Text

Technical Specifications Proposed Change - Permanently Defueled Technical Specifications
	ML20132A200
Person / Time
Site:	Indian Point
Issue date:	04/28/2020
From:	Gaston R; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NL-20-033
	Download: ML20132A200 (288)
	v • d • e

~Enter~ Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 Tel 601-368-5573 Ron Gaston Director, Nuclear Licensing 10 CFR 50.90 NL-20-033 April 28, 2020 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 No. 3 NRC Docket No. 50-286 Renewed Facility Operating License No. DPR-64

References:

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

2) NRC letter to Entergy, "Indian Point Nuclear Generating Unit Nos. 2 and 3 -

Issuance of Amendment Nos. 292 and 267 RE: Changes to Technical Specification Sections 1.1, 'Definitions'; 4.0, 'Design Features'; and 5.0,

'Administrative Controls,' for a Permanently Defueled Condition (EPID L-2019-LLA-0081)," (ADAMS Accession No. ML20071Q717), dated April 10, 2020

3) Entergy Letter to NRC, "Proposed Technical Specifications (TS)

Changes- Indian Point Nuclear Generating Unit 3 TS SR 3.7.7.2 and TS 3.7.6, Required Action A.1," (Letter No. NL-19-093) (ADAMS Accession No. ML19325E913), dated November 21, 2019

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 (FOL) DPR-64 for Indian Point Nuclear Generating Station Unit No. 3 (IP3). This proposed license amendment would revise the IP3 FOL and revise the Technical Specifications (TSs) in Appendix A to Permanently fob DI

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NL-20-033 Page 2 of 3 Defueled Technical Specifications (POTS), the Environmental Technical Specification Requirements in Appendix B of the FOL, 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 IP3 FOL and the Appendices A through C TSs and remove the requirements that would no longer be applicable after IP3 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 IP3 by April 30, 2021. 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 IP3 FOL 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 FOL 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 includes the changes to the IP3 TSs that were approved by the NRC in Reference 2, since these are required to support the permanently shut down and defueled condition. The changes proposed to the IP3 TSs in Reference 3 but not yet approved will not affect the IP3 TSs that will remain in the POTS; thus, they are not included.

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 IP3. Attachment 1 to the Enclosure contains a mark-up of the current FOL, 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, POTS, Appendices B and C TSs, and POTS Bases pages in their entirety.

Entergy requests review and approval of this proposed license amendment by April 30, 2021, with a 90-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 Storage Building has been met, and the license amendment regarding the IP3 TSs affecting the administrative controls for the permanently defueled condition, which was approved by the NRC in Reference 2, has been implemented.

NL-20-033 Page 3 of 3 There are no new regulatory commitments made in this letter.

If you have any questions regarding this submittal, please contact Ms. Mahvash Mirzai, Manager, Regulatory Assurance, at 914-254-7714.

I declare under penalty of perjury that the foregoing is true and correct. Executed on April 28, 2020.

Respectfully, Ron Gaston RWG/cdm/std

Enclosure:

Indian Point Nuclear Generating Station Unit No. 3 - Description and Evaluation of Proposed Changes Attachments to

Enclosure:

1. Indian Point Nuclear Generating Station Unit No. 3 - 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 No. 3 - 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 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-20-033 Indian Point Nuclear Generating Station Unit No. 3 Description and Evaluation of Proposed Changes

Enclosure NL-20-033 Page 1 of 86 Indian Point Nuclear Generating Station Unit 3 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 3 (IP3) no later than April 30, 2021 (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 IP3 Facility Operating License (FOL) and revise the Technical Specifications (TSs) in Appendix A of the FOL tci Permanently Oefueled Technical Specifications (POTS), the Environmental Technical Specification Requirements in Appendix B of the FOL, and the Inter-Unit Fuel Transfer Technical Specifications in Appendix C.

The proposed changes are consistent with the permanent cessation of reactor operation and per"manent defueling of the reactor. The proposed changes would revise certain requirements contained within the IP3 FOL and TSs and remove the requirements that would no longer be applicable after IP3 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 (FHA) in the Fuel Storage Building (FSB) has been met, and the license amendment regarding the IP3 TSs affecting the administrative controls for the permanently defueled condition, which was approved by the NRC in Reference 2, has been implemented.

2. DETAILED DESCRIPTION AND BASIS FOR THE CHANGES This license amendment request modifies the IP3 FOL and revises the IP3 Appendix A TSs into POTS, and IP3 Appendices Band C TSs to comport with a permanently shut down and defueled condition.

