Text

Technical Specifications Proposed Change - Permanently Defueled Technical Specifications
	ML19105B241
Person / Time
Site:	Indian Point
Issue date:	04/15/2019
From:	Halter M; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-19-026
	Download: ML19105B241 (282)
	v • d • e

Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5573 Mandy K. Halter Director, Nuclear Licensing 10 CFR 50.90 NL-19-026 April 15, 2019 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

SUBJECT:

Technical Specifications Proposed Change - Permanently Defueled Technical Specifications Indian Point Nuclear Generating Station Unit 2 NRC Docket No. 50-247 Renewed Facility Operating License No. DPR-26 In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.90, "Application for amendment of license or construction permit," Entergy Nuclear Operations, Inc. (Entergy) is proposing an amendment to Renewed Facility Operating License (OL) DPR-26 for Indian Point Nuclear Generating Station Unit No. 2 (IP2). This proposed license amendment would revise the IP2 OL and revise the Technical Specifications (TSs) in Appendix A to Permanently Defueled Technical Specifications (PDTS), the Environmental Technical Specification Requirements in Appendix B of the OL, and the Inter-Unit Transfer Technical Specifications in Appendix C. The proposed changes are consistent with the permanent cessation of reactor operation and permanent defueling of the reactor. The proposed changes would revise certain requirements contained within the IP2 OL and the Appendices A through C TSs and remove the requirements that would no longer be applicable after IP2 is permanently shut down and defueled.

In Reference 1, Entergy notified the U.S. Nuclear Regulatory Commission (NRC) that it has decided to permanently cease operations of IP2 by April 30, 2020. Once certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel are submitted to the NRC in accordance with 10 CFR 50.82(a)(1)(i) and (ii) and they are docketed, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).

The proposed changes to the IP2 OL and TSs are in accordance with 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5). The proposed changes also include a renumbering of pages in the

NL-19-026 Page 2 of 3 OL and the Appendix A TSs, where appropriate, to condense and reduce the number of pages without affecting the technical content. The Appendix A TSs Table of Contents is also accordingly revised. References 2, 3, and 4 have been submitted to the NRC and are currently in review.

Entergy has reviewed the proposed amendment in accordance with 10 CFR 50.92 and concludes it does not involve a significant hazards consideration.

In accordance with 10 CFR 50.91, copies of this application, with the enclosure, are being provided to the New York State Department of Health and Emergency Management Agency.

The Enclosure to this letter provides a detailed description and evaluation of the proposed changes for IP2. Attachment 1 to the Enclosure contains a mark-up of the current OL, TS, and TS Bases pages. The TS Bases pages are provided for information only. Attachment 2 to the Enclosure contains the retyped Renewed Facility License, PDTS, Appendices B and C TSs, and PDTS Bases pages in their entirety.

Entergy requests review and approval of this proposed license amendment by April 30, 2020, with a 60-day implementation period from the effective date of the amendment. The License Amendment will not be implemented until the certifications required by 10 CFR 50.82(a)(1)(i) have been docketed in accordance with 10 CFR 50.82(a)(2), the decay time requirement established in the analysis of the Fuel Handling Accident in the Fuel Handling Building has been met, the license amendment requests regarding the IP2 TSs affecting the spent fuel storage requirements and administrative controls for the permanently defueled condition submitted in References 2 and 3 have been approved by the NRC and implemented, and the Certified Fuel Handler Training and Retraining Program submitted in Reference 4 has been approved by the NRC.

If you have any questions regarding this submittal, please contact Mr. Robert Walpole, Manager, Regulatory Assurance, at (914) 254-6710.

There are no new regulatory commitments made in this letter.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 15, 2019.

Respectfully, Mandy K. Halter MKH/cdm/std

NL-19-026 Page 3 of 3

Enclosure:

Indian Point Nuclear Generating Station Unit 2 - Description and Evaluation of the Proposed Changes Attachments to

Enclosure:

1. Indian Point Nuclear Generating Station Unit 2 - Mark-up of the Current Facility Operating License, Appendices A through C Technical Specifications, and Appendix A Technical Specifications Bases

2. Indian Point Nuclear Generating Station Unit 2 - Re-typed (Clean) Facility License, Appendix A Permanently Defueled Technical Specifications, Appendices B and C Technical Specifications, and Appendix A Permanently Defueled Technical Specifications Bases

References:

1. Entergy Nuclear Operations, Inc. (Entergy) letter to U.S. Nuclear Regulatory Commission (NRC), "Notification of Permanent Cessation of Power Operations," dated February 8, 2017 (Letter No. NL-17-021)

(ADAMS Accession No. ML17044A004)

2. Entergy letter to NRC, "Indian Point Nuclear Generating Unit No. 2, Proposed License Amendment Regarding Spent Fuel Storage," dated December 11, 2017 (Letter No. NL-17-144) (ADAMS Accession No. ML17354A014)

3. Entergy letter to NRC, "Technical Specifications Proposed Change -

Administrative Controls for Permanently Defueled Condition," dated April 15, 2019 (Letter No. NL-19-013)

4. Entergy letter to NRC, "Request for Approval of a Certified Fuel Handler Training and Retraining Program," dated April 15, 2019 (Letter No.

NL-19-012) cc: NRC Senior Project Manager, NRC NRR DORL Regional Administrator, NRC Region I NRC Senior Resident Inspector, Indian Point Energy Center President and CEO, NYSERDA New York State (NYS) Public Service Commission NYS Department of Health - Radiation Control Program NYS Emergency Management Agency

Enclosure NL-19-026 Indian Point Nuclear Generating Station Unit 2 Description and Evaluation of Proposed Changes

Enclosure NL-19-026 Page 1 of 74 Indian Point Nuclear Generating Station Unit 2 Description and Evaluation of Proposed Changes

1.

SUMMARY

DESCRIPTION On February 8, 2017, Entergy Nuclear Operations, Inc. (Entergy) notified the U.S. Nuclear Regulatory Commission (NRC) that it would permanently cease power operations at Indian Point Nuclear Generating Station Unit No. 2 (IP2) no later than April 30, 2020 (Reference 1). Once certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel are submitted to the NRC in accordance with 10 CFR 50.82(a)(1)(i) and (ii) and they are docketed, the 10 CFR Part 50 license no longer will permit operation of the reactor or placement of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).

This proposed license amendment would revise the IP2 Operating License (OL) and revise the Technical Specifications (TSs) in Appendix A of the OL to Permanently Defueled Technical Specifications (PDTS), the Environmental Technical Specification Requirements in Appendix B of the OL, and the Inter-Unit Fuel Transfer Technical Specifications in Appendix C. The proposed changes are consistent with the permanent cessation of reactor operation and permanent defueling of the reactor. The proposed changes would revise certain requirements contained within the IP2 OL and TSs and remove the requirements that would no longer be applicable after IP2 is permanently shut down and defueled.

The License Amendment will not be implemented until the certifications required by 10 CFR 50.82(a)(1)(i) have been docketed in accordance with 10 CFR 50.82(a)(2), the decay time requirement established in the analysis of the Fuel Handling Accident in the Fuel Handling Building has been met, the License Amendment Requests to modify the IP2 TSs regarding spent fuel storage requirements and administrative controls for the permanently defueled condition submitted in References 2 and 3 have been approved by the NRC and implemented, and the Certified Fuel Handler Training and Retraining Program submitted in Reference 4 has been approved by the NRC.

2. DETAILED DESCRIPTION AND BASIS FOR THE CHANGES The License Amendment Request modifies the IP2 OL and revises the IP2 Appendix A TSs into PDTS, and IP2 Appendices B and C TSs to comport with a permanently shut down and defueled condition.

The proposed changes to the IP2 OL and TSs are in accordance with 10 CFR 50.36(c)(1) through 10 CFR 50.36(c)(5). The proposed changes also include a renumbering of pages in the OL and the Appendix A TSs, where appropriate, to condense and reduce the number of pages without affecting the technical content. The Appendix A TSs Table of Contents is also accordingly revised.

This License Amendment Request assumes that the changes to the IP2 TSs proposed in References 2 and 3 and that the Certified Fuel Handler Training and Retraining Program submitted in Reference 4 have been approved by the NRC and implemented by the site.

General Analysis Applicable to Proposed Change The regulatory requirements related to the content of TSs are promulgated in 10 CFR 50.36, Technical Specifications. As detailed in a subsequent section of this proposed amendment, this regulation lists four criteria to define the scope of equipment and parameters that must be included in TSs. In a permanently defueled condition, the scope of equipment and parameters that must be included in the IP2 TSs is limited to those needed to address the remaining postulated Design

Enclosure NL-19-026 Page 2 of 74 Basis Accidents (DBAs) that will remain applicable to IP2 in the permanently shut down and defueled condition, so that the consequences of the accident are maintained within acceptable limits.

Chapter 14 of the IP2 Updated Final Safety Analysis Report (UFSAR) describes the DBA and transient scenarios applicable to IP2 during power operations. During normal power operations, the forced inlet flow of water through the reactor coolant system (RCS) removes the heat from the reactor by generating steam. The RCS, operating at high temperatures and pressures, transfers this heat through the steam generator tubes to the secondary system. The most severe postulated accidents for nuclear power plants involve damage to the nuclear reactor core and the release of large quantities of fission products to the RCS. Many of the accident scenarios postulated in the UFSAR involve failures or malfunctions of systems which could affect the reactor core.

After the certifications are submitted for permanent cessation of operations and removal of fuel from the IP2 reactor vessel in accordance with 10 CFR 50.82(a)(1)(i) and (ii), and docketed pursuant to 10 CFR 50.82(a)(2), the majority of DBA scenarios postulated in the IP2 UFSAR will no longer be possible. The irradiated fuel will be stored in the Spent Fuel Pit (SFP) and the Independent Spent Fuel Storage Installation (ISFSI) until it is shipped off site in accordance with the schedules to be provided in the Post Shut Down Decommissioning Activities Report (PSDAR) and the Irradiated Fuel Management Plan.

Chapter 14 of the IP2 UFSAR describes the safety analysis aspects of the plant that were evaluated to demonstrate that the plant could be operated safely and that radiological consequences from postulated accidents do not exceed the applicable limits. This chapter is divided into four sections, each dealing with a different behavior category:

1. Core and Coolant Boundary Protection Analysis, Section 14.1 - The incidents presented in Section 14.1 generally have no offsite radiation consequences.

2. Standby Safeguards Analysis, Section 14.2 -The accidents presented in Section 14.2 are more severe and may cause the release of radioactive material to the environment.

3. Rupture of a Reactor Coolant Pipe, Section 14.3 - The accident presented in Section 14.3, the rupture of a reactor coolant pipe, is the worst-case accident and is the primary basis for the design of engineered safety features.

4. Anticipated Transients Without Scram, Section 14.4 - The accidents presented in Section 14.4 were assumed to occur without the benefit of tripping the reactor. While the failure to trip is unlikely, several accidents were evaluated for which credit was not taken for a reactor trip.

Safety analyses are evaluated against regulatory acceptance criteria and are integral of the plant's design and licensing basis. The safety analyses demonstrate the integrity of the fission product barriers, the capability to shut down the reactor and maintain it in a safe shutdown condition, and the capability to prevent or mitigate the consequences of accidents and transients. Systems, Structures, and Components (SSCs) that perform design basis functions are credited in the safety analyses for the purpose of mitigating the transient or accident.

A list of the IP2 UFSAR Chapter 14 transients and DBAs and a determination regarding whether the transient or accident applies to a permanently defueled condition is provided in Table 2-1.

Enclosure NL-19-026 Page 3 of 74 Table 2 IP2 DBAs UFSAR Postulated Accident or Transient Defueled Applicability Section Uncontrolled Rod Cluster Control Assembly Withdrawal 14.1.1 Not Applicable from a Subcritical or Low Power Startup Condition Uncontrolled Rod Cluster Control Assembly Bank 14.1.2 Not Applicable Withdrawal at Power 14.1.3 Incorrect Positioning of Part-Length Rods Not Applicable 14.1.4 Rod Cluster Control Assembly Drop Not Applicable 14.1.5 Chemical and Volume Control System Malfunction Not Applicable 14.1.6 Loss of Reactor Coolant Flow Not Applicable 14.1.7 Startup of an Inactive Reactor Coolant Loop Not Applicable 14.1.8 Loss of External Electrical Load Not Applicable 14.1.9 Loss of Normal Feedwater Not Applicable Excessive Heat Removal Due to Feedwater System 14.1.10 Not Applicable Malfunctions 14.1.11 Excessive Load Increase Incident Not Applicable 14.1.12 Loss of all AC Power to the Station Auxiliaries Not Applicable Likelihood and Consequences of Turbine-Generator Unit 14.1.13 Not Applicable Overspeed 14.2.1 Fuel-Handling Accidents Applicable 14.2.1.1 Fuel-Handling Accident in Fuel-Handling Building Applicable 14.2.1.2 Refueling Accident Inside Containment Not Applicable Applicable - Deemed to 14.2.1.3 Fuel Cask Drop Accident not be Credible - See discussion in (1) below Applicable - Dose dependent on 14.2.2 Accidental Release-Recycle of Waste Liquid volatilized components and is addressed in Section 14.2.3 14.2.3 Accidental Release - Waste Gas Applicable 14.2.4 Steam-Generator Tube Rupture Not Applicable 14.2.5 Rupture of a Steam Pipe Not Applicable Rupture of a Control Rod Mechanism Housing - Rod 14.2.6 Not Applicable Cluster Control Assembly Ejection 14.3 Loss-of-Coolant Accidents Not Applicable 14.4 Anticipated Transients Without Scram Not Applicable (1) Section 14.2.1.3 of the IP2 UFSAR states:

As discussed in Section 9.5.6.4, Control of Heavy Loads Program, and Section 9.5.7.1, FSB 110-Ton Ederer Single Failure Proof Gantry Crane, the IP2 fuel storage building spent fuel cask handling operations are now conducted using a single-failure-proof 110-Ton Ederer Gantry Crane that conforms to the requirements in NUREG-0554 (Single-Failure-Proof Cranes for Nuclear Power Plants, May 1979). The Ederer Gantry Crane performs spent fuel cask handling activities without the necessity of having to postulate the drop of a spent fuel cask.

With the Ederer cranes 110-ton main hoist qualified as single-failure-proof, the crane is used as part of a single-failure-proof handling system for critical lifts as discussed in Revision 1 of

Enclosure NL-19-026 Page 4 of 74 Section 9.1.5, Overhead Heavy Load Handling Systems, Sub-section lll.4.C of NUREG-0800, and a cask drop accident is not a credible event and need not be postulated Performing an evaluation using the analysis assumptions for the fuel-handling accident shows that even with damage to a full core of recently discharged fuel assemblies by a fuel cask dropped into the spent fuel pool, the calculated fuel-handling accident doses would not be exceeded if 90 days had elapsed after shutdown. Since the fuel cask is handled by the single failure proof 110-ton gantry crane approved for use by the NRC in Technical Specification Amendment #244, this accident is not probable.

The analyzed accidents that remain applicable to IP2 in the permanently shut down and defueled condition are the Fuel Handling Accident (FHA) in the Fuel Handling Building (i.e., Fuel Storage Building (FSB)), accidental release-recycle of waste liquid, and the accidental release of waste gas.

The additional discussion of the analyses of these events provided below is based on information from Calculation IP-CALC-19-00003, Post-Permanent Shutdown Analyses of Fuel Handling, Waste Handling, and High Integrity Container Drop Accidents for Indian Point Units 2 and 3. This calculation:

Presents the results of an analysis of the FHA utilizing the Alternate Source Term (AST) methodology described in Regulatory Guide 1.183 that is provided in Calculation IP-CALC-11-00073, AST Analysis of IP2 Fuel Handling Accident in the Fuel Storage Building without FSB Exhaust Fan Operation. This analysis concludes that the dose consequences of the FHA for the Normal case will remain within the licensing basis dose limits without crediting FSB ventilation, the station vent radiation monitors, Control Room isolation, and Control Room filtration assuming 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months of decay time following shut down.

Includes the determination of the dose consequences for a waste gas decay tank rupture accident using a 50,000 Curie (Ci) dose-equivalent Xe-133 waste gas tank activity limit without any credit for mitigating systems.

While the calculation contains an analysis of a High Integrity Container Drop event, this analysis is not credited as part of this PDTS License Amendment Request. It was performed for future purposes that are outside the changes requested by the PDTS License Amendment Request.

Fuel Handling Accident Analysis for the Permanently Shut Down and Defueled Condition An FHA may occur in the FSB during movement of a fuel assembly. The fuel assembly is moved under water and the accident is assumed to occur when one fuel assembly is damaged.

An analysis of the FHA utilizing the AST methodology described in Regulatory Guide 1.183 was previously approved by the NRC in License Amendment No. 211 (Reference 5) on July 27, 2000. It consisted of changes to the TSs which resulted from implementation of an alternate radiological source term as permitted by 10 CFR 50.67 and allowed implementation of plant modifications to the containment air handling systems and the control room air handling systems related to the use of the AST. Later, as part of the IP2 power uprate project, a re-analysis of the FHA was performed utilizing the AST methodology, that is currently the analysis of record as presented in Section 14.2.1.1 of the IP2 UFSAR.

Enclosure NL-19-026 Page 5 of 74 Concurrent with implementation of the PDTS, this UFSAR section will be revised in accordance with 10 CFR 50.59 to reflect the results of the Normal case analyzed in Calculation IP-CALC-11-00073, as summarized in Calculation IP-CALC-19-00003. This FHA analysis utilizes the AST methodology and concludes that the dose consequences of the FHA will remain within the licensing basis dose limits without crediting FSB ventilation, the station vent radiation monitors, Control Room isolation, and Control Room filtration assuming 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months of decay time following shut down. The FHA dose consequences in the IP2 Control Room, Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) were computed using the following:

a) The methodology and assumptions in Regulatory Guide 1.183 b) Appropriate source terms, release pathways, and other assumptions, as described below, c) Post-accident atmospheric dispersion factors, and d) The NRC sponsored code RADTRAD, Rev. 3.03 was used to model the design basis FHA and estimate the dose consequences. The Control Room, EAB and LPZ doses in terms of Total Effective Dose Equivalent (TEDE) were calculated for the FHA.

Fission Product Inventory The fission product inventory in the core is based on full power operation (3216 Megawatt-thermal (MWt) + 2% uncertainty, i.e., 3280.3 MWt). The core inventory of radionuclides of interest at 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months decay are shown in Table 2-2.

Table 2 Core Inventories of Nuclides for Use in Radiological Design-Basis Applications (at 84 Hours of Decay)

Nuclide Activity Nuclide Activity Halogens (Ci) Noble Gases (Ci)

I-130 3.44E+04 Kr-85m 0.00E+00 I-131 6.94E+07 Kr-85 1.10E+06 I-132 6.39E+07 Kr-87 0.00E+00 I-133 1.17E+07 Kr-88 0.00E+00 I-134 0.00E+00 I-135 2.62E+04 Xe-131m 9.85E+05 Xe-133m 2.91E+06 Xe-133 1.36E+08 Xe-135m 4.20E+03 Xe-135 7.83E+05 Xe-138 0.00E+00 Release Fractions and Composition The fission product gap release fractions, for each radionuclide group for the FHA are shown below:

o I-131 0.12 o Kr-85 0.30 o Other iodines and noble gases 0.10

Enclosure NL-19-026 Page 6 of 74 The iodine released from the assembly gap is assumed to be 99.85% elemental and 0.15%

organic.

The overall SFP decontamination factor for iodines is 200.

A value of 285 for the SFP elemental iodine decontamination factor was calculated.

Control Room Dose Consequences The atmospheric dispersion factors associated with the transport of released radioactivity to the Control Room intake are as follows:

0-2 hours - Control Room atmospheric dispersion factor (/Q) - 8.31E-04 seconds/m3 Since releases are assumed to be completed in the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (Regulatory Guide 1.183), no additional time periods are presented.

The Control Room characteristics are as follows:

Free air volume: 102,400 ft3 Filtered makeup: 0 ft3/min Filtered recirculation: 0 ft3/min Unfiltered makeup: 920 ft3/min Unfiltered inleakage: 700 ft3/min The breathing rate was assumed to be 3.5E-04 m3/second for the duration of the accident.

Offsite Dose Consequences Table 2 Offsite /Qs Interval 0-2 hours Interval 0-8 hours EAB /Q 7.5-E-04 LPZ /Q 3.5E-04 (seconds/m3) (seconds/m3)

Table 2 Breathing Rates at EAB and LPZ Receptor Location Time Interval (hours) Breathing Rate (m3/sec)

EAB 0-2 3.5E-04 LPZ 0-8 3.5E-04 LPZ 8 - 24 1.75E-04 LPZ 24 - 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months 2.32E-04

Enclosure NL-19-026 Page 7 of 74 Radiological Consequences The radiological consequences of the postulated FHA are as follows:

Table 2 AST FHA Results (at 84 Hours of Decay)

TEDE Dose Regulatory Location (rem) Limit (rem)

Control Room 3.88 5 EAB 4.2 6.3 LPZ 2.0 6.3 The calculated TEDE values to the Control Room, EAB, and LPZ are less than the limits set forth in 10 CFR 50.67 and Regulatory Guide 1.183.

In addition, after a decay time of at least 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months (30 days) prior to fuel movement, the analysis of the FHA results in an EAB TEDE dose of 0.47 rem, which is less than the Environmental Protection Agency (EPA) Protective Action Guideline recommended threshold for evacuation of 1 rem.

Accidental Release - Waste Gas Section 14.2.3 of the IP2 UFSAR evaluates the accidental release of waste gas. Concurrent with implementation of the PDTS, this UFSAR section will be revised in accordance with 10 CFR 50.59 to reflect the results of Calculation IP-CALC-19-00003, Post-Permanent Shutdown Analyses of Fuel Handling, Waste Handling, and High Integrity Container Drop Accidents for Indian Point Units 2 and 3. This calculation includes the determination of the dose consequences for a waste gas decay tank rupture accident using a 50,000 Ci dose-equivalent Xe-133 waste gas tank activity limit without any credit for mitigating systems.

The waste gas decay tanks receive the radioactive gases from the radioactive liquids from the various laboratories and drains processed by the waste disposal system. The 50,000 Ci dose-equivalent Xe-133 waste gas tank activity assumed in this calculation bounds the current Xe-133 dose-equivalent limit of 29,761 Ci, as well as the administrative Xe-133 dose-equivalent limit of 6,000 Ci.

Other tanks that contain waste gas during operations (the volume control tank and liquid holdup tank) were not considered in this analysis, since gaseous products from these liquid tanks are collected and compressed in the waste gas decay tanks for decay prior to release. Potential liquid waste releases are considered from these tanks; however, any liquid releases are retained in the building or sumps and only volatilized components would be released to the environment. These volatilized components are evaluated as part of the waste gas decay tank accident.

Calculation The RADTRAD model for the waste gas decay tank accident was developed using the following inputs to model an instantaneous ground level release of 50,000 Ci of Xe-133.

3 RADTRAD compartments - Waste Gas Decay Tank (modeled as a volume of 1 ft3),

Environment, and Control Room (volume = 4.72E4 ft3)

Enclosure NL-19-026 Page 8 of 74 Plant power level - 1 MWt - Nominal power level used to release activity of 50,000 Ci of dose equivalent Xe-133 from the associated RADTRAD nuclide inventory file Activity release fractions - noble gas fraction = 1.0 over a duration of 0.00001 hour - This represents an instantaneous release in the RADTRAD release file Assumed high flow rate from Waste Gas Decay Tank volume to the Environment of 1E4 ft3/min to model an instantaneous release from the tank compartment to the environment Control Room Flow Rate o Unfiltered Intake = 1,500 ft3/min o Unfiltered Inleakage = 700 ft3/min o Outflow = 2,200 ft3/min o Recirculation = 0 ft3/min Offsite /Qs: EAB = 1.03E-03 seconds/m3, LPZ = 3.8E-04 seconds/m3 - Note: The IP2

/Qs for the EAB and LPZ are bounded by the Indian Point Nuclear Generating Station Unit No. 3 (IP3) /Qs A bounding /Q was developed for a release from the waste gas decay tanks to the IP2 Control Room. The release point is assumed to be the centerline of the closest large waste gas decay tank. The following inputs were used to develop the Control Room /Q.

Table 2 Control Room Atmospheric Dispersion Parameters (Waste Gas Analysis)

Parameter Value IP3 Control Room intake location (ft) 5783.75 North 1476.0 East Center line location IP3 gas decay 5841.25 North tank #31 (ft) 1552.5 East Lateral dispersion coefficient for a 30 meter 1.5 release, stability class F Vertical dispersion coefficient for a 30 meter 0.85 release, stability class F The wind speed was assumed to be 1 meter/second and the stability class was assumed to be F for calculating a bounding atmospheric dispersion coefficient from the Primary Auxiliary Building (PAB) to the Control Room. These generic meteorology conditions were used to calculate a bounding atmospheric dispersion coefficient. An additional significant conservatism is the implicit assumption that the wind direction is directly toward the Control Room intake at all times.

The IP3 Control Room /Q also does not credit holdup of the activity in the PAB which would delay or disperse activity within the building.

The Control Room intake /Q for IP3 was determined by using these parameters and was established as bounding for IP2. This is conservative as the distance between the IP2 Control Room intake at 6476N - 1373E and the southwest corner of the IP2 PAB at approximately 6550N

- 1476E equates to a tangential distance of approximately 39 meters. With a longer distance and the same direction to the Control Room intake for IP2, the dispersion to the IP2 Control Room would be greater. Therefore, the IP3 Control Room /Q is bounding for the IP2 Control Room.

Enclosure NL-19-026 Page 9 of 74 Radiological Consequences The calculated radiological consequences, following a waste gas decay tank rupture without credit for any mitigating systems or the PAB ventilation system post shutdown, are provided in Table 2-7.

Table 2 Waste Gas Decay Tank Rupture Results with 50,000 Ci Dose Equivalent Xe-133 Limit per Tank Location Whole Body Dose Limit (rem)

(rem)

Control Room 0.77 5.0 EAB 0.30 0.5 LPZ 0.11 0.5 The radiological consequences following a waste gas decay tank rupture are less than the dose consequences following an FHA presented in Table 2-5. They are also less than the 10 CFR 50.67 limit of 5 rem TEDE to the Control Room operators and the 500 mrem EAB and LPZ dose limit following a waste gas tank accident.

Accidental Release - Recycle Waste Liquid Section 14.2.2 of the IP2 UFSAR addresses the accidental release of waste liquid. The discussion concludes that a potential liquid waste release collects in building sumps or is retained in building vaults. It is not released to the environment. As such, the hazard from these releases is derived only from any volatilized components. The volatilized components are what comprise the waste gas accident and were evaluated above. Therefore, a separate liquid-specific release accident evaluation is not required to be performed with regard to removal of supporting systems such as PAB ventilation, station vent radiation monitors, Control Room isolation, and Control Room filtration.

3. REGULATORY EVALUATION IP2 proposes to modify the license conditions and the TSs from Appendices A through C as listed in the following tables. In addition, IP2 is providing a description and basis for each of the proposed changes. to this enclosure contains a mark-up of the current OL, Appendices A, B, and C TSs and Appendix A TSs Bases pages. The proposed changes to the IP2 Appendix A TSs are considered a major rewrite. Thus, the IP2 Appendix A TSs and TSs Bases that are deleted in their entirety are identified as such, but the associated deleted pages are not included in Attachment 1 to this enclosure. In addition, the following administrative changes are not shown in the marked-up (Enclosure, Attachment 1) OL, Appendix A TSs, and Appendix A TSs Bases pages, because they do not affect the technical content of the IP2 OL or Appendix A TSs:

Reformatting (margins, font, tabs, line spacing, etc.) content to create a continuous electronic file; and Renumbering of pages, where appropriate, to condense and reduce the number of pages. of this enclosure provides the re-typed IP2 Facility License, PDTS, PDTS Bases in their entirety, and the affected pages of the Appendices B and C TSs. Since the changes to the

Enclosure NL-19-026 Page 10 of 74 Appendix A TSs and TSs Bases are considered a major rewrite, revision bars are not used. It incorporates the changes to the IP2 Appendix A TSs proposed in References 2 and 3.

Proposed Changes to the IP2 Facility Operating License License Title Current Title Proposed Title Renewed Facility Operating License Renewed Facility Operating License Basis The License Title is modified to eliminate the reference to Operating. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition 1.B Current License Condition 1.B Proposed License Condition 1.B Construction of the Indian Point Nuclear Deleted per Amendment [###]

Generating Unit No. 2 (IP2 or facility) has been substantially completed in conformity with provisional Construction Permit No. CPPR-21, as amended, and the application, as amended, the provisions of the Act and regulations of the Commission; Basis This license condition is proposed for deletion in its entirety. Decommissioning of IP2 is not dependent on the regulations that governed construction of the facility.

License Condition 1.C Current License Condition 1.C Proposed License Condition 1.C The facility will operate in conformity with the The facility will operatebe maintained in application, as amended, the provisions of the conformity with the application, as amended, Act, and the rules and regulations of the the provisions of the Act, and the rules and Commission; regulations of the Commission; Basis This license condition is revised to reflect a more accurate description of the future requirements.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Thus, replacing the verb operate with the verb be maintained will provide accuracy regarding the possession-only 10 CFR Part 50 License.

Enclosure NL-19-026 Page 11 of 74 License Condition 1.D Current License Condition 1.D Current License Condition 1.D There is reasonable assurance (i) that the There is reasonable assurance (i) that the activities authorized by the renewed operating activities authorized by the renewed operating license can be conducted without endangering license can be conducted without endangering the health and safety of the public, and (ii) that the health and safety of the public, and (ii) that such activities will be conducted in compliance such activities will be conducted in compliance with the rules and regulations of the with the rules and regulations of the Commission; Commission; Basis This license condition is revised to reflect that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition 1.H Current License Condition 1.H Proposed License Condition 1.H After weighing the environmental, economic, After weighing the environmental, economic, technical, and other benefits of the facility technical, and other benefits of the facility against environmental costs and considering against environmental costs and considering available alternatives, the issuance of this available alternatives, the issuance of this renewed Facility Operating License No. DPR- renewed Facility Operating License No. DPR-26, subject to the conditions for the protection 26, subject to the conditions for the protection of the environment set forth herein, is in of the environment set forth herein, is in accordance with 10 CFR Part 51, Appendix B, accordance with 10 CFR Part 51, Appendix B, of the Commission's regulations and all of the Commission's regulations and all applicable requirements of said Appendix B applicable requirements of said Appendix B have been satisfied; have been satisfied; Basis This license condition is revised to reflect that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition 2 Current License Condition 2 Proposed License Condition 2 Renewed Facility Operating License No. DPR- Renewed Facility Operating License No. DPR-26, is hereby issued to ENIP2 and ENO to read 26, is hereby issued to ENIP2 and ENO to read as follows: as follows:

Basis This license condition is revised to reflect that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Enclosure NL-19-026 Page 12 of 74 License Condition 2.A Current License Condition 2.A Proposed License Condition 2.A This renewed license applies to the Indian This renewed license applies to the Indian Point Nuclear Generating Unit No. 2, a Point Nuclear Generating Unit No. 2, a pressurized water nuclear reactor and pressurized water nuclear reactor and associated equipment (the facility), which is associated equipment (the facility), which is owned by ENIP2 and operated by ENO. The owned by ENIP2 and operatedmaintained by facility is located in Westchester County, New ENO. The facility is located in Westchester York, on the east bank of the Hudson River in County, New York, on the east bank of the the Village of Buchanan, and is described in the Hudson River in the Village of Buchanan, and Final Facility Description and Safety Analysis is described in the Final Facility Description Report, as supplemented and amended, and andDefueled Safety Analysis Report, as the Environmental Report, as amended. supplemented and amended, and the Environmental Report, as amended.

Basis This license condition is revised to reflect that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

In addition, it is modified to reflect that a Defueled Safety Analysis Report will be prepared to address the permanently shut down and defueled condition.

License Condition 2.B.(1)

Current License Condition 2.B.(1) Proposed License Condition 2.B.(1)

Pursuant to Section 104b of the Act and 10 Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and CFR Part 50, "Licensing of Production and Utilization Facilities, (a) ENIP2 to possess and Utilization Facilities,", (a) ENIP2 to possess and use, and (b) ENO to possess, use, and use, and (b) ENO to possess, and use, and operate, the facility at the designated location in operate the facility at the designated location in Westchester County, New York, in accordance Westchester County, New York, in accordance with the procedures and limitations set forth in with the procedures and limitations set forth in this renewed license; this renewed license; Basis This license condition is revised to reflect that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition 2.B.(2)

Current License Condition 2.B.(2) Proposed License Condition 2.B.(2)

ENO pursuant to the Act and 10 CFR 70, to ENO pursuant to the Act and 10 CFR 70, to receive, possess, and use, at any time special receive, possess, and use, at any time special nuclear material as reactor fuel, in accordance nuclear material that was used as reactor fuel, with the limitations for storage and amounts in accordance with the limitations for storage required for reactor operation, as described in and amounts required for reactor operation, as the Final Facility Description and Safety described in the Final Facility Description Analysis Report, as supplemented and andDefueled Safety Analysis Report, as amended and as described in the supplemented and amended and as described

Enclosure NL-19-026 Page 13 of 74 Commission's authorization through in the Commission's authorization through Amendment No. 75 to this license. Amendment No. 75 to this license.

Basis This license condition is revised to remove the authorization for receipt and use of special nuclear material (SNM) as reactor fuel, eliminate the reference to use of the SNM for reactor operations, and limit the possession of SNM to SNM that was used as reactor fuel at IP2. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As such, IP2 has no need to receive SNM in the form of reactor fuel and cannot use SNM as reactor fuel for reactor operations. The continued authorization to possess SNM "that was used" as reactor fuel is necessary as IP2 currently possesses the reactor fuel that was used for the past operations of the reactor. In addition, it is modified to reflect that a Defueled Safety Analysis Report will be prepared to address the permanently shut down and defueled condition, and eliminate the unnecessary reference to the license amendment number.

License Condition 2.B.(3)

Current License Condition 2.B(3) Proposed License Condition 2.B.(3)

ENO pursuant to the Act and 10 CFR Parts 30, ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use at any 40 and 70, to receive, possess and use at any time any byproduct, source and special nuclear time any byproduct, source and special nuclear material as sealed neutron sources for reactor material as sealed neutron sources that were startup, sealed sources for reactor used for reactor startup, sealed sources that instrumentation and radiation monitoring were used for calibration of reactor equipment calibration, and as fission detectors instrumentation and are used in the in amounts as required; calibration of radiation monitoring equipment calibration, and as fission detectors in amounts as required; Basis This license condition is revised to remove the authorization for receipt and use of byproduct, source, and SNM as sealed neutron sources for reactor startup. The deletion of the authorization to receive and use sources for reactor startup is consistent with the fact that IP2 will no longer be authorized to operate.

The authorization to possess such sources previously used for reactor startup is retained. The continued authorization to possess neutron sources that were used for reactor startup is consistent with the safe storage of byproduct, source, and SNM. The use of sources for radiation monitoring will continue to be required.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). These changes are consistent with the permanently defueled condition.

License Condition 2.B.(5)

Current License Condition 2.B(5) Proposed License Condition 2.B.(5)

ENO pursuant to the Act and 10 CFR Parts 30 ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such and 70, to possess, but not separate, such

Enclosure NL-19-026 Page 14 of 74 byproduct and special nuclear materials as may byproduct and special nuclear materials as may be produced by the operation of the facility. bethat were produced by the operation of the facility.

Basis This license condition is revised to reflect that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition 2.C.(1), Maximum Power Level Current License Condition 2.C.(1) Proposed License Condition 2.C.(1)

ENO is authorized to operate the facility at Deleted per Amendment [###]

steady state reactor core power levels not in excess of 3216 megawatts thermal Basis This license condition is deleted in its entirety to reflect the permanently defueled condition of the facility. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition 2.C.(2), Technical Specifications Current License Condition 2.C.(2) Proposed License Condition 2.C.(2)

The Technical Specifications contained in The Technical Specifications contained in Appendices A, B, and C, as revised through Appendices A, B, and C, as revised through Amendment No. 288, are hereby incorporated Amendment No. 288###, are hereby in the renewed license. ENO shall operate the incorporated in the renewed license. ENO shall facility in accordance with the Technical operate maintain the facility in accordance with Specifications. the Technical Specifications.

Basis This license condition is revised to replace the verb shall operate with the verb shall maintain to better describe the permanently defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

License Condition 2.C.(3)

Current License Condition 2.C.(3) Proposed License Condition 2.C.(3)

The following conditions relate to the Deleted per Amendment [###]

amendment approving the conversion to Improved Standard Technical Specifications:

1. The amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee controlled documentsThe relocation of requirements and detailed information shall

Enclosure NL-19-026 Page 15 of 74 be completed on or before the implementation of this amendment.

2. The following is a schedule for implementing surveillance requirements (SRs)

Basis This license condition is deleted in its entirety. This is a historical license condition, because the activities were completed in accordance with the license condition.

License Condition 2.G Current License Condition 2.G Proposed License Condition 2.G Pursuant to Section 50.60 of 10 CFR Part 50, Deleted per Amendment [###]

paragraph 4 of Provisional Construction Permit No. CPPR-21 allocating quantities of special nuclear material, together with the related estimated schedules contained in Appendix A attached to said provisional construction permit, shall remain effect.

Basis This license condition is deleted in its entirety. It pertains to 10 CFR 50.60 regarding the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Thus, a license condition associated with the protection of the reactor coolant pressure boundary is no longer required.

License Condition 2.K Current License Condition 2.K Proposed License Condition 2.K ENO shall implement and maintain effect all Deleted per Amendment [###]

provisions of the NRC-approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in Safety Evaluations Reports dated November 30, 1977, February 3, 1978, January 31, 1979, October 31, 1980, August 22, 1983, March 30, 1984, October 16, 1984, September 16, 1985, November 13, 1985, March 4, 1987, January 12, 1989, and March 26, 1996. ENO may make changes to the NRC-approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

Enclosure NL-19-026 Page 16 of 74 Basis This license condition is deleted to reflect the permanently defueled condition of the facility. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the fire protection program will be revised to take into account the decommissioning facility conditions and activities. IP2 will continue to utilize the defense-in-depth concept, placing special emphasis on detection and suppression in order to minimize radiological releases to the environment.

This condition, which is based on maintaining an operational fire protection program in accordance with 10 CFR 50.48, with the ability to achieve and maintain safe shut down of the reactor in the event of a fire, will no longer be applicable at IP2. However, many of the elements that are applicable for the operating plant fire protection program continue to be applicable during facility decommissioning. During the decommissioning process, a fire protection program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. However, the regulation is applicable regardless of whether a requirement for a fire protection program is included in the facility license. Therefore, a license condition requiring such a program for a permanently shut down and defueled facility is not needed.

License Condition 2.O, CONTROL ROOM ENVELOPE HABITABILITY Current License Condition 2.O Proposed License Condition 2.O Upon implementation of Amendment No. 258 adopting Deleted per Amendment [###]

TSTF-448, Revision 3 (as supplemented), the determination of control room envelope (CRE) unfiltered air inleakage as required by Technical Specification (TS)

Surveillance Requirement (SR) 3.7.10.4, in accordance with TS 5.5.16.c.(i), the assessment of CRE habitability as required by TS 5.5.16.c.(ii), and the measurement of CRE pressure as required by TS 5.5.16.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.4, in accordance with TS 5.5.16.c.(i), shall be within the next 18 months since the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, TS 5.5.16.c.(ii), shall be within the next 9 months since the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from January 4, 2007, the date of the most recent successful pressure measurement test.

Enclosure NL-19-026 Page 17 of 74 Basis This license condition is deleted in its entirety. The license condition defined requirements of TSTF-448 to assess the Control Room Envelope (CRE) Habitability at the specified frequencies for the first performance of the specific test, assessment, and measurement. This is a historical license condition, because the test, assessment, and measurement were completed in accordance with the schedule specified in the license condition.

