12/22/2009, titled "Indian Point2 RACKLIFE Projections Through 201Oand2012 BADGER Tests with RACKLIFE Version 2.1".

11. NETCO Letter to Giancarlo Delfini from Matt Harris dated 12/12/2012, titled "Update of IP2 RACKLIFE Model - (In partial fulfillment of Entergy Contract 10351857, Change Order No. 1, Task 2A)".

12. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No.

TBD to Facility Operating License No. DPR-26, TBD.

13. Holtec Report Hl-2094289, Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at Indian Point Energy Center, Revision 8.

INDIAN POINT 2 B3.7.13-11 Revision

STC Loading B 3.1.2 B 3.1 INTER-UNIT FUEL TRANSFER B 3.1.2 Shielded Transfer Canister (STC) Loading BASES BACKGROUND As required by plant operations IP3 spent fuel is transferred to the IP2 spent fuel pit in order to maintain adequate fuel storage capacity in the I P3 spent fuel pit. IP3 spent fuel moved to the I P2 spent fuel pit is subsequently transferred to dry cask storage at the IPEC on-site Independent Spent Fuel Storage Installation (ISFSI) as part of spent fuel inventory management in the IP2 spent fuel

• The NRG has issued Amendment 268 for the inter The NRC has issued Inter-unit fuel transfer operations are conducted using the Amendments 268 and TBD.for Shielded Transfer Canister (STC} and the HI-TRAC 1000 transfer the inter-unit transfer of spent cask. The STC is a bolted-lid pressure vessel with an internal fuel nuclear fuel (Refs. 1 and 5) basket that accommodates up to twelve IP3 spent fuel

...___ _ _ _ _ _ _ _ _ __.assemblies. The STC is loaded in the IP3 spent fuel pit, placed into the HI-TRAC transfer cask in the Fuel Storage Building (FSB)

. truck bay, and moved outside the truck bay on air pads or other approved conveyance. The STC/Hl-TRAC assemblage is transported from outside the IP3 FSB truck bay to just outside the IP2 FSB truck bay with a Vertical Cask Transporter (VCT) and moved into the IP2 FSB truck bay on a Low Profile Transporter.

The STC is removed from the HI-TRAC using the cask handling crane and placed into the IP2 spent fuel pit. The STC lid is removed and the IP3 fuel assemblies are moved to their designated IP2 wet storage rack cell locations with the spent fuel bridge crane in accordance with IP2 Appendix A TS LCO 3. 7 .13.

Fuel assemblies to be transferred are selected at IP3 based on the requirements for loading in the STC. The STC fuel loading requirements are such that the fuel selected for transfer to IP2 is suitable for storage in the designated IP2 spent fuel pit storage racks within the limits of IP2 Appendix A TS LCO 3.7.13 and there are open fuel cells available. Fuel move sheets will govern the transfer of the spent fuel from IP3 to IP2.

Table 3.1.2-1 "Minimum Burnup Requirements at Varying Initial Enrichments" is used to classify each assembly as either a Type 1 assembly or a Type 2 assembly based on initial U-235 enrichment and average assembly burnup. This classification is used to determine if, and where, the fuel assembly can be placed in the STC fuel basket. In the STC design, the fuel basket is divided into INDIAN POINT 2 B 3.1.2-1 Revision

STC Loading B 3.1.2 BASES BACKGROUND twelve cells as shown in Figure 3.1.2-1, "Shielded Transfer (continued) Canister Layout (Top View)". Type 2 assemblies are relatively less reactive assemblies and include any assembly that meets the minimum assembly average burnup at a given initial enrichment of Table 3.1.2-1. Type 2 assemblies may be stored in any cell in the STC subject to the* additional restrictions of the LCO. These additional restrictions include post-irradiation cooling time, initial enrichment, allowable average burnup and decay heat of fuel and non fuel hardware as specified in Table 3.1.2-2 "Non Fuel Hardware Post Irradiation Cooling Times and Allowable Average Burnup" and Table 3.1.2-3 "Allowable STC Loading Configurations".

Type 1 assemblies are relatively more reactive assemblies and include any assembly that does not meet t,he minimum assembly average burnup at a given initial enrichment of Table 3.1.2-1.

Type 1 fuel must be placed in the outer cells of the STC subject to the additional restrictions of the LCO. These additional restrictions include post-irradiation cooling time, initial enrichment, allowable average 'burnup and decay heat of fuel and non fuel hardware as specified in Tables 3.1.2-2 and 3.1.2-3.