The proposed changes to the IP3 FOL 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 FOL 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 IP3 TSs that were approved by the NRC in Reference 2 have been implemented by the site. The changes proposed to the IP3 TSs in Reference 3 but not yet approved will not affect the IP3 TSs that will remain in the POTS; thus, they are not included.

Enclosure NL-20-033 Page 2 of 86 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 IP3 TSs is limited to those needed to address the remaining postulated design basis accidents (DBAs) that will remain applicable to IP3 in the permanently shut down and defueled condition, so that the consequences of the accident are maintained within acceptable limits.

Chapter 14 of the IP3 Updated Final Safety Analysis Report (UFSAR) describes the OBA and transient scenarios applicable to IP3 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 IP3 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 OBA scenarios postulated in the IP3 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 provided in the Post Shut Down Decommissioning Activities Report (PSDAR)

(Reference 4).

Chapter 14 of the IP3 UFSAR evaluates the safety aspects of the plant and demonstrates that the plant can be operated safely and that exposures from credible accidents do not exceed the guidelines of 10 CFR 100. This chapter is divided into three 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.

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.

Enclosure NL-20-033 Page 3 of 86 A list of the IP3 UF~AR 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.

Table 2 IP3 DBAs and Events UFSAR Postulated Accident or Transient Defueled Applicability Section Uncontrolled Control Rod Assembly Withdrawal from a 14.1.1 Not Applicable Subcritical Condition 14.1.2 Uncontrolled Control Rod Assembly Withdrawal at Power Not Applicable 14.1.3 Rod Assembly Misalignment Not Applicable 14.1.4 Rod Cluster Control Assembly (RCCA) 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 Inactive Reactor Coolant Loop Not Applicable 14.1.8 Loss of External Electrical Load Not Aoolicable 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 APPiicabie Startup Accidents without Reactor Coolant Pump 14.1.13 Not Applicable Operation 14.1.14 Startup Accident with a Full Pressurizer Not Applicable Applicable - Drop of a fuel assembly onto the floor of the spent fuel pit Not Applicable - Fuel Assembly Stuck Inside the Reactor Vessel Not Applicable - Fuel Assembly or Control Rod Cluster Dropped 14.2.1 Fuel Handling Accidents onto the Floor of the Reactor Cavity Not Applicable - Fuel Assembly Stuck in the Penetration Valve Not Applicable - Fuel Assembly Stuck in the Transfer Carriage or the Carriage Becomes Stuck Not Credible - Fuel

Enclosure NL-20-033 Page 4 of 86 UFSAR Postulated Accident or Transient Defueled Applicability Section Cask Drop Accident Deemed to not be Credible - See discussion in (1) below Applicable - Dose dependent on 14.2.2 Accidental Release 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 APPiicabie 14.2.5 Rupture of a Steam Pipe Not Applicable Rupture of a Control Rod Drive Mechanism Housing 14.2.6 Not Applicable (RCC Assembly Ejection) 14.3 Loss-of-Coolant-Accidents Not Applicable Appendix Consequences of a Turbine Missile at Indian Point 3 Not Applicable 148 (1) Section 14.2.1 of the IP3 UFSAR states:

"As discussed in Section 9.12.4.3, Single Failure Proof Cranes [for] Spent Fuel [Casks], the fuel storage building crane's main hook that handles spent fuel casks has been upgraded to single-failure-proof in accordance with the applicable guidelines of NRC NUREG-0554 (Single-Failure-Proof Cranes for Nuclear Power Plants, May 1979) and the applicable requirements of American Society of Mechanical Engineers ASME NOG-1-2004, Rules [for] Construction of Overhead and Gantry Cranes (top Running Bridge, Multiple Girder) to support spent fuel cask handling activities, without the necessity of having to postulate the drop of a spent fuel cask.

With the crane's main hook qualified as single-failure-proof, and when the [cranes] is used as part of a single-failure-proof handling system for critical lifts as discussed in NRC NUREG-0800, Revision 1 of Section 9.1.5, Overhead Heavy Load Handling Systems,

[Sub-section 111.4.C], a cask drop accident is not a credible event and need not be postulated ... "

The analyzed accidents that remain applicable to IP3 in the permanently shut down and defueled condition are the FHA in the FSB, accidental release 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 includes:

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-11-00074, "AST Analysis of IP3 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.