License Condition 6 Current License Condition 6 Proposed License Condition 6 This renewed license is effective as of the date This renewed license is effective as of the date of its issuance, and shall expire at midnight of its issuance, and shall expire at midnight April April 30, 2024. 30, 2024.until the Commission notifies the licensee in writing that the license is terminated.

Basis After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Thus, this license condition is revised to conform with 10 CFR 50.51, Continuation of license, in that the license authorizes ownership and possession by Entergy until the Commission notifies the licensee in writing that the license is terminated.

ATTACHMENTS AND DATE OF ISSUANCE Current Attachments and Date of Issuance Proposed Attachments and Date of Issuance Appendix A - Technical Specifications Appendix A - Permanently Defueled Technical Specifications Date of Issuance: September 17, 2018 Date of Issuance: September 17, 2018To Be Determined Basis The title of Appendix A is updated to reflect that the Technical Specifications will be retitled as the Permanently Defueled Technical Specifications. The date of issuance is modified to reflect the date that the NRC issues the PDTS which is yet to be determined. These are administrative changes.

APPENDIX A TO FACILITY OPERATING LICENSE DPR-26 Current Title Proposed Title FACILITY OPERATING LICENSE DPR-26 FACILITY OPERATING LICENSE DPR-26 TECHNICAL SPECIFICATIONS AND BASES PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS AND BASES Amendment No. 220 Amendment No. 220###

Enclosure NL-19-026 Page 18 of 74 Basis The License Title is modified to rename the Facility Operating License DPR-26 and the Technical Specifications and Bases as Facility License DPR-26 and Permanently Defueled Technical Specifications and Bases. These changes reflect the upcoming change in status regarding IP2. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

In addition, the amendment number is modified to reflect the amendment number associated with the issuance of the PDTS.

APPENDIX A, TECHNICAL SPECIFICATIONS, TABLE OF CONTENTS Current IP2 TS Basis for Change Table of Contents The Table of Contents is modified to reflect the changes made below.

TS SECTION 1.1, DEFINITIONS TS 1.1, "Definitions," provides defined terms that are applicable throughout the TSs and TSs Bases. A number of the Definitions are proposed to be deleted, because they have no relevance to and no longer apply to the permanently defueled facility status.

Definition Basis for Change ACTUATION LOGIC TEST This definition is proposed for deletion, because the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

AXIAL FLUX DIFFERENCE (AFD) This definition is proposed for deletion, because the term is not used in any PDTS specification. This term is no longer applicable since fuel will be permanently removed from the reactor core.

CHANNEL CALIBRATION This definition is proposed for deletion, because the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

CHANNEL CHECK This definition is proposed for deletion, because the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

Enclosure NL-19-026 Page 19 of 74 CHANNEL OPERATIONAL TEST (COT) This definition is proposed for deletion, because the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

CORE ALTERATION This definition is proposed for deletion, because the term is not used in any PDTS specification. This term is no longer applicable since fuel will be permanently removed from the reactor core.

CORE OPERATING LIMITS REPORT (COLR) This definition is proposed for deletion, because the term is not used in any PDTS specification. TS 5.6.5 that requires the COLR is also proposed for elimination.

DOSE EQUIVALENT I-131 This definition is proposed for deletion, because the term is not used in any PDTS specification. This term is used in current TS 3.4.16 and TS 3.7.14 to express the specific activity limit from a mixture of iodine isotopes contained in reactor coolant and secondary coolant. TS 3.4.16 and TS 3.7.14 are proposed for deletion in the PDTS. The specific activity limit is used as the basis in accident analysis involving coolant releases. Since accident conditions associated with the RCS and secondary coolant system will no longer apply to the permanently shut down and defueled facility, the definition is no longer meaningful.

DOSE EQUIVALENT XE-133 This definition is proposed for deletion, because the term is not used in any PDTS specification. This term is used in current TS 3.4.16 to express the specific activity limit from a mixture of xenon isotopes contained in reactor coolant. TS 3.4.16 is proposed for deletion in the PDTS. The specific activity limit is used as the basis in accident analysis involving coolant releases. Since accident conditions associated with the RCS and secondary coolant system will no longer apply to the permanently shut down and defueled facility, the definition is no longer meaningful.

LEAKAGE This definition is proposed for deletion, because the term is not used in any PDTS specification. Refer to the discussions for the proposed deletion of TS 3.4.13 and TS 5.5.7.

Enclosure NL-19-026 Page 20 of 74 MASTER RELAY TEST This definition is proposed for deletion, because the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

MODE This definition, including Table 1.1-1, is proposed for deletion, because operational MODES are not used in any PDTS specification. MODES as defined in Table 1.1-1 are defined for operating or refueling conditions. This term does not apply to a facility in the permanently defueled condition.

OPERABLE - OPERABILITY This definition is proposed for deletion, because the term is not used in any PDTS specification. There are no systems or components required to be operable in the PDTS, because there are no active systems, structures or components required to function to mitigate any of the remaining DBAs.

PHYSICS TESTS This definition is proposed for deletion, because the term is not used in any PDTS specification. This term does not apply to a facility in the permanently defueled condition.

QUADRANT POWER TILT RATIO (QPTR) This definition is proposed for deletion, because the term is not used in any PDTS specification. This definition only applies to an operating reactor core.

RATED THERMAL POWER (RTP) This definition is proposed for deletion, because the term is not used in any PDTS specification. This term is meaningful only to a reactor authorized to contain fuel and operate at power. It does not apply to a facility in the permanently defueled condition.

SHUTDOWN MARGIN (SDM) This definition is proposed for deletion, because the term is not used in any PDTS specification. This term is meaningful only to a reactor authorized to contain fuel and operate at power. It does not apply to a facility in the permanently defueled condition.

SLAVE RELAY TEST This definition is proposed for deletion, because the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

Enclosure NL-19-026 Page 21 of 74 STAGGERED TEST BASIS This definition is proposed for deletion, because the term is not used in any PDTS specification. This definition applies to the performance of surveillance tests on systems with multiple subsystems or channels. There are no surveillance requirements in the PDTS for operating systems.

THERMAL POWER This definition is proposed for deletion, because the term is not used in any PDTS specification. This term is meaningful only to a reactor authorized to contain fuel and operate at power. It does not apply to a facility in the permanently defueled condition.

TRIP ACTUATING DEVICE OPERATIONAL This definition is proposed for deletion, TEST (TADOT) because the term is not used in any PDTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

TS SECTION 1.2, LOGICAL CONNECTORS TS 1.2, "Logical Connectors," explain the meaning of logical connectors. It is modified to reflect the logical connectors that continue to exist in the TSs.

Current Purpose Proposed Purpose Logical connectors are used in Technical Logical connectors are used in Technical Specifications (TS) to discriminate between, Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Actions, Completion Times, and Surveillances, Frequencies and Frequencies Current Background Proposed Background When logical connectors are used to state a When logical connectors are used to state a Condition, Completion Time, Surveillance, or Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, Frequency, only the first level of logic is used, and the logical connector is left justified with the and the logical connector is left justified with the statement of the Condition, Completion Time, statement of the Condition, Completion Time, Surveillance, or Frequency. Surveillance, or Frequency.

Current Examples Proposed Examples The following examples illustrate the use of The following examples illustrates the use of logical connectors logical connectors Example 1.2-2 Example 1.2-2 is proposed for deletion.

Basis This section is modified to reflect the logical connectors utilized in TS 3.7.12 and TS 3.7.13.

These are the only TSs that utilize logical connectors in the PDTS. These changes are administrative changes.

Enclosure NL-19-026 Page 22 of 74 TS SECTION 1.3, COMPLETION TIMES TS 1.3, Completion Times, establishes the Completion Time convention and provides guidance for its use. It is modified to reflect the permanently shut down and defueled condition and the Completion Times that continue to exist in the PDTS.

Current Background Proposed Background Limiting Conditions for Operation (LCOs) Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring specify minimum requirements for ensuring safe operation of the unit safe operation of the unithandling and storage of spent nuclear fuel Basis The Background section of TS 1.3 is modified to reflect the upcoming change in status regarding IP2. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the primary mission will change from the safe operation of the unit to the safe handling and storage of spent nuclear fuel.

Current Description Proposed Description The Completion Time is the amount of time The Completion Time is the amount of time allowed for completing a Required Action. It is allowed for completing a Required Action. It is referenced to the time of discovery of a referenced to the time of discovery of a situation (e.g., inoperable equipment or variable situation (e.g., inoperable equipment or variable not within limits) that requires entering an not within limits) that requires entering an ACTIONS Condition unless otherwise ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified, providing the unitfacility is in a specified condition stated in the Applicability of MODE or specified condition stated in the the LCO. Required Actions must be completed Applicability of the LCO. Required Actions must prior to the expiration of the specified be completed prior to the expiration of the Completion Time. An ACTIONS Condition specified Completion Time. An ACTIONS remains in effect and the Required Actions Condition remains in effect and the Required apply until the Condition no longer exists or the Actions apply until the Condition no longer unit is not within the LCO Applicability. exists or the unitfacility is not within the LCO Applicability.

If situations are discovered If situations are discovered Basis The Description section of TS 1.3 is modified to reflect the upcoming change in status regarding IP2. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the PDTS will contain no operability requirements for any equipment. In addition, the term facility better represents IP2 in the permanently shut down and defueled condition.

Current Examples Proposed Example The following examples illustrate the use of The following examples illustrates the use of Completion Times with different types of Completion Times with different types of Conditions and changing Conditions. Conditions and changing ConditionsRequired Actions.

Enclosure NL-19-026 Page 23 of 74 Example 1.3-1 Example 1.3-1 is modified to address Example 1.3-2 Completion Times as utilized by TS 3.7.12.

Example 1.3-3 Example 1.3-2 is proposed for deletion.

Example 1.3-4 Example 1.3-3 is proposed for deletion.

Example 1.3-5 Example 1.3-4 is proposed for deletion.

Example 1.3-6 Example 1.3-5 is proposed for deletion.

Example 1.3-7 Example 1.3-6 is proposed for deletion.

Example 1.3-7 is proposed for deletion.

Basis This section is modified to reflect the use of Completion Times that are utilized in TS 3.7.11, TS 3.7.12, and TS 3.7.13. These are the only TSs that have Completion Times in the PDTS. The changes to the Examples section of TS 1.3 are administrative changes.

TS SECTION 1.4, FREQUENCY TS 1.4, Frequency, defines the proper use and application of Frequency requirements. It is modified to reflect the permanently shut down and defueled condition and the Frequencies that continue to exist in the PDTS.

Current Description Proposed Description The "specified Frequency" is referred to The "specified Frequency" is referred to throughout this section and each of the throughout this section and each of the Specifications of Section 3.0.2, Surveillance Specifications of Section 3.0.2, Surveillance Requirement (SR) Applicability. The "specified Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the Frequency" consists of the requirements of the Frequency column of each SR as well as Frequency column of each SR as well as certain Notes in the Surveillance column that certain Notes in the Surveillance column that modify performance requirements. modify performance requirements.

Sometimes special situations dictate when the Sometimes special situations dictate when the requirements of a Surveillance are to be met. requirements of a Surveillance are to be met.

They are "otherwise stated" conditions allowed They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Notes in the Surveillance, as part of the Surveillances, or both. Surveillances, or both.

Situations where a Surveillance could be Situations where a Surveillance could be required (i.e., its Frequency could expire), but required (i.e., its Frequency could expire), but where it is not possible or not desired that it be where it is not possible or not desired that it be preformed until sometime after the associated preformed until sometime after the associated LCO is within its Applicability, represent LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only Frequency) is stated such that it is only

Enclosure NL-19-026 Page 24 of 74 "required" when it can be and should be "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 performed. With an SR satisfied, SR 3.0.4 imposes no restriction. imposes no restriction.

The use of "met" or "performed" in these The use of "met" or "performed" in these instances conveys specific meanings. A instances conveys specific meanings. A Surveillance is "met" only when the acceptance Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the criteria are satisfied. Known failure of the requirements of a Surveillance, even without requirements of a Surveillance, even without a Surveillance specifically being "performed," a Surveillance specifically being "performed,"

constitutes a Surveillance not "met." constitutes a Surveillance not "met."

"Performance" refers only to the requirement to "Performance" refers only to the requirement to specifically determine the ability to meet the specifically determine the ability to meet the acceptance criteria. acceptance criteria.

Some Surveillances contain notes that modify Some Surveillances contain notes that modify the Frequency of performance the Frequency of performance Current Examples Proposed Example The following examples illustrate the various The following examples illustrate the various ways that Frequencies are specified. In these ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3. not shown) is MODES 1, 2, and 3illustrates the type of Frequency statement that Example 1.4-1 appears in the Technical Specifications (TS).

Example 1.4-2 Example 1.4-1 is modified to address Example 1.4-3 Frequencies as utilized by TS 3.7.11.

Example 1.4-4 Example 1.4-2 is proposed for deletion.

Example 1.4-5 Example 1.4-3 is proposed for deletion.

Example 1.4-6 Example 1.4-4 is proposed for deletion.

Example 1.4-5 is proposed for deletion.

Example 1.4-6 is proposed for deletion.

Basis The Description section of TS 1.4 is modified to reflect the upcoming change in status regarding IP2. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the number and types of Surveillance Requirements that remain in the TSs are limited to those in TS 3.7.11, TS 3.7.12, and TS 3.7.13. This section is modified to provide the rules of usage and examples that continue to be applicable for those TSs.

Enclosure NL-19-026 Page 25 of 74 TS SECTION 2.0, SAFETY LIMITS (SLS)DELETED TS Section 2.0 contains safety limits that are necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity from the reactor core and the RCS pursuant to 10 CFR 50.36(c)(1).

TS Section 2.0 is proposed for deletion in its entirety, since the safety limits do not apply to a reactor that is in a permanently defueled condition.

A mark-up is provided to identify the section as deleted, because the TSs will not be renumbered.

Current TS 2.0 Proposed TS 2.0 TS 2.0 Safety Limits (SLs) TS 2.0 Safety Limits (SLs)DELETED TS 2.1, Safety Limits TS 2.1 is proposed for deletion.

TS 2.1.1, Reactor Core SLs TS 2.1.1 is proposed for deletion.

TS 2.1.2, Reactor Coolant System Pressure TS 2.1.2 is proposed for deletion.

SL TS 2.2, Safety Limit Violations TS 2.2 is proposed for deletion.

TS 2.2.1, If SL 2.1.1 is violated TS 2.2.1 is proposed for deletion.

TS 2.2.2, If SL 2.1.2 is violated TS 2.2.2 is proposed for deletion.

Basis TSs 2.0, 2.1 and 2.2 are proposed for deletion in their entirety.

The restrictions of TS 2.1.1 prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant. It is applicable in MODES 1 and 2. Since TS 2.1.1 applies to an operating reactor, its restrictions have no function in the permanently defueled condition.

The restriction of TS 2.1.2 protects the integrity of the RCS from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. It is applicable in MODES 1 through 5, and MODE 6 when the reactor pressure vessel head is on. Since TS 2.1.2 applies to maintaining the RCS pressure, its restriction has no function in the permanently defueled condition.

TS 2.2.1 defines the action to take if SL 2.1.1 is not met. It requires the unit to be placed in MODE 3.

TS 2.2.1 defines the action to take if SL 2.1.2 is not met. If the unit is in MODE 1 or 2, it requires the unit to be placed in MODE 3. If the unit is MODE 3, 4, 5, or 6, it requires compliance to be restored within 5 minutes After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). The safety limits and safety limit violations TSs apply to an operating reactor and the RCS, they have no function in the permanently defueled condition. These specifications do not apply to the safe storage and handling of spent fuel in the SFP.

Enclosure NL-19-026 Page 26 of 74 TS SECTION 3.0, LIMITING CONDITIONS FOR OPERATION (LCO)

TS Section 3.0 contains the general requirements applicable to all Limiting Conditions for Operation (LCOs) and applies at all times unless otherwise stated in a TS. Proposed revisions to these TSs (including those proposed for deletion) are described below. The corresponding TSs Bases are also being revised to reflect these changes.

A mark-up of this section is provided.

Current LCO 3.0.1 Proposed LCO 3.0.1 LCOs shall be met during the MODES or LCOs shall be met during the MODES or other other specified conditions in the specified conditions in the Applicability, except as Applicability, except as provided in LCO provided in LCO 3.0.2, LCO 3.0.7, and LCO 3.0.8.

3.0.2, LCO 3.0.7, and LCO 3.0.8.

Basis MODES as defined in Table 1.1-1 are defined for operating or refueling conditions. MODES are not used in any PDTS specification. Thus, the reference to MODES is deleted, because this term does not apply to a facility in the permanently defueled condition.

In addition, the references to LCOs 3.0.7 and 3.0.8 are deleted to reflect the proposed deletion of those LCOs discussed below.

Current LCO 3.0.2 Proposed LCO 3.0.2 Upon discovery of a failure to meet an LCO, Upon discovery of a failure to meet an LCO, the the Required Actions of the associated Required Actions of the associated Conditions Conditions shall be met, except as provided shall be met, except as provided in LCO 3.0.5 and in LCO 3.0.5 and LCO 3.0.6. LCO 3.0.6.

If the LCO is met or is no longer applicable If the LCO is met or is no longer applicable prior to prior to expiration of the specified expiration of the specified Completion Time(s),

Completion Time(s), completion of the completion of the Required Action(s) is not Required Action(s) is not required unless required unless otherwise stated.

otherwise stated.

Basis LCO 3.0.2 is modified by eliminating the references to LCOs 3.0.5 and 3.0.6. This change reflects the proposed deletion of those LCOs as discussed below.

LCO 3.0.3 This LCO is proposed for deletion.

Basis LCO 3.0.3 provides the actions that must be implemented when an LCO is not met. It is only applicable in MODES 1 through 5. Pursuant to 10 CFR 50.82(a)(2), the facility license for IP2 will no longer authorize operation of the reactor or placement or retention of fuel in the reactor. Thus, references to operating MODES is no longer relevant. Thus, LCO 3.0.3 is no longer applicable in the permanently defueled condition.

Enclosure NL-19-026 Page 27 of 74 LCO 3.0.4 This LCO is proposed for deletion.

Basis LCO 3.0.4 provides limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. Pursuant to 10 CFR 50.82(a)(2), the facility license for IP2 will no longer authorize operation of the reactor or placement or retention of fuel in the reactor. References to operating MODES are no longer relevant. Thus, LCO 3.0.4 is no longer applicable in the permanently defueled condition.

LCO 3.0.5 This LCO is proposed for deletion.

Basis LCO 3.0.5 provides the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS.

The allowance of LCO 3.0.6 to not comply with the requirements of LCO 3.0.2 (i.e., to not comply with the Required Actions) to allow the performance of SRs on equipment declared inoperable or removed from service is no longer required. The remaining permanently defueled TSs ACTIONS do not include requirements to declare equipment inoperable or to remove it from service.

LCO 3.0.6 This LCO is proposed for deletion.

Basis LCO 3.0.6 addresses the actions required for a supported system when the support system LCO is not met. It is proposed for deletion since there are no LCOs for equipment to be operable or in operation in the PDTS.

LCO 3.0.7 This LCO is proposed for deletion.

Basis LCO 3.0.7 pertains to certain special tests and operations required to be performed at various times over the life of the unit. It is proposed for deletion since special tests and operations are not applicable to a permanently defueled facility.

LCO 3.0.8 This LCO is proposed for deletion.

Basis LCO 3.0.8 addresses the actions required when one or more required snubbers are unable to perform their associated support function(s). It is proposed for deletion, because there are no LCOs for equipment to be operable or in operation in the PDTS. Thus, snubbers are not required to support any TS function.

TS SECTION 3.0, SURVEILLANCE REQUIREMENT (SR) APPLICABILITY TS Section 3.0 contains the general requirements applicable to all SRs and applies at all times unless otherwise stated in a TS. Proposed revisions to these TSs are described below. The corresponding TSs Bases are also being revised to reflect these changes.

A mark-up of this section is provided.

Enclosure NL-19-026 Page 28 of 74 Current SR 3.0.1 Proposed SR 3.0.1 SRs shall be met during the MODES or SRs shall be met during the MODES or other other specified conditions in the specified conditions in the Applicability for Applicability for individual LCOs, unless individual LCOs, unless otherwise stated in the otherwise stated in the SRSurveillances SRSurveillances do not have to be performed on do not have to be performed on inoperable inoperable equipment or variables outside equipment or variables outside specified specified limits.

limits.

Basis SR 3.0.1 is modified by deleting the reference to MODES. Pursuant to 10 CFR 50.82(a)(2), the facility license for IP2 will no longer authorize operation of the reactor or placement or retention of fuel in the reactor.

MODES are not used in any PDTS specification. MODES as defined in Table 1.1-1 are for operating or refueling conditions. This term does not apply to a facility in the permanently defueled condition.

In addition, SR 3.0.1 is modified by eliminating the discussion regarding inoperable equipment.

The remaining LCOs do not include any equipment operability requirements.

Current SR 3.0.2 Proposed SR 3.0.2 The specified Frequency for each SR is met The specified Frequency for each SR is met if the if the Surveillance is performed within 1.25 Surveillance is performed within 1.25 times the times the interval specified in the interval specified in the Frequency, as measured Frequency, as measured from the previous from the previous performance or as measured performance or as measured from the time from the time a specified condition of the a specified condition of the Frequency is Frequency is met.

met.

For Frequencies specified as "once," the above For Frequencies specified as "once," the interval extension does not apply.

above interval extension does not apply.

If a Completion Time requires periodic If a Completion Time requires periodic performance on a "once per . . ." basis, the above performance on a "once per . . ." basis, the Frequency extension applies to each performance above Frequency extension applies to each after the initial performance.

performance after the initial performance.

Exceptions to this Specification are stated in the Exceptions to this Specification are stated individual Specifications.

in the individual Specifications.

Basis SR 3.0.2 provides an allowance for extending the frequency for performance of a SR to 1.25 times the interval specified in the frequency to facilitate scheduling or unforeseen problems that may prevent performance during normal intervals. It is proposed for revision to remove conditions for frequencies that do not exist in PDTS LCOs.

Enclosure NL-19-026 Page 29 of 74 Current SR 3.0.4 Proposed SR 3.0.4 Entry into a MODE or other specified Entry into a MODE or other specified condition in condition in the Applicability of an LCO shall the Applicability of an LCO shall only be made only be made when the LCO's when the LCO's Surveillances have been met Surveillances have been met within their within their specified Frequency, except as specified Frequency, except as provided by provided by SR 3.0.3. When an LCO is not met SR 3.0.3. When an LCO is not met due to due to Surveillances not having been met, entry Surveillances not having been met, entry into a MODE or other specified condition in the into a MODE or other specified condition in Applicability shall only be made in accordance with the Applicability shall only be made in LCO 3.0.4.

accordance with LCO 3.0.4.

This provision shall not prevent entry into MODES This provision shall not prevent entry into or other specified conditions in the Applicability MODES or other specified conditions in the that are required to comply with ACTIONS or that Applicability that are required to comply with are part of a shutdown of the unit.

ACTIONS or that are part of a shutdown of the unit.

Basis SR 3.0.4 is modified by deleting the reference to MODES. Pursuant to 10 CFR 50.82(a)(2), the facility license for IP2 will no longer authorize operation of the reactor or placement or retention of fuel in the reactor.

MODES are not used in any PDTS specification. MODES as defined in Table 1.1-1 are for operating or refueling conditions. This term does not apply to a facility in the permanently defueled condition.

In addition, SR 3.0.4 is modified by eliminating the provision that states that it shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit. TS 3.7.11, TS 3.7.12, and TS 3.7.13 are the only remaining TSs with Required Actions, and they do not contain any Required Actions that would require an entry into another specified condition defined in the Applicability of a TS.

In addition, pursuant to 10 CFR 50.82(a)(2), the facility license for IP2 will no longer authorize operation of the reactor or placement or retention of fuel in the reactor. Thus, there will be no ACTIONS that require the shutdown of a unit.

Enclosure NL-19-026 Page 30 of 74 TS SECTION 3.1, REACTIVITY CONTROL SYSTEMS TS Section 3.1 contains requirements to assure and verify operability of reactivity control systems.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). As a result, reactivity control systems will not be required and the requirements in TS Section 3.1 will not apply in the permanently defueled condition.

TS Section 3.1 is proposed for deletion in its entirety. Thus, a mark-up of this TS section is not provided.

TS 3.1.1, SHUTDOWN MARGIN (SDM) TS 3.1.1 is proposed for deletion.

TS 3.1.1 is applicable in MODE 2 with keff < 1.0, and MODES 3 through 5. Thus, the TS will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2.

At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in the applicable MODES and specified conditions will no longer occur. As a result, TS 3.1.1 will not apply in the permanently defueled condition.

TS 3.1.2, Core Reactivity TS 3.1.2 is proposed for deletion.

TS 3.1.2 is applicable in MODES 1 and 2. Thus, the TS will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2. At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel.

Thus, operation in MODES 1 and 2 will no longer occur. As a result, TS 3.1.2 will not apply in the permanently defueled condition.

TS 3.1.3, Moderator Temperature TS 3.1.3 is proposed for deletion.

Coefficient (MTC)

TS 3.1.3 is applicable in MODE 1 and MODE 2 with keff 1.0 for the upper MTC limit and MODES 1, 2, and 3 for the lower MTC limit. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2.

At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in the applicable MODES and specified conditions will no longer occur. As a result, this TS will not apply in the permanently defueled condition.

Enclosure NL-19-026 Page 31 of 74 TS 3.1.4, Rod Group Alignment Limits TS 3.1.4, including Table 3.1.4-1, is proposed for deletion.

TS 3.1.4 is applicable in MODES 1 and 2. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2.

At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in MODES 1 and 2 will no longer occur. As a result, this TS will not apply in the permanently defueled condition.

TS 3.1.5, Shutdown Bank Insertion Limits TS 3.1.5 is proposed for deletion.

TS 3.1.5 is applicable in MODES 1 and 2. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2.

At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in MODES 1 and 2 will no longer occur. As a result, this TS will not apply in the permanently defueled condition.

TS 3.1.6, Control Bank Insertion Limits TS 3.1.6 is proposed for deletion.

TS 3.1.6 is applicable in MODE 1 and MODE 2 with keff with 1.0. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2. At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in the applicable MODES and specified conditions will no longer occur. As a result, this TS will not apply in the permanently defueled condition.

TS 3.1.7, Rod Position Indication TS 3.1.7 is proposed for deletion.

TS 3.1.7 is applicable in MODES 1 and 2. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2.

At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation MODES 1 and 2 will no longer occur. As a result, this TS will not apply in the permanently defueled condition.

Enclosure NL-19-026 Page 32 of 74 TS 3.1.8, PHYSICS TESTS Exceptions - TS 3.1.8 is proposed for deletion.

MODE 2 TS 3.1.8 is applicable during PHYSICS TESTS initiated in MODE 2. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2. At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in the applicable MODE and specified condition will no longer occur.

As a result, this TS will not apply in the permanently defueled condition.

TS SECTION 3.2, POWER DISTRIBUTION LIMITS TS Section 3.2 contains power distribution limits that provide assurance that fuel design criteria are not exceeded and the accident analysis assumptions remain valid.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). As a result, the requirements in TS Section 3.2 will not apply in the permanently defueled condition.

TS Section 3.2 is proposed for deletion in its entirety. Thus, a mark-up of this TS section is not provided.

TS 3.2.1, Heat Flux Hot Channel Factor TS 3.2.1 is proposed for deletion.

(FQ(Z))

TS 3.2.1 is applicable in MODE 1. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2.

At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in MODE 1 will no longer occur. As a result, this TS will not apply in the permanently defueled condition.

TS 3.2.2, Nuclear Enthalpy Rise Hot TS 3.2.2 is proposed for deletion.

Channel Factor (FNH)

TS 3.2.2 is applicable in MODE 1. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2.

At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in MODE 1 will no longer occur. As a result, this TS will not apply in the permanently defueled condition.

Enclosure NL-19-026 Page 33 of 74 TS 3.2.3, AXIAL FLUX DIFFERENCE TS 3.2.3 is proposed for deletion.

(AFD) (Constant Axial Offset Control (CAOC) Methodology) TS 3.2.3 is applicable in MODE 1 with Thermal Power > 15% RTP. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2. At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in the applicable MODE and specified condition will no longer occur.

As a result, this TS will not apply in the permanently defueled condition.

TS 3.2.4, QUADRANT POWER TILT TS 3.2.4 is proposed for deletion.

RATIO (QPTR)

TS 3.2.4 is applicable in MODE 1 with Thermal Power > 50% RTP. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2. At that time, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel. Thus, operation in the applicable MODE and specified condition will no longer occur.

As a result, this TS will not apply in the permanently defueled condition.

TS SECTION 3.3, INSTRUMENTATION TS Section 3.3 contains operability requirements for sensing and control instrumentation required for safe operation of the facility.

After the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2, the 10 CFR Part 50 license will no longer authorize emplacement or retention of fuel in the reactor vessel. The TSs that do not apply in the permanently defueled condition, or for structures, systems, or components that are not needed for accident mitigation in the defueled condition are being proposed for deletion.

TS Section 3.3 is proposed for deletion in its entirety. Thus, a mark-up of this TS section is not provided.

Current IP2 TS Basis for Change TS 3.3.1, Reactor Protection System (RPS) TS 3.3.1, including Table 3.3.1-1, is proposed for Instrumentation deletion.

Dependent on function as defined in Table 3.3.1-1, TS 3.3.1 is applicable in various portions of MODES 1 through 5 or other specified conditions.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in the applicable

Enclosure NL-19-026 Page 34 of 74 MODES and specified conditions will no longer occur. Thus, the RPS will not be required in the permanently defueled condition.

TS 3.3.2, Engineered Safety Feature TS 3.3.2, including Table 3.3.2-1, is proposed for Actuation System (ESFAS) Instrumentation deletion.

Dependent on function as defined in Table 3.3.2-1, TS 3.3.2 is applicable in various portions of MODES 1 through 4 or other specified conditions.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in the applicable MODES and specified conditions will no longer occur. As a result, the ESFAS instrumentation will not be required in the permanently defueled condition.

TS 3.3.3, Post Accident Monitoring (PAM) TS 3.3.3, including Table 3.3.3-1, is proposed for Instrumentation deletion.

TS 3.3.3 is applicable in MODES 1, 2, and 3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 3 will no longer occur. As a result, the PAM instrumentation will not be required in the permanently defueled condition.

TS 3.3.4, Remote Shutdown TS 3.3.4 is proposed for deletion.

TS 3.3.4 is applicable in MODES 1, 2, and 3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 3 will no longer occur. As a result, the remote shutdown functions will not be required in the permanently defueled condition.

TS 3.3.5, Loss of Power (LOP) Diesel TS 3.3.5 is proposed for deletion Generator (DG) Start Instrumentation TS 3.3.5 is applicable in MODES 1 through 4 and when the associated DG is required to be OPERABLE by LCO 3.8.2, AC Sources -

Shutdown.

Enclosure NL-19-026 Page 35 of 74 After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur.

In addition, TS 3.8.2 is proposed for deletion as discussed in the table for Section 3.8. The postulated DBAs and events associated with reactor or power operation analyzed in UFSAR Chapter 14 are no longer applicable in the permanently defueled condition. The analyses of the remaining DBAs (i.e., the FHA in the Fuel Handling Building and the Accident Release of Waste Liquid or Waste Gas) do not rely on alternating current (AC) electrical power sources for accident mitigation (dose consequences are acceptable without relying on any electrically-powered SSCs to remain functional during and following the event). There are no active systems credited as part of the initial conditions of an analysis or as part of the primary success path for mitigation of these events with the IP2 permanently shut down and defueled.

As a result, the LOP DG start instrumentation will not be required in the permanently defueled condition.

TS 3.3.6, Containment Purge System and TS 3.3.6, including Table 3.3.6-1, is proposed for Pressure Relief Line Isolation deletion.

Instrumentation TS 3.3.6 is applicable in MODES 1 through 4 and during movement of recently irradiated fuel assemblies within containment.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur.

In addition, the Bases for TS 3.3.6 defines recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The PDTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

Enclosure NL-19-026 Page 36 of 74 As a result, the containment purge system and pressure relief line isolation instrumentation will not be required in the permanently defueled configuration.

TS 3.3.7, Control Room Ventilation System TS 3.3.7, including Table 3.3.7-1, is proposed for (CRVS) Actuation Instrumentation deletion.

TS 3.3.7 is applicable in MODES 1 through 4 and during movement of recently irradiated fuel assemblies.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur.

In addition, the Bases for TS 3.3.7 defines recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The PDTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

As a result, CRVS actuation instrumentation will not be required in the permanently defueled configuration.

TS SECTION 3.4, REACTOR COOLANT SYSTEM (RCS)

TS Section 3.4 contains requirements that provide for appropriate control of process variables, design features, or operating restrictions needed for appropriate functional capability of RCS equipment required for safe operation of the facility.

After the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2, 10 CFR Part 50 license will no longer authorize emplacement or retention of fuel in the reactor vessel.

The TSs that do not apply in a defueled condition, or for structures, systems, or components that are not needed for accident mitigation in the defueled condition, are being proposed for deletion.

TS Section 3.4 is proposed for deletion in its entirety. Thus, a mark-up of this TS section is not provided.

Current IP2 TS Basis for Change TS 3.4.1, RCS Pressure, Temperature, and TS 3.4.1 is proposed for deletion.

Flow Departure from Nucleate Boiling (DNB) Limits TS 3.4.1 is applicable in MODE 1, with the pressurizer pressure limit not being applicable at specifically defined periods. After the certifications required by 10 CFR 50.82(a)(1) are docketed for

Enclosure NL-19-026 Page 37 of 74 IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 1 will no longer occur. As a result, the RCS pressure, temperature, and low DNB limits are no longer applicable in the permanently defueled condition.

TS 3.4.2, RCS Minimum Temperature for TS 3.4.2 is proposed for deletion.

Criticality TS 3.4.2 is applicable in MODE 1 and MODE 2 with keff 1.0. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in the applicable MODES and specified condition will no longer occur. As a result, the RCS minimum temperature for criticality limit is no longer applicable in the permanently defueled condition.

TS 3.4.3, RCS Pressure and Temperature TS 3.4.3, including Figures 3.4.3-1 and 3.4.3-2, is (P/T) Limits proposed for deletion.

TS 3.4.3 is applicable at all times.

If the requirements are not met in MODE 1, 2, 3, or 4, and the parameter(s) is not restored to within limit or the RCS is determined to not be acceptable for continued operation in accordance with Required Action A.1 or A.2, then the unit is required to be placed in MODE 3 and eventually MODE 5 with RCS pressure < 500 psig.

If the requirements are not met at any time in other than MODE 1, 2, 3, or 4, action is required to determine that the RCS is acceptable for continued operation prior to entering MODE 4 in accordance with Required Action C.2.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 5 will no longer occur. As a result, the RCS P/T limits are no longer applicable in the permanently defueled condition.

Enclosure NL-19-026 Page 38 of 74 TS 3.4.4, RCS Loops - MODES 1 and 2 TS 3.4.4 is proposed for deletion.

TS 3.4.4 is applicable in MODES 1 and 2. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 and 2 will no longer occur.

As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.5, RCS Loops - MODE 3 TS 3.4.5 is proposed for deletion.

TS 3.4.5 is applicable in MODE 3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 3 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.6, RCS Loops - MODE 4 TS 3.4.6 is proposed for deletion.

TS 3.4.6 is applicable in MODE 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.7, RCS Loops - MODE 5, Loops TS 3.4.7 is proposed for deletion.

Filled TS 3.4.7 is applicable in MODE 5 with the RCS loops filled. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 5 with the RCS loops filled will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

Enclosure NL-19-026 Page 39 of 74 TS 3.4.8, RCS Loops - MODE 5, Loops TS 3.4.8 is proposed for deletion.

Not Filled TS 3.4.8 is applicable in MODE 5 with the RCS loops not filled. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 5 with the RCS loops not filled will no longer occur.

As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.9, Pressurizer TS 3.4.9 is proposed for deletion.

TS 3.4.9 is applicable in MODES 1 through 3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 3 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.10, Pressurizer Safety Valves TS 3.4.10 is proposed for deletion.

TS 3.4.10 is applicable in MODES 1 through 3 and MODE 4 with all RCS cold leg temperatures

> 288°F. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 with the specified condition will no longer occur.

As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.11, Pressurizer Power Operated TS 3.4.11 is proposed for deletion.

Relief Valves (PORVs)

TS 3.4.11 is applicable in MODES 1 through 3.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 3 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

Enclosure NL-19-026 Page 40 of 74 TS 3.4.12, Low Temperature Overpressure TS 3.4.12, including Table 3.4.12-1 and Figures Protection (LTOP) 3.4.12-1 through 3.4.12-6, is proposed for deletion.

TS 3.4.12 is applicable in MODE 4 when any RCS cold leg temperature is 288°F, MODE 5, and MODE 6 when the reactor vessel head is on. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in the applicable MODES and specified conditions will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.13, RCS Operational LEAKAGE TS 3.4.13 is proposed for deletion.

TS 3.4.13 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.14, RCS Pressure Isolation Valve TS 3.4.14 is proposed for deletion.

(PIV) Leakage TS 3.4.14 is applicable in MODES 1 through 3 and MODE 4 (except valves in the residual heat removal (RHR) flow path when in, or during the transition to or from, the RHR mode of operation).

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 with the specified condition will no longer occur.

As a result, this TS will not be applicable in a permanently defueled condition.

Enclosure NL-19-026 Page 41 of 74 TS 3.4.15, RCS Leakage Detection TS 3.4.15 is proposed for deletion.

Instrumentation TS 3.4.15 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.16, RCS Specific Activity TS 3.4.16 is proposed for deletion.

TS 3.4.16 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.4.17, Steam Generator (SG) Tube TS 3.4.17 is proposed for deletion.

Integrity TS 3.4.17 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS (ECCS)

TS Section 3.5 contains requirements that provide for appropriate functional capability of ECCS equipment required for mitigation of DBAs or transients so as to protect the integrity of a fission product barrier.

After the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP2, the 10 CFR Part 50 license will no longer authorize emplacement or retention of fuel in the reactor vessel. The TSs that do not apply in a defueled condition, or for structures, systems, or components that are not needed for accident mitigation in the defueled condition, are being proposed for deletion.

TS Section 3.5 is proposed for deletion in its entirety. Thus, a mark-up of this TS section is not provided.

Enclosure NL-19-026 Page 42 of 74 Current IP2 TS Basis for Change TS 3.5.1, Accumulators TS 3.5.1 is proposed for deletion.

TS 3.5.1 is applicable in MODES 1 and 2 and MODE 3 with RCS pressure > 1000 psig. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 3 with the specified condition will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.5.2, ECCS - Operating TS 3.5.2 is proposed for deletion.

TS 3.5.2 is applicable in MODES 1 through 3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 3 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.5.3, ECCS - Shutdown TS 3.5.3 is proposed for deletion.

TS 3.5.3 is applicable in MODE 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.5.4, Refueling Water Storage Tank TS 3.5.4 is proposed for deletion.

(RWST)

TS 3.5.4 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

Enclosure NL-19-026 Page 43 of 74 TS SECTION 3.6, CONTAINMENT SYSTEMS TS Section 3.6 contains requirements that assure the integrity of the containment, depressurization and cooling systems, and containment isolation valves.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). The TSs that do not apply in a defueled condition, or for structures, systems, or components that are not needed for accident mitigation in the defueled condition, are being proposed for deletion.

TS Section 3.6 is proposed for deletion in its entirety. Thus, a mark-up of this TS section is not provided.

Current IP2 TS Basis for Change TS 3.6.1, Containment TS 3.6.1 is proposed for deletion.