Together, the limits on Type 1 and Type 2 fuel assemblies and associated non fuel hardware ensure the criticality, shielding, structural and thermal analyses performed for the STC remain bounding.

To ensure that fuel assemblies selected for transfer can be stored in the IP2 SFP only fuel assemblies with initial average enrichment> 3.2 and< 4.4 wt% U 235 and discharged prior to IP3 Cycle 12 can be placed in the STC basl<et.

Fuel assemblies with an initial enrichment > 5.0 wt% U-235 are not shown on Table 3.1.2-1 and cannot be placed in the STC in accordance with TS 3.1.2.

APPLICABLE The STC has been analyzed for criticality prevention, heat SAFETY rejection capability, shielding, and structural integrity to ensure ANALYSES safe transfer operations from the time that the STC is loaded at .

IP3 to the time it is unloaded at IP2 (~) * , _ I_ _ _ _ ____,

1 Refs. 2 and 6 As required by 10 CFR 50.68 (Ref. 3), if no credit for soluble boron in the STC is taken then, the keff of the STC fuel basket loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

INDIAN POINT 2 B 3.1.2-2 Revision

STC Loading B 3.1.2 BASES ACTIONS assemblies or non fuel hardware from the STC back into the IP3 (continued) spent fuel pool in accordance with Appendix A Technical Specification LCO 3.7.16.

Either action places the fuel in equally safe locations.

The completion time of .Immediately is appropriate because fuel located in the STC may be in an unanalyzed condition and action is required to be initiated and completed without delay to restore the fuel location to an analyzed configuration.

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS This SR verifies by administrative means that the fuel assembly meets the requirements of the STC location in which it is to be placed and that it also meets the requirements for storage in the IP2 spent fuel pit in aooordanoe with IP2 Appendix A TS LCO 3.7.13. This SR ensures the LCO limits for fuel selection and location in the STC and IP2 spent fuel pit are met and the supporting technical analyses remain bounding for all inter-unit transfer operations.

SR 3.1.2.2 This SR verifies by visual inspection that a cell blocker is installed prior to placing a Type 1 fuel assembly and/or non fuel hardware in the STC. This SR ensures that a Type 1 fuel assembly cannot be inserted into STC cells 1, 2, 3, and 4.

REFERENCES 1. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 268 to Facility

5. Safety Evaluation by the Operating License No. DPR-26, July 13, 2012.

Office of Nuclear Reactor Regulation Related to 2. Holtec Report Hl-2094289, Licensing Report on the Inter-Amendment No. TBD to Facility Unit Transfer of Spent Nuclear Fuel at Indian Point Energy Operating License No. DPR-26, Center, Revision 6. -

TBD.

3. 10 CFR 50.68, "Criticality Accident Requirements."

6. Holtec Report Hl-2094289, Licensing Report on the Inter- 4. Double contingency principle of ANSI N16.1-1975, as Unit Transfer of Spent Nuclear specified in the April 14, 1978 NRC letter (Section 1.2) and Fuel at Indian Point Energy \ . implied in the proposed revision to Regulatory Guide 1.13

' Center, Revision 8. (Section 1.4, Appendix A).

INDIAN POINT 2 B 3.1.2-5 Revision

ATTACHMENT 4 TO NL-16-118 Indian Point Unit 3 Proposed Technical Specification Changes (Marked-up)

Affected Appendix C Tech Spec Pages: 3.1.2-1 3.1.2-6 3.1.2-7 (new)

Entergy Nuclear Operations, Inc.

Indian Point Unit 3 Docket No. 50-286

STC Loading .

3.1.2 3.1 INTER-UNIT FUEL TRANSFER 3.1.2 Shielded Transfer Canister {STC) Loading LCO 3.1.2 INTACT FUEL ASSEMBLIES placed into the Shielded Transfer Canister (STC) shall be classified in accordance with Table 3.1.2-1 based on initial enrichment and burnup and shall be restricted based on the following:

a. INTACT FUEL ASSEMBLIES classified as Type 2 may be placed in the STC basket (see Figure 3.1.2-1) with the following restrictions:

1. Post-irradiation cooling time, initial enrichment, and allowable average burnup shall be within the limits for the cell locations as specified in Table 3.1.2-3;

2. Decay heat including NON FUEL HARDWARE::;; 650 Watts (sells 5 through 12)1_1.2 k"'! (any cell~;

3.