Enclosure NL-20-033 Page 5 of 86

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.

An analysis of a High Integrity Container Drop event was performed. However, it is not credited as part of this POTS license amendment request. It was performed for future purposes that are outside the changes requested by the POTS license amendment request.

Analysis of the FHA in the FSB for the Permanently Shut Down and Defueled Condition Concurrent with implementation of the POTS, UFSAR Section 14.2.1 will be revised in accordance with 10 CFR 50.59 to reflect the results of the "Normal" case analyzed in Calculation IP-CALC-11-00074, as summarized in Calculation IP-CALC-19-00003. This FHA analysis utilizes the AST methodology and concludes that the dose consequences of the FHA in the FSB 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 IP3 Control Room, Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) were computed using the following:

The methodology and assumptions in Regulatory Guide 1.183,

Appropriate source terms, release pathways, and other assumptions, as described below,

All fuel pins in the dropped fuel assembly are broken,

Decay time experienced prior to fuel movement = 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months ,

SFP Water Depth = 23 feet,

Post-accident atmospheric dispersion factors, and

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.

Enclosure NL-20-033 Page 6 of 86 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) 1-130 3.41 E+04 Kr-85m 5.62E+01 1-131 6.90E+07 Kr-85 1.11E+06 1-132 6.38E+07 Kr-87 O.OOE+OO 1-133 1.17E+07 Kr-88 O.OOE+OO 1-134 O.OOE+OO 1-135 2.63E+04 Xe-131m 9.71E+05 Xe-133m 2.78E+06 Xe-133 1.36E+08 Xe-135m 4.21E+03 Xe-135 7.86E+05 Xe-138 0-.00E+OO Release Fractions and Composition The fission product gap release fractions, for each radionuclide group for the FHA are shown below:

1-131 0.12

Kr-85 0.30

Other iodines and noble gases 0.1 O 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 Control Room modeling assumptions are:

The Control Room xtQ for 0-2 hours is 1.07E-03 seconds/m 3

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

Control Room volume: 47,200 ft3

Filtered makeup: O ft3/min

Filtered recirculation: 0 ft3/min

Unfiltered makeup: 1,500 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.

Enclosure NL-20-033 Page 7 of 86 Offsite Dose Consequences The offsite modeling assumptions are:

Offsite atmospheric dispersion factors (x/Q) o EAB x/Q is 1.03E-03 seconds/m 3 o LPZ x/Q is 3.8E-04 seconds/m 3

Breathing Rates o EAB for 0-2 hours is 3.5E-04 (consistent with Regulatory Guide 1.183).

o LPZ for 0-8 hours is 3.5E-04 (consistent with Regulatory Guide 1.183).

o LPZ for 8-24 hours is 1. 75E-04 o LPZ for 24-720 hours is 2.32E-04 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 4.91 5 EAB 5.7 6.3 LPZ 2.1 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 IP3 UFSAR evaluates the accidental release of waste gas. Concurrent with implementation of the POTS, 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.

Enclosure NL-20-033 Page 8 of 86 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)

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 xtQs: EAB = 1.03E-03 seconds/m 3 , LPZ = 3.8E-04 seconds/m 3 A bounding xtQ was developed for a release from the waste gas decay tanks to the IP3 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 xtQ.

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.

Enclosure NL-20-033 Page 9 of 86 The IP3 Control Room xtQ also does not credit holdup of the activity in the PAS which would delay or disperse activity within the building.

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 2-5.

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 2-3. 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 of Waste Liquid Section 14.2.2 of the IP3 UFSAR addresses the accidental release of waste liquid. It concludes:

"The incipient hazard from these process or waste liquid releases is derived only from the volatilized components. The releases are described and their effects summarized in Section 14.2.3." Therefore, a separate liquid-specific release accident evaluation is not required to be performed with regard to removal of supporting systems such as PAS ventilation, station vent radiation monitors, Control Room isolation, and Control Room filtration.

3. REGULATORY EVALUATION Indian Point Nuclear Generating Station Unit 3 proposes to modify the license conditions and the TSs from Appendices A through C as listed in the following tables. In addition, IP3 is providing a description and basis for each of the proposed changes. to this enclosure contains a mark-up of the current FOL, Appendices A, B, and C TSs and Appendix A TSs Bases pages. The proposed changes to the IP3 Appendix A TSs are considered a major rewrite. Thus, the IP3 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) FOL, Appendix A TSs, and Appendix A TSs Bases pages, because they do not affect the technical content of the IP3 FOL 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.