TS 3.6.1 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.6.2, Containment Air Locks TS 3.6.2 is proposed for deletion.

TS 3.6.2 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.6.3, Containment Isolation Valves TS 3.6.3 is proposed for deletion.

TS 3.6.3 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

Enclosure NL-19-026 Page 44 of 74 TS 3.6.4, Containment Pressure TS 3.6.4 is proposed for deletion.

TS 3.6.4 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.6.5, Containment Air Temperature TS 3.6.5 is proposed for deletion.

TS 3.6.5 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.6.6, Containment Spray System and TS 3.6.6 is proposed for deletion.

Containment Fan Cooler Unit (FCU)

System TS 3.6.6 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.6.7, Recirculation pH Control System TS 3.6.7 is proposed for deletion.

TS 3.6.7 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.6.8, Not Used TS 3.6.8 is proposed for deletion.

The deletion of this placeholder is an administrative change.

Enclosure NL-19-026 Page 45 of 74 TS 3.6.9, Isolation Valve Seal Water TS 3.6.9 is proposed for deletion.

(IVSW) System TS 3.6.9 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.6.10, Weld Channel and Penetration TS 3.6.10 is proposed for deletion.

Pressurization System (WC&PPS)

TS 3.6.10 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS SECTION 3.7, PLANT SYSTEMSSPENT FUEL PIT REQUIREMENTS TS Section 3.7 provides requirements for the appropriate functional capability of plant equipment required for safe operation of the facility, including the plant being in a defueled condition.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). The TSs that do not apply in a defueled condition, or for structures, systems, or components that are not needed for accident mitigation in the defueled condition, are being proposed for deletion.

TS 3.7 is retitled to reflect that the remaining TS address SFP requirements.

TS 3.7.1 through TS 3.7.10 and TS 3.7.14 are proposed for deletion in their entirety. Thus, mark-ups of these TSs are not provided.

TS 3.7.11 provides the limit regarding the SFP water level. It will be retained in the PDTS, and modified to eliminate the reference to LCO 3.0.3.

TS 3.7.12 provides requirements regarding the SFP boron concentration. It will be retained in the PDTS, and modified to eliminate the reference to LCO 3.0.3.

TS 3.7.13 provides the limits for storing fuel assemblies in the SFP. It will be retained in the PDTS, and modified to eliminate the reference to LCO 3.0.3.

Mark-ups of TS 3.7.11, TS 3.7.12, and TS 3.7.13 are provided in Attachment 1 to this enclosure.

Enclosure NL-19-026 Page 46 of 74 Current IP2 TS Basis for Change TS 3.7, PLANT SYSTEMS Proposed TS 3.7, PLANT SYSTEMSSPENT FUEL PIT REQUIREMENTS The TS section is proposed to be retitled to reflect that the remaining TSs in the section deal with SFP requirements in a permanently shut down and defueled facility. This is an administrative change.

TS 3.7.1, Main Steam Safety Valves TS 3.7.1, including Tables 3.7.1-1 and 3.7.1-2, is (MSSVs) proposed for deletion.

TS 3.7.1 is applicable in MODES 1 through 3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 3 with the specified condition will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.7.2, Main Steam Isolation Valves TS 3.7.2 is proposed for deletion.

(MSIVs) and Main Steam Check Valves (MSCVs) TS 3.7.2 is applicable in MODE 1 and MODES 2 and 3 except when all MSIVs are closed. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 3 with the specified condition will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.7.3, Main Feedwater Isolation TS 3.7.3 is proposed for deletion.

TS 3.7.3 is applicable in MODES 1 through 3 except when the flow path is isolated by a closed and deactivated automatic valve or a closed manual valve. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 3 with the specified condition will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

Enclosure NL-19-026 Page 47 of 74 TS 3.7.4, Atmospheric Dump Valves TS 3.7.4 is proposed for deletion.

(ADVs)

TS 3.7.4 is applicable in MODES 1 through 3 and MODE 4 when the SG is relied upon for heat removal. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 with the specified condition will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.7.5, Auxiliary Feedwater (AFW) TS 3.7.5 is proposed for deletion.

System TS 3.7.5 is applicable in MODES 1 through 3 and MODE 4 when the SG is relied upon for heat removal. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 with the specified condition will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.7.6, Condensate Storage Tank (CST) TS 3.7.6 is proposed for deletion.

TS 3.7.6 is applicable in MODES 1 through 3 and MODE 4 when the SG is relied upon for heat removal. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 with the specified condition will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.7.7, Component Cooling Water (CCW) TS 3.7.7 is proposed for deletion.

System TS 3.7.7 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

Enclosure NL-19-026 Page 48 of 74 TS 3.7.8, Service Water System (SWS) TS 3.7.8 is proposed for deletion.

TS 3.7.8 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.7.9, Ultimate Heat Sink (UHS) TS 3.7.9 is proposed for deletion.

TS 3.7.9 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.7.10, Control Room Ventilation TS 3.7.10 is proposed for deletion.

System (CRVS)

TS 3.7.10 is applicable in MODES 1 through 4, and during movement of recently irradiated fuel assemblies.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur.

In addition, the Bases for TS 3.7.10 defines recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The PDTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

As a result, the CRVS will not be required in the permanently defueled configuration.

Enclosure NL-19-026 Page 49 of 74 TS 3.7.11, Spent Fuel Pit Water Level TS 3.7.11 is retained in the PDTS. It is administratively changed from affecting plant systems to addressing spent fuel pit requirements to comport with the remaining PDTS Section 3.7 LCOs. In addition, the NOTE in Required Action A.1 (LCO 3.0.3 is not applicable) is proposed to be deleted to conform to the deletion of TS LCO 3.0.3 as previously proposed.

TS 3.7.12, Spent Fuel Pit Boron TS 3.7.12 is retained in the PDTS. It is Concentration administratively changed from affecting plant systems to addressing spent fuel pit requirements to comport with the remaining PDTS Section 3.7 LCOs. In addition, the NOTE in Required Action A.1 (LCO 3.0.3 is not applicable) is proposed to be deleted to conform to the deletion of TS LCO 3.0.3 as previously proposed.

TS 3.7.13, Spent Fuel Pit Storage TS 3.7.13, including Figures 3.7.13-1 through 3.7.13-5, is retained in the PDTS. It is administratively changed from affecting plant systems to addressing spent fuel pit requirements to comport with the remaining PDTS Section 3.7 LCOs. In addition, the NOTE in Required Action A.1 (LCO 3.0.3 is not applicable) is proposed to be deleted to conform to the deletion of TS LCO 3.0.3 as previously proposed.

TS 3.7.14, Secondary Specific Activity TS 3.7.14 is proposed for deletion.

TS 3.7.14 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS SECTION 3.8, ELECTRICAL POWER SYSTEMS TS Section 3.8 contains operability requirements that provide for appropriate functional capability of plant electrical equipment required for safe operation of the facility, including the plant being in a defueled condition.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

The DBAs and transients analyzed in UFSAR Chapter 14 will no longer be applicable in the permanently defueled condition, with the exception of the FHA in the Fuel Handling Building and the Accidental Releases of Waste Liquid or Waste Gas. There are no active systems credited as

Enclosure NL-19-026 Page 50 of 74 part of the initial conditions of these analyses or as part of the primary success path for mitigation of these events with the IP2 permanently shut down and defueled.

TS Section 3.8 is proposed for deletion in its entirety. Thus, a mark-up of this TS section is not provided.

Current IP2 TS Basis for Change TS 3.8.1, AC Sources - Operating TS 3.8.1 is proposed for deletion.

TS 3.8.1 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.8.2, AC Sources - Shutdown TS 3.8.2 is proposed for deletion.

TS 3.8.2 is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 5 and 6 will no longer occur.

In addition, the Bases for TS 3.8.2 defines recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The PDTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

As a result, this TS will not be applicable in the permanently defueled configuration.

TS 3.8.3, Diesel Fuel Oil and Starting Air TS 3.8.3 is proposed for deletion.

TS 3.8.3 is applicable when the associated Diesel Generator (DG) is required to be OPERABLE.

TS 3.8.1 and TS 3.8.2 provide the OPERABILITY requirements regarding the DGs. As previously discussed, these TSs are proposed for deletion.

Thus, TS 3.8.3 is not included in the PDTS because the TSs that it supports are no longer

Enclosure NL-19-026 Page 51 of 74 required after IP2 is permanently shut down and defueled.

TS 3.8.4, DC Sources - Operating TS 3.8.4 is proposed for deletion.

TS 3.8.4 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.8.5, DC Sources - Shutdown TS 3.8.5 is proposed for deletion.

TS 3.8.5 is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 5 and 6 will no longer occur.

In addition, the Bases for TS 3.8.5 defines recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The PDTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

As a result, this TS will not be applicable in the permanently defueled configuration.

TS 3.8.6, Battery Parameters TS 3.8.6 is proposed for deletion.

TS 3.8.6 is applicable when the associated DC electrical power subsystems are required to be OPERABLE.

TS 3.8.4 and TS 3.8.5 provide the OPERABILITY requirements regarding the DC sources. As previously discussed, these TSs are proposed for deletion. Thus, TS 3.8.6 is not included in the PDTS because the TSs that it supports are no longer required after IP2 is permanently shut down and defueled.

Enclosure NL-19-026 Page 52 of 74 TS 3.8.7, Inverters - Operating TS 3.8.7 is proposed for deletion.

TS 3.8.7 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.8.8, Inverters - Shutdown TS 3.8.8 is proposed for deletion.

TS 3.8.8 is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 5 and 6 will no longer occur.

In addition, the Bases for TS 3.8.8 defines recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The PDTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

As a result, this TS will not be applicable in the permanently defueled configuration.

TS 3.8.9, Distribution Systems - Operating TS 3.8.9 is proposed for deletion.

TS 3.8.9 is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 1 through 4 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.8.10, Distribution Systems - TS 3.8.10 is proposed for deletion.

Shutdown TS 3.8.10 is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies.

Enclosure NL-19-026 Page 53 of 74 After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODES 5 and 6 will no longer occur.

In addition, the Bases for TS 3.8.10 defines recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The PDTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

As a result, this TS will not be applicable in the permanently defueled configuration.

TS SECTION 3.9, REFUELING OPERATIONS TS Section 3.9 contains requirements that provide for appropriate functional capability of parameters and equipment that are required for mitigation of DBAs during refueling operations (moving irradiated fuel to or from the reactor core).

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

The DBAs and transients analyzed in UFSAR Chapter 14 will no longer be applicable in the permanently defueled condition, with the exception of the FHA in the Fuel Handling Building and the Accidental Releases of Waste Liquid or Waste Gas. There are no active systems credited as part of the initial conditions of these analyses or as part of the primary success path for mitigation of these events with the IP2 permanently shut down and defueled.

TS Section 3.9 is proposed for deletion in its entirety. Thus, a mark-up of this TS section is not provided.

Current IP2 TS Basis for Change TS 3.9.1, Boron Concentration TS 3.9.1 is proposed for deletion.

TS 3.9.1 is applicable in MODE 6. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 6 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

Enclosure NL-19-026 Page 54 of 74 TS 3.9.2, Nuclear Instrumentation TS 3.9.2 is proposed for deletion.

TS 3.9.2 is applicable in MODE 6. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 6 will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.9.3, Containment Penetrations TS 3.9.3 is proposed for deletion.

TS 3.9.3 is applicable during movement of recently irradiated fuel assemblies within containment. The Bases for TS 3.9.3 defines recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The PDTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.9.4, Residual Heat Removal (RHR) TS 3.9.4 is proposed for deletion.

and Coolant Circulation - High Water Level TS 3.9.4 is applicable in MODE 6 with the water level 23 feet above the top of the reactor vessel flange. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 6 with the specified condition will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.9.5, Residual Heat Removal (RHR) TS 3.9.5 is proposed for deletion.

and Coolant Circulation - Low Water Level TS 3.9.5 is applicable in MODE 6 with the water level < 23 feet above the top of the reactor vessel flange. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Thus, operation in MODE 6 with the specified condition will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition.

Enclosure NL-19-026 Page 55 of 74 TS 3.9.6, Refueling Cavity Water Level TS 3.9.6 is proposed for deletion.

TS 3.9.6 is applicable during movement of irradiated fuel assemblies within containment.

The IP2 10 CFR 50 facility license will no longer authorize use of the facility for power operation or emplacement or retention of fuel in the reactor vessel as provided in 10 CFR Part 50.82(a)(2).

Thus, the deletion of this TS is appropriate.

TS Section 4.0, Design Features Currently, TS Section 4.0, Design Features, provides information and design requirements associated with plant systems.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). The TSs that do not apply in a defueled condition, or for structures, systems, or components that are not needed for accident mitigation in the defueled condition, are being proposed for deletion.

TS 4.1 is modified and TS 4.3.1.2 is proposed for deletion.

Current IP2 TS Basis for Change Current TS 4.1 Proposed TS 4.1 For the purpose of satisfying 10 CFR For the purpose of satisfying 10 CFR Part 20, Part 20, the Restricted Area is the same the Restricted Area is the same as the Exclusion as the Exclusion Area shown in UFSAR, Area shown in UFSAR the Defueled Safety Figure 2.2-2. Analysis Report (DSAR), Figure 2.2-2.

The UFSAR is proposed to be retitled as the Defueled Safety Analysis Report (DSAR). The DSAR is the document that will be maintained in accordance with 10 CFR 50.59 and 10 CFR 50.71(e) and remain applicable to IP2 in the permanently shut down and defueled condition.

Current TS 4.3.1.2 TS 4.3.1.2 is proposed for deletion.

The new fuel storage racks are designed After the certifications required by 10 CFR and shall be maintained with 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). IP2 will never acquire new fuel again. Thus, this TS is not applicable in the permanently shut down and defueled condition.

Enclosure NL-19-026 Page 56 of 74 TS Section 5.0, Administrative Controls TS Section 5.0 establishes the requirements associated with staffing, training, procedures, programs and reporting requirements. This section is proposed to be revised to include only those administrative requirements needed for safe storage and movement of fuel in the SFP.

The UFSAR is proposed to be retitled as the DSAR and the references to UFSAR in Section 5.0 are replaced with DSAR. The DSAR is the document that will be maintained in accordance with 10 CFR 50.59 and 10 CFR 50.71(e) and remain applicable to IP2 in the permanently shut down and defueled condition.

Current TS 5.2.1 Proposed TS 5.2.1

a. Lines of authority, responsibility, and a. Lines of authority, responsibility, and communication shall be defined and communication shall be defined and established throughout highest established throughout highest management management levels, intermediate levels, levels, intermediate levels, and all and all decommissioning organization decommissioning organization positions.

positions. These relationships shall be These relationships shall be documented and documented and updated, as updated, as appropriate, in organization appropriate, in organization charts, charts, functional descriptions of functional descriptions of departmental departmental responsibilities and responsibilities and relationships, and job relationships, and job descriptions for key descriptions for key personnel positions, personnel positions, or in equivalent forms of or in equivalent forms of documentation. documentation. These requirements These requirements including the facility- including the facility-specific titles of those specific titles of those personnel fulfilling personnel fulfilling the responsibilities of the the responsibilities of the positions positions delineated in these Technical delineated in these Technical Specifications shall be documented in the Specifications shall be documented in the UFSARDSAR, UFSAR, Basis TS 5.2.1 will be retained, but modified by replacing the reference to the UFSAR in TS 5.2.1.a with a reference to the DSAR.

Following the permanent shut down and defueling of IP2, the IP2 UFSAR will be updated to reflect this condition. The document will be retitled as the DSAR. Thus, this proposed change is an administrative change.

Current TS 5.4.1 Proposed TS 5.4.1

a. The procedures applicable to the safe a. The procedures applicable to the safe storage storage of nuclear fuel recommended in of nuclear fuel recommended in Regulatory Regulatory Guide 1.33, Revision 2, Guide 1.33, Revision 2, Appendix A, February Appendix A, February 1978 except as 1978 except as provided in the quality provided in the quality assurance assurance program described or referenced in program described or referenced in the the Updated FSARDSAR; Updated FSAR;

d. Fire Protection Program

d. Fire Protection Program implementation. implementationDeleted.

Enclosure NL-19-026 Page 57 of 74 Basis Following the permanent shut down and defueling of IP2, the IP2 UFSAR will be updated to reflect this condition. The document will be retitled as the DSAR. Thus, the proposed change to TS 5.4.1.a is an administrative change.

TS 5.4.1.d is proposed for deletion. This is consistent with the proposed deletion of License Condition 2.K. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the fire protection program will be revised to take into account the decommissioning facility conditions and activities.

IP2 will continue to utilize the defense-in-depth concept, placing special emphasis on detection and suppression in order to minimize radiological releases to the environment. During the decommissioning process, a fire protection program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. However, the regulation is applicable regardless of whether the TSs contain a requirement to establish, implement, and maintain procedures for a fire protection program. Therefore, a TS requirement for fire protection program procedures for a permanently shut down and defueled facility is not needed.

TS 5.5.2, Primary Coolant Sources Outside The title for TS 5.5.2 is deleted.

Containment This is an administrative change, because the TS was previously deleted in another license amendment.

TS 5.5.3, Radioactive Effluent Controls TS 5.5.3 will be retained, but TS 5.5.3.d, TS Program 5.5.3.h, and TS 5.5.3.i are proposed to be modified to replace unit with unit/facility. Since IP2 will be permanently shut down before Indian Point Unit 3 (IP3) is permanently shut down, IP2 will be permanently defueled while IP3 will still be in operation. This is an administrative change to better reflect the status of IP2 and IP3.

TS 5.5.4, Component Cyclic or Transient TS 5.5.4 is proposed for deletion.

Limit After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Operation in MODES 1 through 6 will never occur again. Thus, this TS is not applicable in the permanently shut down and defueled condition.

Enclosure NL-19-026 Page 58 of 74 TS 5.5.5, Reactor Coolant Pump Flywheel TS 5.5.5 is proposed for deletion.

Inspection Program After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Thus, the reactor coolant pumps will no longer perform a function in the permanently shut down and defueled state.

TS 5.5.5 is proposed for deletion to be consistent with the proposed deletion of TS 3.4.4 through TS 3.4.8. These TSs provide the operability requirements for the RCS loops. Given their proposed deletion, there is no need to maintain this support program.

TS 5.5.6, Inservice Testing Program TS 5.5.6 is proposed for deletion.

TS 5.5.6 provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. In the permanently shut down and defueled condition, there are no longer any ASME Code class pumps and valves that remain in operation and are relied upon to mitigate a DBA. As such, the inservice testing program will no longer be relevant in the permanently shut down and defueled condition.

TS 5.5.7, Steam Generator (SG) Program TS 5.5.7 is proposed for deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the SG will no longer perform a function in the permanently shut down and defueled state.

TS 3.4.17 provides the requirements to ensure SG tube integrity in MODES 1 through 4. It is proposed for deletion. Thus, the proposed deletion of this supporting TS program is appropriate.

Enclosure NL-19-026 Page 59 of 74 TS 5.5.8, Secondary Water Chemistry TS 5.5.8 is proposed for deletion.

Program After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, there will be no need to monitor secondary water chemistry to inhibit steam generator tube degradation in the permanently defueled condition. Thus, the deletion of this TS is appropriate.

TS 5.5.9, Ventilation Filter Testing Program TS 5.5.9 is proposed for deletion.

(VFTP)

As previously discussed, TS 3.7.10, Control Room Ventilation System (CRVS), is proposed for deletion. Thus, this support program is not required in the permanently shut down and defueled condition.

TS 5.5.11, Diesel Fuel Oil Testing Program TS 5.5.11 is proposed for deletion.

As previously discussed, TS 3.8.1, TS 3.8.2, and TS 3.8.3 are proposed for deletion. These TSs define the operability requirements regarding the DGs. Thus, this support program is not required in the permanently shut down and defueled condition.

TS 5.5.12, Technical Specifications (TS) TS 5.5.12 will be retained, but modified by Bases Control Program replacing the references to the updated UFSAR and UFSAR in TS 5.5.12.b.2 and TS 5.5.12.c with references to the DSAR.

Following the permanent shut down and defueling of IP2, the IP2 UFSAR will be updated to reflect this condition. The document will be retitled as the DSAR. Thus, this proposed change is an administrative change.

TS 5.5.13, Safety Function Determination TS 5.5.13 is proposed for deletion.

Program (SFDP)

This program was established to ensure loss of safety function is detected and appropriate actions taken. The LCOs remaining in the PDTS do not rely on the operability of any active equipment or systems to satisfy the LCO.

Because 10 CFR 50.82(a)(2) prohibits operation of the plant or placing fuel in the reactor vessel, there is no longer a need for redundant systems. Therefore, the requirements of the SFDP, which directs cross train checks of multiple and redundant safety systems, no longer apply.

Enclosure NL-19-026 Page 60 of 74 Additionally, the SFDP is invoked in LCO 3.0.6, which is being deleted in its entirety as previously discussed. Thus, the SFDP is not needed in a permanently shut down and defueled condition.

TS 5.5.14, Containment Leakage Rate TS 5.5.14 is proposed for deletion.

Testing Program The IP2 10 CFR 50 facility license will no longer authorize use of the facility for power operation or emplacement or retention of fuel in the reactor vessel as provided in 10 CFR Part 50.82(a)(2).

Containment integrity is not credited in the analysis of the accidents that remain credible in the permanently defueled condition. In addition, TS 3.6.1 through TS 3.6.6 regarding the containment systems are proposed for deletion.

Thus, the deletion of this TS is appropriate.

TS 5.5.15, Battery Monitoring and TS 5.5.15 is proposed for deletion.

Maintenance Program As previously discussed, TS 3.8.4, TS 3.8.5, and TS 3.8.6 are proposed for deletion. These TSs define the operability requirements regarding the DC sources. Thus, this support program is not required in the permanently shut down and defueled condition.

TS 5.5.16, Control Room Envelope TS 5.5.16 is proposed for deletion.

Habitability Program As previously discussed, TS 3.7.10, Control Room Ventilation System (CRVS), is proposed for deletion. Thus, this support program is not required in the permanently shut down and defueled condition.

TS 5.6.4, Not Used The placeholder for TS 5.6.4 is proposed for deletion. This is an administrative change to reflect reorganization of the TS.

TS 5.6.5, Core Operating Limits Report TS 5.6.5 is proposed for deletion.

(COLR)

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, there will no longer be a need to establish core operating limits. Thus, this TS will not be applicable in the permanently defueled condition.

TS 5.6.6, Post Accident Monitoring Report TS 5.6.6 is proposed for deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize

Enclosure NL-19-026 Page 61 of 74 operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

TS 3.3.3 provides the operability requirements for the PAM instrumentation. It was proposed for deletion. Given that the reporting requirements in Conditions B and F of LCO 3.3.3 are proposed for deletion, the proposed deletion of the TS 5.6.6 reporting details is appropriate.

TS 5.6.7, Steam Generator Tube Inspection TS 5.6.7 is proposed for deletion.

Report After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the SG will no longer perform a function in the permanently shut down and defueled state.

TS 3.4.17 provides the requirements to ensure SG tube integrity in MODES 1 through 4. It is proposed for deletion. In addition, TS 5.5.7, Steam Generator (SG) Program, is proposed for deletion. Thus, the proposed deletion of this supporting TS program is appropriate.

APPENDIX B TO FACILITY OPERATING LICENSE Current Title Proposed Title APPENDIX B TO FACILITY OPERATING APPENDIX B TO FACILITY OPERATING LICENSE LICENSE Basis Appendix B is modified by replacing the reference to Facility Operating License with a reference to Facility License. This change reflects the upcoming change in status regarding IP2. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Current Section 1.0 Proposed Section 1.0 The Environmental Protection Plan (EPP) is to The Environmental Protection Plan (EPP) is to provide for protection of environmental values provide for protection of environmental values during construction and operation of the during construction and operationhandling and nuclear facility. The principal objectives of the storage of spent fuel and maintenance of the EPP are as follows: nuclear facility. The principal objectives of the EPP are as follows:

Enclosure NL-19-026 Page 62 of 74 (1) Verify that the plant is operated in an (1) Verify that the plant is operatedfacility is environmentally acceptable manner, as maintained in an environmentally established by the FES and other NRC acceptable manner, as established by the environmental impact assessments. FES and other NRC environmental impact assessments.

(2) Coordinate NRC requirements and maintain consistency with other Federal, State and (2) Coordinate NRC requirements and maintain local requirements for environmental consistency with other Federal, State and protection. local requirements for environmental protection.

(3) Keep NRC informed of the environmental effects of facility construction and operation (3) Keep NRC informed of the environmental and of actions taken to control those effects. effects of handling and storage of spent fuel and maintenance of the facility Environmental concerns identified in the FES construction and operation and of actions which relate to water quality matters are taken to control those effects.

regulated by way of the licensee's SPDES permit. Environmental concerns identified in the FES which relate to water quality matters are regulated by way of the licensee's SPDES permit.

Basis The proposed changes to Section 1.0 replace a reference to construction and operation with a reference to handling and storage of spent fuel and maintenance and a reference to plant is operated with facility is maintained, and a reference to facility construction and operation with handling and storage of spent fuel and maintenance of the facility. These proposed changes reflect the revised mission of the facility in the permanently shut down and defueled condition.

Current Section 3.1 Proposed Section 3.1 The licensee may make changes in station The licensee may make changes in design or operation or perform tests or stationfacility design or operation or perform experimentsChanges in plant design or tests or experimentsChanges in plantfacility operation or performance of tests or design or operation or performance of tests or experiments experiments A proposed change, test, or experiment A proposed change, test, or experiment shall(2) a significant change in effluents or shall(2) a significant change in effluents or power level; power level; Basis The proposed changes to Section 3.1 replace references to station and plant with references to facility. These proposed changes reflect the revised mission of the facility in the permanently shut down and defueled condition.

The proposed change to Section 3.1 to eliminate the reference to power level reflects the permanently shut down and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

Enclosure NL-19-026 Page 63 of 74 Current Section 3.3 Proposed Section 3.3 Changes in plant design or operation and Changes in plantfacility design or operation and Basis The proposed change to Section 3.3 replaces the reference to plant with a reference to facility.

This proposed change reflects the revised mission of the facility in the permanently shut down and defueled condition.

Current Section 4.1 Proposed Section 4.1 Any occurrence of an unusual or important Any occurrence of an unusual or important event that indicates or could result in significant event that indicates or could result in significant environmental impact causally related to plant environmental impact causally related to plant operation shall be recorded and operationthe handling and storage of spent fuel and maintenance of the facility shall be recorded and Basis The proposed change to Section 4.1 replaces the reference to plant operation with a reference to the handling and storage of spent fuel and maintenance of the facility. This proposed change reflects the revised mission of the facility in the permanently shut down and defueled condition.

Current Section 4.2 Proposed Section 4.2 The currently applicable Biological Opinion The currently applicable Biological Opinion concludes that continued operation of IP2 and concludes that continued operation of IP2 and IP3 is not likely to jeopardize the continued IP3 is not likely to jeopardize the continued existence of the listed species or to adversely existence of the listed species or to adversely affect the designated critical habitat of those affect the designated critical habitat of those species. species. This Biological Opinion conservatively bounds the conditions that will occur in the permanently shut down and defueled condition.

Basis The proposed change to Section 4.2 concludes that the Biological Opinion rendered during the evaluation of the continued operation of IP2 and IP3 conservatively bounds the conditions that will occur in the permanently shut down and defueled condition. This addition clarifies that the permanent shut down and defueling of IP2 will not impact the Biological Opinion regarding the shortnose sturgeon in an adverse manner when compared to the continued operation of IP2 and IP3.

Current Section 5.2 Proposed Section 5.2 Records and logs relative to the environmental Records and logs relative to the environmental aspects of plant operation shall be made and aspects of previous plant operation and the retained in a manner convenient for review and handling and storage of spent fuel and inspection. These records and logs shall be maintenance of the facility shall be made and made available to the NRC on request. retained in a manner convenient for review and

Enclosure NL-19-026 Page 64 of 74 inspection. These records and logs shall be Records of modifications to plant structures, made available to the NRC on request.

systems and components determined to potentially affect the continued protection of the Records of modifications to plantfacility environment shall be retained for the life of the structures, systems and components plant. All other records, data and logs relating determined to potentially affect the continued to this EPP shall be retained for five years or, protection of the environment shall be retained where applicable, in accordance with the for the life of the plantfacility. All other records, requirements of other agencies. data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.

Basis The proposed changes to Section 5.2 clarify that the reference to plant operation refers to plant operations previous to the permanent shut down, and includes a reference to the handling and storage of spent fuel and maintenance of the facility. In addition, references to plant are replaced with facility. These proposed changes reflect the revised mission of the facility in the permanently shut down and defueled condition.

Current Section 5.4.1 Proposed Section 5.4.1 (b) A list of all changes in station design or (b) A list of all changes in stationfacility design operation, tests, and experiments made in or operation, tests, and experiments made accordance with subsection 3.1 which in accordance with subsection 3.1 which involved a potentially significant unreviewed involved a potentially significant unreviewed environmental issue environmental issue Basis The proposed change to Section 5.4.1 replaces the reference to station with a reference to facility. This proposed change reflects the revised mission of the facility in the permanently shut down and defueled condition.

Current Section 5.4.2 Proposed Section 5.4.2 The report shall (1) describe, analyze, and The report shall (1) describe, analyze, and evaluate the event, including extent and evaluate the event, including extent and magnitude of the impact and plant operating magnitude of the impact and plant operating characteristics, (2) characteristicsfacility conditions, (2)

Basis The proposed change to Section 5.4.2 replaces a reference to plant operating characteristics with facility conditions. This proposed change reflects the revised mission of the facility in the permanently shut down and defueled condition.

Enclosure NL-19-026 Page 65 of 74 APPENDIX C TO FACILITY OPERATING LICENSE Current Title for Part 1 Proposed Title for Part 1 APPENDIX C TO FACILITY OPERATING APPENDIX C TO FACILITY OPERATING LICENSE LICENSE Current Header for Part 1 Proposed Header for Part 1 Facility Operating License Facility Operating License Current Title for Part 2 Proposed Title for Part 2 APPENDIX C TO FACILITY OPERATING APPENDIX C TO FACILITY OPERATING LICENSE LICENSE Basis Appendix C is modified by replacing the references to Facility Operating License with Facility License. These changes reflect the upcoming change in status regarding IP2. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2).

The mark-ups of the Appendix A TSs Bases and re-typed versions of the PDTS Bases are provided for information only. Upon approval of this amendment, changes to the Appendix A TSs Bases will be incorporated in accordance with TS 5.5.12, Technical Specifications (TS) Bases Control Program.

3.1 APPLICABLE REGULATORY REQUIREMENT/CRITERIA 10 CFR 50.82, Termination of License The 10 CFR 50.82(a)(1) paragraph requires that when a licensee has determined to permanently cease operations the licensee shall, within 30 days, submit a written certification to the NRC, consistent with the requirements of 10 CFR 50.4(b)(8), and once fuel has been permanently removed from the reactor vessel, the licensee shall submit a written certification to the NRC that meets the requirements of 10 CFR 50.4(b)(9). On February 8, 2017, Entergy notified the NRC that IP2 would permanently cease operations no later than April 30, 2020 (Reference 1). Entergy recognizes that approval of these proposed changes is contingent upon the submittal of the certifications required by 10 CFR 50.82(a)(1).

The 10 CFR 50.82(a)(2) paragraph states: Upon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, or when a final legally effective order to permanently cease operations has come into effect, the 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel."

10 CFR 50.36, Technical Specifications In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TSs. In doing so, the Commission placed emphasis on those matters related to

Enclosure NL-19-026 Page 66 of 74 the prevention of accidents and mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TSs those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity. (Statement of Consideration, Technical Specification for Facility Licenses; Safety Analysis Reports, 33 FR 18610 (December 17, 1968))

Pursuant to 10 CFR 50.36, TSs are required to include items in the following five categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) LCOs; (3) SRs; (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a facilities' TSs.

These criteria, which were subsequently codified in changes to Section 36 of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.36) (60 FR 36953), also pertain to the TS requirements for safe storage of spent fuel. A general discussion of these considerations is provided below to address the existing LCOs.

Criterion 1 of 10 CFR 50.36(c)(2)(ii)(A) states that TS LCOs must be established for installed instrumentation that is used to detect, and indicate in the Control Room, a significant abnormal degradation of the reactor coolant pressure boundary. Since no fuel will be present in the reactor or RCS at IP2 in the permanently shut down and defueled condition, this criterion is not applicable.

Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) states that TS LCOs must be established for a process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The purpose of this criterion is to capture those process variables that have initial values assumed in the DBA and transient analyses, and which are monitored and controlled during power operation. While this criterion was developed for operating reactors, there are some DBAs which continue to apply to a facility authorized only to handle, store, and possess nuclear fuel. The scope of DBAs applicable to a facility with a reactor that is permanently shut down and defueled is markedly reduced from those postulated for an operating plant. The applicable DBAs for IP2 in the permanently defueled condition are discussed in more detail within this license amendment request.

Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) states that TS LCOs must be established for SSCs that are part of the primary success path and which function or actuate to mitigate a DBA or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The intent of this criterion is to capture into the TSs only those SSCs that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion), so that the plant response to DBAs and transients limits the consequences of these events to within the appropriate acceptance criteria. While there are no transients that will continue to apply to IP2, there are still DBAs that will continue to apply to a facility authorized only to handle, store, and possess nuclear fuel. The scope of DBAs applicable to a facility with a reactor that is permanently shut down and defueled is markedly reduced from those postulated for an operating plant. The scope of DBAs that will be applicable to IP2 is discussed in more detail within this license amendment request.

Enclosure NL-19-026 Page 67 of 74 Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) states that TS LCOs must be established for SSCs that operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The intent of this criterion is that risk insights and operating experience be factored into the establishment of TS LCOs. All of the accident sequences that previously dominated risk at IP2 will no longer be applicable after the reactor is in the permanently shut down and defueled condition.

Addressing administrative controls, 10 CFR 50.36(c)(5) states that they ...are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner. This license amendment request is proposing changes to the Administrative Controls section, with conforming changes proposed to additional sections, consistent with the pending decommissioning status of the plant. This request applies the principles identified in 10 CFR 50.36(c)(6), Decommissioning, for a facility which has submitted certifications required by 50.82(a)(1) and proposes changes to the Administrative Controls appropriate for the IP2 permanently defueled condition. As 10 CFR 50.36(c)(6) states, this type of change should be considered on a case-by-case basis.

The 10 CFR 50.36(c)(6), Decommissioning, provisions apply only to nuclear power reactor facilities that have submitted the certifications required by 10 CFR 50.82(a)(1). For such facilities, TSs involving safety limits, limiting safety system settings, and limiting control system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls will be developed on a case-by-case basis.

This proposed amendment deletes the portions of the previous IP2 TSs that are no longer applicable to a permanently defueled facility while modifying the remaining portions to correspond to the permanently shut down and defueled condition.

10 CFR 50.48(f), Fire Protection During Decommissioning The 10 CFR 50.48(f) paragraph states, in part, that: Licensees that have submitted the certifications required under 10 CFR 50.82(a)(1) shall maintain a fire protection program to address the potential for fires that could cause the release or spread of radioactive materials (i.e., that could result in a radiological hazard)...

(1) The objectives of the fire protection program are to (i) Reasonably prevent these fires from occurring; (ii) Rapidly detect, control, and extinguish those fires that do occur and that could result in a radiological hazard; and (iii) Ensure that the risk of fire-induced radiological hazards to the public environment and plant personnel is minimized.

(2) The licensee shall assess the fire protection program on a regular basis. The licensee shall revise the plan as appropriate throughout the various stages of facility decommissioning.

(3) The licensee may make changes to the fire protection program without NRC approval if these changes do not reduce the effectiveness of fire protection for facilities, systems,

Enclosure NL-19-026 Page 68 of 74 and equipment that could result in a radiological hazard, taking into account the decommissioning plant conditions and activities.

10 CFR 50.51, Continuation of License The 10 CFR 50.51(b) paragraph states: Each license for a facility that has permanently ceased operations, continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the Commission notifies the licensee in writing that the license is terminated. During such period of continued effectiveness the licensee shall:

(1) Take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition, and (2) Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC regulations and the provisions of the specific 10 CFR part 50 license for the facility.

10 CFR 50, Appendix A, General Design Criteria (GDC) for Nuclear Power Plants Section 1.3 of the IP2 UFSAR states:

The detailed results of the evaluation of Indian Point Unit 2 compliance with the then current General Design Criteria established by the Nuclear Regulatory Commission (NRC) in 10 CFR 50 Appendix A, were submitted to the NRC by Con Edison on August 11, 1980 [Reference 6]. Commission concurrence was received on January 19, 1982.

IP2 design and licensing basis for fuel storage and handling and radiological controls is detailed in the UFSAR and other plant-specific licensing basis documents.

10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors The 10 CFR 50.46(a)(1)(i) paragraph states: This section does not apply to a nuclear power reactor facility for which the certifications required under 10 CFR 50.82(a)(1) have been submitted.

10 CFR 50.62, Requirements for Reduction of Risk from Anticipated Transients Without Scram (ATWS) Events for Light-Water-Cooled Nuclear Power Plants The 10 CFR 50.62(a) paragraph states: The requirements of this section apply to all commercial light-water-cooled nuclear power plants, other than nuclear power reactor facilities for which the certifications required under § 50.82(a)(1) have been submitted.

Enclosure NL-19-026 Page 69 of 74 Design Basis Accidents (DBAs)

Section 14 of the IP2 UFSAR describes the DBA scenarios that are applicable during plant operations. After certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel are submitted to the NRC in accordance with 10 CFR 50.82(a)(1)(i) and (ii) and they are docketed for IP2, the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). With the reactor in a permanently shut down and defueled condition, the facility mission changes. The primary mission is now the safe storage and handling of irradiated fuel. In this condition, the spectrum of credible accidents is much smaller than for an operational plant. Therefore, most of the accident scenarios postulated in UFSAR Section 14 will no longer be applicable after IP2 is in the permanently defueled condition. The only remaining DBAs will be the FHA in the Fuel Handling Building and the accidental release of waste liquid or waste gas. This license amendment request includes additional discussion regarding the analyses of these accidents.

3.2 NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Pursuant to 10 CFR 50.92, Entergy has reviewed the proposed changes and concludes that the changes do not involve a significant hazards consideration since the proposed changes satisfy the criteria in 10 CFR 50.92(c). These criteria require that operation of the facility in accordance with the proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; (2) create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in a margin of safety.

On February 8, 2017, Entergy notified the U.S. Nuclear Regulatory Commission (NRC) that it would permanently cease power operations at Indian Point Generating Station, Unit No. 2 (IP2) no later than April 30, 2020 (Reference 1). After the certifications for permanent cessation of operations and permanent fuel removal from the reactor vessel are docketed for IP2, the 10 CFR Part 50 license for IP2 will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, in accordance with 10 CFR 50.82(a)(2).

This proposed license amendment would revise the IP2 Operating License (OL), revise the Technical Specifications (TSs) in Appendix A of the OL to Permanently Defueled Technical Specifications (PDTS), and revise the Environmental Technical Specification Requirements in Appendix B of the OL, and the Inter-Unit Fuel Transfer Technical Specifications in Appendix C. The proposed changes are consistent with the permanent cessation of reactor operation and permanent defueling of the reactor. The proposed changes would revise certain requirements contained within the IP2 OL and TSs and remove the requirements that would no longer be applicable after IP2 is permanently shut down and defueled.