FUEL HARDWARE::;; 9.621 k ;

4. Post-irradiation cooling time and the maximum average burnup of NON FUEL HARDWARE shall be within the cell lqcations and limits specified in Table 3.1.2-2. In accordance with Table 3.1.2-2 RCCAs and Hafnium Flux Suppressors cannot be placed in locations 5, 6, 7, 8, 9, 10, 11, 12 of the STC basket.

-NOTE-If one or more Type 1 fuel assemblies are in the STC, cells 1, 2, 3, AND 4 must be empty, with a cell blocker installed that prevents inserting fuel assemblies and/or NON-FUEL HARDWARE.

b. INTACT FUEL ASSEMBLIES classified as Type 1 or Type 2 may be placed in locations 5, 6, 7, 8, 9, 10, 11, 12 of the STC basket (see Figure 3.1.2-1) with the following restrictions:

1. Post-irradiation cooling time, initial enrichment, and allowable average burnup shall be within the limits for the cell locations as specified in Table 3.1.2-3;

2. Decay heat including NON FUEL HARDWARE::;; 650 'Nattsl1.2 kWt

3. Post-irradiation cooling time and the maximum average burnup of NON FUEL HARDWARE shall be within the cell locations and limits specified in Table 3.1.2-2. In accordance with Table 3.1.2-2 RCCAs and Hafnium Flux Suppressors cannot be placed in locations 5, 6, 7, 8, 9, 1O, 11, 12 of the STC basket.

e. Only INTACT FUEL ASSEMBLIES Nith initial average enriehment > 3.2 1

and < 4.4 vlt% U 235 and diseharged prior to I P3 Cyele 12 shall be plaeed in the STC basket.

INDIAN POINT 3 3.1.2-1 Amendment No. TBD

l STC Loading 3.1.2 Table 3.1.2-3l(Sheet 1 of 2) I Allowable STC Loading Configurations Configuration(c) Cells 1 2 3 4(a)(b) Cells 5, 6, 7, 8, 9, 10, 11, 12(a)(b)

Burnup :::;; 55,000 MWD/MTU Burnup :::;; 40,000 MWD/MTU 1 Cooling time ;::: 1O years Cooling time ;::: 25 years Initial Enrichment ;::: 3.4 wt% U-235 Initial Enrichment ;::: 2.3 wt% U-235 Burnup :::;; 45,000 MWD/MTU Burnup :::;; 45,000 MWD/MTU 2 Cooling time ;::: 1O years Cooling time;::: 20 years Initial Enrichment;::: 3.2 wt% U-235 Initial Enrichment;::: 3.2 wt% U-235 Burnup :::;; 55,000 MWD/MTU Burnup :::;; 45,000 MWD/MTU 3 Cooling time ;::: 1O years . Cooling time ;::: 20 years Initial Enrichment ;::: 3.4 wt% U-235 Initial Enrichment;::: 3.2 wt% U-235 Burnup :::;; 45,000 MWD/MTU Burnup :::;; 40,000 MWD/MTU 4 Cooling time ;::: 1O years Cooling time ;::: 12 years Initial Enrichment;::: 3.6 wt% U-235 Initial Enrichment;::: 3.2 wt% U-235 Burnup :::;; 45,000 MWD/MTU Burnup :::;; 40,000 MWD/MTU 5 Cooling time ;::: 14 years Cooling time ;::: 12 years Initial Enrichment;::: 3.4 wt% U-235 Initial Enrichment;::: 3.2 wt% U-235 Burnup :::;; 45,000 MWD/MTU Burnup :::;; 40,000 MWD/MTU 6 Cooling time ;::: 20 years Cooling time ;::: 20 years Initial Enrichment;::: 3.2 wt% U-235 Initial Enrichment;::: 2.3 wt% U-235 INDIAN POINT 3 3.1.2-6 Amendment No. TBD

STC Loading 3.1.2 Table 3.1.2-3 (Sheet 2 of 2)

Allowable STC Loading Configurations Configuration(c) Cells 1 2 3 4(a)(b) Cells 5, 6, 7, 8, 9, 10, 11, 12(a)(b)