Enclosure NL-20-033 Page 10 of 86 of this enclosure provides the re-typed IP3 Facility License, POTS, POTS Bases in their entirety, and the affected pages of the Appendices B and C TSs. Since the changes to the Appendix A TSs and TSs Bases are considered a major rewrite, revision bars are not used. It incorporates the changes to the IP3 Appendix A TSs approved by the NRC in Reference 2.

Proposed Changes to the IP3 Facility Operating License License Title Current Title Proposed Title Renewed Facility Operating License Renewed Facility GpeFatiRg 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 IP3, 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 1.8 Current License Condition 1.8 Proposed License Condition 1.8 The facility will operate in conformity with the The facility will opemtebe maintained in application, the provisions of the Act, and the conformity with the application, the provisions rules and regulations of the Commission; of the Act, and the rules and 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 IP3, 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.

License Condition 2 Current License Condition 2 Proposed License Condition 2 Accordingly, Renewed Facility Operating Accordingly, Renewed Facility GperatiRg License No. OPR-64, is hereby issued to License No. OPR-64, is hereby issued to ENIP3 and ENO to read as follows: ENIP3 and ENO to read 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 IP3, 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.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

Enclosure NL-20-033 Page 11 of 86 Point Nuclear Generating Unit No. 3, a Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and pressurized water nuclear reactor and associated equipment (the facility), owned by associated equipment (the facility), owned by ENIP3 and operated by ENO. The facility is ENIP3 and operatedmaintained by ENO. The located in Westchester County, New York, on facility is located in Westchester County, New the east bank of the Hudson River in the Village York, on the east bank of the Hudson River in of Buchanan, and is described in the "Final the Village of Buchanan, and is described in the Facility Description and Safety Analysis "Final Facility Description andDefueled Safety Report," as supplemented and amended, and Analysis Report," as supplemented and the Environmental Report, as amended. 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 IP3, 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 2.8.(1)

Current License Condition 2.8.(1) Progosed License Condition 2.8.(1)

Pursuant to Section 104b of the Act and Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," (a) ENIP3 to possess and Utilization Facilities," (a) ENIP3 to possess and use, and (b) ENO to possess, use, and use, and (b) ENO to possess, and use,---aoo 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 procequres 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 IP3, 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 2.8.(2)

Current License Condition 2.8.(2) Progosed License Condition 2.8.(2)

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

Enclosure NL-20-033 Page 12 of 86 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 IP3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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, IP3 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 IP3 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.

License Condition 2.8.(3)

Current License Condition 2.8(3) Proposed License Condition 2.8.(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 time, any byproduct, source and special nuclear material as sealed neutron sources for nuclear material as sealed neutron sources reactor startup, sealed sources for reactor that were used for reactor startup, sealed instrumentation and radiation monitoring sources that were used for calibration of equipment calibration, and as fission detectors reactor instrumentation and are used in the in amounts as required; calibration of radiation monitoring equipment calibration, and that were used 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 and fission detectors. The deletion of the authorization to receive and use sources for reactor startup is consistent with the fact that IP3 will no longer be authorized to operate. The deletion of the authorization to receive and use fission detectors is consistent with the fact that IP3 complies with the criteria of 10 CFR 50.68(b) in lieu of maintaining a monitoring system capable of detecting criticality in the spent fuel pit as described in Section 9.5 of the IP3 Updated Final Safety Analysis Report.

The authorization to possess such sources previously used for reactor startup and fission detectors is retained. The continued authorization to possess neutron sources that were used for reactor startup and fission detectors 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 IP3, 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.

Enclosure NL-20-033 Page 13 of 86 License Condition 2.8.(4)

Current License Co ndition 2.8(4) Proposed License Condition 2.8.(4)

ENO pursuant to th e Act and 10 CFR Parts 30, ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to recei ve, possess, and use in 40 and 70, to receive, possess, and use in amounts as require d any byproduct, source or amounts as required any byproduct, source or special nuclear material without restriction to special nuclear material without restriction to chemical or physica I form, for sample analysis chemical or physical form, for sample analysis or instrument calibration; or associated with or instrument calibration; or associated with radioactive apparat us or com onents. radioactive ap aratus or com onents-:-;

Basis This license conditi on is revised by making a grammatical correction. The ending period is replaced with a se mi-colon.