The existing IP2 Appendix A TSs contain Limiting Conditions for Operation (LCOs) that provide for appropriate functional capability of equipment required for safe operation of the facility, including the plant being in a defueled condition. Since the safety function related to safe storage and management of irradiated fuel at an operating plant is similar to the corresponding function at a permanently defueled facility, the existing Appendix A TSs provide an appropriate level of control. However, the majority of the existing TSs are only applicable with the reactor in an operational mode. LCOs and associated Surveillance Requirements (SRs) that will not apply in the permanently defueled condition are being

Enclosure NL-19-026 Page 70 of 74 proposed for deletion. The remaining portions of the Appendix A TSs are being proposed for revision and incorporation as the PDTS to provide a continuing acceptable level of safety which addresses the reduced scope of postulated design basis accidents (DBAs) associated with a defueled facility.

The discussion below addresses each 10 CFR 50.92(c) no significant hazards consideration criterion and demonstrates that the proposed amendment does not constitute a significant hazard.

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 not take effect until IP2 has permanently ceased operation, entered a permanently defueled condition, met the decay requirements established in the analysis of the Fuel Handling Accident (FHA),

implemented NRC approved License Amendments regarding fuel storage requirements and administrative controls for the permanently defueled condition, and received NRC approval of the Certified Fuel Handler Training and Retraining Program. The proposed amendment would modify the IP2 OL and TSs in Appendices A through C by deleting the portions of the OL and TSs that are no longer applicable to a permanently defueled facility, while modifying other portions to correspond to the permanently defueled condition.

These proposed changes are consistent with the criteria set forth in 10 CFR 50.36 for the contents of TSs.

Section 14 of the IP2 Updated Final Safety Analysis Report (UFSAR) describes the DBA and transient scenarios applicable to IP2 during power operations. After the reactor is in a permanently defueled condition, the spent fuel pit (SFP) and its cooling systems will be dedicated only to spent fuel storage. In this condition, the spectrum of credible accidents will be much smaller than for an operational plant. After the certifications are docketed for IP2 in accordance with 10 CFR 50.82(a)(1), and the consequent removal of authorization to operate the reactor or to place or retain fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2), the majority of the accident scenarios previously postulated in the UFSAR will no longer be possible and will be removed from the UFSAR under the provisions of 10 CFR 50.59.

The deletion of TS definitions and rules of usage and application requirements that will not be applicable in a defueled condition has no impact on facility structures, systems, and components (SSCs) or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shut down and defueled status of IP2 has no impact on the remaining applicable DBAs.

The removal of LCOs or SRs that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents do not affect the applicable DBAs previously evaluated since these DBAs are no longer applicable in the permanently defueled condition. The safety functions involving core reactivity control,

Enclosure NL-19-026 Page 71 of 74 reactor heat removal, reactor coolant system (RCS) inventory control, and containment integrity are no longer applicable at IP2 as a permanently shut down and defueled facility. The analyzed accidents involving damage to the RCS, main steam lines, reactor core, and the subsequent release of radioactive material will no longer be possible at IP2.

After IP2 permanently ceases operation, the future generation of fission products will cease and the remaining source term will decay. The radioactive decay of the irradiated fuel following shut down of the reactor will have reduced the consequences of the FHA below those previously analyzed.

The SFP water level, boron concentration, and fuel storage TSs are retained to preserve the current requirements for safe storage of irradiated fuel. SFP cooling and make-up related equipment and support equipment (e.g., electrical power systems) are not required to be continuously available since there will be sufficient time to effect repairs, establish alternate sources of make-up flow, or establish alternate sources of cooling in the event of a loss of cooling and make-up flow to the SFP.

The deletion and modification of provisions of the administrative controls of the Appendix A TSs and the non-radiological environmental protection requirements in Appendix B do not directly affect the design of SSCs necessary for safe storage of irradiated fuel or the methods used for handling and storage of such fuel in the SFP. The changes do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shut down and defueled condition of the reactor.

The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses.

Additionally, the occurrence of postulated accidents associated with reactor operation will no longer be credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents.

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

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 changes to the IP2 OL and Appendices A through C TSs have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The removal of TSs that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents, cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor will be permanently shut down and defueled and IP2 will no longer be authorized to operate the reactor.

Enclosure NL-19-026 Page 72 of 74 The proposed deletion and modification of requirements of the IP2 OL and Appendices A through C TSs do not affect systems credited in the accidents that remain applicable at IP2 in the permanently defueled condition. The proposed OL and TSs will continue to require proper control and monitoring of safety significant parameters and activities.

The Appendix A TSs regarding SFP water level, boron concentration, and fuel storage are retained to preserve the current requirements for safe storage of irradiated fuel. The restriction on the SFP water level is fulfilled by normal operating conditions and preserves initial conditions assumed in the analyses of the postulated DBA.

The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (fuel cladding and spent fuel cooling). Since extended operation in a defueled condition will be the only operation allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident.

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

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

Response: No.

Because the 10 CFR Part 50 license for IP2 will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP2 as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation are no longer credible. The only remaining credible accidents are the FHA and the accidental release of waste liquids or waste gas. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact the remaining DBAs.

The proposed amendment would modify the IP2 OL and TSs in Appendices A through C by deleting the portions of the OL and TSs that are no longer applicable to a permanently defueled facility, while modifying other portions to correspond to the permanently defueled condition. The requirements that are proposed to be deleted from the IP2 OL and Appendix A TSs are not credited in the existing accident analyses for the remaining DBAs; and as such, do not contribute to the margin of safety associated with the accident analyses.

Postulated DBAs involving the reactors will no longer be possible because the reactor will be permanently shut down and defueled and IP2 will no longer be authorized to operate the reactor.

The Appendix A TSs regarding SFP water level, boron concentration, and fuel storage are retained to preserve the current requirements for safe storage of irradiated fuel.

Enclosure NL-19-026 Page 73 of 74 Therefore, the proposed amendment does not involve a significant reduction in the margin of safety.

Based on the above, Entergy concludes that the proposed amendment 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.

3.3 PRECEDENT The proposed changes to the IP2 OL and Appendices A through C TSs are consistent with the intent of the license and accompanying PDTS issued to facilities that have been permanently shut down and defueled: (1) Fort Calhoun Station, for which an amendment was issued on March 6, 2018 (Reference 7); (2) Oyster Creek Nuclear Generating Station, for which an amendment was issued on October 26, 2018 (Reference 8); (3) San Onofre Nuclear Generating Station, Units 2 and 3, for which an amendment was issued on July 17, 2015 (Reference 9); (4) Crystal River Nuclear Plant, Unit 3, for which an amendment was issued on September 4, 2015 (Reference 10).

3.4 CONCLUSION

Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

4. ENVIRONMENTAL CONSIDERATIONS This amendment request meets the eligibility criteria for categorical exclusion from environmental review set forth in 10 CFR 51.22(c)(9) as follows:

(i) The amendment involves no significant hazards consideration.

As described in Section 3.2 of this evaluation, the proposed amendment involves no significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite.

The proposed amendment does not involve any physical alterations to the facility configuration that could lead to a change in the type or amount of effluent release offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

The proposed amendment does not involve a significant increase individual or cumulative occupational radiation exposure.

Based on the above, Entergy concludes that the proposed change meets the eligibility criteria for categorical exclusion as set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

Enclosure NL-19-026 Page 74 of 74

5. REFERENCES

1. Letter, Entergy Nuclear Operations, Inc., to NRC, Notification of Permanent Cessation of Power Operations, dated February 8, 2017 (Letter NL-17-021) (ADAMS Accession No. ML17044A004)

2. Letter, Entergy Nuclear Operations, Inc., to NRC, Indian Point Nuclear Generating Unit No. 2, Proposed License Amendment Regarding Spent Fuel Storage, dated December 11, 2017 (Letter Number NL-17-144) (ADAMS Accession No. ML17354A014)

3. Letter, Entergy Nuclear Operations, Inc., to NRC, Technical Specifications Proposed Change - Administrative Controls for Permanently Defueled Condition," dated April 15, 2019 (Letter Number: NL-19-013)

4. Letter, Entergy Nuclear Operations, Inc., to NRC, Request for Approval of a Certified Fuel Handler Training and Retraining Program," dated April 15, 2019 (Letter Number:

NL-19-012)

5. Letter, NRC to Entergy Nuclear Operations, Inc, Indian Point Nuclear Generating Unit No. 2 - Re: Issuance of Amendment Affecting Containment Air Filtration, Control Room Air Filtration, and Containment Integrity during Fuel Handling Operations (TAC No. MA6955),

dated July 27, 2000 (ADAMS Accession No. ML003727636)

6. Letter from P. Zarakas, Con Edison, to H. Denton, NRC,

Subject:

Actions Taken to Comply with NRC Confirmatory Order of February 11, 1980, dated August 11, 1980

7. Letter, NRC to Omaha Public Power District, Fort Calhoun Station, Unit 1 - Issuance of Amendment Re: Revised Technical Specifications to Align to Those Requirements for Decommissioning (CAC No. MF9567), dated March 6, 2018 (ADAMS Accession No. ML18010A087)

8. Letter, NRC to Exelon Nuclear, Oyster Creek Nuclear Generating Station - Issuance of Amendment Re: License Amendment Request for Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition (EPID L-2017-LLA-0395), dated October 26, 2018 (ADAMS Accession No. ML18227A338)

9. Letter, NRC to Southern California Edison Company, San Onofre Nuclear Generating Station, Units 2 and 3 - Issuance of Amendment for Permanently Shutdown and Defueled Operating License and Technical Specifications (TAC Nos. MF3774 and MF3775), dated July 17, 2015 (ADAMS Accession No. ML15139A390)

10. Letter, NRC to Crystal River Nuclear Plant, Crystal River Unit 3 Nuclear Generating Plant -

Issuance of Amendment for Permanently Shutdown and Defueled Operating License and Technical Specifications (TAC No. MF3089), dated September 4, 2015 (ADAMS Accession No. ML15224B286)

Enclosure, Attachment 1 NL-19-026 Indian Point Nuclear Generating Station Unit 2 Mark-up of the Current Facility Operating License, Appendices A through C Technical Specifications, and Appendix A Technical Specifications Bases

ENTERGY NUCLEAR INDIAN POINT 2, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-247 INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 RENEWED FACILITY OPERATING LICENSE Renewed License No. DPR-26

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

A. The application for a renewed license filed by Entergy Nuclear Indian Point 2, LLC (ENIP2) (the licensee) and Entergy Nuclear Operations, Inc. (ENO) (operator), for Indian Point Nuclear Generating Unit No. 2 at the Indian Point Energy Center (IPEC) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 Deleted per CFR Chapter I; Amendment

[###]; B. Construction of the Indian Point Nuclear Generating Unit No. 2 (IP2 or facility) has been substantially completed in conformity with provisional Construction Permit No.

CPPR-21, as amended, and the application, as amended, the provisions of the Act and the rules and regulations of the Commission; be maintained C. The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; D. There is reasonable assurance: (i) that the activities authorized by this renewed operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; E. ENO is technically and financially qualified and ENIP2 is financially qualified to engage in the activities authorized by this renewed license in accordance with the rules and regulations of the Commission; F. ENIP2 and ENO have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; G. The issuance of this renewed license will not be inimical to the common defense and security or to the health and safety of the public;

H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental costs and considering available alternatives, the issuance of this renewed Facility Operating License No. DPR-26, subject to the conditions for the protection of the environment set forth herein, is in accordance with 10 CFR Part 51, Appendix B, of the Commission's regulations and all applicable requirements of said Appendix B have been satisfied; I. The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this renewed license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70, including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31; and J. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1); and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by this renewed license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations.

2. Renewed Facility Operating License No. DPR-26 is hereby issued to ENIP2 and ENO to read as follows: Defueled maintained A. This renewed license applies to the Indian Point Nuclear Generating Unit No. 2, a pressurized water nuclear reactor and associated equipment (the facility), which is owned by ENIP2 and operated by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the "Final Facility Description and Safety Analysis Report", as supplemented and amended, and the Environmental Report, as amended.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses: ,

and (1) Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities", (a) ENIP2 to possess and use, and (b) ENO to possess, use and operate, the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; Defueled that was used (2) ENO pursuant to the Act and 10 CFR Part 70, to receive, Amdt. 75 possess, and use, at any time special nuclear material as 1-11-82 reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Facility Description and Safety Analysis Report, as supplemented and amended and as described in the Commission's authorization through Amendment No. 75 to this license.

that were used calibration of that were used (3) ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to Amdt. 42 receive, possess and use, at any time any byproduct, source 10-17-78 are used in the and special nuclear material as sealed neutron sources for calibration of reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to Amdt. 42 receive, possess, and use in amounts as required any 10-17-78 byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to Amdt. 220 possess, but not separate, such byproduct and special 09-06-01 nuclear materials as may be produced by the operation of the facility. that were C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Level Deleted per ENO is authorized to operate the facility at steady state Amdt. 241 Amendment reactor core power levels not in excess of 3216 megawatts 10-27-04

[###] thermal (2) Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 288, are hereby incorporated in the renewed license. ENO shall operate the facility in accordance with the Technical Specifications. maintain (3) The following conditions relate to the amendment approving the conversion Deleted per to Improved Standard Technical Specifications:

Amendment

[###] 1. This amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee controlled documents as described in Table R, "Relocated Technical Specifications from the CTS," and Table LA, "Removed Details and Less Restrictive Administrative Changes to the CTS" attached to the NRC staff's Safety Evaluation enclosed with this amendment. The relocation of requirements and detailed information shall be completed on or before the implementation of this amendment.

2. The following is a schedule for implementing surveillance requirements (SRs):

For SRs that are new in this amendment, the first performance is due at the end of the first surveillance interval that begins on the date of implementation of this amendment.

For SRs that existed prior to this amendment whose intervals of performance are being reduced, the first reduced surveillance interval begins upon completion of the first surveillance performed after the date of implementation of this amendment.

For SRs that existed prior to this amendment that have modified acceptance criteria, the first performance is due at the end of the first surveillance interval that began on the date the surveillance was last performed prior to the date of implementation of this amendment.

For SRs that existed prior to this amendment whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to the date of implementation of this amendment.

D. (1) Deleted per Amdt. 82, 12-11-82.

(2) Deleted per Amendment 238.

E. Deleted per Amdt. 71, dated 8-5-81, effective 5-14-81.

F. This renewed license is also subject to appropriate conditions by the New York State Department of Environmental Conservation in its letter granting a Section 401 certification under the Federal Water Pollution Control Act amendments of 1972.

G. Pursuant to Section 50.60 of 10 CFR Part 50, paragraph 4 of Provisional Construction Permit No. CPPR-21 allocating quantities of special nuclear material, Deleted per together with the related estimated schedules contained in Appendix A attached to Amendment [###] said provisional construction permit, shall remain in effect.

H. ENO shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1 for the Indian Point Energy Center, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Physical Security, Training and Qualification, and Safeguards Contingency Plan, Revision 0," and was submitted by letter dated October 14, 2004, as supplemented by letter dated May 18, 2006.

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

ENO shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The ENO CSP was approved by License Amendment No. 266, as supplemented by changes approved by License Amendment Nos. 279, 284, and 286.

ENO has been granted Commission authorization to use "stand alone preemption authority" under Section 161A of the Atomic Energy Act, 42 U.S.C. 2201a with respect to the weapons described in Section II supplemented with Section Ill of Attachment 1 to its application submitted by letter dated August 20, 2013, as supplemented by letters dated November 21, 2013, and July 24, 2014, and citing letters dated April 27, 2011, and January 4, 2012. ENO shall fully implement and maintain in effect the provisions of the Commission-approved authorization.

I. Deleted per Amdt. 133, 7-6-88.

J. Deleted per Amdt. 133, 7-6-88.

K. ENO shall implement and maintain in effect all provisions of the NRC-approved fire Deleted per protection program as described in the Updated Final Safety Analysis Report for the Amendment facility and as approved in Safety Evaluations Reports dated November 30, 1977,

[###] February 3, 1978, January 31, 1979, October 31, 1980, August 22, 1983, March 30, 1984, October 16, 1984, September 16, 1985, November 13, 1985, March 4, 1987, January 12, 1989, and March 26, 1996. ENO may make changes to the NRC-approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

L. Deleted per Amendment 238 M. Deleted per Amendment 238 N. Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy

(c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders O. Control Room Envelope Habitability Deleted per Upon implementation of Amendment No. 258 adopting TSTF-448, Revision 3 (as Amendment supplemented), the determination of control room envelope (CRE) unfiltered air

[###] inleakage as required by Technical Specification (TS) Surveillance Requirement (SR) 3.7.10.4, in accordance with TS 5.5.16.c.(i), the assessment of CRE habitability as required by TS 5.5.16.c.(ii), and the measurement of CRE pressure as required by TS 5.5.16.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.10.4, in accordance with TS 5.5.16.c.(i),

shall be within the next 18 months since the time period since the most recent successful tracer gas test is greater than 6 years.

(b) The first performance of the periodic assessment of CRE habitability, TS 5.5.16.c.(ii), shall be within the next 9 months since the time period since the most recent successful tracer gas test is greater than 3 years.

(c) The first performance of the periodic measurement of CRE pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from January 4, 2007, the date of the most recent successful pressure measurement test.

P. ENO may transfer IP3 spent fuel to the IP2 spent fuel pit subject to the conditions listed in Appendix C. ENO is further authorized to transfer IP3 spent fuel into NRC approved storage casks for onsite storage by ENO and Entergy Nuclear Indian Point 3, LLC.

Q. License Renewal License Conditions (1) The information in the UFSAR supplement, submitted pursuant to 10 CFR 54.21(d) and as revised during the license renewal application review process, and licensee commitments as listed in Appendix A of the Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Units 2 and 3, (SER) and supplements to the SER, are collectively the License Renewal UFSAR Supplement. The UFSAR Supplement is henceforth part of the UFSAR, which will be updated in accordance with 10 CFR 50.71(e). As such, the licensee may make changes to the programs, activities, and commitments described in the UFSAR Supplement, provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, Changes, Tests, and Experiments, and otherwise complies with the requirements in that section.

(2) The License Renewal UFSAR Supplement, as defined in license condition Q(1) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).

6. This renewed license is effective as of the date of issuance, and shall expire at midnight April 30, 2024.

FOR THE NUCLEAR REGULATORY COMMISSION until the Commission notifies the

/RA/ licensee in writing that the license is terminated PERMANENTLY Ho K. Nieh, Director DEFUELED Office of Nuclear Reactor Regulation Attachments:

Appendix A - Technical Specifications Appendix B - Environmental Technical Specification Requirements Appendix C - Inter-Unit Fuel Transfer Technical Specifications Date of Issuance: September 17, 2018 To Be Determined Amendment No. 289

APPENDIX A TO FACILITY OPERATING LICENSE DPR-26 FOR ENTERGY NUCLFAR INDIAN romr 2, LLC AND ElffERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING PLANT UNIT NO. 2 DOCKET NO. 50-247 TECHNICAL SPECIFICATIONS AND BASES PERMANENTLY DEFUELED

-~

Amendment No- ~

Facility Operating License No. DPR-26 Appendix A - Technical Specifications TABLE OF CONTENTS 1.0 USE AND APPLICATION PERMANENTLY 1.1 Definitions DEFUELED 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency DELETED 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.2 Reactor Coolant System Pressure SL 2.2 Safety Limit Violations 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) 3.1.2 Core Reactivity 3.1.3 Moderator Temperature Coefficient (MTC) 3.1.4 Rod Group Alignment Limits 3.1.5 Shutdown Bank Insertion Limits 3.1.6 Control Bank Insertion Limits 3.1.7 Rod Position Indication 3.1.8 PHYSICS TESTS Exceptions - MODE 2 3.2 POWER DISTRIBUTION LIMITS 3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNH) 3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Constant Axial Offset Control (CAOC) Methodology) 3.2.4 QUADRANT POWER TILT RATIO (QPTR) 3.3 INSTRUMENTATION 3.3.1 Reactor Protection System (RPS) Instrumentation 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation 3.3.3 Post Accident Monitoring (PAM) Instrumentation 3.3.4 Remote Shutdown 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation 3.3.6 Containment Purge System and Pressure Relief Line Isolation Instrumentation 3.3.7 Control Room Ventilation System (CRVS) Actuation Instrumentation Indian Point 2 i Amendment No. 251

Facility Operating License No. DPR-26 Appendix A - Technical Specifications PERMANENTLY TABLE OF CONTENTS DEFUELED 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits 3.4.2 RCS Minimum Temperature for Criticality 3.4.3 RCS Pressure and Temperature (P/T) Limits 3.4.4 RCS Loops - MODES 1 and 2 3.4.5 RCS Loops - MODE 3 3.4.6 RCS Loops - MODE 4 3.4.7 RCS Loops - MODE 5, Loops Filled 3.4.8 RCS Loops - MODE 5, Loops Not Filled 3.4.9 Pressurizer 3.4.10 Pressurizer Safety Valves 3.4.11 Pressurizer Power Operated Relief Valves (PORVs) 3.4.12 Low Temperature Overpressure Protection (LTOP) 3.4.13 RCS Operational LEAKAGE 3.4.14 RCS Pressure Isolation Valve (PIV) Leakage 3.4.15 RCS Leakage Detection Instrumentation 3.4.16 RCS Specific Activity 3.4.17 Steam Generator (SG) Tube Integrity 3.5 EMERGENCY CORE COOLING SYSTEM (ECCS) 3.5.1 Accumulators 3.5.2 ECCS - Operating 3.5.3 ECCS - Shutdown 3.5.4 Refueling Water Storage Tank (RWST) 3.6 CONTAINMENT SYSTEMS 3.6.1 Containment 3.6.2 Containment Air Locks 3.6.3 Containment Isolation Valves 3.6.4 Containment Pressure 3.6.5 Containment Air Temperature 3.6.6 Containment Spray System and Containment Fan Cooler Unit (FCU)

System 3.6.7 Recirculation pH Control System 3.6.8 Not Used 3.6.9 Isolation Valve Seal Water (IVSW) System 3.6.10 Weld Channel and Penetration Pressurization System (WC&PPS)

Indian Point 2 ii Amendment No. 251

SPENT FUEL PIT Facility Operating License No. DPR-26 REQUIREMENTS Appendix A - Technical Specifications TABLE OF CONTENTS PERMANENTLY DEFUELED 3.7 PLANT SYSTEMS 3.7.1 Main Steam Safety Valves (MSSVs) 3.7.2 Main Steam Isolation Valves (MSIVs) and Main Steam Check Valves (MSCVs) 3.7.3 Main Feedwater Isolation 3.7.4 Atmospheric Dump Valves (ADVs) 3.7.5 Auxiliary Feedwater (AFW) System 3.7.6 Condensate Storage Tank (CST) 3.7.7 Component Cooling Water (CCW) System 3.7.8 Service Water System (SWS) 3.7.9 Ultimate Heat Sink (UHS) 3.7.10 Control Room Ventilation System (CRVS) 3.7.11 Spent Fuel Pit Water Level 3.7.12 Spent Fuel Pit Boron Concentration 3.7.13 Spent Fuel Pit Storage 3.7.14 Secondary Specific Activity 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating 3.8.2 AC Sources - Shutdown 3.8.3 Diesel Fuel Oil and Starting Air 3.8.4 DC Sources - Operating 3.8.5 DC Sources - Shutdown 3.8.6 Battery Parameters 3.8.7 Inverters - Operating 3.8.8 Inverters - Shutdown 3.8.9 Distribution Systems - Operating 3.8.10 Distribution Systems - Shutdown 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration 3.9.2 Nuclear Instrumentation 3.9.3 Containment Penetrations 3.9.4 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level 3.9.6 Refueling Cavity Water Level 4.0 DESIGN FEATURES 4.1 Site Location 4.2 Deleted 4.3 Fuel Storage Indian Point 2 iii Amendment No.

Facility Operating License No. DPR-26 Appendix A - Technical Specifications TABLE OF CONTENTS PERMANENTLY DEFUELED 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.2.1 Onsite and Offsite Organizations 5.2.2 Facility Staff 5.3 Facility Staff Qualifications 5.4 Procedures 5.5 Programs And Manuals 5.5.1 Deleted Offsite Dose Calculation Manual (ODCM) 5.5.2 Primary Coolant Sources Outside Containment 5.5.3 Deleted Radioactive Effluent Controls Program 5.5.4 Deleted Component Cyclic or Transient Limit 5.5.5 Reactor Coolant Pump Flywheel Inspection Program Deleted 5.5.6 Inservice Testing Program 5.5.7 Deleted Steam Generator (SG) Program 5.5.8 Deleted Secondary Water Chemistry Program 5.5.9 Deleted Ventilation Filter Testing Program (VFTP) 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5.11 Deleted Diesel Fuel Oil Testing Program 5.5.12 Technical Specification (TS) Bases Control Program 5.5.13 Safety Function Determination Program (SFDP) 5.5.14 Containment Leakage Rate Testing Program 5.5.15 Battery Monitoring and Maintenance Program 5.5.16 Control Room Envelope Habitability Program 5.6 Reporting Requirements 5.6.1 Not Used 5.6.2 Annual Radiological Environmental Operating Report 5.6.3 Radioactive Effluent Release Report 5.6.4 Not Used 5.6.5 CORE OPERATING LIMITS REPORT (COLR) 5.6.6 Post Accident Monitoring Report 5.6.7 Steam Generator Tube Inspection Report 5.7 High Radiation Area Indian Point 2 iv Amendment No.

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions

- NOTE -

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.

AXIAL FLUX DIFFERENCE AFD shall be the difference in normalized flux signals between (AFD) the top and bottom halves of a two section excore neutron detector.

CERTIFIED FUEL HANDLER A CERTIFIED FUEL HANDLER is an individual who complies with the provisions of the CERTIFIED FUEL HANDLER Training and Retraining Program required by TS 5.3.2.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status INDIAN POINT 2 1.1 - 1 Amendment No.

Definitions 1.1 1.1 Definitions derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL A COT shall be the injection of a simulated or actual signal into TEST (COT) the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints required for channel OPERABILITY such that the setpoints are within the necessary range and accuracy. The COT may be performed by means of any series of sequential, overlapping, or total channel steps.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS The COLR is the unit specific document that provides cycle REPORT (COLR) specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present. If a specific isotope is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT I-131 shall be performed using Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988.

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE Equivalent XE-133 shall be performed using effective dose conversion factors for air submersion listed INDIAN POINT 2 1.1 - 2 Amendment No.

Definitions 1.1 1.1 Definitions in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank,

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE, or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);

b. Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE, and

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each required master relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required master relay. The MASTER RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

INDIAN POINT 2 1.1 - 3 Amendment No.

Definitions 1.1 1.1 Definitions NON-CERTIFIED OPERATOR A NON-CERTIFIED OPERATOR is a non-licensed operator who complies with the qualification requirements of Specification 5.3.1, but is not a CERTIFIED FUEL HANDLER.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in UFSAR Chapter 13, Tests and Operations,

b. Authorized under the provisions of 10 CFR 50.59, or

c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore detector RATIO (QPTR) calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the reactor (RTP) coolant of 3216 MWt.

INDIAN POINT 2 1.1 - 4 Amendment No.

Definitions 1.1 1.1 Definitions SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM, and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each required slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay.

The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device and OPERATIONAL TEST verifying the OPERABILITY of all devices in the channel required (TADOT) for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping, or total channel steps.

Indian Point 2 1.1 - 5 Amendment No. 238

Definitions Table 1.1-1 (page 1 of 1)

MODES AVERAGE REACTIVITY % RATED REACTOR COOLANT MODE TITLE CONDITION THERMAL (a) TEMPERATURE (keff) POWER

(°F) 1 Power Operation 0.99 >5 NA 2 Startup 0.99 5 NA 3 Hot Standby < 0.99 NA 350 (b) 4 Hot Shutdown < 0.99 NA 350 > Tavg > 200 (b) 5 Cold Shutdown < 0.99 NA 200 (c) 6 Refueling NA NA NA (a) Excluding decay heat.

(b) All reactor vessel head closure bolts fully tensioned.

(c) One or more reactor vessel head closure bolts less than fully tensioned.

INDIAN POINT 2 1.1 - 6 Amendment No. 238

Logical Connectors 1.2 1.0 USE AND APPLICATION 1.2 Logical Connectors PURPOSE The purpose of this section is to explain the meaning of logical connectors.

and Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies. The only logical connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings.

BACKGROUND Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action). The successive levels of logic are identified by additional digits of the Required Action number and by successive indentations of the logical connectors.

When logical connectors are used to state a Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Condition, Completion Time, Surveillance, or Frequency.

EXAMPLES The following examples illustrate the use of logical connectors.

s EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Verify . . .

AND A.2 Restore . . .

In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.

INDIAN POINT 2 1.2 - 1 Amendment No. 238

Logical Connectors 1.2 1.2 Logical Connectors EXAMPLES (continued)

EXAMPLE 1.2-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Trip . . .

OR A.2.1 Verify . . .

AND A.2.2.1 Reduce . . .

OR A.2.2.2 Perform . . .

OR A.3 Align . . .

This example represents a more complicated use of logical connectors.

Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector OR and the left justified placement. Any one of these three Actions may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND. Required Action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The indented position of the logical connector OR indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed.

INDIAN POINT 2 1.2 - 2 Amendment No. 238

Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.

handling and storage of spent nuclear fuel BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s).

DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires facility entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability. facility If situations are discovered that require entry into more than one Condition at a time within a single LCO (multiple Conditions), the Required Actions for each Condition must be performed within the associated Completion Time.

When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the situation that required entry into the Condition.

Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically stated. The Required Actions of the Condition continue to apply to each additional failure, with Completion Times based on initial entry into the Condition.

However, when a subsequent train, subsystem, component, or variable expressed in the Condition is discovered to be inoperable or not within limits, the Completion Time(s) may be extended. To apply this Completion Time extension, two criteria must first be met. The subsequent inoperability:

a. Must exist concurrent with the first inoperability and INDIAN POINT 2 1.3 - 1 Amendment No. 238

Completion Times 1.3 1.3 Completion Times DESCRIPTION (continued)

b. Must remain inoperable or not within limits after the first inoperability is resolved.

The total Completion Time allowed for completing a Required Action to address the subsequent inoperability shall be limited to the more restrictive of either:

a. The stated Completion Time, as measured from the initial entry into the Condition, plus an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or

b. The stated Completion Time as measured from discovery of the subsequent inoperability.

The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re-entry into the Condition (for each train, subsystem, component, or variable expressed in the Condition) and separate tracking of Completion Times based on this re-entry. These exceptions are stated in individual Specifications.

The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase "from discovery . . ." Example 1.3-3 illustrates one use of this type of Completion Time. The 10 day Completion Time specified for Conditions A and B in Example 1.3-3 may not be extended.

s EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.

Required Actions INDIAN POINT 2 1.3 - 2 Amendment No. 238

Completion Times 1.3 1.3 Completion Times Suspend movement of fuel EXAMPLES (continued) assemblies in the Spent Fuel Pit. Initiate action to restore EXAMPLE 1.3-1 Spent Fuel Pit boron concentration to within limit.

ACTIONS A A CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Spent Fuel Pit boron Action and A Immediately concentration not associated AND Completion Immediately within limit Time not met. B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months A

Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered. A A

The Required Actions of Condition B are to be in MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months AND in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . A total of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is allowed for reaching MODE 3 and a total of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (not 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months ) is allowed for reaching MODE 5 from the time that Condition B was entered. If MODE 3 is reached within 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , the time allowed for reaching MODE 5 is the next 33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months because the total time allowed for reaching MODE 5 is 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

If Condition B is entered while in MODE 3, the time allowed for reaching MODE 5 is the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

immediately suspend movement of fuel assemblies in the Spent Fuel Pit and initiate action to restore Spent Fuel Pit boron concentration to within limit.

INDIAN POINT 2 1.3 - 3 Amendment No. 238

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One pump A.1 Restore pump to 7 days inoperable. OPERABLE status.

B. Required B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action and associated AND Completion Time not met. B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months When a pump is declared inoperable, Condition A is entered. If the pump is not restored to OPERABLE status within 7 days, Condition B is also entered and the Completion Time clocks for Required Actions B.1 and B.2 start. If the inoperable pump is restored to OPERABLE status after Condition B is entered, Condition A and B are exited, and therefore, the Required Actions of Condition B may be terminated.

When a second pump is declared inoperable while the first pump is still inoperable, Condition A is not re-entered for the second pump. LCO 3.0.3 is entered, since the ACTIONS do not include a Condition for more than one inoperable pump. The Completion Time clock for Condition A does not stop after LCO 3.0.3 is entered, but continues to be tracked from the time Condition A was initially entered.

While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has not expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition A.

INDIAN POINT 2 1.3 - 4 Amendment No. 238

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued)

While in LCO 3.0.3, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time for Condition A has expired, LCO 3.0.3 may be exited and operation continued in accordance with Condition B. The Completion Time for Condition B is tracked from the time the Condition A Completion Time expired.

On restoring one of the pumps to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first pump was declared inoperable. This Completion Time may be extended if the pump restored to OPERABLE status was the first inoperable pump. A 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months extension to the stated 7 days is allowed, provided this does not result in the second pump being inoperable for > 7 days.

INDIAN POINT 2 1.3 - 5 Amendment No. 238

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Restore Function X 7 days Function X train to OPERABLE train status. AND inoperable.

10 days from discovery of failure to meet the LCO B. One B.1 Restore Function Y 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Function Y train to OPERABLE train status. AND inoperable.

10 days from discovery of failure to meet the LCO C. One C.1 Restore Function X 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Function X train to OPERABLE train status.

inoperable.

OR AND C.2 Restore Function Y 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months One train to OPERABLE Function Y status.

train inoperable.

When one Function X train and one Function Y train are inoperable, Condition A and Condition B are concurrently applicable. The Completion Times for Condition A and Condition B are tracked separately for each train starting from the time each train was declared inoperable and the Condition was entered. A separate Completion Time is established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered).

INDIAN POINT 2 1.3 - 6 Amendment No. 238

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued)

If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected train was declared inoperable (i.e., initial entry into Condition A).

The Completion Times of Conditions A and B are modified by a logical connector with a separate 10 day Completion Time measured from the time it was discovered the LCO was not met. In this example, without the separate Completion Time, it would be possible to alternate between Conditions A, B, and C in such a manner that operation could continue indefinitely without ever restoring systems to meet the LCO. The separate Completion Time modified by the phrase "from discovery of failure to meet the LCO" is designed to prevent indefinite continued operation while not meeting the LCO. This Completion Time allows for an exception to the normal "time zero" for beginning the Completion Time "clock." In this instance, the Completion Time "time zero" is specified as commencing at the time the LCO was initially not met, instead of at the time the associated Condition was entered.

INDIAN POINT 2 1.3 - 7 Amendment No. 238

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Restore valve(s) to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months valves OPERABLE inoperable. status.

B. Required B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action and associated AND Completion Time not met. B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months A single Completion Time is used for any number of valves inoperable at the same time. The Completion Time associated with Condition A is based on the initial entry into Condition A and is not tracked on a per valve basis.

Declaring subsequent valves inoperable, while Condition A is still in effect, does not trigger the tracking of separate Completion Times.

Once one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but continues from the time the first valve was declared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status was the first inoperable valve.

The Condition A Completion Time may be extended for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provided this does not result in any subsequent valve being inoperable for

> 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (including the extension) expires while one or more valves are still inoperable, Condition B is entered.

INDIAN POINT 2 1.3 - 8 Amendment No. 238

Completion Times

~

1 3 Completion Times E.X/\MPLES (eontinued)

EXAMPLE 1.3 5 ACTIONS on;

______ _____ -N . parable valve.

Separate Condition entry .is allo... t i ed for eaeh mo CONDITION REQUIRED ACTION COMPLETIO N TIIVIE A. One or more A.1 Restore valve to "al11es OPER/\BLE i~o~erable. status.

B. Required B.1 Be in MODE 3. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Aetion and assoeiated ANG Con=ipletion Time not met. 8.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of

" ONS Table is a metho~ . modifying how ~he ho*"' the Completion Nole aiaa,e _the. n G~ked. lflhis FRe!had al FR_ee,~;:ilia; : Iha Nale would GaFRpleliaA T1FRe IS Ira plioalale ealy la a spee,~*,7h ACTIONS Talale.

Time is traoked was_~P rather than at the top o e.'

appear iA Iha! Gaae,1,aa . fer eash iaaperalale II s Cond1t1on " to be entered separate!~ . I11,i,en a valve 1s The Nate a - ~ TiFRes lraeked "" a Per,F *a Ive basis. n t rts re

. GaFRpleliea TiFRe &a .

"al**e aad GaR1plel1ea diliaA A is ealered aad ,ts d'l'aA A is ealered fer

*

lal Cea -~ n

a lal Goa 11 n h deelared iaapera~ e' e deelared iaapera e, ~ 8 are Ira eked fer eae If subsequent val¥es Completion Times start an eaei, **l*e aad separ .

~

II Iha CaR1plel1aA

. Time assooiated *.v1th . >>al>>e iA CaA a * ~ lotion Times fer Iha! Yalve. II !he Ca p 8Adili9A B is ealeree d*t* " O*p1res, HOA n assoo1a. led

~e"'k,,"'

Caadijiaa Bis ealered11 iA CaadijiaA A e*p,re, _C TiRJes start aad are t 11al es -- lotion - *

wijh sulase~ueA- * ::al**e aAd separate GaRlp iAle GaAdiliaA B IS separately fer ea~~ l~e

II a ,alve Illa! eaused *1ee Iha! ,alve.

traeked for each : ~~E status Condition B is e*1 restored le OPER,, ' . I . aAd lraekiA~

s*insepara ee the tNote lti le Condition en ry ly of in thi~ e*a~ple e Con=iplet1on Times, ~==~l=t~o:Tin=ie 0*tensions do not app.

INDIAN POINT 2 1.3 9 Amendment No. ~8i

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One channel A.1 Perform Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months inoperable. SR 3.x.x.x.

OR A.2 Reduce 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months THERMAL POWER to 50% RTP.

B. Required B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action and associated Completion Time not met.

Entry into Condition A offers a choice between Required Action A.1 or A.2.

Required Action A.1 has a "once per" Completion Time, which qualifies for the 25% extension, per SR 3.0.2, to each performance after the initial performance. The initial 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months interval of Required Action A.1 begins when Condition A is entered and the initial performance of Required Action A.1 must be complete within the first 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months interval. If Required Action A.1 is followed, and the Required Action is not met within the Completion Time (plus the extension allowed by SR 3.0.2), Condition B is entered. If Required Action A.2 is followed and the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is not met, Condition B is entered.

If after entry into Condition B, Required Action A.1 or A.2 is met, Condition B is exited and operation may then continue in Condition A.

INDIAN POINT 2 1.3 - 10 Amendment No. 238

Completion Times 1.3 1.3 Completion Times EXAMPLES (continued)

EXAMPLE 1.3-7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One A.1 Verify affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months subsystem subsystem inoperable. isolated.

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 Restore 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months subsystem to OPERABLE status.

B. Required B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action and associated Completion AND Time not met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action A.1 has two Completion Times. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time begins at the time the Condition is entered and each "Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter" interval begins upon performance of Required Action A.1.

If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or any subsequent 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months interval from the previous performance (plus the extension allowed by SR 3.0.2), Condition B is entered. The Completion Time clock for Condition A does not stop after Condition B is entered, but continues from the time Condition A was initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance with Condition A, provided the Completion Time for Required Action A.2 has not expired.

IMMEDIATE When "Immediately" is used as a Completion Time, the Required Action COMPLETION TIME should be pursued without delay and in a controlled manner.

INDIAN POINT 2 1.3 - 11 Amendment No. 238

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0.2, Surveillance Requirement (SR)

Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR as well as certain Notes in the Surveillance column that modify performance requirements.

Sometimes special situations dictate when the requirements of a Surveillance are to be met. They are "otherwise stated" conditions allowed by SR 3.0.1. They may be stated as clarifying Notes in the Surveillance, as part of the Surveillances, or both.

Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be preformed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.

The use of "met" or "performed" in these instances conveys specific meanings. A Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requirement to specifically determine the ability to meet the acceptance criteria.