Burnup :5 45,000 MWD/MTU Burnup :5 45,000 MWD/MTU 7 Cooling time ;:: 1O years Cooling time ;:: 12 years Initial Enrichment;:: 3.2 wt% U-235 Initial Enrichment;:: 3.2 wt% U-235 Burnup :5 55,000 MWD/MTU Burnup :5 55,000 MWD/MTU 8 Cooling time ;:: 1O years Cooling time ;:: 15 years Initial Enrichment;:: 3.4 wt% U-235 Initial Enrichment;:: 3.4 wt% U-235 Burnup :5 55,000 MWD/MTU* Burnup :5 45,000 MWD/MTU 9 Cooling time ;:: 11 years Cooling time ;:: 12 years Initial Enrichment;:: 3.4 wt% U-235 Initial Enrichment;:: 3.2 wt% U-235 Burnup :5 45,000 MWD/MTU Burnup :5 55,000 MWD/MTU 10 _Cooling time ;:: 1O years Cooling time ;:: 15 years Initial Enrichment ;:: 3.2 wt% U-235 - Initial Enrichment;:: 3.4 wt% U-235 Burnup :5 45,000 MWD/MTU Burnup :5 45,000 MWD/MTU 11 Cooling time ;:: 6 years Cooling time ;:: 14 years Initial Enrichment ;:: 3.2 wt% U-235 Initial Enrichment;:: 3.2 wt% U-235 Burnup :5 60,000 MWD/MTU Burnup :5 50,000 MWD/MTU 12 Cooling time ;:: 9 years Cooling time ;:: 14 years Initial Enrichment ;:: 4.2 wt% U-235 Initial Enrichment;:: 3.6 wt% U-235 (a) Initial enrichment is the assembly average enrichment. Natural or enriched uranium blankets are not considered in determining the fuel asseml;:>ly average enrichment for comparison to the minimum allowed initial average enrichment.

(b) Rounding to one decimal place to determine initial enrichment is permitted.

(c} FLJel with five middle lnconel spacers are limited to cells 1, 2, 3, and 4 for all loading configurations except loading configuration 6 which allows fuel with lnconel spacers in all cells.

INDIAN POINT 3 3.1.2-7 Amendment No. TBD

ATTACHMENT 5 TO NL-16-118 Indian Point Unit 3 Technical Specification Bases Changes (Marked-up)

(for information only)

Affected Appendix C Tech Spec Bases Pages: B 3.1.2-1 B 3.1.2-2 B 3.1.2-5 Entergy Nuclear Operations, Inc.

Indian Point Unit 3 Docket No. 50-286

STC Loading B 3.1.2 B 3.1 INTER-UNIT FUEL TRANSFER B 3.1.2 Shielded Transfer Canister (STC) Loading BASES BACKGROUND As required by plant operations IP3 spent fuel is transferred to the IP2 spent fuel pit in order to maintain adequate fuel storage capacity in the IP3 spent fuel pit. IP3 spent fuel moved to the IP2 spent fuel pit is subsequently transferred) to dry cask storage at the IPEC on-site Independent Spent Fuel Storage Installation (ISFSI) as part of spent fuel inventory management in the IP2

---

~:::;::: :t{~~.:~ndment 246 Im tlm inter The NRC has issued Inter-unit fuel transfer operations are conducted using the Amendments 246 and TBD for Shielded Transfer Canister (STC) and the HI-TRAC 1000 transfer the inter-unit transfer of spent cask. The STC is a bolted-lid pressure vessel with an internal fuel nuclear fuel (Refs. 1 and 5) basket that accommodates up to twelve I P3 spent fuel

'-------------~assemblies. The STC is loaded in the IP3 spent fuel pit, placed into the HI-TRAC transfer cask in the Fuel Storage Building (FSB) truck bay, and moved outside the truck bay on air pads or other approved conveyance. The STC/Hl-TRAC assemblage is transported from outside the IP3 FSB truck bay to just outside the IP2 FSB truck bay with a Vertical Cask Transporter (VCT) and moved into the IP2 FSB truck bay on a Low Profile Transporter.

The STC is removed from the HI-TRAC using the cask handling crane and placed into the IP2 spent fuel pit. The STC lid is removed and the IP3 fuel assemblies are moved to their designated IP2 wet storage rack cell locations with the spent fuel bridge crane in accordance with IP2 Appendix A TS LCO 3.7.13.