License Condition 2.8.(5)

Current License Co ndition 2.8(5) Proposed License Condition 2.8.(5)

ENO pursuant to th e 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 byproduct and spec ial 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 conditi on is revised to reflect that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or place mentor 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 Co ndition 2.C.(1) Proposed License Condition 2.C.(1)

ENO is authorized t o operate the facility at Deleted per Amendment['###]

steady state reacto r core power levels not in excess of 3216 megawatts thermal (100% of rated power).

Basis This license conditi on is deleted in its entirety to reflect the permanently defueled condition of the facility. After the ce rtifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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-20-033 Page 14 of 86 License Condit ion 2.C.(2), Technical Specifications Current License Condition 2.C.(2) Pro12osed License Condition 2.C.(2)

The Technical Specifications containe din The Technical Specifications contained in Appendices A, B, and C, as revised th rough Appendices A, ~. and C, as revised through Amendment No. 266, are hereby inco rporated Amendment No. 266###, are hereby in the renewed license. ENO shall op erate 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 rep lace 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 IP3, 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).

Lie ense Condition 2.C.(3)

Current License Condition 2.C.(3) Pro12osed License Condition 2.C.(3)

(DELETED) None Basis The historical reference to a deleted Ii cense condition is deleted in its entirety. This is an administrative change.

Lie ense Condition 2.C.(4)

Current License Condition 2.C.(4) Pro12osed License Condition 2.C.(4)

(DELETED) None Basis The historical reference to a deleted Ii cense condition is deleted in its entirety. This is an administrative change.

License Condition 2.H Current License Condition 2.H Pro12osed License Condition 2.H ENO shall implement and maintain in effect all Deleted per Amendment[###]

provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for Indian Point Nucle ar Generating Unit No. 3 and as approvedin NRC fire protection safety evaluations (SEs ) dated September 21, 1973, March 6, 1979, May 2, 1980, November 18, 1982, December 30, 1982, February 2, 1984, April 16, 1984 January 7, 1987, September 9, 1988, October' 21, 1991, April 20, 1994, January 5, 1995, and su lements thereto, sub*ect to the fo llowinQ

Enclosure NL-20-033 Page 15 of 86 provision:

ENO may make changes to the 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.

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 IP3, 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. Indian Point Nuclear Generating Station Unit 3 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 IP3. 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 2.0 Current License Condition 2.0 Pro12osed License Condition 2.0 Evaluation, status and schedule for completion Deleted per Amendment['###]

of balance of plant modifications as outlined in letter dated February 12, 1983, shall be forwarded to the NRC by January 1, 1984.

Basis The license condition is deleted in its entirety. It refers to a historical obligation that was previously met. The removal of this license condition is an administrative change.

License Condition 2.AA Current License Condition 2.AA Pro12osed License Condition 2.AA The following conditions relate to the Deleted per Amendment[###]

amendment approving the conversion to Improved Standard Technical Specifications:

1. This amendment authorizes the relocation of certain Technical Specification

Enclosure NL-20-033 Page 16 of 86 requirements and detailed information to licensee-controlled documents ... 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) ...

Basis The license condition is deleted in its entirety. It refers to a historical license condition associated with the conversion to the Improved Standard Technical Specifications. This license condition was previously met. Thus, the removal of this license condition is an administrative change.

License Condition 2.AB Current License Condition 2.AB Pro12osed License Condition 2.AB With the reactor critical, Entergy shall maintain Deleted per Amendment[###]

the reactor coolant system cold leg at a temperature (Tco1ct) greater than or equal to 525 °F. Entergy shall maintain a record of the cumulative time that the plant is operated with the reactor critical while Tcolct is below 525 °F.

Upon determination by Entergy that the cumulative time of plant operation with the reactor critical while Tcolct is below 525 °F has exceeded one (1) year, Entergy must:

(a) within one (1) month, inform the NRC, in writing, and (b) within six (6) months submit the results of an analysis of the impact of the operation with Tcolct below 525 °Fon the pressurized thermal shock reference temperature (RTpts).

Basis The license condition is deleted in its entirety. It refers to operations with the reactor critical.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.AD, CONTROL ROOM ENVELOPE HABITABILITY Current License Condition 2.AD Pro12osed License Condition 2.AD Upon implementation of Amendment No. 239 Deleted per Amendment[###]

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

~~

Enclosure NL-20-033 Page 17 of 86 Surveillance Requirement (SR) 3. 7.11.4, 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 3.7.11.4, in accordance with TS 5.5.16.c.(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from February 1, 2005, the date of the most recent successful tracer gas test, as stated in the June 28, 2005, letter response to Generic Letter 2003-01.