Some Surveillances contain notes that modify the Frequency of performance or the conditions during which the acceptance criteria must be satisfied. For these Surveillances, the MODE-entry restrictions of SR 3.0.4 may not apply. Such a Surveillance is not required to be performed prior to entering a MODE or other specified condition in the Applicability of the associated LCO if any of the following three conditions are satisfied:

INDIAN POINT 2 1.4 - 1 Amendment No. 238

Frequency 1.4 1.4 Frequency DESCRIPTION (continued)

1. The Surveillance is not required to be met in the MODE or other specified condition to be entered: or

2. The Surveillance is required to be met in the MODE or other specified condition to be entered, but has been performed within the specified Frequency (i.e., it is current) and is known not to be failed; or

3. The Surveillance is required to be met, but not performed, in the MODE or other specified condition to be entered, and is known not to be failed.

Examples 1.4-3, 1.4-4, 1.4-5, and 1.4-6 discuss these special situations.

EXAMPLES The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3.

illustrates the type of Frequency statement that appears in the Technical Specifications (TS).

INDIAN POINT 2 1.4 - 2 Amendment No. 238

Frequency 1.4 1.4 Frequency EXAMPLES (continued)

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Verify level is within limits Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. The measurement of this interval continues at all facility times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the unit is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the unit is in a MODE or other specified condition in the Applicability of the LCO, and the performance of the Surveillance is not otherwise modified (refer to Example 1.4-3), then SR 3.0.3 becomes applicable. facility facility If the interval as specified by SR 3.0.2 is exceeded while the unit is not in a MODE or other specified condition in the Applicability of the LCO for which performance of the SR is required, then SR 3.0.4 becomes applicable. The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3, prior to entry into the MODE or other specified condition or the LCO is considered not met (in accordance with SR 3.0.1) and LCO 3.0.4 becomes applicable.

INDIAN POINT 2 1.4 - 3 Amendment No. 252

Frequency 1.4 1.4 Frequency EXAMPLES (continued)

EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits. Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after 25% RTP AND 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter Example 1.4-2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to 25% RTP, the Surveillance must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the 25% extension allowed by SR 3.0.2. "Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If reactor power decreases to < 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.

INDIAN POINT 2 1.4 - 4 Amendment No. 238

FFequeney 4:4 4:-4 FFOquenoy EX,*\MPLES (oontinued)

EXAMPLE 1.<1 3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Not . ** NOTi -

rnqu,rnd

,. 26% RTP. to be po FfOFmed until 12 hOUFS a#eF PeFfoFm channel adjustment 7 days

!"e in!eNal eenlim1es . ***Reiff -

etween peFfOFmanees' n OF OF not the unit operntion .IS< 26% RTP

/\s the Note modifies the .

eonstFUed to be part of the .~equ,~~d peFforn:ianee of the S . .

be e*oeeded wAile e . speeilled F,eqseney." SA sn.e1llanee, ii is pewe, ,eaoAes 2e~e,al1en ,s < 26% RTP IA is Nol esld IAe 7 day in!erval slill eensiderod k, b; ~TP le pelfe,,., IAe S~Neillan e allows 12 AesFS oiler

,I the SsFYeillanee ,*. pe-<>Fffled within lhe "speeified ;e. The SsNeillanee is allowed by SR 3.9 ~e,e ""'. pelferffled wilhin lhe 7 d:.qseney." TheFclero, oensliMe a failsro ~/~:";;~al, bs_l epe,alien was < 2l~p~"; lhe eo!ensien 1:/:;i SR 3.g.4 866\JFS ** *he BF fail a,e le ffleellhe LG *

p. ,1 wesld RBI fflel, p,e,ided ep',;,ral~*~*:ng1ng MODES' e11Cn w~A ~s~, ne *iela!ien el oes not e:>Eeeed 12 h . ay FFequeney not Once the un**

BSFS Wilh pewe, 26% RTP 1~ rnaohes 26% Rl *

~he SuFVeillanee. If the S~F' *e11P, 12 ho urn would be allo*"ed f inleFYal, !Aero we aid I A ~~, a nee were nel pelfe,,., d ~~;

speeified Froqsene. en be a fa1ls,e le peffe,,., a e :*.*IA1n U.1s 12 Rear

~* eefflpleling y, and the prni.*isions of SR Sur.e,llance within the 3.0.3 would apply.

INDIAN POINT 2 1.<1 6 Amendment No. 238

FFequeney 4-4 4-4 P:Fequeney EXL\MPLES (eontinued)

EXAMPLE 1.4 4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Only . NOTE -

FCqu1FCd to be met in MODE 1 VeFifv, a leak g e Fates aFC v.*ithin limits 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months E>mmple1.44s ha,e le loe I' e.*..

,~es Iha! !he Fe~UifeFH . -

lhe FFe~ue:: :t~i!h~ URil_i* iR MODE /'::e~R:*.SuFYeillaaee de Rel El<8FHl'IO 1.4 1. H&"e'~e,e1llaRse SORliAues al all li:al FHeasuFeFHeAI '"'

e*se!'lieR le lhe *" ,e,,. the Nole eeaslijules "."* as dese,iloed iR SuFYeillaAee 'NeFe nl'~IIE'891hly el lhis SuNeillaA aA elhe,wise slaled" e><leRSiOR allowed 1, ~~:ern,,.,ed wilhiR lhe 24 ::~, The,eleFe, ii lhe Be RO !oil UFO el !he ~R 3.Q.2)' lnff lhe UR ii was ROI iA M~RleNal (l'IUS lhe el SR 3 g 4 es ROF fa1luFe le A1eel lhe LGO ~ DE 1, lheFe would F,e~ueRoy 0UFS ,reR

e*eeeded ...h .shaRgiRg MODES ~-*e Tt-leF

~~-Fe, f RO YielalieR MODE 1. PFiO I ' l'Fe'tl<led lhe MODE .~

A Wllh lhe 24 heUF F '

eRleaag MO

,e~ueRoy we,e Rel A1el), SR ";i~DE,.~ (assuA1iRg aage ..,.,

aga;~

iFC sati I

f .;h;,*;hA1ade iRle e 24 houF s yIng the SR.

INDIAN POINT 2 1.4 e Amendment No. 238

Frequency 4-:4 4-:4 Frequenoy EXAMPLES (oontinued)

EXAMPLE 1.4 5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Gnly . . NOTE!

required to 13 e perforrned in MGDE 1.

_ Perforrn r n60p 1ete oycle of the valve 7 days T

-orh*a inl*FYal *eAliAS es,/.h~ther (the assurn ,., or not the unit . . .

p*flermaAe*s . ** n pplleabili!y el lh* associatedep*ral1eA ISLCO)

IA MOElE 1, 2 bet'lt'een

~A,s the is construed to be iA!eFYal b* ****** :.t.~..~

Note rnodifies the r .

"sp**ili**

equ1red perforrnanee oft th* Fre~**==y~**e;;*~llaAee, lh* Nole eAlry iAle aA* aper f " . I* *P*ra!1eA is Rel iA MOEl. **I* 11,e 7 **Y liisFYeillaAee is sl~I ieA IA MOElE:ii 2 aA* a le P*;;.rm E 1, !!,is Nole allows Fre~seAcy" ii eeAS1*ered le b* P*" th* :iiSFYOlllaAee. The Iii . eemplete* p . -er"'** "'llhin lh "

b:

sFYe1llaAee were Rel ,. ~r,er le *RleriAg MOElE ; ~ e speeilie*

SR a. g .4 ee6"rs .,,h epera!1eA was Rel iA

R er fa1lsre le "'**!!he L' M;~ *""'"*"""

by SR 3.Q.2) iAIOFYai" """". wi!hiA th* 7 *ay {piss I. Th*relere, if !he eeAs!Rsle a leilsre ell~

E 1, ~ allewe*

wesld Rel FAe!, pre,ided .;~~::::AglAg MOElE:ii, ***A wRh7h~*7~s~, AO YielalieA el oes not result in entn* . t a~ Frequency not 0 . ' IA 8 MOElE 1 nee the unit reaches MGD . .

perforrned within its s . E 1, the requ1reA=ient for lii*iv*illaAe* ha* b**:**1fi*d Fre~**"*Y appli*s aA* ~~* .

iisFYelllOAee le b*

pr,er le *Al*riAg MOEl~*fler"'*** If lh* :iisFY*illaAe:~~1* re~wre Iha! lh*

iisFV*illaA** 'lo'ilhiA lh* 1,. lh*r* wesl* lh*A b* a i'"'re AO! p*flerFA**

wesl* apply. sp*e1fi** Fre~**Acy, aAd lh e pro,IsIons SR a.a.a

l*re le ofp*fle"" e INDl,A,N POINT 2 1.4 7 Arnendrnent No. 238

rrcqucney

+.-4 EXAMPLE 1.4 6 SURVEILLANCE REQUIREMENTS SURVEILLANCE rREQUENCY NOTE Not required to be rnet in MODE 3 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Verifv, rpara n cter .is within lirnits E>mrnplc 1*4 6 spce1f1es.. that t hai.<c to be rnct whil . -~c requircrncnts of th . .

asseeialed LGO . : the URI! is IR MODE a ~he

is SuR1e1llaRee de Rot F is ~ODES 1 2 assuff!ed

l'l'li I, T requeRey el this Su * .~ " aRd 3). The iR!ewal n ----oa- 1-,ty efthe E"""'l'le 1.4 1. He--:-~1llaRee eeRtiRues et a1i'u,.,:easureff!eRI fer the e*oe1>tieR to the

wr,er, __lhe Nete eeRslitute ~,' as deseAlled iR SuFYeillaRee ""<ere nl'~ ieall11ily al this Sue*eill." aR etherwise stated" no p r t ' nee +h e><leRsieR allewed I,* SR- e7!rmed withiR the 24 he* :-erelere, ii the 0

RO failure ellhe SR~ / g .2)' a Rd the UR ii was iR 11.11c:'~ iRleFYal (l'I us !he SR 3.Q.4 eeours --*h er;:11lure le meellhe LCO T' E a, !here weuld Ile 24 heur FrequeR;. :R ehaRg1Rg MODES le eRie, ~erefere, Re ,ielotieR el eRIFy iRle MODE y2 *~eeded, 1>re,ided the MODE~~ ODE a, eYeR with !he 24 heur FrequeRG . :., r,er le eRleFiRg MODE 2 ( aRge dees Rel result iR

'1 v.erc not rnet), SR a.0.4 would assu_rn1ng ?gain that the require satisfying the SR INDIAN POIN+ 2 1.4 8 Arnendrnent No. 238

Deleted SLs DELETED 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the limits specified in the COLR. The departure from nucleate boiling ratio (DNBR) shall be maintained 1.17 for the WRB-1 DNB correlations.

2.1.2 Reactor Coolant System Pressure SL In MODES 1, 2, 3, 4, and 5 and MODE 6 when reactor vessel head is on, the RCS pressure shall be maintained 2735 psig.

2.2 SAFETY LIMIT VIOLATIONS 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, 5 or 6, restore compliance within 5 minutes.

INDIAN POINT 2 2.0 - 1 Amendment No. 238

LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7, and LCO 3.0.8.

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated.

LCO 3.0.3 When an LCO is not met and the associated ACTIONS are not met, an associated ACTION is not provided, or if directed by the associated ACTIONS, the unit shall be placed in a MODE or other specified condition in which the LCO is not applicable. Action shall be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit, as applicable, in:

a. MODE 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months ,

b. MODE 4 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months , and

c. MODE 5 within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

Exceptions to this Specification are stated in the individual Specifications.

Where corrective measures are completed that permit operation in accordance with the LCO or ACTIONS, completion of the actions required by LCO 3.0.3 is not required.

LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4.

LCO 3.0.4 When an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made:

a. When the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time; INDIAN POINT 2 3.0 - 1 Amendment No. 245

LCO Applicability 3.0 3.0 LCO Applicability LCO 3.0.4 (continued)

b. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate; exceptions to this Specification are stated in the individual Specifications, or

c. When an allowance is stated in the individual value, parameter, or other Specification.

This Specification shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

LCO 3.0.5 Equipment removed from service or declared inoperable to comply with ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to LCO 3.0.2 for the system returned to service under administrative control to perform the testing required to demonstrate OPERABILITY.

LCO 3.0.6 When a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered. Only the support system LCO ACTIONS are required to be entered. This is an exception to LCO 3.0.2 for the supported system. In this event, an evaluation shall be performed in accordance with Technical Specification 5.5.13, "Safety Function Determination Program (SFDP)." If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

When a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

INDIAN POINT 2 3.0 - 2 Amendment No. 238

LCO Applicability 3.0 3.0 LCO Applicability LCO 3.0.7 Test Exception LCO 3.1.8 allows specified Technical Specification (TS) requirements to be changed to permit performance of special tests and operations. Unless otherwise specified, all other TS requirements remain unchanged. Compliance with Test Exception LCOs is optional. When a Test Exception LCO is desired to be met but is not met, the ACTIONS of the Test Exception LCO shall be met. When a Test Exception LCO is not desired to be met, entry into a MODE or other specified condition in the Applicability shall be made in accordance with the other applicable Specifications.

LCO 3.0.8 When one or more required snubbers are unable to perform their associated support function(s), any affected supported LCO(s) are not required to be declared not met solely for this reason if risk is assessed and managed, and:

a. the snubbers not able to perform their associated support function(s) are associated with only one train or subsystem of a multiple train or subsystem supported system or are associated with a single train or subsystem supported system and are able to perform their associated support function within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; or

b. the snubbers not able to perform their associated support function(s) are associated with more than one train or subsystem of a multiple train or subsystem supported system and are able to perform their associated support function within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

At the end of the specified period the required snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met.

INDIAN POINT 2 3.0 - 3 Amendment No. 245

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per . . ."

basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

INDIAN POINT 2 3.0 - 4 Amendment No. 238

SR Applicability 3.0 3.0 SR Applicability SR 3.0.4 Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

INDIAN POINT 2 3.0 - 5 Amendment No. 238

SPENT FUEL PIT Spent Fuel Pit Water Level REQUIREMENTS 3.7.11 3.7 3.7.11 Spent Fuel Pit Water Level LCO 3.7.11 The Spent Fuel Pit water level shall be~ 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the Spent Fuel Pit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent Fuel Pit water A.1 level not within limit. NOTI!!

LGO a.a.a is not ap pIieab IC.-----------------------

Suspend movement of Immediately irradiated fuel assemblies in the Spent Fuel Pit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7 .11 .1 Verify the Spent Fuel Pit water level is ~ 23 ft above 7 days the top of the irradiated fuel assemblies seated in the storage racks.

INDIAN POINT 2 3.7.11 - 1 Amendment No. ~

SPENT FUEL PIT Spent Fuel Pit Boron Concentration REQUIREMENTS 3.7.12 3.7 3.7.12 Spent Fuel Pit Boron Concentration LCO 3.7.12 The Spent Fuel Pit boron concentration shall be ~ 2000 ppm.

APPLICABILITY: When fuel assemblies are stored in the Spent Fuel Pit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A Spent Fuel Pit boron NOTI!

concentration not within LCO 3.0.3 is not limit. a pp Iioa ble.______________________ _

A.1 Suspend movement of fuel Immediately assemblies in the Spent Fuel Pit.

A.2 Initiate action to restore Immediately Spent Fuel Pit boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Verify the Spent Fuel Pit boron concentration is 7 days within limit.

INDIAN POINT 2 3.7.12 - 1 Amendment No. ~

SPENT FUEL PIT Spent Fuel Pit Storage REQUIREMENTS 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 may be 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 4.4 w/o U235 and discharged prior to IP3 Cycle 12 shall be stored in the Spent Fuel Pit. IP3 fuel assemblies V43 and V48 are not approved for storage 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. 287

Spent Fuel Pit Storage 3.7.13 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 -----------------------------------

LCO not met. - NOTE -

LCO 3.0.3 is not applicable.

Initiate action to move the Immediately noncomplying fuel assembly to an acceptable location.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify by administrative means that the IP2 fuel Prior to storing the assembly has been classified in accordance with fuel assembly in Figure 3.7.13-1, Figure 3.7.13-2, Figure 3.7.13-3, or the Spent Fuel Pit.

Figure 3.7.13-4 and meets the requirements for the intended storage location.

OR Verify by administrative means that the IP3 fuel Prior to storing the assembly meets the requirements for the intended fuel assembly in the Spent Fuel Pit.

storage location.

INDIAN POINT 2 3.7.13 - 2 Amendment No. 268

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location Indian Point 2 is located on the East bank of the Hudson River at Indian Point, Village of Buchanan, in upper Westchester County, New York. The site is approximately 24 miles north of the New York City boundary line. The nearest city is Peekskill which is 2.5 miles northeast of Indian Point.

The minimum distance from the reactor center line to the boundary of the site exclusion area and the outer boundary of the low population zone, as defined in 10 CFR 100.3, is 520 meters and 1100 meters, respectively. For the purpose of satisfying 10 CFR Part 20, the Restricted Area is the same as the Exclusion Area shown in UFSAR, Figure 2.2-2.

4.2 Deleted the Defueled Safety Analysis Report (DSAR) 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent, INDIAN POINT 2 4.0 - 1 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued)

b. keff < 1.0 if fully flooded with unborated water, and

c. Each fuel assembly classified based on initial enrichment, burnup, cooling time and number of Integral Fuel Burnable Absorbers (IFBA) rods with individual fuel assembly storage location within the spent fuel storage rack restricted as required by Technical Specification 3.7.13.

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent, and poisons, if necessary, to meet the limit for keff,

b. keff 0.95 if fully flooded with unborated water, and

c. A 20.5 inch center to center distance between fuel assemblies placed in the storage racks to meet the limit for keff.

4.3.2 Drainage The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pit below a nominal elevation of 88 feet, 6 inches.

4.3.3 Capacity The spent fuel pit is designed and shall be maintained with a storage capacity limited to no more than 269 fuel assemblies in Region I and 1105 fuel assemblies in Region II.

INDIAN POINT 2 4.0 - 2 Amendment No. 238

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for facility staff and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear fuel.

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all decommissioning organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements including the facility-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the UFSAR, DSAR

b. The plant manager shall be responsible for overall safe maintenance of the facility and shall have control over those onsite activities necessary for storage and maintenance of nuclear fuel.

c. The corporate officer with direct responsibility for IP2 shall have corporate responsibility for the safe storage and handling of nuclear fuel and shall take any measures needed to ensure acceptable performance of the staff in maintaining and providing technical support to the facility to ensure safe management of nuclear fuel, and

d. The individuals who train the CERTIFIED FUEL HANDLERS, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

5.2.2 Facility Staff The facility staff organization shall include the following:

a. Each duty shift shall be composed of at least one shift manager and one NON-CERTIFIED OPERATOR. The NON-CERTIFIED OPERATOR position may be filled by a CERTIFIED FUEL HANDLER.

INDIAN POINT 2 5.2 - 1 Amendment No.

Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The procedures applicable to the safe storage of nuclear fuel recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 except as provided in the quality assurance program described or referenced in the Updated FSAR; DSAR

b. Deleted;

c. Quality assurance for effluent and environmental monitoring; Deleted

d. Fire Protection Program implementation;

e. All programs specified in Technical Specification 5.5; and

f. Personnel radiation protection consistent with the requirements of 10 CFR 20.

INDIAN POINT 2 5.4 - 1 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.2 Primary Coolant Sources Outside Containment Deleted Deleted 5.5.3 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

INDIAN POINT 2 5.5 - 2 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Radioactive Effluent Controls Program (continued)

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001 - 20.2402,

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I,

/facility

e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days,

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I, INDIAN POINT 2 5.5 - 3 Amendment No. 250

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Radioactive Effluent Controls Program (continued)

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:

1. For noble gases: a dose rate 500 mrem/yr to the whole body and a dose rate 3000 mrem/yr to the skin and

2. For iodine-131, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ,

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I,

/facility

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, tritium, and all radionuclides in particulate form with half lives

> 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I, and

/facility INDIAN POINT 2 5.5 - 4 Amendment No. 250

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Radioactive Effluent Controls Program (continued)

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Deleted Controls Program surveillance frequency.

5.5.4 Component Cyclic or Transient Limit This program provides controls to track the UFSAR, Section 4.1, cyclic and transient Deleted occurrences to ensure that components are maintained within the design limits.

5.5.5 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel using ultrasonic methods. The program shall include inspection frequencies and acceptance criteria. The inspection frequency will ensure that each reactor coolant Deleted pump flywheel is inspected at 20-year intervals.

5.5.6 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Required Frequencies for Addenda terminology for performing inservice testing inservice testing activities activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days INDIAN POINT 2 5.5 - 5 Amendment No. 267

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.6 Inservice Testing Program (continued)

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities,

c. The provisions of SR 3.0.3 are applicable to inservice testing activities, and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

5.5.7 Steam Generator (SG) Program Deleted A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following:

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the as found condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The as found condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged, to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down), all anticipated transients included in the design specification and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads INDIAN POINT 2 5.5 - 6 Amendment No. 281

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 150 gpd per SG.

3. The operational LEAKAGE performance criterion is specified in LCO 3.4.13, RCS Operational LEAKAGE.

c. Provisions for SG tube plugging criteria. Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following SG tube alternate plugging criteria shall be applied as an alternative to the preceding criteria.

Tubes with service-induced flaws located greater than 18.9 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 18.9 inches below the top of the tubesheet shall be plugged upon detection.

d. Provisions for SG tube inspections. Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from 18.9 inches below the top of the tubesheet on the hot leg side to 18.9 inches below the top of the tubesheet on the cold leg side, and that may satisfy the applicable tube plugging criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. A degradation assessment shall be performed to determine the type and INDIAN POINT 2 5.5 - 7 Amendment No. 281

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued) location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG installation.

2. After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period.

Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.

a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period; b) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and c) During the remaining life of the SGs, inspect 100% of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.

INDIAN POINT 2 5.5 - 8 Amendment No. 281

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued)

3. If crack indications are found in any SG tube, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-line indication is not associated with a crack(s),

then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

INDIAN POINT 2 5.5 - 8a Amendment No. 281

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Secondary Water Chemistry Program Deleted This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. The program shall include:

a. Identification of a sampling schedule for the critical variables and control points for these variables,

b. Identification of the procedures used to measure the values of the critical variables,

c. Identification of process sampling points.

d. Procedures for the recording and management of data,

e. Procedures defining corrective actions for all off control point chemistry conditions, and

f. A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events, which is required to initiate corrective action.

5.5.9 Ventilation Filter Testing Program (VFTP)

Deleted A program shall be established to implement the following required testing of the Control Room Ventilation System (CRVS) in accordance with Regulatory Guide 1.52, Revision 2, March 1978, and ANSI N510-1975. Tests described in Technical Specifications 5.5.9.a, 5.5.9.b, 5.5.9.c and 5.5.9.d shall be performed:

1) Within 31 days after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation since the last test (requires performance of 5.5.9.c only);

2) After 24 months of standby service;

3) After each complete or partial replacement of the HEPA filter train or charcoal adsorber filter;

4) After any structural maintenance on the system housing that could alter system integrity; and

5) After painting, fire, or chemical release in any ventilation zone communicating with the system while it is in operation.

INDIAN POINT 2 5.5 - 9 Amendment No. 238

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Ventilation Filter Testing Program.

The Required testing shall:

a. Demonstrate that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Position C.5.c of Regulatory Guide 1.52, Revision 2, March 1978, and ANSI N510-1975, while operating the system at ambient conditions and at a flow rate of 2000 cfm +/-10%.

b. Demonstrate that an inplace test of the charcoal adsorber shows a penetration and system bypass < 0.05% when tested in accordance with Regulatory Position C.5.d of Regulatory Guide 1.52, Revision 2, March 1978, and ANSI N510-1975, while operating the system at ambient conditions and at a flow rate of 2000 cfm +/-10%.

c. Demonstrate that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than 2.5% when tested in accordance with ASTM D3803-1989 at a temperature of 30C (86°F) and a relative humidity of 95%, and a face velocity of 0.203 m/sec (40 ft/min).

d. Demonstrate that the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than 6 inches water gauge when tested in accordance with Regulatory Guide 1.52, Revision 2, and N510-1975 at the system flowrate of 2000 cfm (+/- 10%).

5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure. The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures.

INDIAN POINT 2 5.5 - 10 Amendment No. 265

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion),

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents, and

c. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.11 Diesel Fuel Oil Testing Program Deleted A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established for the onsite DG fuel oil storage tanks and the DG reserve fuel oil storage tanks. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with applicable ASTM Standards. The purpose of the program is to establish the following:

a. Verification of the acceptability of new fuel oil for use prior to addition to the DG fuel oil onsite storage tanks by determining that the fuel oil has:

1. Relative density within the limits of 0.83 to 0.89;

2. Kinematic viscosity within the limits of 1.8 to 5.8; and INDIAN POINT 2 5.5 - 11 Amendment No. 238

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Diesel Fuel Oil Testing Program (continued)

3. A clear and bright appearance with proper color.

b.1 Verification of the acceptability of the fuel oil in the onsite storage tanks and the reserve storage tanks every 92 days by verifying that the properties of the fuel oil in the tanks, other than those addressed in item a, are within limits for ASTM 2D fuel oil. The sampling technique for the reserve storage tanks may deviate from ASTM D270-1975 in that only a bottom sample is required; or b.2 Verification of the acceptability of each new fuel addition made subsequent to the last verification made in accordance with item b.1 by verifying, within 31 days following the addition, that the properties of the new fuel oil, other than those properties addressed in item a, are within limits for ASTM 2D fuel oil.

c. Verification every 92 days that total particulate concentration of the fuel oil in the onsite and reserve storage tanks is less than or equal to 10 mg/l when tested in accordance with ASTM D-2276, Method A-2 or A-3. The sampling technique for the reserve storage tanks may deviate from ASTM D270-1975 in that only a bottom sample is required.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Program testing frequencies.

5.5.12 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

DSAR

1. A change in the TS incorporated in the license or

2. A change to the updated UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

DSAR INDIAN POINT 2 5.5 - 12 Amendment No. 238

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Technical Specifications (TS) Bases Control Program (continued)

d. Proposed changes that meet the criteria of Technical Specification 5.5.12b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.13 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6. The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected,

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists,

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities, and

d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, and assuming no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable, or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable, or

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

INDIAN POINT 2 5.5 - 13 Amendment No. 238

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Function Determination Program (SFDP) (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

5.5.14 Containment Leakage Rate Testing Program

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with NEI 94-01, Revision 2A, Industry Guidelines for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, October 2008.

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, P a , is assumed to be the containment design pressure of 47 psig.

c. The maximum allowable containment leakage rate, L a , at P a , and 271°F shall be 0.1% of containment steam air weight per day.

d. Leakage rate acceptance criteria:

1. Containment leakage rate acceptance criterion is 1.0 L a . During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 L a for the Type B and C tests and 0.75 L a for Type A tests.

2. Air lock testing acceptance criteria shall be established to ensure that limits for Type B and C testing in Technical Specification 5.5.14.d.1 are met.

INDIAN POINT 2 5.5 - 14 Amendment No. 283

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program (continued)

3. Isolation Valve Seal Water System leakage rate acceptance criteria is 14,700 cc/hour.

e. Acceptance criterion for leakage into containment from isolation valves sealed with the service water system is 0.36 gpm per fan cooler unit when pressurized at 1.1 Pa. This limit protects the internal recirculation pumps from flooding during the 12-month period of post accident recirculation.

f. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

g. Nothing in these Technical Specifications shall be construed to modify the testing Frequencies required by 10 CFR 50, Appendix J.

5.5.15 Battery Monitoring and Maintenance Program This program provides for battery restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications, or of the battery manufacturer including the following:

a. Actions to restore battery cells with float voltage < 2.13 V, and

b. Actions to equalize the test battery cells that had been discovered with electrolyte level below the minimum established design limit.

5.5.16 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Ventilation System (CRVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge.

The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

INDIAN POINT 2 5.5 - 15 Amendment No. 258

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Control Room Envelop Habitability Program (continued)

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

d. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CRVS, operating at the flow rate required by the VFTP, at a Frequency of 24 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.

e. The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analysis of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

INDIAN POINT 2 5.5 - 16 Amendment No. 258

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.3 Radioactive Effluent Release Report

- NOTE -

A single submittal may be made for a multiple unit/facility station. The submittal shall combine sections common to all units/facilities at the station; however, for units/facilities with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit/facility.

The Radioactive Effluent Release Report covering the operation of the unit/facility in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit/facility. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

5.6.4 Not Used 5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1. Technical Specification 2.1, Safety Limits (SL);

2. Technical Specification 3.1.1, SHUTDOWN MARGIN (SDM);

3. Technical Specification 3.1.3, Moderator Temperature Coefficient (MTC);

4. Technical Specification 3.1.5, Shutdown Bank Insertion Limits;

5. Technical Specification 3.1.6, Control Bank Insertion Limits;

6. Technical Specification 3.2.1, Heat Flux Hot Channel Factor (FQ(Z));

7. Technical Specification 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor; INDIAN POINT 2 5.6 - 2 Amendment No.

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. Technical Specification 3.2.3, Axial Flux Difference (AFD);

9. Technical Specification 3.3.1, Reactor Protection System Instrumentation;

10. Technical Specification 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits; and

11. Technical Specification 3.9.1, Boron Concentration.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. WCAP-9272-P-A, Westinghouse Reload Safety Evaluation Methodology, July 1985;

2. WCAP-8385, Power Distribution Control and Load Following Procedures - Topical Report, September 1974;

3. T.M. Anderson to K. Kniel (NRC) January 31, 1980 -

Attachment:

Operation and Safety Analysis Aspects of an Improved Load Follow Package;

4. NUREG-0800, Standard Review Plan, US Nuclear Regulatory Commission, Section 4.3, Nuclear Design, July 1981, including Branch Technical Position CPB 4.3-1,Westinghouse Constant Axial Offset Control (CAOC), Rev. 2, July 1981;

5. WCAP-11397-P-A, Revised Thermal Design Procedure, April 1989;

6. WCAP-16009-P-A, Realistic Large Break LOCA Evaluation Methodology Using Automated Statistical Treatment of Uncertainty Method (ASTRUM), M. E. Nissley, et al., January 2005.

7. WCAP-8745-P-A, Design Bases for the Thermal Overpower Delta-T and Thermal Overtemperature Delta-T Trip Functions, September 1986; INDIAN POINT 2 5.6 - 3 Amendment No. 248

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. WCAP-12610-P-A, "VANTAGE+ Fuel Assembly Reference Core Report,"

April 1995;

9. WCAP-10079-P-A, "NOTRUMP, A Nodal Transient Small Break and General Network Code," August 1985;

10. WCAP-10054-P-A, "Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code," August 1985;

11. WCAP-10054-P-A, Addendum 2, Revision 1, "Addendum to the Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code: Safety Injection Into the Broken Loop and Cosi Condensation Model,"

July 1997;

12. WCAP-13749-P-A, "Safety Evaluation Supporting the Conditional Exemption of the Most Negative EOL Moderator Temperature Coefficient Measurement," March 1997;

13. WCAP-16045-P-A, "Qualification of the Two-Dimensional Transport Code PARAGON," August 2004; and

14. WCAP-10965-P-A, "ANC: A Westinghouse Advanced Nodal Computer Code," September 1986.

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided to the NRC upon issuance for each reload cycle.

5.6.6 Post Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.3, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6. 7 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.7, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Degradation mechanisms found,

c. Nondestructive examination techniques utilized for each degradation mechanism, INDIAN POINT 2 5.6 -4 Amendment No. 285

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Steam Generator Tube Inspection Report (continued)

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,

e. Number of tubes plugged during the inspection outage for each degradation mechanism,

f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each SG,

g. The results of condition monitoring, including the results of tube pulls and in-situ testing,

h. The primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report,

i. The calculated accident leakage rate from the portion of the tubes below 18.9 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident leakage rate from the most limiting accident is less than 1.75 times the maximum primary to secondary leakage rate, the report should describe how it was determined, and

j. The results of monitoring for tube displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.

INDIAN POINT 2 5.6 - 5 Amendment No. 281

APPENDIX B TO FACILITY OPERATING LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 2, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNITS NUMBER 1 AND 2 ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION PLAN FACILITY LICENSES NO. DPR-5 AND DPR-26 DOCKET NUMBERS 50-3 AND 50-247 Renewed License Nos. DPR-5 and DPR-26

handling and storage of spent fuel and maintenance 1.0 Objectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) is to provide for protection of environmental values during construction and operation of the nuclear facility. The principal objectives of the EPP are as follows:

facility is maintained (1) Verify that the plant is operated in an environmentally acceptable manner, as established by the FES and other NRC environmental impact assessments.

(2) Coordinate NRC requirements and maintain consistency with other Federal, State and local requirements for environmental protection.

handling and storage of spent fuel and maintenance of the (3) Keep NRC informed of the environmental effects of facility construction and operation and of actions taken to control those effects.

Environmental concerns identified in the FES which relate to water quality matters are regulated by way of the licensee's SPDES permit.

1-1 Renewed License Nos. DPR-5 and DPR-26

3.0 Consistency Requirements 3.1 Plant Design and Operation facility The licensee may make changes in station design or operation or perform tests or experiments affecting the environment provided such changes, tests or experiments do not involve an unreviewed environmental question, and do not involve a change in the Environmental Protection Plan. 1 Changes in plant design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of the EPP. Activities governed by Section 3.3 are not subject to the requirements of this section.

facility Before engaging in additional construction or operational activities which may affect the environment, the licensee shall prepare and record an environmental evaluation of such activity.

When the evaluation indicates that such activity involves an unreviewed environmental question, the licensee shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation. When such activity involves a change in the Environmental Protection Plan, such activity and change to the Environmental Protection Plan may be implemented only in accordance with an appropriate license amendment as set forth in Section 5.3.

A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the final environmental statement 1

This provision does not relieve the licensee of the requirements of 10 CFR 50.59.

3-1 Renewed License Nos. DPR-5 and DPR-26

(FES) or final supplemental environmental impact statement (FSEIS), as modified by the staff's testimony to the Atomic Safety and Licensing Board, supplements to the FES, or FSEIS environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board, or the Commission; or (2) a significant change in effluents or power level; or (3) a matter not previously reviewed and evaluated in the documents specified in (1) of this subsection which may have a significant adverse environmental impact.

The licensee shall maintain records of changes in facility design or operation and of tests and experiments carried out pursuant to this subsection. These records shall include a written evaluation which provides a basis for the determination that the change, test, or experiment does not involve an unreviewed environmental question nor constitute a decrease in the effectiveness of the EPP to meet the objectives specified in Section 1.0. The licensee shall include as part of its Annual Environmental Protection Plan Report (per subsection 5.4.1) brief descriptions, analyses, interpretations, and evaluations of such changes, tests and experiments.

3.2 Reporting Related to the SPDES Permit Violations of the SPDES permit shall be reported to the NRC by submittal of copies of the reports required by the SPDES permit.

Changes and additions to the SPDES permit shall be reported to the NRC within 30 days following the date the change is approved. If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted.

3-2 Renewed License Nos. DPR-5 and DPR-26

The NRC shall be notified of the changes to the effective SPDES permit proposed by the licensee by providing NRC with a copy of the proposed change at the same time it is submitted to the permitting agency. The notification of a licensee-initiated change shall include a copy of the requested revision submitted to the permitting agency. The licensee shall provide the NRC a copy of the application for renewal of the SPDES permit at the same time the application is submitted to the permitting agency.

3.3 Changes Required for Compliance with Other Environmental Regulations facility Changes in plant design or operation and performance of tests or experiments which are required to achieve compliance with other Federal, State, or local environmental regulations are not subject to the requirements of Section 3.1.

3-3 Renewed License Nos. DPR-5 and DPR-26

the handling and storage of spent fuel and maintenance of the facility 4.0 Environmental Conditions 4.1 Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and promptly reported to the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone, telegraph or facsimile transmissions followed by a written report per subsection 5.4.2. The following are examples: excessive bird impaction events, onsite plant or animal disease outbreaks, unusual mortality or occurrence of any species protected by the Endangered Species Act of 1973, unusual fish kills, unusual increase in nuisance organisms or conditions, and unanticipated or emergency discharge of waste water or chemical substances.

No routine monitoring programs are required to implement this condition.

4.2 Environmental Monitoring In accordance with Section 7(a) of the Endangered Species Act, the National Marine Fisheries Service (NMFS) issued a Biological Opinion related to the continued operation of IP2 and IP3 that pertains to shortnose sturgeon (Acipenser brevirostrum) and Atlantic sturgeon (Acipenser oxyrinchus oxyrinchus). The Biological Opinion includes an Incidental Take Statement with Reasonable and Prudent Measures that the NMFS has determined to be necessary or appropriate to minimize the amount or extent of incidental take and associated Terms and Conditions, which are non-discretionary and implement the Reasonable and Prudent Measures. The currently applicable Biological Opinion concludes that continued operation of IP2 and IP3 is not likely to jeopardize the continued existence of the listed species or to adversely affect the designated critical habitat of those species.

This Biological Opinion conservatively bounds the conditions that will occur in the permanently shut down and defueled condition.

4-1 Renewed License Nos. DPR-5 and DPR-26

5.0 Administrative Procedures 5.1 Review and Audit The licensee shall provide for review and audit of compliance with the Environmental Protection Plan. The audits shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure utilized to achieve the independent review and audit function and results of the audit activities shall be maintained and made available for inspection.

and the handling and storage of spent fuel and maintenance of the facility 5.2 Records Retention previous Records and logs relative to the environmental aspects of plant operation shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to the NRC on request.

facility facility Records of modifications to plant structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the plant. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.

5.3 Changes in Environmental Protection Plan Requests for changes in the Environmental Protection Plan shall include an assessment of the environmental impacts of the proposed change and a supporting justification. Implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes 5-1 Renewed License Nos. DPR-5 and DPR-26

5.4 Plant Reporting Requirements 5.4.1 Routine Reports An Annual Environmental Protection Plan Report describing implementation of this EPP for the previous year shall be submitted to the NRC prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following issuance of the operating license. The period of the first report shall begin with the date of issuance of the operating license.

The Annual Environmental Protection Plan Report shall include:

(a) A list of EPP noncompliances and the corrective actions taken to remedy them.

facility (b) A list of all changes in station design or operation, tests, and experiments made in accordance with subsection 3.1 which involved a potentially significant unreviewed environmental issue.

(c) A list of nonroutine reports submitted in accordance with subsection 5.4.2.

(d) A list of all reports submitted in accordance with the SPDES permit.

In the event that some results are not available by the report due date, the report shall be submitted noting and explaining the missing results. The missing data shall be submitted as soon possible in a supplementary report.

5-3 Renewed License Nos. DPR-5 and DPR-26

facility conditions 5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of a nonroutine event. The report shall (1) describe, analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (2) describe the probable cause of the event, (3) indicate the action taken to correct the reported event, (4) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (5) indicate the agencies notified and their preliminary responses.

Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided a copy of such report at the same time it is submitted to the other agency.

5-4 Renewed License Nos. DPR-5 and DPR-26

APPENDIX C TO FACILITY OPERATING LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 2, LLC (ENIP2)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 2 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART I: SPENT FUEL TRANSFER CANISTER AND TRANSFER CASK SYSTEM FACILITY LICENSE NO. DPR-26 DOCKET NO. 50-247 Amendment No. 268

Facility Operating License Appendix C - Inter-Unit Fuel Transfer Technical Specifications SPENT FUEL SHIELDED TRANSFER CANISTER AND TRANSFER CASK SYSTEM

1.0 DESCRIPTION

The spent fuel transfer system consists of the following components: (1) a spent fuel shielded transfer canister (STC), which contains the fuel; (2) a transfer cask (HI-TRAC 100D) (hereafter referred to as HI-TRAC), which contains the STC during transfer operations; and (3) a bottom missile shield.