Fuel assemblies to be transferred are selected at I P3 based on the requirements for loading in the STC. The STC fuel loading requirements are such that the fuel selected for transfer to IP2 is suitable for storage in the designated IP2 spent fuel pit storage racks within the limits of IP2 Appendix A TS LCO 3.7.13 and there are open fuel cells available. Fuel move sheets will govern the transfer of the spent fuel from IP3 to IP2.

Table 3.1.2-1 "Minimum Burnup Requirements at Varying Initial

• Enrichments" is used to classify each assembly as either a Type 1 assembly or a Type 2 assembly based on initial U-235 enrichment and average assembly burnup. This classification is used to determine if, and where, the fuel assembly can be placed in the STC fuel basket. In the STC design, the fuel basket is divided into INDIAN POINT 3 B 3.1.2-1 Revision

STC Loading B 3.1.2 BASES BACKGROUND twelve cells as shown in Figure 3.1.2-1, "Shielded Transfer (continued) Canister Layout (Top View)". Type 2 assemblies are relatively less reactive assemblies and include any assembly that meets the minimum assembly average burnup at a given initial enrichment of Table 3.1.2-1. Type 2 assemblies m_ay be stored in any cell in the STC subject to the additional restrictions of the LCO. These additional restrictions include post-irradiation cooling time, initial enrichment, allowable average burnup and decay heat of fuel and non fuel hardware as specified in Table 3.1.2-2 "Non Fuel Hardware Post Irradiation Cooling Times and Allowable Average Burnup" and Table 3.1.2-3 "Allowable STC Loading Configurations".

Type 1 assemblies are relatively more reactive assemblies and include any assembly that does not meet the minimum assembly average burnup at a given initial enrichment of Table 3.1.2-1.

Type 1 fuel must be placed in the outer cells of the STC subject to the additional restrictions of the LCO. These additional restrictions include post-irradiation cooling time, initial enrichment, allowable average burnup and decay heat of fuel and non fuel hardware as specified in Tables 3.1.2-2 and 3.1.2-3.

Together, the limits on Type 1 and Type 2 fuel assemblies and associated non fuel hardware ensure the criticality, shielding, structural and thermal analyses performed for the STC remain bounding.

To ensure that fuel assemblies selected for transfer can be stored in the IP2 SFP only fuel assemblies with initial average enrichment> 3.2 and< 4.4 vl1:% U 235 and discharged prior to IP3 Cycle 12 can be placed in the STC basl<et.

Fuel assemblies with an initial enrichment > 5.0 wt% U-235 are not shown on Table 3.1.2-1 and cannot be placed in the STC in accordance with TS 3.1.2.

APPLICABLE The STC has been analyzed for criticality prevention, heat SAFETY rejection capability, shielding, and structural integrity to ensure ANALYSES safe transfer operations from the time that the STC is loaded at IP3 to the time it is unloade_d at IP2 (~).-1---- - - - - - ,

1 Refs. 2 and 6 As required by 10 CFR 50.68 (Ref. 3), if no credit for soluble boron in the STC is taken then, the kett of the STC fuel basket loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

INDIAN POINT 3 B 3.1.2-2 Revision

STC Loading B 3.1.2 BASES ACTIONS assemblies or non fuel hardware from the STC back into the IP3 (continued) spent fuel pool in accordance with Appendix A Technical Specification LCO 3. 7 .16.

Either action places the fuel in equally safe locations.

The completion time of "Immediately is appropriate because fuel located in the STC may be in an unanalyzed condition and action is required to be initiated and completed without delay to restore the fuel location to an analyzed configuration.

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS This SR verifies by administrative means that the fuel assembly meets the requirements of the STC location .in which it is to be placed and that it also meets the requirements for storage in the IP2 spent fuel pit in accordance with IP2 Appendix A TS LCO 3.7.13. This SR ensures the LCO limits for fuel selection and location in the STC and IP2 spent fuel pit *are met and the supporting technical analyses remain bounding for all inter-unit transfer operations.

SR 3.1.2.2 This SR verifies by visual inspection that a cell blocker is installed prior to placing a Type 1 fuel assembly and/or non fuel hardware in the STC. This SR ensures that a Type 1 fuel assembly cannot be inserted into STC cells 1, 2, 3, and 4.