(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 June 18, 2007, the date of the most recent successful pressure measurement test.

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 2.AF.(2).C, CONTROL ROOM ENVELOPE HABITABILITY Current License Condition 2.AF.(2).c Pro12osed License Condition 2.AF.(2).c The licensee shall notify the NRC in writing None 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.

Basis This license condition is deleted in its entirety. It is an obligation to notify the NRC within a specified time period regarding License Renewal activities that were required to be completed prior to the period of extended operation. This is a historical license condition, because the license condition was met in accordance with the schedule specified in the license condition. The removal of this license condition is an administrative change.

Enclosure NL-20-033 Page 18 of 86 License Condition 3 Current License Condition 3 Proposed License Condition 3 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 sl=lall e*J:liFe at FRiEIAi§!l=lt ,A,13Fil April 30, 2025. 30, 2026.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 IP3, 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 Defue/ed Technical Specifications ...

Date of Issuance: September 17, 2018 Date of Issuance: Se13teFRbeF 17, 2018 To 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 POTS which is yet to be determined. These are administrative changes.

APPENDIX A TO FACILITY QP!iRA+ING LICENSE DPR-64 Current Title Proposed Title FACILITY OPERATING LICENSE DPR-64 ... FACILITY OPERATING LICENSE DPR-64 ...

TECHNICAL SPECIFICATIONS AND PERMANENTLY DEFUELED TECHNICAL BASES ... SPECIFICATIONS AND BASES ...

Amendment No. 203 Amendment No. 2:Wt###

Basis The License Title is modified to rename the "Facility Operating License DPR-64" and the "Technical Specifications and Bases" as "Facility License DPR-64" and "Permanently Defueled Technical Specifications and Bases." These changes reflect the upcoming change in status regarding IP3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no lonQer authorize operation of the reactor or placement or retention

Enclosure NL-20-033 Page 19 of 86 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 POTS.

APPENDIX A, TECHNICAL SPECIFICATIONS, TABLE OF CONTENTS Current IP3 TS Basis for Change Table of Contents The Table of Contents is modified to reflect the changes made below.

TECHNICAL SPECIFICATION SECTION 1.1, DEFINITIONS Technical Specification 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 POTS 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 POTS 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 POTS 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 POTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

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

Enclosure NL-20-033 Page 20 of 86 CORE AL TERATl ON This definition is proposed for deletion, because the term is not used in any POTS 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 POTS specification. Technical Specification 5.6.5 that requires the COLR is also proposed for elimination.

DOSE EQUIVALENT 1-131

This definition is proposed for deletion, because the term is not used in any POTS specification. This term is used in current TS 3.4.16 and TS 3. 7 .17 to express t~e specific activity limit from a mixture of iodine isotopes contained in reactor coolant and secondary coolant. Technical Specification 3.4.16 and TS 3.7.17 are proposed for deletion in the POTS.

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 meaninoful.

DOSE EQUIVALENT XE-133 This definition is proposed for deletion, because the term is not used in any POTS 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. Technical Specification 3.4.16 is proposed for deletion in the POTS. 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.

La This definition is proposed for deletion, because the term is not used in any POTS.

Technical Specification 5.5.15 that refers to La is also proposed for elimination. La is the maximum allowable primary containment leakage rate. Since accident conditions occurring inside the primary containment will no longer apply to the permanently shut down and defueled facility, the definition is no longer meaningful.

Enclosure NL-20-033 Page 21 of 86 LEAKAGE This definition is proposed for deletion, because the term is not used in any POTS specification. Refer to the discussions for the proposed deletion of TS 3.4.13 and TS 5.5.8.

MASTER RELAY TEST This definition is proposed for deletion, because the term is not used in any POTS 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 POTS 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 POTS specification. There are no systems or components required to be operable in the POTS, 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 POTS 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 POTS 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 POTS 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 (SOM) This definition is proposed for deletion, because the term is not used in any POTS 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 POTS specification. There is no instrumentation credited in the analysis of the accidents that remain credible in the permanently defueled condition.

Enclosure NL-20-033 Page 22 of 86 STAGGERED TEST BASIS This definition is proposed for deletion, because the term is not used in any POTS specification. This definition applies to the performance of surveillance tests on systems with multiple subsystems or channels. There are no surveillance requirements in the POTS for operatino systems.