The STC and HI-TRAC are designed to transfer irradiated nuclear fuel assemblies from the Indian Point 3 (IP3) spent fuel pit to the Indian Point 2 (IP2) spent fuel pit. A fuel basket within the STC holds the fuel assemblies and provides criticality control. The shielded transfer canister provides the confinement boundary, water retention boundary, gamma radiation shielding, and heat rejection capability. The HI-TRAC provides a water retention boundary, protection of the STC, gamma and neutron radiation shielding, and heat rejection capability. The STC contains up to 12 fuel assemblies.

The STC is the confinement system for the fuel. It is a welded, multi-layer steel and lead cylinder with a welded base-plate and bolted lid. The inner shell of the canister forms an internal cylindrical cavity for housing the fuel basket. The outer surface of the canister inner shell is buttressed with lead and steel shells for radiation shielding. The minimum thickness of the steel, lead and steel shells relied upon for shielding starting with the innermost shell are 3/4 inch steel, 2 3/4 inch lead and 3/4 inch steel, respectively. The canister closure incorporates two O-ring seals to ensure its confinement function. The confinement system consists of the canister inner shell, bottom plate, top flange, top lid, top lid O-ring seals, vent port seal and cover plate, and drain port seal and coverplate. The fuel basket, for the transfer of 12 Pressurized Water Reactor (PWR) fuel assemblies, is a fully welded, stainless steel, honeycomb structure with neutron absorber panels attached to the individual storage cell walls under stainless steel sheathing.

The maximum gross weight of the fully loaded STC is 40 tons.

The HI-TRAC is a multi-layer steel and lead cylinder with a bolted bottom (or pool) and top lid.

For the fuel transfer operation the HI-TRAC is fitted with a solid top lid, an STC centering assembly, and a bottom missile shield. The inner shell of the transfer cask forms an internal cylindrical cavity for housing the STC. The outer surface of the cask inner shell is buttressed with intermediate lead and steel shells for radiation shielding. The minimum thickness of the steel, lead and steel shells relied upon for shielding starting with the innermost shell are 3/4 inch steel, 2 inch lead and 1 inch steel, respectively. An outside shell called the water jacket contains water for neutron shielding, with a minimum thickness of 5. The HI-TRAC bottom and top lids incorporate a gasket seal design to ensure its water confinement function. The water confinement system consists of the HI-TRAC inner shell, bottom lid, top lid, top lid seal, bottom lid seal, vent port seal, vent port cap and bottom drain plug.

The HI-TRAC provides a water retention boundary, protection of the STC, gamma and neutron radiation shielding, and heat rejection capability. The bottom missile shield is attached to the bottom of the HI-TRAC and provides tornado missile protection of the pool lid bolted joint. The HI-TRAC can withstand a tornado missile in other areas without the need for additional shielding. The STC centering assembly provides STC position control within the HI-TRAC and also acts as an internal impact limiter in the event of a non-mechanistic tipover accident.

INDIAN POINT 2 1 Amendment 268

Facility Operating License Appendix C - Inter-Unit Fuel Transfer Technical Specifications 2.0 CONDITIONS 2.1 OPERATING PROCEDURES Written operating procedures shall be prepared for cask handling, loading, movement, surveillance, maintenance, and recovery from off normal conditions such as crane hang-up. The written operating procedures shall be consistent with the technical basis described in Chapter 10 of the Licensing Report (Holtec International Report HI-2094289).

2.2 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM Written cask acceptance tests and maintenance program shall be prepared consistent with the technical basis described in Chapter 8 of the Licensing Report (Holtec International Report HI-2094289).

2.3 PRE-OPERATIONAL TESTING AND TRAINING EXERCISE A training exercise of the loading, closure, handling/transfer, and unloading, of the equipment shall be conducted prior to the first transfer. The training exercise shall not be conducted with irradiated fuel. The training exercise may be performed in an alternate step sequence from the actual procedures, but all steps must be performed. The training exercise shall include, but is not limited to the following:

a) Moving the STC into the IP3 spent fuel pool.

b) Preparation of the HI-TRAC for STC loading.

c) Selection and verification of specific fuel assemblies and non-fuel hardware to ensure type conformance.

d) Loading specific assemblies and placing assemblies into the STC (using a single dummy fuel assembly), including appropriate independent verification.

e) Remote installation of the STC lid and removal of the STC from the spent fuel pool.

f) Placement of the STC into the HI-TRAC with the STC centering assembly.

g) STC closure, establishment of STC water level with steam, verification of STC water level, STC leakage testing, and operational steps required prior to transfer, as applicable.

h) Establishment and verification of HI-TRAC water level.

i) Installation of the HI-TRAC top lid.

j) HI-TRAC closure, leakage testing, and operational steps required prior to transfer, as applicable.

k) Movement of the HI-TRAC with STC from the IP3 fuel handling building to the IP2 fuel handling building along the haul route with designated devices.

l) Moving the STC into the IP2 spent fuel pool.

m) Manual crane operations for bare STC movements including demonstration of recovery from a crane hang-up with the STC suspended from the crane.

INDIAN POINT 2 2 Amendment 268

APPENDIX C TO FACILITY OPERATING LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 2, LLC (ENIP2)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 2 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART II: TECHNICAL SPECIFICATIONS FACILITY LICENSE NO. DPR-26 DOCKET NO. 50-247 Amendment No. 268

Facility Operating License No. DPR-26 Appendix A - Technical Specifications DELETED PERMANENTLY DEFUELED TABLE OF CONTENTS B.2.0 SAFETY LIMITS (SLs)

B.2.1.1 Reactor Core SLs B.2.1.2 Reactor Coolant System Pressure SL B.3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY B.3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY B.3.1 REACTIVITY CONTROL SYSTEMS B.3.1.1 SHUTDOWN MARGIN (SDM)

B.3.1.2 Core Reactivity B.3.1.3 Moderator Temperature Coefficient (MTC)

B.3.1.4 Rod Group Alignment Limits B.3.1.5 Shutdown Bank Insertion Limits B.3.1.6 Control Bank Insertion Limits B.3.1.7 Rod Position Indication B.3.1.8 PHYSICS TESTS Exceptions - MODE 2 B.3.2 POWER DISTRIBUTION LIMITS B.3.2.1 Heat Flux Hot Channel Factor (FQ(Z))

B.3.2.2 Nuclear Enthalpy Rise Hot Channel Factor (FNH)

B.3.2.3 AXIAL FLUX DIFFERENCE (AFD) (Constant Axial Offset Control (CAOC) Methodology)

B.3.2.4 QUADRANT POWER TILT RATIO (QPTR)

B.3.3 INSTRUMENTATION B.3.3.1 Reactor Protection System (RPS) Instrumentation B.3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation B.3.3.3 Post Accident Monitoring (PAM) Instrumentation B.3.3.4 Remote Shutdown B.3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation B.3.3.6 Containment Purge System and Pressure Relief Line Isolation Instrumentation B.3.3.7 Control Room Ventilation System (CRVS) Actuation Instrumentation B.3.4 REACTOR COOLANT SYSTEM (RCS)

B.3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits B.3.4.2 RCS Minimum Temperature for Criticality B.3.4.3 RCS Pressure and Temperature (P/T) Limits B.3.4.4 RCS Loops - MODES 1 and 2 B.3.4.5 RCS Loops - MODE 3 B.3.4.6 RCS Loops - MODE 4 B.3.4.7 RCS Loops - MODE 5, Loops Filled B.3.4.8 RCS Loops - MODE 5, Loops Not Filled B.3.4.9 Pressurizer INDIAN POINT 2 i Revision 2

Facility Operating License No. DPR-26 Appendix A - Technical Specifications PERMANENTLY TABLE OF CONTENTS DEFUELED B.3.4.10 Pressurizer Safety Valves B.3.4.11 Pressurizer Power Operated Relief Valves (PORVs)

B.3.4.12 Low Temperature Overpressure Protection (LTOP)

B.3.4.13 RCS Operational LEAKAGE B.3.4.14 RCS Pressure Isolation Valve (PIV) Leakage B.3.4.15 RCS Leakage Detection Instrumentation B.3.4.16 RCS Specific Activity B.3.4.17 Steam Generator (SG) Tube Integrity B.3.5 EMERGENCY CORE COOLING SYSTEM (ECCS)

B.3.5.1 Accumulators B.3.5.2 ECCS - Operating B.3.5.3 ECCS - Shutdown B.3.5.4 Refueling Water Storage Tank (RWST)

B.3.6 CONTAINMENT SYSTEMS B.3.6.1 Containment B.3.6.2 Containment Air Locks B.3.6.3 Containment Isolation Valves B.3.6.4 Containment Pressure B.3.6.5 Containment Air Temperature B.3.6.6 Containment Spray System and Containment Fan Cooler Unit (FCU)

System B.3.6.7 Recirculation pH Control System SPENT FUEL PIT B.3.6.8 Not Used REQUIREMENTS B.3.6.9 Isolation Valve Seal Water (IVSW) System B.3.6.10 Weld Channel and Penetration Pressurization System (WC&PPS)

B.3.7 PLANT SYSTEMS B.3.7.1 Main Steam Safety Valves (MSSVs)

B.3.7.2 Main Steam Isolation Valves (MSIVs) and Main Steam Check Valves (MSCVs)

B.3.7.3 Main Feedwater Isolation B.3.7.4 Atmospheric Dump Valves (ADVs)

B.3.7.5 Auxiliary Feedwater (AFW) System B.3.7.6 Condensate Storage Tank (CST)

B.3.7.7 Component Cooling Water (CCW) System B.3.7.8 Service Water System (SWS)

B.3.7.9 Ultimate Heat Sink (UHS)

B.3.7.10 Control Room Ventilation System (CRVS)

B.3.7.11 Spent Fuel Pit Water Level B.3.7.12 Spent Fuel Pit Boron Concentration B.3.7.13 Spent Fuel Pit Storage B.3.7.14 Secondary Specific Activity INDIAN POINT 2 ii Revision 2

Facility Operating License No. DPR-26 Appendix A - Technical Specifications TABLE OF CONTENTS B.3.8 ELECTRICAL POWER SYSTEMS B.3.8.1 AC Sources - Operating B.3.8.2 AC Sources - Shutdown B.3.8.3 Diesel Fuel Oil and Starting Air B.3.8.4 DC Sources - Operating B.3.8.5 DC Sources - Shutdown B.3.8.6 Battery Parameters B.3.8.7 Inverters - Operating B.3.8.8 Inverters - Shutdown B.3.8.9 Distribution Systems - Operating B.3.8.10 Distribution Systems - Shutdown B.3.9 REFUELING OPERATIONS B.3.9.1 Boron Concentration B.3.9.2 Nuclear Instrumentation B.3.9.3 Containment Penetrations B.3.9.4 Residual Heat Removal (RHR) and Coolant Circulation - High Water Level B.3.9.5 Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level B.3.9.6 Refueling Cavity Water Level INDIAN POINT 2 iii Revision 2

LCO Applicability B 3.0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES 2 LCOs LCO 3.0.1 through LCO 3.0.8 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

facility LCO 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the MODES or other specified conditions of the Applicability statement of each Specification).

LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:

a. Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification and

b. Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.

There are two basic types of Required Actions. The first type of Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore an inoperable system or component to OPERABLE status or to restore variables to within specified limits. If this type of Required Action is not completed within the specified Completion Time, a shutdown may be required to place the unit in a MODE or condition in which the Specification is not applicable. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.) The second type of Required Action specifies the remedial measures that permit continued operation of the unit that is not further restricted by the Completion Time. In this case, compliance with the Required Actions provides an acceptable level of safety for continued operation.

INDIAN POINT 2 B 3.0 - 1 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.2 (continued)

Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.

The nature of some Required Actions of some Conditions necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual LCO's ACTIONS specify the Required Actions where this is the case. An example of this is in LCO 3.4.3, "RCS Pressure and Temperature (P/T)

Limits."

The Completion Times of the Required Actions are also applicable when a system or component is removed from service intentionally. The reasons for intentionally relying on the ACTIONS include, but are not limited to, performance of Surveillances, preventive maintenance, corrective maintenance, or investigation of operational problems. Entering ACTIONS for these reasons must be done in a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Additionally, if intentional entry into ACTIONS would result in redundant equipment being inoperable, alternatives should be used instead.

Doing so limits the time both subsystems/trains of a safety function are inoperable and limits the time conditions exist which may result in LCO 3.0.3 being entered. Individual Specifications may specify a time limit for performing an SR when equipment is removed from service or bypassed for testing. In this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment remains removed from service or bypassed.

When a change in MODE or other specified condition is required to comply with Required Actions, the unit may enter a MODE or other specified condition in which another Specification becomes applicable. In this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable, and the ACTIONS Condition(s) are entered.

LCO 3.0.3 LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and either:

a. An associated Required Action and Completion Time is not met and no other Condition applies or INDIAN POINT 2 B 3.0 - 2 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.3 (continued)

b. The condition of the unit is not specifically addressed by the associated ACTIONS. This means that no combination of Conditions stated in the ACTIONS can be made that exactly corresponds to the actual condition of the unit. Sometimes, possible combinations of Conditions are such that entering LCO 3.0.3 is warranted; in such cases, the ACTIONS specifically state a Condition corresponding to such combinations and also that LCO 3.0.3 be entered immediately.

This Specification delineates the time limits for placing the unit in a safe MODE or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LCO and its ACTIONS.

It is not intended to be used as an operational convenience that permits routine voluntary removal of redundant systems or components from service in lieu of other alternatives that would not result in redundant systems or components being inoperable.

Upon entering LCO 3.0.3, 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months is allowed to prepare for an orderly shutdown before initiating a change in unit operation. This includes time to permit the operator to coordinate the reduction in electrical generation with the load dispatcher to ensure the stability and availability of the electrical grid. The time limits specified to reach lower MODES of operation permit the shutdown to proceed in a controlled and orderly manner that is well within the specified maximum cooldown rate and within the capabilities of the unit, assuming that only the minimum required equipment is OPERABLE. This reduces thermal stresses on components of the Reactor Coolant System and the potential for a plant upset that could challenge safety systems under conditions to which this Specification applies. The use and interpretation of specified times to complete the actions of LCO 3.0.3 are consistent with the discussion of Section 1.3, Completion Times.

A unit shutdown required in accordance with LCO 3.0.3 may be terminated and LCO 3.0.3 exited if any of the following occurs:

a. The LCO is now met,

b. A Condition exists for which the Required Actions have now been performed, or

c. ACTIONS exist that do not have expired Completion Times. These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited.

INDIAN POINT 2 B 3.0 - 3 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.3 (continued)

The time limits of LCO 3.0.3 allow 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months for the unit to be in MODE 5 when a shutdown is required during MODE 1 operation. If the unit is in a lower MODE of operation when a shutdown is required, the time limit for reaching the next lower MODE applies. If a lower MODE is reached in less time than allowed, however, the total allowable time to reach MODE 5, or other applicable MODE, is not reduced. For example, if MODE 3 is reached in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , then the time allowed for reaching MODE 4 is the next 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months , because the total time for reaching MODE 4 is not reduced from the allowable limit of 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months . Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a lower MODE of operation in less than the total time allowed.

In MODES 1, 2, 3, and 4, LCO 3.0.3 provides actions for Conditions not covered in other Specifications. The requirements of LCO 3.0.3 do not apply in MODES 5 and 6 because the unit is already in the most restrictive Condition required by LCO 3.0.3. The requirements of LCO 3.0.3 do not apply in other specified conditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications sufficiently define the remedial measures to be taken.

Exceptions to LCO 3.0.3 are provided in instances where requiring a unit shutdown, in accordance with LCO 3.0.3, would not provide appropriate remedial measures for the associated condition of the unit. An example of this is in LCO 3.7.11, "Spent Fuel Pit Water Level." LCO 3.7.11 has an Applicability of "During movement of irradiated fuel assemblies in the spent fuel pit." Therefore, this LCO can be applicable in any or all MODES. If the LCO and the Required Actions of LCO 3.7.11 are not met while in MODE 1, 2, or 3, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of LCO 3.7.11 of "Suspend movement of irradiated fuel assemblies in the spent fuel pit" is the appropriate Required Action to complete in lieu of the actions of LCO 3.0.3.

These exceptions are addressed in the individual Specifications.

LCO 3.0.4 LCO 3.0.4 establishes limitations on changes in MODES or other specified conditions in the Applicability when an LCO is not met. It allows placing the unit in a MODE or other specified condition stated in that Applicability (e.g.,

the Applicability desired to be entered) when unit conditions are such that the requirements of the LCO would not be met, in accordance with LCO 3.0.4.a, LCO 3.0.4.b, or LCO 3.0.4.c.

INDIAN POINT 2 B 3.0 - 4 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.4 (continued)

LCO 3.0.4.a allows entry into a MODE or other specified condition in the Applicability with the LCO not met when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. Compliance with Required Actions that permit continued operation of the unit for an unlimited period of time in a MODE or other specified condition provides an acceptable level of safety for continued operation. This is without regard to the status of the unit before or after the MODE change. Therefore, in such cases, entry into a MODE or other specified condition in the Applicability may be made in accordance with the provisions of the Required Actions.

LCO 3.0.4.b allows entry into a MODE or other specified condition in the Applicability with the LCO not met after performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the MODE or other specified condition in the Applicability, and establishment of risk management actions, if appropriate.

The risk assessment may use quantitative, qualitative, or blended approaches, and the risk assessment will be conducted using the plant program, procedures, and criteria in place to implement 10 CFR 50.65(a)(4),

which requires that risk impacts of maintenance activities to be assessed and managed. The risk assessment, for the purposes of LCO 3.0.4 (b),

must take into account all inoperable Technical Specification equipment regardless of whether the equipment is included in the normal 10 CFR 50.65(a)(4) risk assessment scope. The risk assessments will be conducted using the procedures and guidance endorsed by Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." Regulatory Guide 1.182 endorses the guidance in Section 11 of NUMARC 93-01, "Industry Guideline for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants." These documents address general guidance for conduct of the risk assessment, quantitative and qualitative guidelines for establishing risk management actions, and example risk management actions. These include actions to plan and conduct other activities in a manner that controls overall risk, increased risk awareness by shift and management personnel, actions to reduce the duration of the condition, actions to minimize the magnitude of risk increases (establishment of backup success paths or compensatory measures), and determination that the proposed MODE change is acceptable. Consideration should also be given to the probability of completing restoration such that the requirements of the LCO would be met prior to the expiration of ACTIONS Completion Times that would require exiting the Applicability.

INDIAN POINT 2 B 3.0 - 5 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.4 (continued)

LCO 3.0.4.b may be used with single, or multiple systems and components unavailable. NUMARC 93-01 provides guidance relative to consideration of simultaneous unavailability of multiple systems and components.

The results of the risk assessment shall be considered in determining the acceptability of entering the MODE or other specified condition in the Applicability, and any corresponding risk management actions. The LCO 3.0.4.b risk assessments do not have to be documented.

The Technical Specifications allow continued operation with equipment unavailable in MODE 1 for the duration of the Completion Time. Since this is allowable, and since in general the risk impact in that particular MODE bounds the risk of transitioning into and through the applicable MODES or other specified conditions in the Applicability of the LCO, the use of the LCO 3.0.4.b allowance should be generally acceptable, as long as the risk is assessed and managed as stated above. However, there is a small subset of systems and components that have been determined to be more important to risk and use of the LCO 3.0.4.b allowance is prohibited. The LCOs governing these system and components contain Notes prohibiting the use of LCO 3.0.4.b by stating that LCO 3.0.4.b is not applicable.

LCO 3.0.4.c allows entry into a MODE or other specified condition in the Applicability with the LCO not met based on a Note in the Specification which states LCO 3.0.4.c is applicable. These specific allowances permit entry into MODES or other specified conditions in the Applicability when the associated ACTIONS to be entered do not provide for continued operation for an unlimited period of time and a risk assessment has not been performed. This allowance may apply to all the ACTIONS or to a specific Required Action of a Specification. The risk assessments performed to justify the use of LCO 3.0.4.b usually only consider systems and components. For this reason, LCO 3.0.4.c is typically applied to Specifications which describe values and parameters (e.g., RCS specific Activity), and may be applied to other Specifications based on NRC plant-specific approval.

The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

INDIAN POINT 2 B 3.0 - 6 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.4 (continued)

The provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of LCO 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5.

Upon entry into a MODE or other specified condition in the Applicability with the LCO not met, LCO 3.0.1 and LCO 3.0.2 require entry into the applicable Conditions and Required Actions until the Condition is resolved, until the LCO is met, or until the unit is not within the Applicability of the Technical Specification.

Surveillances do not have to be performed on the associated inoperable equipment (or on variables outside the specified limits), as permitted by SR 3.0.1. Therefore, utilizing LCO 3.0.4 is not a violation of SR 3.0.1 or SR 3.0.4 for any Surveillances that have not been performed on inoperable equipment. However, SRs must be met to ensure OPERABILITY prior to declaring the associated equipment OPERABLE (or variable within limits) and restoring compliance with the affected LCO.

LCO 3.0.5 LCO 3.0.5 establishes the allowance for restoring equipment to service under administrative controls when it has been removed from service or declared inoperable to comply with ACTIONS. The sole purpose of this Specification is to provide an exception to LCO 3.0.2 (e.g., to not comply with the applicable Required Action(s)) to allow the performance of required testing to demonstrate either:

a. The OPERABILITY of the equipment being returned to service or

b. The OPERABILITY of other equipment.

The administrative controls ensure the time the equipment is returned to service in conflict with the requirements of the ACTIONS is limited to the time absolutely necessary to perform the required testing to demonstrate OPERABILITY. This Specification does not provide time to perform any other preventive or corrective maintenance.

INDIAN POINT 2 B 3.0 - 7 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.5 (continued)

An example of demonstrating the OPERABILITY of the equipment being returned to service is reopening a containment isolation valve that has been closed to comply with Required Actions and must be reopened to perform the required testing.

An example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to prevent the trip function from occurring during the performance of required testing on another channel in the other trip system. A similar example of demonstrating the OPERABILITY of other equipment is taking an inoperable channel or trip system out of the tripped condition to permit the logic to function and indicate the appropriate response during the performance of required testing on another channel in the same trip system.

LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for support systems that have an LCO specified in the Technical Specifications (TS). This exception is provided because LCO 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported system LCO be entered solely due to the inoperability of the support system. This exception is justified because the actions that are required to ensure the unit is maintained in a safe condition are specified in the support system LCO's Required Actions. These Required Actions may include entering the supported system's Conditions and Required Actions or may specify other Required Actions.

When a support system is inoperable and there is an LCO specified for it in the TS, the supported system(s) are required to be declared inoperable if determined to be inoperable as a result of the support system inoperability.

However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to do so by the support system's Required Actions. The potential confusion and inconsistency of requirements related to the entry into multiple support and supported systems' LCOs' Conditions and Required Actions are eliminated by providing all the actions that are necessary to ensure the unit is maintained in a safe condition in the support system's Required Actions.

INDIAN POINT 2 B 3.0 - 8 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.6 (continued)

However, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system. This may occur immediately or after some specified delay to perform some other Required Action. Regardless of whether it is immediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry into Conditions and Required Actions for a supported system, the applicable Conditions and Required Actions shall be entered in accordance with LCO 3.0.2.

Technical Specification 5.5.13, "Safety Function Determination Program (SFDP)," ensures loss of safety function is detected and appropriate actions are taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other limitations, remedial actions, or compensatory actions may be identified as a result of the support system inoperability and corresponding exception to entering supported system Conditions and Required Actions. The SFDP implements the requirements of LCO 3.0.6.

Cross train checks to identify a loss of safety function for those support systems that support multiple and redundant safety systems are required.

The cross train check verifies that the supported systems of the redundant OPERABLE support system are OPERABLE, thereby ensuring safety function is retained. A loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable,

b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable, or

c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.

If this evaluation determines that a loss of safety function exists, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

INDIAN POINT 2 B 3.0 - 9 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.6 (continued)

This loss of safety function does not require the assumption of additional single failures or loss of offsite power. Since operations is being restricted in accordance with the ACTIONS of the support system, any resulting temporary loss of redundancy or single failure protection is taken into account. Similarly, the ACTIONS for inoperable offsite circuit(s) and inoperable diesel generator(s) provide the necessary restriction for cross train inoperabilities. This explicit cross train verification for inoperable AC electrical power sources also acknowledges that supported system(s) are not declared inoperable solely as a result of inoperability of a normal or emergency electrical power source (refer to the definition of OPERABILITY).

When a loss of safety function is determined to exist, and the SFDP requires entry into the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists, consideration must be given to the specific type of function affected. Where a loss of function is solely due to a single Technical Specification support system (e.g., loss of automatic start due to inoperable instrumentation, or loss of pump suction source due to low tank level) the appropriate LCO is the LCO for the support system. The ACTIONS for a support system LCO adequately addresses the inoperabilities of that system without reliance on entering its supported system LCO. When the loss of function is the result of multiple support systems, the appropriate LCO is the LCO for the supported system.

LCO 3.0.7 There are certain special tests and operations required to be performed at various times over the life of the unit. These special tests and operations are necessary to demonstrate select unit performance characteristics, to perform special maintenance activities, and to perform special evolutions.

Test Exception LCO 3.1.8 allows specified Technical Specification (TS) requirements to be changed to permit performances of these special tests and operations, which otherwise could not be performed if required to comply with the requirements of these TS. Unless otherwise specified, all the other TS requirements remain unchanged. This will ensure all appropriate requirements of the MODE or other specified condition not directly associated with or required to be changed to perform the special test or operation will remain in effect.

INDIAN POINT 2 B 3.0 - 10 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.7 (continued)

The Applicability of a Test Exception LCO represents a condition not necessarily in compliance with the normal requirements of the TS.

Compliance with Test Exception LCOs is optional. A special operation may be performed either under the provisions of the appropriate Test Exception LCO or under the other applicable TS requirements. If it is desired to perform the special operation under the provisions of the Test Exception LCO, the requirements of the Test Exception LCO shall be followed.

LCO 3.0.8 LCO 3.0.8 establishes conditions under which systems are considered to remain capable of performing their intended safety function when associated snubbers are not capable of providing their associated support function(s).

This LCO states that the supported system is not considered to be inoperable solely due to one or more snubbers not capable of performing their associated support function(s). This is appropriate because a limited length of time is allowed for maintenance, testing, or repair of one or more snubbers not capable of performing their associated support function(s) and appropriate compensatory measures are specified in the snubber requirements, which are located outside of the Technical Specifications (TS) under licensee control. The snubber requirements do not meet the criteria in 10 CFR 50.36(c)(2)(ii), and, as such, are appropriate for control by the licensee.

If the allowed time expires and the snubber(s) are unable to perform their associated support function(s), the affected supported systems LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2.

LCO 3.0.8.a applies when one or more snubbers are not capable of providing their associated support function(s) to a single train or subsystem of a multiple train or subsystem supported system or to a single train or subsystem supported system. LCO 3.0.8.a allows 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the snubber(s) before declaring the supported system inoperable. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function and due to the availability of the redundant train of the supported system.

INDIAN POINT 2 B 3.0 - 11 Revision 1

LCO Applicability B 3.0 BASES LCO 3.0.8 (continued)

LCO 3.0.8.b applies when one or more snubbers are not capable of providing their associated support function(s) to more than one train or subsystem of a multiple train or subsystem supported system. LCO 3.0.8.b allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to restore the snubber(s) before declaring the supported system inoperable. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function.

LCO 3.0.8 requires that risk be assessed and managed. Industry and NRC guidance on the implementation of 10 CFR 50.65(a)(4) (the Maintenance Rule) does not address seismic risk. However, use of LCO 3.0.8 should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issues are properly addressed. The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function.

INDIAN POINT 2 B 3.0 - 12 Revision 1

SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the MODES or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.

Systems and components are assumed to be OPERABLE when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that systems or components are OPERABLE when:

a. The systems or components are known to be inoperable, although still meeting the SRs; or

b. The requirements of the Surveillance(s) are known not to be met between required Surveillance performances.

facility Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs associated with a test exception are only applicable when the test exception is used as an allowable exception to the requirements of a Specification.

Unplanned events may satisfy the requirements (including applicable acceptance criteria) for a given SR. In this case, the unplanned event may be credited as fulfilling the performance of the SR. This allowance includes those SRs whose performance is normally precluded in a given MODE or other specified condition.

INDIAN POINT 2 B 3.0 - 13 Revision 1

to restore variables within their SR Applicability variables that are specified limits. B 3.0 outside their specified limits BASES SR 3.0.1 (continued)

Surveillances, including Surveillances invoked by Required Actions, do not have to be performed on inoperable equipment because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2, prior to returning equipment to OPERABLE status.

Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPERABLE. This includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance with SR 3.0.2. Post maintenance testing may not be possible in the current MODE or other specified conditions in the Applicability due to the necessary unit parameters not having been established. In these situations, the equipment may be considered OPERABLE provided testing has been satisfactorily completed to the extent possible and the equipment is not otherwise believed to be incapable of performing its function. This will allow operation to proceed to a MODE or other specified condition where other necessary post maintenance tests can be completed.

Some examples of this process are:

a. Auxiliary feedwater (AFW) pump turbine maintenance during refueling that requires testing at steam pressures > 800 psi. However, if other appropriate testing is satisfactorily completed, the AFW System can be considered OPERABLE. This allows startup and other necessary testing to proceed until the plant reaches the steam pressure required to perform the testing.

b. High pressure safety injection (HPI) maintenance during shutdown that requires system functional tests at a specified pressure. Provided other appropriate testing is satisfactorily completed, startup can proceed with HPI considered OPERABLE. This allows operation to reach the specified pressure to complete the necessary post maintenance testing.

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per . . ."

interval.

INDIAN POINT 2 B 3.0 - 14 Revision 1

SR Applicability B 3.0 facility BASES SR 3.0.2 (continued)

SR 3.0.2 permits a 25% extension of the interval specified in the Frequency.

This extension facilitates Surveillance scheduling and considers plant operating conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs. The exceptions to SR 3.0.2 are those Surveillances for which the 25%

extension of the interval specified in the Frequency does not apply. These exceptions are stated in the individual Specifications. The requirements of regulations take precedence over the TS. An example of where SR 3.0.2 does not apply is in the Containment Leakage Rate Testing Program. This program establishes testing requirements and Frequencies in accordance with the requirements of regulations. The TS cannot in and of themselves extend a test interval specified in the regulations. As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodic Completion Time that requires performance on a "once per ..." basis. The 25% extension applies to each performance after the initial performance.

The initial performance of the Required Action, whether it is a particular Surveillance or some other remedial action, is considered a single action with a single Completion Time. One reason for not allowing the 25%

extension to this Completion Time is that such an action usually verifies that no loss of function has occurred by checking the status of redundant or diverse components or accomplishes the function of the inoperable a equipment in an alternative manner.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as an operational convenience to extend Surveillance intervals (other than those consistent with refueling intervals) or periodic Completion Time intervals beyond those specified.

SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

INDIAN POINT 2 B 3.0 - 15 Revision 1

SR Applicability B 3.0 BASES SR 3.0.3 (continued)

This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance. facility The basis for this delay period includes consideration of unit conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

When a Surveillance with a Frequency based not on time intervals, but upon specified unit conditions, operating situations, or requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance with 10 CFR 50, Appendix J, as modified by approved exemptions, etc.) is discovered to not have been performed when specified, SR 3.0.3 allows for the full delay period of up to the specified Frequency to perform the Surveillance. However, since there is not a time interval specified, the missed Surveillance should be performed at the first reasonable opportunity.

SR 3.0.3 provides a time limit for, and allowances for the performance of, Surveillances that become applicable as a consequence of MODE changes imposed by Required Actions. a facility facility Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals. While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable facility opportunity should include consideration of the impact on plant risk (from delaying the Surveillance as well as any plant configuration changes required or shutting the plant down to perform the Surveillance) and impact on any analysis assumptions, in addition to unit conditions, planning, availability of personnel, and the time required to perform the Surveillance.

This risk impact should be managed through the program in place to implement 10 CFR 50.65(a)(4) and its implementation guidance, NRC Regulatory Guide 1.182, "Assessing and Managing Risk Before Maintenance Activities at Nuclear Power Plants." This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk INDIAN POINT 2 B 3.0 - 16 Revision 1

SR Applicability B 3.0 BASES SR 3.0.3 (continued) management action up to and including plant shutdown. The missed Surveillance should be treated as an emergent condition as discussed in the Regulatory Guide. The risk evaluation may use quantitative, qualitative, or blended methods. The degree of depth and rigor of the evaluation should be commensurate with the importance of the component. Missed Surveillances for important components should be analyzed quantitatively. If the results of the risk evaluation determine the risk increase is significant, this evaluation should be used to determine the safest course of action. All missed Surveillances will be placed in the Corrective Action Program.

If a Surveillance is not completed within the allowed delay period, then the equipment is considered inoperable or the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is inoperable, or the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1. variables ensure safe handling and storage of spent fuel SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability.

This Specification ensures that system and component OPERABILITY requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring systems or components to OPERABLE status before entering an associated MODE or other specified condition in the Applicability.

variables within specified limits A provision is included to allow entry into a MODE or other specified condition in the Applicability when an LCO is not met due to Surveillance not being met in accordance with LCO 3.0.4.

INDIAN POINT 2 B 3.0 - 17 Revision 1

SR Applicability B 3.0 variables that are outside their specified limits a variable is outside its BASES specified limit SR 3.0.4 (continued)

However, in certain circumstances, failing to meet an SR will not result in SR 3.0.4 restricting a MODE change or other specified condition change.

When a system, subsystem, division, component, device, or variable is inoperable or outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that surveillances do not have to be performed on inoperable equipment. When equipment is inoperable, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditions of the Applicability. However, since the LCO is not met in this instance, LCO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specified condition changes. SR 3.0.4 does not restrict changing MODES or other specified conditions of the Applicability when a Surveillance has not been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3.

The provisions of SR 3.0.4 shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. In addition, the provisions of SR 3.0.4 shall not prevent changes in MODES or other specified conditions in the Applicability that result from any unit shutdown. In this context, a unit shutdown is defined as a change in MODE or other specified condition in the Applicability associated with transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3, MODE 3 to MODE 4, and MODE 4 to MODE 5.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the MODE or other specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after entering the LCO's Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note, as not required (to be met or performed) until a particular event, condition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

INDIAN POINT 2 B 3.0 - 18 Revision 1

SPENT FUEL PIT Spent Fuel Pit Water Level REQUIREMENTS B 3.7.11 B 3.7 PLANT SYSTEMS B 3.7.11 Spent Fuel Pit Water Level BASES BACKGROUND The minimum water level in the spent fuel pit meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the spent fuel pit including the cooling and cleanup system is given in the UFSAR, Section 9.3 (Ref. 1). The assumptions of the fuel handling accident are given in Reference 2.

APPLICABLE The minimum water level in the spent fuel pit meets the assumptions of the SAFETY fuel handling accident evaluated. The radiological consequence analysis for ANALYSES a fuel handling accident in the fuel storage building assumes that a fuel assembly is dropped and damaged during refueling. The analysis methodology is consistent with Regulatory Guide 1.183 (Rev. 3).

Activity released from the damaged assembly is released to the outside atmosphere through the fuel-handling building ventilation system to the plant vent. No credit is taken for removal of iodine by filters, nor is credit taken for isolation of release paths. The activity released from the damaged assembly is assumed to be released to the environment over a 2-hour period. The fuel assembly fission product inventory is based on the assumption that the subject fuel assembly has been operated at 1.7 times core average power (and thus has 1.7 times the average fuel assembly fission product inventory). The decay time used in the analysis is 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . In accordance with this LCO, it is assumed that there is a minimum of 23 feet of water above the spent fuel racks. With this water depth, the decontamination factor (DF) of 500 specified by Reference 3 for elemental iodine would apply.

The DF was reduced to 400 for conservatism because the fuel rod pressure may exceed the assumption of 1200 psig (but would be less than 1500 psig).

The DF for organic iodine and noble gases was 1.0. The elemental iodine DF was reduced further to 285 in order to be consistent with Reference 3 guidance that the overall iodine DF be equal to 200. Since Reference 3 specifies that the 0.15% iodine is in the organic form, the limit of 200 for the overall iodine DF required that the DF for elemental iodine be 285.

INDIAN POINT 2 B 3.7.11 - 1 Revision 1

Spent Fuel Pit Water Level B 3.7.11 BASES APPLICABLE SAFETY ANALYSES (continued)

With 23 ft of water over the damaged fuel, the assumptions of Reference 3 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be

< 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

At Indian Point 2, the radiological consequence analyses for the fuel handling accident demonstrate compliance with the dose acceptance criterion in Reference 4.

The spent fuel pit water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii).

LCO The spent fuel pit water level is required to be 23 ft over the top of irradiated fuel assemblies seated in the storage racks. The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 2). As such, it is the minimum required for fuel storage and movement within the spent fuel pit.

The spent fuel pit minimum required level of 23 feet corresponds to an elevation of 92 feet, 2 inches.

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel pit, because the potential for a release of fission products exists.

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel pit water level is lower than the required level, the movement of irradiated fuel assemblies in the spent fuel pit is immediately suspended to INDIAN POINT 2 B 3.7.11 - 2 Revision 1

Spent Fuel Pit Water Level B 3.7.11 BASES ACTIONS (continued) a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not preclude movement of a fuel assembly to a safe position.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.11.1 REQUIREMENTS This SR verifies sufficient spent fuel pit water is available in the event of a fuel handling accident. The water level in the spent fuel pit must be checked periodically. The 7 day Frequency is appropriate because the volume in the spent fuel pit is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

During refueling operations, the level in the spent fuel pit is in equilibrium with the refueling canal, and the level in the refueling cavity is checked daily in accordance with SR 3.9.6.1.

REFERENCES 1. UFSAR, Section 9.3.

2. WCAP-16157, Indian Point Unit Stretch Power Uprating Licensing Report, January 2004.

3. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

4. NUREG-0800, Standard Review Plan, US Nuclear Regulatory Commission, Section 15.0.1, Radiological Consequences Analysis Using Alternative Source Terms, Rev. 0, July 2000.

INDIAN POINT 2 B 3.7.11 - 3 Revision 1

Spent Fuel Pit Boron Concentration SPENT FUEL PIT B 3.7.12 REQUIREMENTS B 3.7 PLANT SYSTEMS B 3.7.12 Spent Fuel Pit Boron Concentration BASES BACKGROUND The Spent Fuel Pit (SFP) is used to store spent fuel removed from the reactor and new fuel ready for insertion into the reactor. The SFP has been evaluated to meet the requirements of option (b) of 10 CFR 50.68, Criticality Accident Requirements (Ref. 1). IP2 compliance with 10 CFR 50.68(b)(4) was confirmed by an analysis documented in Northeast Technology Corporation report 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. 2). This analysis demonstrated 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 the following requirements are met:

a) Spent Fuel Pit boron concentration is maintained within the limits of LCO 3.7.12, Spent Fuel Pit Boron Concentration, whenever fuel is stored in the SFP; and, b) Fuel assembly storage location within the Spent Fuel Pit is restricted in accordance with LCO 3.7.13, Spent Fuel Pit Storage, based on the fuel assemblys initial enrichment, burnup, decay of Plutonium-241 (i.e., cooling time), and number of Integral Fuel Burnable Absorbers (IFBA) rods.

A detailed description of how this combination of minimum boron concentration and restrictions on fuel assembly storage location is presented in the Bases for LCO 3.7.13.