REFERENCES 1. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 246 to Facility

5. Safety Evaluation by the Operating License No. DPR-64, July 13, 2012.

Office of Nuclear Reactor Regulation Related to 2. Holtec Report Hl-2094289, Licensing Report on the Inter-Amendment No. TBD to Facility Unit Transfer of Spent Nuclear Fuel at Indian Point Energy Operating License No. DPR-64, Center, Revision 6.

TBD.

3. 10 CFR 50.68, "Criticality Accident Requirements."

6. Holtec Report Hl-2094289, Licensing Report on the Inter- 4. Double contingency principle of ANSI N 16.1-1975, as Unit Transfer of Spent Nuclear specified in the April 14, 1978 NRG letter (Section 1.2) and Fuel at Indian Point Energy \ implied in the proposed revision to Regulatory Guide 1 .13 Center, Revision R * (Section 1.4, Appendix A).

INDIAN POINT 3 B 3.1.2-5 Revision

ENCLOSURE 4 TO NL-16-118 Affidavits executed pursuant to 10 CFR 2.390 governing the proprietary information included in the Holtec reports and evaluations.

Entergy Nuclear Operations, Inc.

Indian Point Units 2 and 3 Docket Nos. 50-247 and 50-286

Westinghouse Affidavit Entergy Nuclear Operations, Inc.

Indian Point Units 2 and 3 Docket Nos. 50-247 and 50-286

Westinghouse Non-Proprielmy Class 3 0 e Westinghouse Electric Company 1000 Wastlllghouse Orlva Cranberry Township, Pannsylvanln 16056 USA U.S. Nuclear Regulatory Commission Dheel tel: (4 l2} 374*4643 Document Control Desk Direct fm: (724) 940*8560 l 1555 Rockville Pike e*mail: gresh11ja@westmghouse.com Rookvme. MD 20852 CAW*164501 November 17. 20!6

.A.UUCATIONFQR WITHHOLDINQ PROP.RIW'ARX JNFORMATIONffiOM PUBLIC DISClOStJR&

Subject:

Report HI-2094289. "'Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at the Indian Point Energy Center,0 Revision:~. (l'roprietruy)

The Application for Withholding Proprietary Infommtion from Public Disclosure is submitted by Westinghouse Electric Company LLC C'WesliRghouse"). pursuant to the provisions of paragraph (b){l)

, of: Section Z.390 of the Commission's regulations. It conaains commercial straregic informatlon proprietaxy to Westinghouse and customarily held in confidence*

The proprietary infommt!on on pages 4-24, 4-33 and 4*36 of Report IDb2094289 fur which withholding is being requested is further Identified i.n Affidavit CAW-16-4507 signed by the owner of the propriewy infonnation, Westinghouse .Electri.c Compnny LLC. The Affidavit. which accompzmies this lelter, sets forth the basis on which the infommtion may be whhheld from public disclosure by the Commission ond addresses with specificity lhe considerations listed in paragraph (b)(4) of IO~ Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Entergy Noolear Operations. Inc.

Conespondence with respect to the proprietary aspects of !he Application for Withholding or the Westinghouse Affidavit should reference CAW-16-4507. and shoold be addressed co James A. Gresham, Manager, Regulatory Compliance, Westinghouse Eieclfic Company. 1000 Westinghouse Drive. Building 3 Suire 3IO, Cranberry Township, Permsylvania 16066.

© 2016 Westinghouse .Electric Company LLC. All Rights Reserved.

CAW-16-4507 AEfIDAVIT COM:MONWEALTH OFPENNSYLVANIA:

COT)NTY OF BUTLER:

l, James A. Gresham, am authorized to eKecute this Affidavit on behnlf of Westinghouse Electric Company LLC ("Westinghouse"), and that the avem:ients of .fuct set forth in this Affidavit nre m.ie and correct to Rhe best of my knowledge. inf'onnaiion, and belief.

James A. Gresham, Manager Regulatory Ccmpti1mce

CAW~Hi-4507 t

.1 (I) I am Manager. Regulntory Compliance, Westinghouse Electric Company LLC ("Westinghouse"). iI Md as sooh. r have been specifically delegated the function of reviewing the proprietary lnformation sought to be withheld from public disclosure in connection with nuclear power plant licensing and rwe making proce~lngs, and am auU1orized to apply fo! its withholding cm behalf of Westinghouse.

(2) I am making ibis Affidavit in conformance with the provisions of JO CFR Section 2.390 of the Commission*s regulations and in conjunction with the Westb11ghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit (3) I have personal knowledge of the criteria and procedures udli2ed by Westinghouse in designating information ns n tmde secrelt privileged or as conftdentiat commercial or financial information.