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

TECHNICAL SPECIFICATION SECTION 1.2, LOGICAL CONNECTORS Technical Specification 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 ... _ ..... ,J t:".-.--

-**- ---=--

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, ei=

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, ~pc----=--*.

Current Examples Proposed Examples EXAMPLES The following examples illustrate EXAMPLES The following examples the use of logical connectors ... illustrates the use of 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. 7.15. This is the only TS that utilizes logical connectors in the POTS. These changes are administrative changes.

Enclosure NL-20-033 Page 23 of 86 TECHNICAL SPECIFICATION SECTION 1.3, COMPLETION TIMES Technical Specification 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 POTS.

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 ...

The Background section of TS 1.3 is modified to reflect the upcoming change in status regarding IP3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 handlinQ and storaQe 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 discovery of a situation (e.g., referenced to the discovery of a situation (e.g.,

inoperable equipment or variable not within inoperable equipment or variable not within limits) that requires entering an ACTIONS limits) that requires entering an ACTIONS Condition unless otherwise specified, providing Condition unless otherwise specified, providing the unit is in a MODE or specified condition the t!Rftfacility is in a MODE or specified stated in the Applicability of the LCO. condition stated in the Applicability of the LCO.

Unless otherwise specified, ... Unless otherwise speoified ...

Required Actions must be completed prior to Required Actions must be completed prior to the expiration of the specifie*d Completion Time. the expiration of the specified Completion Time.

An ACTIONS Condition remains in effect and An ACTIONS Condition remains in effect and the Required Actions apply until the Condition the Required Actions apply until the Condition no longer exists or the unit is not within the no longer exists or the YfHtfacility is not within LCO Applicability. the LCO Applicability.

If situations are discovered ... If situations are dissevered ...

The Description section of TS 1.3 is modified to reflect the upcoming change in status regarding IP3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 POTS will contain no operability requirements for any equipment. In addition, the term facility better represents IP3 in the permanently shut down and defueled condition.

Enclosure NL-20-033 Page 24 of 86 Current Examples Proposed Example EXAMPLES The following examples illustrate EXAMPLES The following examples the use of Completion Times with different illustrates the use of Completion Times with types of Conditions and changing Conditions. different types of Conditions and shanging ConditionsRequired Actions.

Example 1.3-1 ... Example 1.3-1 is modified to address Completion Times as utilized by TS 3.7.15.

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

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

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

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

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

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

This section is modified to reflect the use of Completion Times that are utilized in TS 3.7.14, TS 3.7.15, and TS 3.7.16. These are the only TSs that have Completion Times in the POTS.

The changes to the Examples section of TS 1.3 are administrative changes.

TECHNICAL SPECIFICATION SECTION 1.4, FREQUENCY Technical Specification 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 POTS.

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, Surveillance Specifications of Section 3.0, 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 sertain Notes in the Surveillanse solumn that modify performance requirements. modify performanse requirements.

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 performed until sometime after the associated LCO is within its A licabilit , re resent

Enclosure NL-20-033 Page 25 of 86 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 "required" when it can be and should be "required" 'Nhen 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. im----- nn Current Examples Proposed Example EXAMPLES The following examples illustrate EXAMPLES The following examples illustrate the various ways that Frequencies are the various ways that Frequencies are specified. In these examples, the Applicability specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, 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).

Example 1.4-1 ... Example 1.4-1 is modified to address an example of a Frequency that is utilized by TS 3.7.14.

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

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

Technical Specification 1.4 is modified to reflect the upcoming change in status regarding IP3.

This includes modifications to the description section, and to the examples. These proposed changes are administrative changes that reflect the changes to the other TSs and the remaining requirements.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 3.7.14, TS 3.7.15, and TS 3.7.16.

This section is modified to provide the rules of usage and examples that continue to be applicable for those TSs.

Example 1.4-1 is modified to address an example of a Frequency thatis utilized by TS 3.7.14.

This includes the elimination of the references to the term "operational," inoperable equipment, Modes, Example 1.4-3, and LCO 3.0.4, and replacing the term "unit" with "facility." Examples 1.4-2 and 1.4-3 are eliminated.

TECHNICAL SPECFICATION SECTION 2.0, SAFETY LIMITS (SLS)DELETED Technical Specification 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 ).

Technical Specification Section 2.0 is proposed for deletion in its entirety, since the safety limits do not a I to a reactor that is in a ermanentl defueled condition.