APPLICABLE NET-173-01, Criticality Analysis for Soluble Boron and Burnup Credit in the SAFETY Con Edison Indian Point Unit No. 2 Spent Fuel Storage Racks (Ref. 2)

ANALYSES evaluated non-accident conditions in the SFP including the affects of the projected boraflex degradation through the year 2006. Based upon BADGER testing in calendar years 2003, 2006 and 2010 and RACKLIFE code projections, the validity of the criticality and boron dilution analysis documented in References 2, 3, 5 and 6 can be extended through the end of the current license (September 28, 2013). Reference 7 allowed BADGER testing to be 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. Reference 2 determined that if storage location requirements in this LCO are met then the INDIAN POINT 2 B 3.7.12 - 1 Revision 2

Spent Fuel Pit Boron Concentration B 3.7.12 BASES LCO The Spent Fuel Pit boron concentration is required to be 2000 ppm. The specified concentration of dissolved boron in the Spent Fuel Pit preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference 2. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the Spent Fuel Pit.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the Spent Fuel Pit.

ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the Spent Fuel Pit is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies.

If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable. If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.12.1 REQUIREMENTS This SR verifies that the concentration of boron in the Spent Fuel Pit is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of Spent Fuel Pit water is expected to take place over such a short period of time.

INDIAN POINT 2 B 3.7.12 - 3 Revision 2

SPENT FUEL PIT Spent Fuel Pit Storage REQUIREMENTS B 3.7.13 B 3.7 Plant Systems B 3.7.13 Spent Fuel Pit Storage BASES BACKGROUND An issue has been identified with the degradation of boraflex used in the spent fuel pool to meet the licensing basis. To address this degradation Procedure 0-NF-203, Attachment 3, Transfer Form Checklist, discusses the administrative controls used to mitigate the effects of the boraflex degradation.

IP2 Fuel Assemblies The Spent Fuel Pit (SFP) is used to store spent fuel removed from the reactor and new fuel ready for insertion into the reactor. Spent fuel racks (SFRs) are erected on the SFP floor to hold the fuel assemblies. The SFRs have been evaluated to meet the requirements of option (b) of 10 CFR 50.68, Criticality Accident Requirements (Ref. 1) when: 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 in accordance with LCO 3.7.13, Spent Fuel Pit Storage, based on the fuel assemblys initial enrichment, burnup, decay of Plutonium-241 (i.e., cooling time), and number of Integral Fuel Burnable Absorbers (IFBA) rods.

In 1990, Spent Fuel Pit storage capacity was increased from 980 fuel assemblies to 1376 fuel assemblies by the installation of high-density racks that reduced the distance between adjacent fuel assemblies. This was possible because the k-effective of the SFP was maintained within the limits of 10 CFR 50.68(b) (Ref. 1) by the following: 1) the use of boraflex absorber panels (i.e., neutron absorbers) between SFR cells; and, 2) restrictions on fuel assembly storage location within the SFP based on initial enrichment and burnup. The original design of the high density racks met the requirements of 10 CFR 50.68(b) without crediting soluble boron.

The use of high-density SFRs that depend on boraflex absorber panels between cells requires that IP2 adhere to a long-term inspection program to monitor the performance of the boraflex panels. Requirements for the boraflex inspection program are specified in IP2 Amendment 150 (Ref. 2) and Generic Letter 96-04, Boraflex Degradation in Spent Fuel Pool Storage Racks (Ref. 3).

INDIAN POINT 2 B 3.7.13 - 1 Revision 7

Spent Fuel Pit Storage B 3.7.13 BASES LCO (continued) loading configuration (1 out of every two cells with every other cell vacant) and fuel assemblies meet the criteria of Figure 3.7.13-4.)

d. Fuel assemblies that satisfy requirements of Figure 3.7.13-4 may be stored as follows: 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 of Figure 3.7.13-5.

As shown on Figure 3.7.13-4, the maximum initial enrichment that can be stored in Region 1-2 with when there are no IFBA rods is 4.50 w/o U235.

Figure 3.7.13-4 does not provide any allowance from the minimum required IFBA rods based on the decay of Pu241.

The six peripheral cells may be used to store fuel that meets the requirements for storage in any location in the SFP (i.e., meets requirements for storage in Region 1-1, 1-2, 2-1 or 2-2). 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 canisters that are loaded in Module H in the southeast corner of the SFP. Module H is in the upper right corner of the SFP in Figure 3.7.13-5.

IP3 Fuel Assemblies This LCO establishes restrictions on fuel assembly storage location within the SFP to ensure that the requirements of 10 CFR 50.68 are met.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in the Spent Fuel Pit.

ACTIONS A.1 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the configuration of fuel assemblies stored in the Spent Fuel Pit is not in accordance with the rules established by LCO 3.7.13, the immediate action is to initiate action to make the necessary fuel assembly movement(s)

INDIAN POINT 2 B 3.7.13 - 9 Revision 7

Spent Fuel Pit Storage B 3.7.13 BASES ACTIONS (continued) to bring the configuration into compliance with the rules established by LCO 3.7.13.

If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation.

Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.13.1 REQUIREMENTS This SR verifies by administrative means that the IP2 fuel assembly has been classified based on initial enrichment, burnup, cooling time and number of Integral Fuel Burnable Absorbers (IFBA) rods in the fuel assembly in accordance with Figure 3.7.13-1, Figure 3.7.13-2, Figure 3.7.13-3, or Figure 3.7.13-4 and that the fuel assembly meets the requirements for the intended storage location defined on Figure 3.7.13-5. This SR also verifies by administrative means that the IP3 fuel assembly meets the requirements for storage in the IP2 SFP. This administrative verification must be completed prior to placing any fuel assembly in the SFP. This SR ensures that this LCO and Specification 4.3.1.1 will be met after the fuel assembly is inserted in the SFP.

REFERENCES 1. 10 CFR 50.68, Criticality Accident Requirements.

2. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 150 to Facility Operating License No.

DPR-26, April 19, 1990.

3. Generic Letter 96-04, Boraflex Degradation in Spent Fuel Pool Storage Racks.

4. Northeast Technology Corporation report NET-173-01, Criticality Analysis for Soluble Boron and Burnup Credit in the Con Edison Indian Point Unit No. 2 Spent Fuel Storage Racks.

5. Northeast Technology Corporation report NET-173-02, Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis.

6. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 227 to Facility Operating License No.

DPR-26, May 29, 2002.

INDIAN POINT 2 B 3.7.13 - 10 Revision 7

Enclosure, Attachment 2 NL-19-026 Indian Point Nuclear Generating Station Unit 2 Re-typed (Clean) Facility License, Appendix A Permanently Defueled Technical Specifications, Appendices B and C Technical Specifications, and Appendix A Permanently Defueled Technical Specifications Bases

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 2, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-247 INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 RENEWED FACILITY LICENSE Renewed License No. DPR-26

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

A. The application for a renewed license filed by Entergy Nuclear Indian Point 2, LLC (ENIP2) (the licensee) and Entergy Nuclear Operations, Inc. (ENO) (operator), for Indian Point Nuclear Generating Unit No. 2 at the Indian Point Energy Center (IPEC) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. Deleted per Amendment[###];

C. The facility will be maintained in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; D. There is reasonable assurance: (i) that the activities authorized by this renewed license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; E. ENO is technically and financially qualified and ENIP2 is financially qualified to engage in the activities authorized by this renewed license in accordance with the rules and regulations of the Commission; F. ENIP2 and ENO have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; G. The issuance of this renewed license will not be inimical to the common defense and security or to the health and safety of the public; Amendment No.

H. After weighing the environmental, economic, technical, and other benefits of the facility against environmental costs and considering available alternatives, the issuance of this renewed Facility License No. DPR-26, subject to the conditions for the protection of the environment set forth herein, is in accordance with 10 CFR Part 51, Appendix B, of the Commission's regulations and all applicable requirements of said Appendix B have been satisfied; I. The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this renewed license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70, including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31; and J. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1); and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by this renewed license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations.

2. Renewed Facility License No. DPR-26 is hereby issued to ENIP2 and ENO to read as follows:

A. This renewed license applies to the Indian Point Nuclear Generating Unit No. 2, a pressurized water nuclear reactor and associated equipment (the facility), which is owned by ENIP2 and maintained by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the Defueled Safety Analysis Report, as supplemented and amended, and the Environmental Report, as amended.

B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

(1) Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," (a) ENIP2 to possess and use, and (b) ENO to possess and use, the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; (2) ENO pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear material that was used as reactor fuel, in accordance with the limitations for storage, as described in the Defueled Safety Analysis Report, as supplemented and amended.

Amendment No.

(3) ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use, at any time any byproduct, source and special nuclear material as sealed neutron sources that were used for reactor startup, sealed sources that were used for reactor instrumentation and are used in the calibration of radiation monitoring equipment, and as fission detectors in amounts as required; Amdt. 42 (4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to 10-17-78 receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials that were produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Deleted per Amendment [###]

(2) Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. [###], are hereby incorporated in the renewed license. ENO shall maintain the facility in accordance with the Technical Specifications.

(3) Deleted per Amendment [###]

Amendment No.

D. (1) Deleted per Amdt. 82, 12-11-82.

(2) Deleted per Amendment 238.

E. Deleted per Amdt. 71, dated 8-5-81, effective 5-14-81.

F. This renewed license is also subject to appropriate conditions by the New York State Department of Environmental Conservation in its letter granting a Section 401 certification under the Federal Water Pollution Control Act amendments of 1972.

G. Deleted per Amendment [###]

H. ENO shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans1 for the Indian Point Energy Center, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: "Physical Security, Training and Qualification, and Safeguards Contingency Plan, Revision 0," and was submitted by letter dated October 14, 2004, as supplemented by letter dated May 18, 2006.

1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.

Amendment No.

ENO shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The ENO CSP was approved by License Amendment No. 266, as supplemented by changes approved by License Amendment Nos. 279, 284, and 286.

ENO has been granted Commission authorization to use "stand alone preemption authority" under Section 161A of the Atomic Energy Act, 42 U.S.C. 2201a with respect to the weapons described in Section II supplemented with Section Ill of Attachment 1 to its application submitted by letter dated August 20, 2013, as supplemented by letters dated November 21, 2013, and July 24, 2014, and citing letters dated April 27, 2011, and January 4, 2012. ENO shall fully implement and maintain in effect the provisions of the Commission-approved authorization.

I. Deleted per Amdt. 133, 7-6-88.

J. Deleted per Amdt. 133, 7-6-88.

K. Deleted per Amendment [###]

L. Deleted per Amendment 238 M. Deleted per Amendment 238 N. Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(a) Fire fighting response strategy with the following elements:

1. Pre-defined coordinated fire response strategy and guidance

2. Assessment of mutual aid fire fighting assets

3. Designated staging areas for equipment and materials

4. Command and control

5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:

1. Protection and use of personnel assets

2. Communications

3. Minimizing fire spread

4. Procedures for implementing integrated fire response strategy

5. Identification of readily-available pre-staged equipment

6. Training on integrated fire response strategy Amendment No.

(c) Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders O. Deleted per Amendment [###]

P. ENO may transfer IP3 spent fuel to the IP2 spent fuel pit subject to the conditions listed in Appendix C. ENO is further authorized to transfer IP3 spent fuel into NRC approved storage casks for onsite storage by ENO and Entergy Nuclear Indian Point 3, LLC.

Q. License Renewal License Conditions (1) The information in the UFSAR supplement, submitted pursuant to 10 CFR 54.21(d) and as revised during the license renewal application review process, and licensee commitments as listed in Appendix A of the Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Units 2 and 3, (SER) and supplements to the SER, are collectively the License Renewal UFSAR Supplement. The UFSAR Supplement is henceforth part of the UFSAR, which will be updated in accordance with 10 CFR 50.71(e). As such, the licensee may make changes to the programs, activities, and commitments described in the UFSAR Supplement, provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, Changes, Tests, and Experiments, and otherwise complies with the requirements in that section.

(2) The License Renewal UFSAR Supplement, as defined in license condition Q(1) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).

Amendment No.

a. The licensee shall implement those new programs and enhancements to existing programs no later than the date specified in the License Renewal UFSAR Supplement.

b. The licensee shall complete those activities no later than the date specified in the License Renewal UFSAR Supplement.

c. The licensee shall notify the NRC in writing within 30 days after having accomplished item (2)a above and include the status of those activities that have been or remain to be completed in item (2)b above.

3. On the closing date of the transfer of the license, Con Edison shall transfer to ENIP2 all of the accumulated decommissioning trust funds for IP2 and such additional funds to be deposited in the decommissioning trust for IP2 such that the total amount transferred for Indian Point Nuclear Generating Unit No. 1 (IP1) and IP2 is no less than $430,000,000.

Furthermore, ENIP2 shall either (a) establish a provisional trust for decommissioning funding assurance for IP1 and IP2 in an amount no less than $25,000,000 (to be updated as required under applicable NRC regulations, unless otherwise approved by the NRC) or (b) obtain a surety bond for an amount no less than $25,000,000 (to be updated as required under applicable NRC regulations, unless otherwise approved by the NRC). The total decommissioning funding assurance provided for IP2 by the combination of the decommissioning trust and the provisional trust or surety bond at the time of transfer of the licenses shall be at a level no less than the amounts calculated pursuant to, and required under, 10 CFR 50.75. The provisional trust and surety bond shall be subject to or be consistent with the following requirements, as applicable:

(a) Deleted (b) Provisional Trust:

(i) The provisional trust agreement must be in a form acceptable to the NRC.

(ii) Investments in the securities or other obligations of Entergy Corporation or its affiliates, subsidiaries, successors, or assigns are and shall be prohibited.

Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are and shall be prohibited.

(iii) The provisional trust agreement must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee unless the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment. The provisional trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.

(iv) The provisional trust agreement must provide that the agreement cannot be amended in any material respect, or terminated, without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

Amendment No.

(v) The appropriate section of the provisional trust agreement shall state that the trustee, investment advisor, or anyone else directing the investments made in the trust shall adhere to a "prudent investor" standard, as specified in 18 CFR 35.32(a)(3) of the Federal Energy Regulatory Commission's regulations.

(vi) Use of assets in the provisional trust, in the first instance, shall be limited to the expenses related to decommissioning IP2 or IP1 as defined by the NRC in its regulations and issuances, and as provided in this license and any amendments thereto.

(c) Surety Bond (i) The surety bond agreement must be in a form acceptable to the NRC and be in accordance with all applicable NRC regulations.

(ii) The surety company providing any surety bond obtained to comply with the requirements of the Order approving the transfer shall be one of those listed by the U.S. Department of the Treasury in the most recent edition of Circular 570 and shall have a coverage limit sufficient to cover the amount of the surety bond.

(iii) ENIP2 shall establish a standby trust to receive funds from the surety bond, if a surety bond is obtained, in the event that ENIP2 defaults on its funding obligations for the decommissioning of IP2. The standby trust agreement must be in a form acceptable to the NRC, and shall conform with all conditions otherwise applicable to the decommissioning trust agreement, and with all conditions that would be applicable to the provisional trust above, if established.

(iv) The surety agreement must provide that the agreement cannot be amended in any material respect, or terminated, without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

4. Deleted

5. ENIP2 and ENO shall take no action to cause Entergy Global Investments, Inc., or Entergy International Ltd. LLC or their parent companies to void, cancel, or modify the

$55 million contingency commitment to provide funding for the IP1 and IP2 plants as represented in the application without the prior written consent of the Director of the Office of Nuclear Reactor Regulation.

Amendment No.

6. This renewed license is effective as of the date of issuance, and until the Commission notifies the licensee in writing that the license is terminated.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Ho K. Nieh, Director Office of Nuclear Reactor Regulation Attachments:

Appendix A - Permanently Defueled Technical Specifications Appendix B - Environmental Technical Specification Requirements Appendix C - Inter-Unit Fuel Transfer Technical Specifications Date of Issuance: To be Determined Amendment No.

APPENDIX A TO FACILITY LICENSE DPR-26 FOR ENTERGY NUCLEAR INDIAN POINT 2, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING PLANT UNIT NO. 2 DOCKET NO. 50-247 PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS AND BASES Amendment No. ###

FACILITY LICENSE No. DPR-26 Appendix A - Permanently Defueled Technical Specifications Table of Contents 1.0 USE AND APPLICATION 1.1 Definitions 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency 2.0 DELETED 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.7 SPENT FUEL PIT REQUIREMENTS 3.7.11 Spent Fuel Pit Water Level 3.7.12 Spent Fuel Pit Boron Concentration 3.7.13 Spent Fuel Pit Storage 4.0 DESIGN FEATURES 4.1 Site Location 4.2 Deleted 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.2.1 Onsite and Offsite Organizations 5.2.2 Facility Staff 5.3 Facility Staff Qualifications 5.4 Procedures 5.5 Programs And Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) 5.5.2 Deleted 5.5.3 Radioactive Effluent Controls Program 5.5.4 Deleted 5.5.5 Deleted 5.5.6 Deleted 5.5.7 Deleted 5.5.8 Deleted 5.5.9 Deleted 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5.11 Deleted 5.5.12 Technical Specification (TS) Bases Control Program 5.5.13 Deleted 5.5.14 Deleted 5.5.15 Deleted 5.5.16 Deleted Indian Point 2 i Amendment No.

FACILITY LICENSE No. DPR-26 Appendix A - Permanently Defueled Technical Specifications Table of Contents (continued) 5.6 Reporting Requirements 5.6.1 Not Used 5.6.2 Annual Radiological Environmental Operating Report 5.6.3 Radioactive Effluent Release Report 5.7 High Radiation Area Indian Point 2 ii Amendment No.

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions

- NOTE -

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.

Term Definition ACTIONS ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.

CERTIFIED FUEL HANDLER A CERTIFIED FUEL HANDLER is an individual who complies with the provisions of the CERTIFIED FUEL HANDLER Training and Retraining Program required by TS 5.3.2.

NON-CERTIFIED OPERATOR A NON-CERTIFIED OPERATOR is a non-licensed operator who complies with the qualification requirements of Specification 5.3.1, but is not a CERTIFIED FUEL HANDLER.

Indian Point 2 1.1 - 1 Amendment No.

Logical Connectors 1.2 1.0 USE AND APPLICATION 1.2 Logical Connectors PURPOSE The purpose of this section is to explain the meaning of logical connectors.

Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Required Actions and Surveillances. The only logical connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings.

BACKGROUND Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action). The successive levels of logic are identified by additional digits of the Required Action number and by successive indentations of the logical connectors.

When logical connectors are used to state a Surveillance, only the first level of logic is used, and the logical connector is left justified with the statement of the Surveillance.

EXAMPLE The following example illustrates the use of logical connectors.

EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Verify . . .

AND A.2 Restore . . .

In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.

Indian Point 2 1.2 - 1 Amendment No.

Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.

BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe handling and storage of spent nuclear fuel. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s).

DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the facility is in a specified condition stated in the Applicability of the LCO. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the facility is not within the LCO Applicability.

EXAMPLE The following example illustrates the use of Completion Times with different Required Actions.

EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent Fuel Pit A.1 Suspend movement Immediately boron of fuel assemblies in concentration the Spent Fuel Pit not within limit AND A.2 Initiate action to Immediately restore Spent Fuel Pit boron concentration to within limits.

Indian Point 2 1.3 - 1 Amendment No.

Completion Times 1.3 1.3 Completion Times EXAMPLE (continued)

Condition A has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition A is entered.

The Required Actions of Condition A are to immediately suspend movement of fuel assemblies in the Spent Fuel Pit and initiate action to restore Spent Fuel Pit boron concentration to within limit.

IMMEDIATE When "Immediately" is used as a Completion Time, the Required COMPLETION TIME Action should be pursued without delay and in a controlled manner.

Indian Point 2 1.3 - 2 Amendment No.

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0.2, Surveillance Requirement (SR)

Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR.

The use of "met" or "performed" in these instances conveys specific meanings. A Surveillance is "met" only when the acceptance criteria are satisfied. Known failure of the requirements of a Surveillance, even without a Surveillance specifically being "performed," constitutes a Surveillance not "met." "Performance" refers only to the requirement to specifically determine the ability to meet the acceptance criteria.

EXAMPLE The following example illustrates the type of Frequency statement that appears in the Technical Specifications (TS).

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify level is within limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Example 1.4-1 contains the type of SR encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval.

Although the Frequency is stated as 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when a variable is outside specified limits, or the facility is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is Indian Point 2 1.4 - 1 Amendment No.

Frequency 1.4 1.4 Frequency EXAMPLE (continued) exceeded while the facility is in a specified condition in the Applicability of the LCO, then SR 3.0.3 becomes applicable.

If the interval as specified by SR 3.0.2 is exceeded while the facility is not in a specified condition in the Applicability of the LCO for which performance of the SR is required, then SR 3.0.4 becomes applicable.

The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3, prior to entry into the specified condition or the LCO is considered not met (in accordance with SR 3.0.1).

Indian Point 2 1.4 - 2 Amendment No.

DELETED 2.0 2.0 DELETED Indian Point 2 2.0 - 1 Amendment No.

LCO APPLICABILITY 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the specified conditions in the Applicability, except as provided in LCO 3.0.2.

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated.

Indian Point 2 3.0 - 1 Amendment No.

SR APPLICABILITY 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater.

This delay period is permitted to allow performance of the Surveillance.

A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

SR 3.0.4 Entry into a specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3.

Indian Point 2 3.0 - 2 Amendment No.

Spent Fuel Pit Water Level 3.7.11 3.7 SPENT FUEL PIT REQUIREMENTS 3.7.11 Spent Fuel Pit Water Level LCO 3.7.11 The Spent Fuel Pit water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the Spent Fuel Pit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent Fuel Pit water A.1 Suspend movement of Immediately level not within limit. irradiated fuel assemblies in the Spent Fuel Pit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify the Spent Fuel Pit water level is 23 ft above 7 days the top of the irradiated fuel assemblies seated in the storage racks.

Indian Point 2 3.7.11 - 1 Amendment No.

Spent Fuel Pit Boron Concentration 3.7.12 3.7 SPENT FUEL PIT REQUIREMENTS 3.7.12 Spent Fuel Pit Boron Concentration LCO 3.7.12 The Spent Fuel Pit boron concentration shall be 2000 ppm.

APPLICABILITY: When fuel assemblies are stored in the Spent Fuel Pit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent Fuel Pit boron A.1 Suspend movement of fuel Immediately concentration not within assemblies in the Spent limit. Fuel Pit.

AND A.2 Initiate action to restore Immediately Spent Fuel Pit boron concentration to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Verify the Spent Fuel Pit boron concentration is within 7 days limit.

Indian Point 2 3.7.12 - 1 Amendment No.

Spent Fuel Pit Storage 3.7.13 3.7 SPENT FUEL PIT REQUIREMENTS 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 may be 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 4.4 w/o U235 and discharged prior to IP3 Cycle 12 shall be stored in the Spent Fuel Pit. IP3 fuel assemblies V43 and V48 are not approved for storage in the Spent Fuel Pit.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Initiate action to move the Immediately LCO not met. noncomplying fuel assembly to an acceptable location.

Indian Point 2 3.7.13 - 1 Amendment No.

Spent Fuel Pit Storage 3.7.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify by administrative means that the IP2 fuel Prior to storing the assembly has been classified in accordance with fuel assembly in Figure 3.7.13-1, Figure 3.7.13-2, Figure 3.7.13-3, or the Spent Fuel Pit.

Figure 3.7.13-4 and meets the requirements for the intended storage location.

OR Verify by administrative means that the IP3 fuel Prior to storing the assembly meets the requirements for the intended fuel assembly in storage location. the Spent Fuel Pit.

Indian Point 2 3.7.13 - 2 Amendment No.

Spent Fuel Pit Storage 3.7.13 Cooling 5 Yr Cooling 10 Yr. Cooling 15Yr Cooling 20 Yr. Coollng Acceptable for Storage In Region 2-1 Not Acceptable for Storage In Region 2-1 10 0

1 2 3 4 5 Initial Enrichment, w/o U235 Figure 3.7.13-1 IP2 Fuel Assembly Limiting Burnup and Cooling Time versus Initial Enrichment:

Acceptable for Storage in Any Location in Region 2-1, Region 2-2, Region 1-2 or Region 1-1 Indian Point 2 3.7.13-3 Amendment No.

Spent Fuel Pit Storage 3.7.13 Burnup, GWD/MTU Initial Enrichment, w/o U235 Figure 3.7.13-2 IP2 Fuel Assembly Limiting Burnup and Cooling Time versus Initial Enrichment:

Acceptable for Storage in Any Location in Region 2-2, Region 1-2 or Region 1-1 Indian Point 2 3.7.13 - 4 Amendment No.

Spent Fuel Pit Storage 3.7.13 40

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~ 30 0

(9 Aa:eptable for Storage

a. In Region 1-1

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Not Acx:eptable for Storage 10 In Region 1-1 0

1 2 3 4 Initial Enrichment, w/o U 235 Figure 3.7.13-3 IP2 Fuel Assembly Limiting Burnup versus Initial Enrichment:

Acceptable for Storage in Any Location in Region 1-2, Region 1-1, or in locations designated as "peripheral" cells in Region 2-2 Indian Point 2 3.7.13-5 Amendment No.

Spent Fuel Pit Storage 3.7.13 Number of IFBA Rods Initial Enrichment, w/o U235 Figure 3.7.13-4 IP2 Fuel Assembly Minimum number of IFBA rods versus Initial Enrichment:

1) Acceptable for Storage in Any Location in Region 1-2, or

2) Acceptable for Storage in a checkerboard loading configuration in Region 1-1, or

3) Acceptable for Storage in locations designated as peripheral cells in Region 2-2 Indian Point 2 3.7.13 - 6 Amendment No.

Spent Fuel Pit Storage 3.7.13 J

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PERIPHERAL CELL Figure 3.7.13-5 Spent Fuel Pit Rack Layout Indian Point 2 3.7.13- 7 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location Indian Point 2 is located on the East bank of the Hudson River at Indian Point, Village of Buchanan, in upper Westchester County, New York. The site is approximately 24 miles north of the New York City boundary line. The nearest city is Peekskill which is 2.5 miles northeast of Indian Point.

The minimum distance from the reactor center line to the boundary of the site exclusion area and the outer boundary of the low population zone, as defined in 10 CFR 100.3, is 520 meters and 1100 meters, respectively. For the purpose of satisfying 10 CFR Part 20, the Restricted Area is the same as the Exclusion Area shown in the Defueled Safety Analysis Report (DSAR), Figure 2.2-2.

4.2 Deleted 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent,

b. keff < 1.0 if fully flooded with unborated water, and

c. Each fuel assembly classified based on initial enrichment, burnup, cooling time and number of Integral Fuel Burnable Absorbers (IFBA) rods with individual fuel assembly storage location within the spent fuel storage rack restricted as required by Technical Specification 3.7.13.

4.3.2 Drainage The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pit below a nominal elevation of 88 feet, 6 inches.

4.3.3 Capacity The spent fuel pit is designed and shall be maintained with a storage capacity limited to no more than 269 fuel assemblies in Region I and 1105 fuel assemblies in Region II.

Indian Point 2 4.0 - 1 Amendment No.

Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The plant manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.

The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

5.1.2 The shift manager (SM) shall be responsible for the shift command function.

Indian Point 2 5.1 - 1 Amendment No.

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for facility staff and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear fuel.

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all decommissioning organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements including the facility-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications shall be documented in the DSAR,

b. The plant manager shall be responsible for overall safe maintenance of the facility and shall have control over those onsite activities necessary for storage and maintenance of nuclear fuel.

c. The corporate officer with direct responsibility for IP2 shall have corporate responsibility for the safe storage and handling of nuclear fuel and shall take any measures needed to ensure acceptable performance of the staff in maintaining and providing technical support to the facility to ensure safe management of nuclear fuel, and

d. The individuals who train the CERTIFIED FUEL HANDLERS, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

5.2.2 Facility Staff The facility staff organization shall include the following:

a. Each duty shift shall be composed of at least one shift manager and one NON-CERTIFIED OPERATOR. The NON-CERTIFIED OPERATOR position may be filled by a CERTIFIED FUEL HANDLER.

At least one person qualified to stand watch in the control room (NON-CERTIFIED OPERATOR or CERTIFIED FUEL HANDLER) shall be present in the control room when nuclear fuel is stored in the spent fuel pool.

Indian Point 2 5.2 - 1 Amendment No.

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.2 Facility Staff (continued)

b. Shift crew composition may be less than the minimum requirement of 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements and all of the following conditions are met:

1) No fuel movements are in progress;

2) No movement of loads over fuel are in progress; and

3) No unmanned shift positions during shift turnover shall be permitted while the shift crew is less than the minimum.

c. An individual qualified in radiation protection procedures shall be on site during fuel handling operations and during movement of heavy loads over the fuel storage racks. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

d. Not Used.

e. The shift manager shall be a CERTIFED FUEL HANDLER.

f. Deleted.

Indian Point 2 5.2 - 2 Amendment No.

Facility Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Facility Staff Qualifications 5.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1-1978 for comparable positions with exceptions specified in the IPEC Quality Assurance Program Manual (QAPM).

5.3.2 An NRC approved training and retraining program for CERTIFIED FUEL HANDLERS shall be maintained.

Indian Point 2 5.3 - 1 Amendment No.

Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The procedures applicable to the safe storage of nuclear fuel recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 except as provided in the quality assurance program described or referenced in the DSAR;

b. Deleted;

c. Quality assurance for effluent and environmental monitoring;

d. Deleted;

e. All programs specified in Technical Specification 5.5; and

f. Personnel radiation protection consistent with the requirements of 10 CFR 20.

Indian Point 2 5.4 - 1 Amendment No.

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program, and

b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Technical Specification 5.6.2 and Technical Specification 5.6.3.

c. Licensee initiated changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

a. Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s) and

b. A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations,

2. Shall become effective after the approval of the plant manager, and

3. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

Indian Point 2 5.5 - 1 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.2 Deleted 5.5.3 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001 - 20.2402,

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit/facility to unrestricted areas, conforming to 10 CFR 50, Appendix I,

e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days.

Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days,

f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I,

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be in accordance with the following:

1. For noble gases: a dose rate 500 mrem/yr to the whole body and a dose rate 3000 mrem/yr to the skin and Indian Point 2 5.5 - 2 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Radioactive Effluent Controls Program (continued)

2. For iodine-131, tritium, and all radionuclides in particulate form with half-lives greater than 8 days: a dose rate 1500 mrem/yr to any organ,

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit/facility to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I,

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit/facility to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I, and

j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary, due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluent Controls Program surveillance frequency.

5.5.4 Deleted 5.5.5 Deleted 5.5.6 Deleted 5.5.7 Deleted 5.5.8 Deleted 5.5.9 Deleted 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, "Postulated Radioactive Release due to Waste Gas System Leak or Failure.

The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, "Postulated Radioactive Release due to Tank Failures.

Indian Point 2 5.5 - 3 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program (continued)

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion),

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents, and

c. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.11 Deleted 5.5.12 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or

2. A change to the DSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the DSAR.

Indian Point 2 5.5 - 4 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.12 Technical Specifications (TS) Bases Control Program (continued)

d. Proposed changes that meet the criteria of Technical Specification 5.5.12b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

Indian Point 2 5.5 - 5 Amendment No.

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Not Used 5.6.2 Annual Radiological Environmental Operating Report

- NOTE -

A single submittal may be made for a multiple unit/facility station. The submittal should combine sections common to all units/facilities at the station.

The Annual Radiological Environmental Operating Report covering the operation of the unit/facility during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

A full listing of the information to be contained in the Annual Radiological Environmental Operating Report is provided in the ODCM.

5.6.3 Radioactive Effluent Release Report

- NOTE -

A single submittal may be made for a multiple unit/facility station. The submittal shall combine sections common to all units/facilities at the station; however, for units/facilities with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit/facility.

The Radioactive Effluent Release Report covering the operation of the unit/facility in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit/facility. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

Indian Point 2 5.6 - 1 Amendment No.

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.

b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following facility radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously displays radiation dose rates in the area; or

2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or Indian Point 2 5.7 - 1 Amendment No.

High Radiation Area 5.7 5.7 High Radiation Area 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation (continued)

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:

1. All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or his or her designee.

2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.

Indian Point 2 5.7 - 2 Amendment No.

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued)

b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.

c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following facility radiation protection procedures for entry to, exit from, and work in such areas.

d. Each individual or group entering such an area shall possess:

1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or

2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.

Indian Point 2 5.7 - 3 Amendment No.

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued)

4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.

e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

Indian Point 2 5.7 - 4 Amendment No.

APPENDIX B TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 2, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNITS NUMBER 1 AND 2 ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION PLAN FACILITY LICENSES NO. DPR-5 AND DPR-26 DOCKET NUMBERS 50-3 AND 50-247 Renewed License Nos. DPR-5 and DPR-26

INDIAN POINT NUCLEAR GENERATING PLANT UNITS 1 AND 2 ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION PLAN TABLE OF CONTENTS Section Page 1.0 Objectives of the Environmental Protection Plan . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 2.0 Environmental Protection Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 3.0 Consistency Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.1 Plant Design and Operation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-1 3.2 Reporting Related to SPDES Permit. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3-2 3.3 Changes Required for Compliance with Other Environmental Regulations. . . . . . 3-3 4.0 Environmental Conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.1 Unusual or Important Environmental Events . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 4.2 Environmental Monitoring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-1 5.0 Administrative Procedures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.1 Review and Audit . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.2 Records Retention . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.3 Changes in Environmental Protection Plan . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-1 5.4 Plant Reporting Requirements . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-3 Renewed License Nos. DPR-5 and DPR-26

1.0 Objectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) is to provide for protection of environmental values during handling and storage of spent fuel and maintaince of the nuclear facility. The principal objectives of the EPP are as follows:

(1) Verify that the facility is maintained in an environmentally acceptable manner, as established by the FES and other NRC environmental impact assessments.

(2) Coordinate NRC requirements and maintain consistency with other Federal, State and local requirements for environmental protection.

(3) Keep NRC informed of the environmental effects of handling and storage of spent fuel and maintenance of the facility and of actions taken to control those effects.

Environmental concerns identified in the FES which relate to water quality matters are regulated by way of the licensee's SPDES permit.

1-1 Renewed License Nos. DPR-5 and DPR-26

2.0 Environmental Protection Issues In the FES-OL for Unit 2 dated September 1972 and the FES-OL for Unit 3 dated February 1975, the staff considered the environmental impacts associated with operation of the Indian Point Nuclear Generating Plant. Certain environmental issues were identified which required study or license conditions to resolve environmental concerns and to assure adequate protection of the environment. The Appendix B Environmental Technical Specifications issued with the licenses included monitoring programs and other requirements to protect water quality and aquatic biota during plant operation with once-through cooling. Appendix B ETS amendments, No. 30 Unit 1 and No. 70 Unit 2, issued on July 11, 1979, included monitoring and other requirements to address the following non-radiological aquatic protection issues:

(1) Controlled release of thermal discharges (ETS Sections 2.1, 3.1, 2.2.2, 3.2.2, and 4.1.1.a).

(2) Controlled release of non-radioactive chemical discharges (ETS Sections 2.3 and 3.3).

(3) Controlled intake flow velocity to limit impingement of organisms on intake structures (ETS Sections 2.2.1 and 3.2.1).

(4) Monitoring of aquatic biota in the Hudson River to evaluate effects of once-through operation (ETS Section 4.1.2). These issues were fully addressed in the SPDES permit with effective date of May 14, 1981 and were deleted from the ETS with Amendment No.

30 for Unit 1 and Amendment No. 70 for Unit 2.

2-1 Renewed License Nos. DPR-5 and DPR-26

Aquatic issues are now addressed by the effluent limitations, monitoring requirements and other conditions in or annexed to the effective SPDES permit issued by the Department of Environmental Conservation of the State of New York (DEC). The NRC will therefore rely on the DEC for regulation of matters involving water quality and aquatic biota and in the case of federally listed sturgeon, decisions made by the National Marine Fisheries Service (NMFS) under authority of the Endangered Species Act, for requirements pertaining to aquatic monitoring.

2-2 Renewed License Nos. DPR-5 and DPR-26

3.0 Consistency Requirements 3.1 Plant Design and Operation The licensee may make changes in facility design or operation or perform tests or experiments affecting the environment provided such changes, tests or experiments do not involve an unreviewed environmental question, and do not involve a change in the Environmental Protection Plan. 1 Changes in facility design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of the EPP. Activities governed by Section 3.3 are not subject to the requirements of this section.

Before engaging in additional construction or operational activities which may affect the environment, the licensee shall prepare and record an environmental evaluation of such activity.

When the evaluation indicates that such activity involves an unreviewed environmental question, the licensee shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation. When such activity involves a change in the Environmental Protection Plan, such activity and change to the Environmental Protection Plan may be implemented only in accordance with an appropriate license amendment as set forth in Section 5.3.

A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns (1 a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the final environmental statement 1

This provision does not relieve the licensee of the requirements of 10 CFR 50.59.

3-1 Renewed License Nos. DPR-5 and DPR-26

(FES) or final supplemental environmental impact statement (FSEIS), as modified by the staff's testimony to the Atomic Safety and Licensing Board, supplements to the FES, or FSEIS environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board, or the Commission; or (2) a significant change in effluents; or (3) a matter not previously reviewed and evaluated in the documents specified in (1) of this subsection which may have a significant adverse environmental impact.

The licensee shall maintain records of changes in facility design or operation and of tests and experiments carried out pursuant to this subsection. These records shall include a written evaluation which provides a basis for the determination that the change, test, or experiment does not involve an unreviewed environmental question nor constitute a decrease in the effectiveness of the EPP to meet the objectives specified in Section 1.0. The licensee shall include as part of its Annual Environmental Protection Plan Report (per subsection 5.4.1) brief descriptions, analyses, interpretations, and evaluations of such changes, tests and experiments.

3.2 Reporting Related to the SPDES Permit Violations of the SPDES permit shall be reported to the NRC by submittal of copies of the reports required by the SPDES permit.

Changes and additions to the SPDES permit shall be reported to the NRC within 30 days following the date the change is approved. If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted.

3-2 Renewed License Nos. DPR-5 and DPR-26

The NRC shall be notified of the changes to the effective SPDES permit proposed by the licensee by providing NRC with a copy of the proposed change at the same time it is submitted to the permitting agency. The notification of a licensee-initiated change shall include a copy of the requested revision submitted to the permitting agency. The licensee shall provide the NRC a copy of the application for renewal of the SPDES permit at the same time the application is submitted to the permitting agency.

3.3 Changes Required for Compliance with Other Environmental Regulations Changes in facility design or operation and performance of tests or experiments which are required to achieve compliance with other Federal, State, or local environmental regulations are not subject to the requirements of Section 3.1.

3-3 Renewed License Nos. DPR-5 and DPR-26

4.0 Environmental Conditions 4.1 Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to the handling and storage of spent fuel and maintenance of the facility shall be recorded and promptly reported to the NRC within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by telephone, telegraph or facsimile transmissions followed by a written report per subsection 5.4.2. The following are examples: excessive bird impaction events, onsite plant or animal disease outbreaks, unusual mortality or occurrence of any species protected by the Endangered Species Act of 1973, unusual fish kills, unusual increase in nuisance organisms or conditions, and unanticipated or emergency discharge of waste water or chemical substances.

No routine monitoring programs are required to implement this condition.

4.2 Environmental Monitoring In accordance with Section 7(a) of the Endangered Species Act, the National Marine Fisheries Service (NMFS) issued a Biological Opinion related to the continued operation of IP2 and IP3 that pertains to shortnose sturgeon (Acipenser brevirostrum) and Atlantic sturgeon (Acipenser oxyrinchus oxyrinchus). The Biological Opinion includes an Incidental Take Statement with Reasonable and Prudent Measures that the NMFS has determined to be necessary or appropriate to minimize the amount or extent of incidental take and associated Terms and Conditions, which are non-discretionary and implement the Reasonable and Prudent Measures. The currently applicable Biological Opinion concludes that continued operation of IP2 and IP3 is not likely to jeopardize the continued existence of the listed species or to adversely affect the designated critical habitat of those species. This Biological Opinion conservatively bounds the conditions that will occur in the permanently shut down and defueld condition.