(4) Pursmmt to the provisions of paragraph (b/(4) of Section 2.390 of the Commission's regulations, the foJlowing is furnished for consideration by the Commission in determining whether the information s01.1ght to be withheld from public disclosure should be withheld.

(i} The information sought to be withheld from public disclosure is owned and has been held in confi~noo by Westinghouse.

(ii) The information is of a type customarily beld in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a mtiom11 basis for determining I.he types of information customarily held in confidence by it and, in that coruiecUon, utilizes a system to determine when and whet.tier to hold certain types of information in confidence. The !.lpplication of that system and ~ substance of that system constitute We5tinghouse policy and provide the rational basis required.

Under lhmt system, iruormaaion is held in confidence if it falls in one or more of severnl types, the release of which' might result in the loss of an existing or potential competitive advantage. as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool. method, elc.) where prevention ofits use by 1111y of

4 , CAW-164507 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data. including test data, relative ao a process (or component, structure, tool, method, etc.). the application of which data secures a competitive economic advanlage, e.g.. by optimization or i~proved marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture. shipment, installation. assmimce of guulity, or licensing a similar product.

(d) It reveals cDst or price information, production capacities, budget levels, or commeroinl strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects ofJiast, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(iH)

• There are sound policy reasons behind the Westinghouse system which foclude "the following:

(a) Tbe use of such information by Westinghouse gives Westinghouse a competitive advi:mtage over its competitors. It is, therefore, witl1held from disc1.osure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such infommUon is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resoorees at-our expense.

5 ., CAW-16-4507 (d) Each component of proprietary information pertinent to a. particutnr competitive advantage is potemiatiy ns valuable as the total competitive advantage. If

• compeli!:Ors acquire compone:nts of proprietary information, any one component may be the key to the el'ltire puzzle, thereby 'depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of promimmce of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in reseureh and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv) The information is being transmitted to the Commission in i::onfideru:e and. under the provisions of 10 CFR Section 2.390, is to be received in confidence by the Commission.

(v) The information sought to be protected is not avnilab!e in public sources or available information has not been previously employed in the aame original mrumer or method to tlte best of our knowledge and belief.

(vi) The proprietary informaticm sought to be withhold in this submittal Is that which is appropriately marked in Report m-2094289. pages 4-24, 4-33, and 4*36 (Proprietary).

for submittal to the Commission, being transmitted by Entergy Nuclear Operations, Inc.

letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary informatlon as submitted by Westinghouse is that associated with the inter-unit lramfer of spent nuclear fuel between Indian Point Units 2 and 3, and may be used only for that purpose.

• (a) This information is part of thnt which wm ennble Westinghouse to:

(i) Assist customers In obtaining licensing changes.

CAW-16-4507 (ii} Assist customers in analyzing the spent fuel pool nnd alisorber panels to ensure criticality does not occur.

(b) Further. thisJnfor.rootion lms substantial commemlni vniue ns follows:

(i) Westinghouse plans to sell the t.1se of similar information to its cusmmers for the purpose of assisting in obroining license changes.

(ii) Westinghouse can sell support and defense of speht fuel pool criticalily analyses.

(iii) The information requested to be withheld reveals ;he distinguishing nspects of a methodology which was devefoped by Westinghouse.

Public diselosure of this proprietary information is likely to cause substant~l harm to the competitive position of Westinghouse because it would enhnm:e the ability of competitors to provide similar ttechnicml evaluation justifications and licensing; defense services for commercial power reactors without commem.mmte expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right lo use the infommtion.

The development of the technology described in part by the information is the result of applying the results of many years of experience in im intensive Westinghouse effort and lhe expenditure of a considerable sum of money.

In order for competitors of Westinghouse lo duplkatB lhis information, similar technlc~

programs would have to be performed and a significant mnnpower effort, having the requisite talent and experience, wou1d have to be expended.

Further the deponent snyeth not.

PROPRJETARY INFORMATION NOTICE Transmitted herewith are "proprietary ruld non-proprietary versions* or a document, furnished to the NRC assooiated with the inleMmit tr1msfer of spent fuel between Indian Point Units 2 and 3, and may be used only for that purpose.

In order to conform to the requirements of l 0 CFR 2.390 of the Commission* s regulations concerning the protection of proprietary information so submitted to the NRC, ~he infonnation which fa proprietary *in the proprietary versions is contained wlthin bracketst and where the proprietary information bas been deleted in the oon*proprietary versions. only the brackets remain (the information tlmt was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicaled in both versions by means of iower case lell.ers (a) through (f)

, located ns a supe~cript immediately fuUow111g !he brackets enck.ising each item of information bein~f identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse custornafily holds in confidence idenlified in Sections (4)(il){a}

through (4)(ii)(f) of the Affidavit accompanying ~his transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The .NRC is permitted to make tht'; number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as weU as the issuance.

denial, amendment. transfer. renewal, modification, s~pensh:m, revocation, or violation of a license, permit, order, or regulation subject to the requirements o( 10 CFR 2.390 regarding restriciions on public disc1osure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports. the NRC is pennilled to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing Jn the appropriate docket files in the public document room in Washington, DC and in locru public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this pl.ll'pose. Copies made by the NRC must include the.copyright notice in nll instances nnd the proprietary notice if the original was Identified as proprietary.

Holtec Affidavit Entergy Nuclear Operations, Inc.

Indian Point Units 2 and 3

. Docket Nos. 50-247 and 50-286

U.S. Nuclear Regulatory Commission ATI'N: Document Control Desk AF.FIDA vrr PURSUANT TO l 0 CFR 2.390 I, Kimberly Manzione, being duly sworn, depose and state as foUows:

(1) [have reviewed the in.formation described in paragraph (2) which is sought

. to be withheld, and am authorized to apply for its withholding.

(2) The infonnation sought to be withheld is information provided in LICENSING REPORT ON THE INTER-UNIT TRANSFER of SPENT NUCLEAR FUEL at THE INDIAN POINT ENERGY CENTER, Revision

8. This document contains Holtec Proprietary information.

(3) In making this application for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act (11FOIA11 ), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a.)(4), 2.390(a)(4), and 2.390(b)(l) for 11trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Ex.emption 4). The material for which exemption from disclosure is here sought is all "confidential commercial infonnation",

and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear

~egulato01 Commission, 975F2d871 (DC Cir. 1992), and Public Citizen Health Research GrouR v. FDA~ 704F2dl280 (DC Cir. l 983).

J ofS

U.S. Nuclear Regulatory Commission A1TN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method1 or apparatus, jncluding supporting data and analyses; where prevention of its use by Holtec*s competitol".s without license from Holtec International constitutes a competitive economic advantage over other companies;

b. . Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive ppsition in the design, manufacture, shipment, installatiori, assurance of quality, or licensing of a similar product.

c. Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of HoUec International, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future Holtcc International customer-funded development plans and programs of potential commercial value to Holtec International;

e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The infonnation sought lo be withheld is considered lo be proprietary for the reasons set forth in paragraphs 4.a, 4.b, and 4.e above.

{5) The information sought to be withheld is being submitted to the NRC in confidence. The infonnation (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The iafonnation sought to be withheld has, to the best of my knowledge and belief, consistent1y been he1d in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any* required transmittals to the NRC) have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for 2of5

U.S. Nuclear Regulatory Commission

. ATIN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR l.390 maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized discJosure~ are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Holtec International is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Holtoo International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the infonnation, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodoJogies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec Intemational's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtec's competitor to copy our technology and offer it for sale in competition with our company, causing us financial injury.

3 of5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AF.FIDAVDT .PURSUANT TO 10 CFR 2.390 (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm Lo Holtec Intemational 1s competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of HoJtec lntemational's comprehensive spent fuel storage technology basef and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to delennine and apply the appropriate evaluation process.

The research1 development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but jt clearly is .

substantial.

Holtec Intemational's competitive advantage wilt be lost if its competitors

.are abJe to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this infonnation to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical too1s.

4of5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 STATE OF NEW JERSEY }

) ss; COUNTY OF BURLINGTON )

Kimberly Manzione) being duly sworn, deposes and says:

That she has read the foregoing ~ffidavit and the matters stated therein are true and correct to the best of her knowledge, infonnation, and belief.

Executed at Marlton, New Jersey, this 28th day of Novembert 20 J6.

~~

Kimberly Manzione Licensing Manager Holtec International Subscribed and sworn before me this 28th day of November, 2016 .

. 5of5