4-1 Renewed License Nos. DPR-5 and DPR-26

Entergy shall adhere to the requirements within the Incidental Take Statement of the currently applicable Biological Opinion. Changes to the Biological Opinion, including the Incidental Take Statement, Reasonable and Prudent Measures, and Terms and Conditions contained therein, must be preceded by consultation between the NRC, as the authorizing agency, and the NMFS.

4-2 Renewed License Nos. DPR-5 and DPR-26

5.0 Administrative Procedures 5.1 Review and Audit The licensee shall provide for review and audit of compliance with the Environmental Protection Plan. The audits shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure utilized to achieve the independent review and audit function and results of the audit activities shall be maintained and made available for inspection.

5.2 Records Retention Records and logs relative to the environmental aspects of previous plant operation and the handling and storage of spent fuel and maintenance of the facility shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to the NRC on request.

Records of modifications to facility structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the facility. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.

5.3 Changes in Environmental Protection Plan Requests for changes in the Environmental Protection Plan shall include an assessment of the environmental impacts of the proposed change and a supporting justification. Implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes 5-1 Renewed License Nos. DPR-5 and DPR-26

in the form of a license amendment incorporating the appropriate revision to the Environmental Protection Plan.

5-2 Renewed License Nos. DPR-5 and DPR-26

5.4 Plant Reporting Requirements 5.4.1 Routine Reports An Annual Environmental Protection Plan Report describing implementation of this EPP for the previous year shall be submitted to the NRC prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following issuance of the operating license. The period of the first report shall begin with the date of issuance of the operating license.

The Annual Environmental Protection Plan Report shall include:

(a) A list of EPP noncompliances and the corrective actions taken to remedy them.

(b) A list of all changes in facility design or operation, tests, and experiments made in accordance with subsection 3.1 which involved a potentially significant unreviewed environmental issue.

(c) A list of nonroutine reports submitted in accordance with subsection 5.4.2.

(d) A list of all reports submitted in accordance with the SPDES permit.

In the event that some results are not available by the report due date, the report shall be submitted noting and explaining the missing results. The missing data shall be submitted as soon possible in a supplementary report.

5-3 Renewed License Nos. DPR-5 and DPR-26

5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of a nonroutine event. The report shall (1) describe, analyze, and evaluate the event, including extent and magnitude of the impact and facility conditions, (2) describe the probable cause of the event, (3) indicate the action taken to correct the reported event, (4) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (5) indicate the agencies notified and their preliminary responses.

Events reportable under this subsection which also require reports to other Federal, State or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided a copy of such report at the same time it is submitted to the other agency.

5-4 Renewed License Nos. DPR-5 and DPR-26

APPENDIX C TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 2, LLC (ENIP2)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 2 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART I: SPENT FUEL TRANSFER CANISTER AND TRANSFER CASK SYSTEM FACILITY LICENSE NO. DPR-26 DOCKET NO. 50-247 Amendment No.

Facility License Appendix C - Inter-Unit Fuel Transfer Technical Specifications SPENT FUEL SHIELDED TRANSFER CANISTER AND TRANSFER CASK SYSTEM

1.0 DESCRIPTION

The spent fuel transfer system consists of the following components: (1) a spent fuel shielded transfer canister (STC), which contains the fuel; (2) a transfer cask (HI-TRAC 100D) (hereafter referred to as HI-TRAC), which contains the STC during transfer operations; and (3) a bottom missile shield.

The STC and HI-TRAC are designed to transfer irradiated nuclear fuel assemblies from the Indian Point 3 (IP3) spent fuel pit to the Indian Point 2 (IP2) spent fuel pit. A fuel basket within the STC holds the fuel assemblies and provides criticality control. The shielded transfer canister provides the confinement boundary, water retention boundary, gamma radiation shielding, and heat rejection capability. The HI-TRAC provides a water retention boundary, protection of the STC, gamma and neutron radiation shielding, and heat rejection capability. The STC contains up to 12 fuel assemblies.

The STC is the confinement system for the fuel. It is a welded, multi-layer steel and lead cylinder with a welded base-plate and bolted lid. The inner shell of the canister forms an internal cylindrical cavity for housing the fuel basket. The outer surface of the canister inner shell is buttressed with lead and steel shells for radiation shielding. The minimum thickness of the steel, lead and steel shells relied upon for shielding starting with the innermost shell are 3/4 inch steel, 2 3/4 inch lead and 3/4 inch steel, respectively. The canister closure incorporates two O-ring seals to ensure its confinement function. The confinement system consists of the canister inner shell, bottom plate, top flange, top lid, top lid O-ring seals, vent port seal and cover plate, and drain port seal and coverplate. The fuel basket, for the transfer of 12 Pressurized Water Reactor (PWR) fuel assemblies, is a fully welded, stainless steel, honeycomb structure with neutron absorber panels attached to the individual storage cell walls under stainless steel sheathing.

The maximum gross weight of the fully loaded STC is 40 tons.

The HI-TRAC is a multi-layer steel and lead cylinder with a bolted bottom (or pool) and top lid.

For the fuel transfer operation the HI-TRAC is fitted with a solid top lid, an STC centering assembly, and a bottom missile shield. The inner shell of the transfer cask forms an internal cylindrical cavity for housing the STC. The outer surface of the cask inner shell is buttressed with intermediate lead and steel shells for radiation shielding. The minimum thickness of the steel, lead and steel shells relied upon for shielding starting with the innermost shell are 3/4 inch steel, 2 % inch lead and 1 inch steel, respectively. An outside shell called the water jacket contains water for neutron shielding, with a minimum thickness of 5. The HI-TRAC bottom and top lids incorporate a gasket seal design to ensure its water confinement function. The water confinement system consists of the HI-TRAC inner shell, bottom lid, top lid, top lid seal, bottom lid seal, vent port seal, vent port cap and bottom drain plug.

The HI-TRAC provides a water retention boundary, protection of the STC, gamma and neutron radiation shielding, and heat rejection capability. The bottom missile shield is attached to the bottom of the HI-TRAC and provides tornado missile protection of the pool lid bolted joint. The HI-TRAC can withstand a tornado missile in other areas without the need for additional shielding. The STC centering assembly provides STC position control within the HI-TRAC and also acts as an internal impact limiter in the event of a non-mechanistic tipover accident.

INDIAN POINT 2 1 Amendment

Facility License Appendix C - Inter-Unit Fuel Transfer Technical Specifications 2.0 CONDITIONS 2.1 OPERATING PROCEDURES Written operating procedures shall be prepared for cask handling, loading, movement, surveillance, maintenance, and recovery from off normal conditions such as crane hang-up. The written operating procedures shall be consistent with the technical basis described in Chapter 10 of the Licensing Report (Holtec International Report HI-2094289).

2.2 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM Written cask acceptance tests and maintenance program shall be prepared consistent with the technical basis described in Chapter 8 of the Licensing Report (Holtec International Report HI-2094289).

2.3 PRE-OPERATIONAL TESTING AND TRAINING EXERCISE A training exercise of the loading, closure, handling/transfer, and unloading, of the equipment shall be conducted prior to the first transfer. The training exercise shall not be conducted with irradiated fuel. The training exercise may be performed in an alternate step sequence from the actual procedures, but all steps must be performed. The training exercise shall include, but is not limited to the following:

a) Moving the STC into the IP3 spent fuel pool.

b) Preparation of the HI-TRAC for STC loading.

c) Selection and verification of specific fuel assemblies and non-fuel hardware to ensure type conformance.

d) Loading specific assemblies and placing assemblies into the STC (using a single dummy fuel assembly), including appropriate independent verification.

e) Remote installation of the STC lid and removal of the STC from the spent fuel pool.

f) Placement of the STC into the HI-TRAC with the STC centering assembly.

g) STC closure, establishment of STC water level with steam, verification of STC water level, STC leakage testing, and operational steps required prior to transfer, as applicable.

h) Establishment and verification of HI-TRAC water level.

i) Installation of the HI-TRAC top lid.

j) HI-TRAC closure, leakage testing, and operational steps required prior to transfer, as applicable.

k) Movement of the HI-TRAC with STC from the IP3 fuel handling building to the IP2 fuel handling building along the haul route with designated devices.

l) Moving the STC into the IP2 spent fuel pool.

m) Manual crane operations for bare STC movements including demonstration of recovery from a crane hang-up with the STC suspended from the crane.

INDIAN POINT 2 2 Amendment

Facility License Appendix C - Inter-Unit Fuel Transfer Technical Specifications APPENDIX C TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 2, LLC (ENIP2)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 2 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART II: TECHNICAL SPECIFICATIONS FACILITY LICENSE NO. DPR-26 DOCKET NO. 50-247 Amendment No.

Facility License No. DPR-26 Appendix A - Permanently Defueled Technical Specifications Bases TABLE OF CONTENTS B.2.0 DELETED B.3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY B.3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY B.3.7 SPENT FUEL PIT REQUIREMENTS B.3.7.11 Spent Fuel Pit Water Level B.3.7.12 Spent Fuel Pit Boron Concentration B.3.7.13 Spent Fuel Pit Storage Indian Point 2 i Revision

LCO APPLICABILITY B 3.0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCOs LCO 3.0.1 through LCO 3.0.2 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

LCO 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the facility is in the specified conditions of the Applicability statement of each Specification).

LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that:

a. Completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification and

b. Completion of the Required Actions is not required when an LCO is met within the specified Completion Time, unless otherwise specified.

Completing the Required Actions is not required when an LCO is met or is no longer applicable, unless otherwise stated in the individual Specifications.

Indian Point 2 B 3.0 - 1 Revision

SR Applicability B 3.0 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SRs SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during the specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.

Surveillances do not have to be performed when the facility is in a specified condition for which the requirements of the associated LCO are not applicable.

Surveillances do not have to be performed on variables that are outside their specified limits because the ACTIONS define the remedial measures that apply. Surveillances have to be met and performed in accordance with SR 3.0.2 to restore variables within their specified limits.

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances.

SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers facility conditions that may not be suitable for conducting the Surveillance (e.g., other ongoing Surveillance or maintenance activities).

The 25% extension does not significantly degrade the reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs.

The provisions of SR 3.0.2 are not intended to be used repeatedly merely as a convenience to extend Surveillance intervals beyond those specified.

Indian Point 2 B 3.0 - 2 Revision

SR Applicability B 3.0 BASES SR 3.0.3 SR 3.0.3 establishes the flexibility to defer an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

This delay period provides adequate time to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance.

The basis for this delay period includes consideration of facility conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the requirements.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as a convenience to extend Surveillance intervals. While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity. The determination of the first reasonable opportunity should include consideration of the impact on facility risk (from delaying the Surveillance as well as any facility configuration changes required to perform the Surveillance) and impact on any analysis assumptions, in addition to facility conditions, planning, availability of personnel, and the time required to perform the Surveillance.

All missed Surveillances will be placed in the Corrective Action Program.

If a Surveillance is not completed within the allowed delay period, then the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable LCO Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.

Indian Point 2 B 3.0 - 3 Revision

SR Applicability B 3.0 Bases SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a specified condition in the Applicability.

This Specification ensures that variable limits are met before entry into specified conditions in the Applicability for which these variables ensure safe handling and storage of spent fuel. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring variables within specified limits before entering an associated specified condition in the Applicability.

However, in certain circumstances, failing to meet an SR will not result in SR 3.0.4 restricting a specified condition change. When a variable is outside its specified limits, the associated SR(s) are not required to be performed, per SR 3.0.1, which states that surveillances do not have to be performed on variables that are outside their specified limits. When a variable is outside its specified limit, SR 3.0.4 does not apply to the associated SR(s) since the requirement for the SR(s) to be performed is removed. Therefore, failing to perform the Surveillance(s) within the specified Frequency does not result in an SR 3.0.4 restriction to changing specified conditions of the Applicability. SR 3.0.4 does not restrict changing specified conditions of the Applicability when a Surveillance has not been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance.

A Surveillance that could not be performed until after entering the LCO's Applicability, would have its Frequency specified such that it is not "due" until the specific conditions needed are met. Alternately, the Surveillance may be stated in the form of a Note, as not required (to be met or performed) until a particular event, condition, or time has been reached.

Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequency.

Indian Point 2 B 3.0 - 4 Revision

Spent Fuel Pit Water Level B 3.7.11 B 3.7 SPENT FUEL PIT REQUIREMENTS B 3.7.11 Spent Fuel Pit Water Level BASES BACKGROUND The minimum water level in the spent fuel pit meets the assumptions of iodine decontamination factors following a fuel handling accident. The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel.

A general description of the spent fuel pit including the cooling and cleanup system is given in the UFSAR, Section 9.3 (Ref. 1). The assumptions of the fuel handling accident are given in Reference 2.

APPLICABLE The minimum water level in the spent fuel pit meets the assumptions of SAFETY the fuel handling accident evaluated. The radiological consequence ANALYSES analysis for a fuel handling accident in the fuel storage building assumes that a fuel assembly is dropped and damaged during refueling. The analysis methodology is consistent with Regulatory Guide 1.183 (Rev. 3).

Activity released from the damaged assembly is released to the outside atmosphere through the fuel-handling building ventilation system to the plant vent. No credit is taken for removal of iodine by filters, nor is credit taken for isolation of release paths. The activity released from the damaged assembly is assumed to be released to the environment over a 2-hour period. The fuel assembly fission product inventory is based on the assumption that the subject fuel assembly has been operated at 1.7 times core average power (and thus has 1.7 times the average fuel assembly fission product inventory). The decay time used in the analysis is 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . In accordance with this LCO, it is assumed that there is a minimum of 23 feet of water above the spent fuel racks. With this water depth, the decontamination factor (DF) of 500 specified by Reference 3 for elemental iodine would apply. The DF was reduced to 400 for conservatism because the fuel rod pressure may exceed the assumption of 1200 psig (but would be less than 1500 psig). The DF for organic iodine and noble gases was 1.0. The elemental iodine DF was reduced further to 285 in order to be consistent with Reference 3 guidance that the overall iodine DF be equal to 200. Since Reference 3 specifies that the 0.15% iodine is in the organic form, the limit of 200 for the overall iodine DF required that the DF for elemental iodine be 285.

Indian Point 2 B 3.7.11 - 1 Revision

Spent Fuel Pit Water Level B 3.7.11 BASES APPLICABLE SAFETY ANALYSES (continued)

With 23 ft of water over the damaged fuel, the assumptions of Reference 3 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks. In the case of a single bundle dropped and lying horizontally on top of the spent fuel racks, however, there may be < 23 ft of water above the top of the fuel bundle and the surface, indicated by the width of the bundle. To offset this small nonconservatism, the analysis assumes that all fuel rods fail, although analysis shows that only the first few rows fail from a hypothetical maximum drop.

At Indian Point 2, the radiological consequence analyses for the fuel handling accident demonstrate compliance with the dose acceptance criterion in Reference 4.

The spent fuel pit water level satisfies Criteria 2 and 3 of 10 CFR 50.36(c)(2)(ii).

LCO The spent fuel pit water level is required to be 23 ft over the top of irradiated fuel assemblies seated in the storage racks. The specified water level preserves the assumptions of the fuel handling accident analysis (Ref. 2). As such, it is the minimum required for fuel storage and movement within the spent fuel pit.

The spent fuel pit minimum required level of 23 feet corresponds to an elevation of 92 feet, 2 inches.

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel pit, because the potential for a release of fission products exists.

ACTIONS A.1 When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring. When the spent fuel pit water level is lower than the required level, the movement of irradiated fuel assemblies in the spent fuel pit is immediately suspended to a safe position. This action effectively precludes the occurrence of a fuel handling accident. This does not preclude movement of a fuel assembly to a safe position.

Indian Point 2 B 3.7.11 - 2 Revision

Spent Fuel Pit Water Level B 3.7.11 SURVEILLANCE SR 3.7.11.1 REQUIREMENTS This SR verifies sufficient spent fuel pit water is available in the event of a fuel handling accident. The water level in the spent fuel pit must be checked periodically. The 7 day Frequency is appropriate because the volume in the spent fuel pit is normally stable. Water level changes are controlled by plant procedures and are acceptable based on operating experience.

REFERENCES 1. UFSAR, Section 9.3.

2. WCAP-16157, Indian Point Unit Stretch Power Uprating Licensing Report, January 2004.

3. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, July 2000.

4. NUREG-0800, Standard Review Plan, US Nuclear Regulatory Commission, Section 15.0.1, Radiological Consequences Analysis Using Alternative Source Terms, Rev. 0, July 2000.

Indian Point 2 B 3.7.11 - 3 Revision

Spent Fuel Pit Boron Concentration B 3.7.12 B 3.7 SPENT FUEL PIT REQUIREMENTS B 3.7.12 Spent Fuel Pit Boron Concentration BASES BACKGROUND The Spent Fuel Pit (SFP) is used to store spent fuel removed from the reactor and new fuel ready for insertion into the reactor. The SFP has been evaluated to meet the requirements of option (b) of 10 CFR 50.68, Criticality Accident Requirements (Ref. 1). IP2 compliance with 10 CFR 50.68(b)(4) was confirmed by an analysis documented in Northeast Technology Corporation report 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. 2). This analysis demonstrated 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 the following requirements are met:

a) Spent Fuel Pit boron concentration is maintained within the limits of LCO 3.7.12, Spent Fuel Pit Boron Concentration, whenever fuel is stored in the SFP; and, b) Fuel assembly storage location within the Spent Fuel Pit is restricted in accordance with LCO 3.7.13, Spent Fuel Pit Storage, based on the fuel assemblys initial enrichment, burnup, decay of Plutonium-241 (i.e., cooling time), and number of Integral Fuel Burnable Absorbers (IFBA) rods.

A detailed description of how this combination of minimum boron concentration and restrictions on fuel assembly storage location is presented in the Bases for LCO 3.7.13.

APPLICABLE NET-173-01, Criticality Analysis for Soluble Boron and Burnup Credit in SAFETY the Con Edison Indian Point Unit No. 2 Spent Fuel Storage Racks ANALYSES (Ref. 2) evaluated non-accident conditions in the SFP including the affects of the projected boraflex degradation through the year 2006.

Based upon BADGER testing in calendar years 2003, 2006 and 2010 and RACKLIFE code projections, the validity of the criticality and boron dilution analysis documented in References 2, 3, 5 and 6 can be extended through the end of the current license (September 28, 2013).

Reference 7 allowed BADGER testing to be 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.

Reference 2 determined that if storage location requirements in this LCO are met then the Indian Point 2 B 3.7.12 - 1 Revision

Spent Fuel Pit Boron Concentration B 3.7.12 BASES APPLICABLE SAFETY ANALYSES (continued)

SFP will have a keff of 0.95 if filled with a soluble boron concentration of 786 ppm and will have a keff of < 1.0 if filled with unborated water.

Reference 2 also evaluated credible abnormal occurrences in accordance with ANSI/ANS-57.2-1983. This evaluation considered the effects of the following: a) a dropped fuel assembly or an assembly placed alongside a rack; b) a misloaded fuel assembly; and, c) abnormal heat loads.

Reference 2 determined that the SFP will maintain a keff of 0.95 under the worst-case accident scenario if the SFP is filled with a soluble boron concentration of 1495 ppm.

NET-173-02, Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis (Ref. 3) evaluated postulated unplanned SFP boron dilution scenarios assuming an initial SFP boron concentration within the limits of LCO 3.7.12. The evaluation considered various scenarios by which the SFP boron concentration may be diluted and the time available before the minimum boron concentration necessary to ensure subcriticality for the non-accident condition (i.e. it is not assumed an assembly is misloaded concurrent with the Spent Fuel Pit dilution event). Reference 3 determined that an unplanned or inadvertent event that could dilute the SFP boron concentration from 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.

References 2 and 3 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. Based upon BADGER testing in calendar years 2003, 2006 and 2010 and RACKLIFE code projections, the validity of the criticality and boron dilution analysis documented in References 2, 3, 5 and 6 can be extended through the end of the current license (September 28, 2013). Reference 7 allowed BADGER testing to be 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. These compensatory measures for boraflex degradation in the SFP were evaluated by the NRC in Reference 4.

The concentration of dissolved boron in the spent fuel pit satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

Indian Point 2 B 3.7.12 - 2 Revision

Spent Fuel Pit Boron Concentration B 3.7.12 BASES LCO The Spent Fuel Pit boron concentration is required to be 2000 ppm. The specified concentration of dissolved boron in the Spent Fuel Pit preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference 2. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the Spent Fuel Pit.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the Spent Fuel Pit.

ACTIONS A.1 and A.2 When the concentration of boron in the Spent Fuel Pit is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies.

SURVEILLANCE SR 3.7.12.1 REQUIREMENTS This SR verifies that the concentration of boron in the Spent Fuel Pit is within the required limit. As long as this SR is met, the analyzed accidents are fully addressed. The 7 day Frequency is appropriate because no major replenishment of Spent Fuel Pit water is expected to take place over such a short period of time.

Indian Point 2 B 3.7.12 - 3 Revision

Spent Fuel Pit Boron Concentration B 3.7.12 BASES REFERENCES 1. 10 CFR 50.68, Criticality Accident Requirements.

2. Northeast Technology Corporation report NET-173-01, Criticality Analysis for Soluble Boron and Burnup Credit in the Con Edison Indian Point Unit No. 2 Spent Fuel Storage Racks.

3. Northeast Technology Corporation report NET-173-02, Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis.

4. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 227 to Facility Operating License No.

DPR-26, May 29, 2002.

5. NETCO Letter to M. R. Hansler from E. Lindquist, Northeast Technology Corp. dated 12/19/06, Subject - Reference 2 and 3 extension.

6. NETCO Letter to Floyd Gumble from Matt Harris dated 12/22/2009, titled Indian Point 2 RACKLIFE Projections Through 2010 and 2012 BADGER Tests with RACKLIFE Version 2.1

7. 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).

Indian Point 2 B 3.7.12 - 4 Revision

Spent Fuel Pit Storage B 3.7.13 B 3.7 SPENT FUEL PIT REQUIREMENTS B 3.7.13 Spent Fuel Pit Storage BASES BACKGROUND An issue has been identified with the degradation of boraflex used in the spent fuel pool to meet the licensing basis. To address this degradation Procedure 0-NF-203, Attachment 3, Transfer Form Checklist, discusses the administrative controls used to mitigate the effects of the boraflex degradation.

IP2 Fuel Assemblies The Spent Fuel Pit (SFP) is used to store spent fuel removed from the reactor and new fuel ready for insertion into the reactor. Spent fuel racks (SFRs) are erected on the SFP floor to hold the fuel assemblies. The SFRs have been evaluated to meet the requirements of option (b) of 10 CFR 50.68, Criticality Accident Requirements (Ref. 1) when: 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 in accordance with LCO 3.7.13, Spent Fuel Pit Storage, based on the fuel assemblys initial enrichment, burnup, decay of Plutonium-241 (i.e., cooling time), and number of Integral Fuel Burnable Absorbers (IFBA) rods.

In 1990, Spent Fuel Pit storage capacity was increased from 980 fuel assemblies to 1376 fuel assemblies by the installation of high-density racks that reduced the distance between adjacent fuel assemblies. This was possible because the k-effective of the SFP was maintained within the limits of 10 CFR 50.68(b) (Ref. 1) by the following: 1) the use of boraflex absorber panels (i.e., neutron absorbers) between SFR cells; and, 2) restrictions on fuel assembly storage location within the SFP based on initial enrichment and burnup. The original design of the high density racks met the requirements of 10 CFR 50.68(b) without crediting soluble boron.

The use of high-density SFRs that depend on boraflex absorber panels between cells requires that IP2 adhere to a long-term inspection program to monitor the performance of the boraflex panels. Requirements for the boraflex inspection program are specified in IP2 Amendment 150 (Ref. 2) and Generic Letter 96-04, Boraflex Degradation in Spent Fuel Pool Storage Racks (Ref. 3).

Indian Point 2 B 3.7.13 - 1 Revision

Spent Fuel Pit Storage B 3.7.13 BASES BACKGROUND (continued)

During an inspection of the SFRs in 2000, it was determined that the assumptions regarding the boraflex panels used in the criticality analysis for the SFP were no longer valid because of thinning and gaps in the boraflex panels. This degradation of the boraflex panels between SFR cells required that IP2 adopt the use of soluble boron option in 10 CFR 50.68(b)(4) which specifies that:

If credit is taken for soluble boron, 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.

Based on the results of an inspection and analysis that conservatively projected the condition of the boraflex panels through the end of 2006, IP2 compliance with 10 CFR 50.68(b)(4) was confirmed by an analysis documented in Northeast Technology Corporation report 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).

Based upon BADGER testing in calendar years 2003, 2006 and 2010 and RACKLIFE code projections, the validity of the criticality 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. This analysis demonstrated that 10 CFR 50.68(b)(4) will be met for all normal and credible accident scenarios if the following requirements are met:

a) Spent Fuel Pit boron concentration is maintained within the limits of LCO 3.7.12, Spent Fuel Pit Boron Concentration, whenever fuel is stored in the SFP; and, b) Fuel assembly storage location within the Spent Fuel Pit is restricted in accordance with LCO 3.7.13, Spent Fuel Pit Storage, based on the fuel assemblys initial enrichment, burnup, decay of Plutonium-241 (i.e., cooling time), and number of Integral Fuel Burnable Absorbers (IFBA) rods.

Fuel assembly storage location within the Spent Fuel Pit is an essential element for the validity of the analysis because the storage racks in the Indian Point 2 B 3.7.13 - 2 Revision

Spent Fuel Pit Storage B 3.7.13 BASES BACKGROUND (continued) areas designated Region 1 have a different design than the storage racks in the areas designated Region 2. These design differences have a significant impact on criticality calculations. Additionally, each of the two regions is sub-divided into two parts based on the extent of the boraflex degradation. Therefore, the SFP is divided into four distinct regions based on rack design and boraflex degradation. Figure 3.7.13-5 identifies the four regions as Region 1-1, Region 1-2, Region 2-1 and Region 2-2.

Additionally, selected cells located on the perimeter of Region 2-2 have higher neutron leakage rates than other cells in the Region and are designated as peripheral cells.

Each SFP region and sub-region is shown in Figure 3.7.13-5 and is described below beginning with the region that can be used to store only the least reactive fuel and ending with the region that must be used to store the most reactive fuel.

Region 2, consisting of nine racks that use the egg-crate design, can store 1105 fuel assemblies and two failed fuel canisters. Region 2 racks consist of boxes welded into a "checkerboard" array with a storage location in each square. One Boraflex absorber panel is held to one side of each cell wall by picture frame sheathing. Region 2 racks were originally designed to store fuel assemblies that have undergone significant burnup (e.g., 5.0 weight percent (w/o) U235 with a burnup of at least 40,900 megawatt days per metric ton (MWD/MT)) or fuel assemblies with a relatively low initial enrichment and low burnup (i.e., 1.764 w/o U235 at zero burnup).

Region 2 is subdivided into two regions (Region 2-1 and Region 2-2):

Region 2-1 is assumed to have sustained a 100% loss of Boraflex (i.e., none of the boraflex in the panels is assumed to be available). Figure 3.7.13-1 shows the fuel assembly criteria that will meet the requirements of 10 CFR 50.68(b)(4) if stored in Region 2-1. As shown on Figure 3.7.13-1, the maximum initial enrichment that can be stored in Region 2-1 with no burnup is 1.06 w/o U235. Figure 3.7.13-1 shows an allowance permitting storage of fuel assemblies with higher initial enrichments based on the reactivity reduction due to the cumulative burnup of the fuel assembly in the core and the decay of Pu241 (expressed as cooling time) after a fuel assembly is discharged.

Region 2-2 is assumed to have sustained only a 30% loss of Boraflex (i.e., 70% of the boraflex in the panels is assumed to be available). Figure 3.7.13-2 shows the fuel assembly criteria that will meet the requirements of 10 CFR 50.68(b)(4) if stored in Indian Point 2 B 3.7.13 - 3 Revision

Spent Fuel Pit Storage B 3.7.13 BASES BACKGROUND (continued)

Region 2-2. As shown on Figure 3.7.13-2, the maximum initial enrichment that can be stored in Region 2-2 with no burnup is 1.80 w/o U235. Additionally, Figure 3.7.13-2 shows an allowance permitting storage of fuel assemblies with higher initial enrichments based on the reactivity reduction due to the cumulative burnup of the fuel assembly in the core and the decay (expressed as cooling time) of Pu241 after a fuel assembly is discharged.

Region 1, consisting of three racks that use the flux trap design, can store 269 new or irradiated fuel assemblies. The flux trap design used in Region 1 uses spacer plates in the axial direction to separate the cells.

Boraflex absorber panels are held in place adjacent to each side of the cell by picture-frame sheathing. The spacer plates between cells form a flux trap between the boraflex absorber panels. Region 1 racks were originally designed to store new fuel with enrichments up to 5.0 w/o U235.

Region 1 is subdivided into two regions (Region 1-1 and Region 1-2):

Region 1-1 is assumed to have sustained a 100% loss of Boraflex (i.e., none of the boraflex in the panels is assumed to be available). Figure 3.7.13-3 shows the fuel assembly criteria that will meet the requirements of 10 CFR 50.68(b)(4) if stored in Region 1-1. As shown on Figure 3.7.13-3, the maximum initial enrichment that can be stored in Region 1-1 with no burnup is 1.95 w/o U235. Additionally, Figure 3.7.13-3 shows an allowance permitting storage of fuel assemblies with higher initial enrichments based on the reactivity reduction due to the cumulative burnup of the fuel assembly in the core. Figure 3.7.13-3 does not provide any allowance from the minimum required fuel assembly burnup based on the decay of Pu241.

(Fuel assemblies that do not meet the criteria in Figure 3.7.13-3 may be stored in Region 1-1 if the following two conditions are met: a) the fuel assemblies are stored in a checkerboard loading configuration (1 out of every two cells with every other cell vacant); and, b) fuel assemblies meet the criteria of Figure 3.7.13-4.)

Region 1-2 is assumed to have sustained a 50% loss of Boraflex (i.e., 50% of the boraflex in the panels is assumed to be available). Region 1-2 can accommodate unirradiated fuel up to 5.0 w/o U235 assuming the presence of a minimum number of IFBA rods as specified in Figure 3.7.13-4. As shown on Figure 3.7.13-4, the maximum initial enrichment that can be stored in Region 1-2 Indian Point 2 B 3.7.13 - 4 Revision

Spent Fuel Pit Storage B 3.7.13 BASES BACKGROUND (continued) when there are no IFBA rods is 4.50 w/o U235. Figure 3.7.13-4 does not provide any allowance from the minimum required IFBA rods based on the decay of PU241.

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 peripheral cells.

IP3 Fuel Assemblies The SFP is also used to store spent fuel transferred from the IP3 SFP.

The IP3 fuel assembly storage location is also restricted in accordance with LCO 3.7.13 that limits IP3 fuel assemblies to Region 1-2 of the IP2 SFP. The NRC has issued Amendments 268 and 287 for the inter-unit transfer of spent nuclear fuel (Refs. 8 and 12). These Amendments are based on evaluations conducted for each aspect of the inter-unit transfer of fuel as documented in References 9 and 13.

APPLICABLE IP2 Fuel Assemblies 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 assemblys initial enrichment, burnup, decay of Pu241 (i.e., cooling time) and number of Integral Fuel Burnable Absorbers (IFBA) rods.

Indian Point 2 B 3.7.13 - 5 Revision

Spent Fuel Pit Storage B 3.7.13 BASES APPLICABLE SAFETY ANALYSES (continued)

Reference 4 evaluated non-accident conditions in the SFP including the affects of the projected boraflex degradation through the year 2006.

Based upon BADGER testing in calendar years 2003, 2006 and 2010 and RACKLIFE code projections, the validity of the criticality 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. Reference 4 determined that if storage location requirements in this LCO are met then the SFP will have a keff of 0.95 if filled with a soluble boron concentration of 786 ppm and will have a keff of < 1.0 if filled with unborated water.

Reference 4 also evaluated credible abnormal occurrences in accordance with ANSI/ANS-57.2-1983. This evaluation considered the effects of the following: a) a dropped fuel assembly or an assembly placed alongside a rack; b) a misloaded fuel assembly; and, c) abnormal heat loads.

Reference 4 determined that the SFP will maintain a keff of 0.95 under the worst-case accident scenario if the SFP is filled with a soluble boron concentration of 1495 ppm.

Therefore, reference 4 confirmed that the requirements in 10 CFR 50.68, Criticality Accident Requirements, (Ref. 1) will be met for both normal SFP operation and credible abnormal occurrences if:

a) Spent Fuel Pit boron concentration is maintained within the limits of LCO 3.7.12, Spent Fuel Pit Boron Concentration, whenever fuel is stored in the SFP; and, b) Fuel assembly storage location within the Spent Fuel Pit is restricted in accordance with LCO 3.7.13, Spent Fuel Pit Storage, based on the fuel assemblys initial enrichment, burnup, decay of Plutonium-241 (i.e., cooling time), and number of Integral Fuel Burnable Absorbers (IFBA) rods.

Reference 5 evaluated postulated unplanned SFP boron dilution scenarios assuming an initial SFP boron concentration within the limits of LCO 3.7.12. The evaluation considered various scenarios by which the SFP boron concentration may be diluted and the time available before the minimum boron concentration necessary to ensure subcriticality for the non-accident condition (i.e. it is not assumed an assembly is misloaded concurrent with the Spent Fuel Pit dilution event). Reference 5 determined that an unplanned or inadvertent event that could dilute the Indian Point 2 B 3.7.13 - 6 Revision

Spent Fuel Pit Storage B 3.7.13 BASES APPLICABLE SAFETY ANALYSES (continued)

SFP boron concentration from 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 2010 and RACKLIFE code projections, the validity of the criticality 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 13, 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. Only intact fuel assemblies with initial average enrichment

< 4.4 w/o U235, and discharged prior to IP3 Cycle 12.

c. IP3 fuel assemblies V43 and V48 are not approved for storage in the Spent Fuel Pit.

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 B 3.7.13 BASES LCO IP2 Fuel Assemblies This LCO establishes restrictions on fuel assembly storage location within the SFP to ensure that the requirements of 10 CFR 50.68 are met. This LCO requires that each fuel assembly stored in the Spent Fuel Pit is 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, that fuel assembly storage location within the Spent Fuel Pit is restricted to Regions identified in Figure 3.7.13-5 as follows:

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

As shown on Figure 3.7.13-1, the maximum initial enrichment that can be stored in Region 2-1 with no burnup is 1.06 w/o U235. Additionally, Figure 3.7.13-1 shows an allowance permitting storage of fuel assemblies with higher initial enrichments based on the reactivity reduction due to the cumulative burnup of the fuel assembly in the core and the decay of Pu241 after a fuel assembly is discharged (expressed as cooling time).

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.

As shown on Figure 3.7.13-2, the maximum initial enrichment that can be stored in Region 2-2 with no burnup is 1.80 w/o U235. Additionally, Figure 3.7.13-2 shows an allowance permitting storage of fuel assemblies with higher initial enrichments based on the reactivity reduction due to the cumulative burnup of the fuel assembly in the core and the decay (expressed as cooling time) of Pu241 after a fuel assembly is discharged.

c. Fuel assemblies that satisfy requirements of Figure 3.7.13-3 may be stored in any location in Region 1-2 or Region 1-1.

As shown on Figure 3.7.13-3, the maximum initial enrichment that can be stored in Region 1-1 with no burnup is 1.95 w/o U235. Additionally, Figure 3.7.13-3 shows an allowance permitting storage of fuel assemblies with higher initial enrichments based on the reactivity reduction due to the cumulative burnup of the fuel assembly in the core. Figure 3.7.13-3 does not provide any allowance from the minimum required fuel assembly burnup based on the decay of Pu241.

Indian Point 2 B 3.7.13 - 8 Revision

Spent Fuel Pit Storage B 3.7.13 BASES LCO (continued)

(Fuel assemblies that do not meet the criteria in Figure 3.7.13-3 may be stored in Region 1-1 if the fuel assemblies are stored in a checkerboard loading configuration (1 out of every two cells with every other cell vacant) and fuel assemblies meet the criteria of Figure 3.7.13-4.)

d. Fuel assemblies that satisfy requirements of Figure 3.7.13-4 may be stored as follows: 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 of Figure 3.7.13-5.

As shown on Figure 3.7.13-4, the maximum initial enrichment that can be stored in Region 1-2 when there are no IFBA rods is 4.50 w/o U235.

Figure 3.7.13-4 does not provide any allowance from the minimum required IFBA rods based on the decay of Pu241.

The six peripheral cells may be used to store fuel that meets the requirements for storage in any location in the SFP (i.e., meets requirements for storage in Region 1-1, 1-2, 2-1 or 2-2). 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 canisters that are loaded in Module H in the southeast corner of the SFP. Module H is in the upper right corner of the SFP in Figure 3.7.13-5.

IP3 Fuel Assemblies This LCO establishes restrictions on fuel assembly storage location within the SFP to ensure that the requirements of 10 CFR 50.68 are met.

APPLICABILITY This LCO applies whenever any fuel assembly is stored in the Spent Fuel Pit.

Indian Point 2 B 3.7.13 - 9 Revision

Spent Fuel Pit Storage B 3.7.13 BASES ACTIONS A.1 When the configuration of fuel assemblies stored in the Spent Fuel Pit is not in accordance with the rules established by LCO 3.7.13, the immediate action is to initiate action to make the necessary fuel assembly movement(s) to bring the configuration into compliance with the rules established by LCO 3.7.13.

SURVEILLANCE SR 3.7.13.1 REQUIREMENTS This SR verifies by administrative means that the IP2 fuel assembly has been classified based on initial enrichment, burnup, cooling time and number of Integral Fuel Burnable Absorbers (IFBA) rods in the fuel assembly in accordance with Figure 3.7.13-1, Figure 3.7.13-2, Figure 3.7.13-3, or Figure 3.7.13-4 and that the fuel assembly meets the requirements for the intended storage location defined on Figure 3.7.13-5.

This SR also verifies by administrative means that the IP3 fuel assembly meets the requirements for storage in the IP2 SFP. This administrative verification must be completed prior to placing any fuel assembly in the SFP. This SR ensures that this LCO and Specification 4.3.1.1 will be met after the fuel assembly is inserted in the SFP.

REFERENCES 1. 10 CFR 50.68, Criticality Accident Requirements.

2. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 150 to Facility Operating License No.

DPR-26, April 19, 1990.

3. Generic Letter 96-04, Boraflex Degradation in Spent Fuel Pool Storage Racks.

4. Northeast Technology Corporation report NET-173-01, Criticality Analysis for Soluble Boron and Burnup Credit in the Con Edison Indian Point Unit No. 2 Spent Fuel Storage Racks.

5. Northeast Technology Corporation report NET-173-02, Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis.

6. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 227 to Facility Operating License No.

DPR-26, May 29, 2002.

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

Indian Point 2 B 3.7.13 - 10 Revision

Spent Fuel Pit Storage B 3.7.13 BASES REFERENCES (continued)

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 HI-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 Point 2 RACKLIFE Projections Through 2010 and 2012 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 287 to Facility Operating License No.

DPR-26, December 22, 2017.

13. Holtec Report HI-2094289, Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at Indian Point Energy Center, Revision 10.

Indian Point 2 B 3.7.13 - 11 Revision

v t e Letter Sequence Request
EPID:L-2019-LLA-0079, TSTF-448 (Approved, Closed)
Initiation Request, Request, Request Acceptance Administration Questions Answers Results Approval Other: ML20122A262, ML21113A200
MONTHYEAR2019-05-20 [Table View]

NL-19-026, Technical Specifications Proposed Change - Permanently Defueled Technical Specifications

Text

SUBJECT:

Enclosure:

Enclosure:

References:

SUMMARY

3.4 CONCLUSION

Subject:

Attachment:

1.0 DESCRIPTION

1.0 DESCRIPTION

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