Enclosure NL-20-033 Page 26 of 86 A mark-up is provided to identify the section as deleted, because the TSs will not be renumbered.

Current TS 2.1 Proposed TS 2.1 TS 2.0 SLs TS 2.0 ShsDELETED TS 2.1.1, Reactor Core SLs Technical Specification 2.1.1 is proposed for deletion.

TS 2.1.2, RCS Pressure SL Technical Specification 2.1.2 is proposed for deletion.

TS 2.2, SL Violations Technical Specification 2.2 is proposed for deletion.

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

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

Basis Technical Specifications 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 over-pressurization 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.

Technical Specification 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. It is deleted, because SL 2.1.1 is deleted.

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 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . If the unit is MODE 3, 4, 5, or 6, it requires compliance to be restored within 5 minutes. It is deleted, because SL 2.1.2 is deleted.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 the reactor core 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-20-033 Page 27 of 86 TECHNICAL SPECIFICATION SECTION 3.0, LIMITING CONDITIONS FOR OPERATION (LCO) APPLICABILITY Technical Specification 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 POTS 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.e ...

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 IP3 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-20-033 Page 28 of 86 LCO 3.0.4 I 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. LCO 3.0.4 is not proposed for inclusion in the POTS since all actions in the remaining TS (i.e., TS 3.7.14, TS 3.7.15, and TS 3.7.16) have a completion time of "Immediately." This makes LCO 3.0.4 unnecessary. Thus, LCO 3.0.4 is no longer applicable in the permanently defueled condition.

LCO 3.0.5 I 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 3.0.5 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 I 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 POTS.

LCO 3.0.7 I 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 I 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 POTS. Thus, snubbers are not required to support any TS function.

Enclosure NL-20-033 Page 29 of 86 TECHNICAL SPECFICATION SECTION 3.0, SURVEILLANCE REQUIREMENT (SR) APPLICABILITY Technical Specification 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.

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 SR. .. Surveillances SR ... Surveillances 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.

SR 3.0.1 is modified by deleting the reference to MODES. Pursuant to 10 CFR 50.82(a)(2), the facility license for IP3 will no longer authorize operation of the reactor or placement or retention of fuel in the reactor.

MODES are not used in any POTS 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.

Enclosure NL-20-033 Page 30 of 86 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 eF as meas1:1Fes performance or as measured from the time frnm the time a specifies censitien ef the a specified condition of the Frequency is Fmq1:1ency is met.

met.

For Frequencies specified as "once," the FeF FFeq1:1encies specifies as "ence," the aeeve above interval extension does not apply. intOFval extensien sees net apply.

If a Completion Time requires periodic If a Gempletien +ime Feq1:1iFes peFiesic performance on a "once per ... " basis, the peFfeFmance en a "ence peF ... " easis, tile aee 1.ie above Frequency extension applies to each FFeq1:1ency extensien applies te eacn peFfeFmance performance after the initial performance. afteF the initial peFfeFmance.

Exceptions to this Specification are stated Exceptiens te tnis Specificatien aFe states in the in the individual Specifications. insii.1is1:1al Specificatiens.

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 POTS TSs.

Current SR 3.0.3 Proposed SR 3.0.3 If it is discovered that a Surveillance was If it is discovered that a Surveillance was not not performed within its specified performed within its specified Frequency, then Frequency, then compliance with the compliance with the requirement to declare the requirement to declare the LCO not met LCO not met may be delayed, from the time of may be delayed, from the time of discovery, 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 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 specified Frequency, whichever is greater. delay period is permitted to allow performance of This delay period is permitted to allow the Surveillance. +no selay peFies is enly performance of the Surveillance. The delay applicaele wnen theFe is a Feasenaele expectatien period is only applicable when there is a tile s1:1P1eillance 111ill ee met 1.vnen peFfeFmes. A reasonable expectation the surveillance risk evaluation shall be performed for any will be met when performed. A risk Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and evaluation shall be performed for any the risk impact shall be managed 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.

Enclosure NL-20-033 Page 31 of 86 SR 3.0.3 provides an allowance to delay declaring an LCO not met, when a surveillance is not performed within its required frequency. This requirement is revised by removing language that was included within License Amendment No. 266 to the IP3 Facility Operating License to incorporate TSTF-529 (Reference 5). This language identified that the delay permitted by the specification was only applicable when there is a reasonable expectation that the surveillance will be met when performed. This requirement is not necessary in the permanently shut down and defueled condition, because SR 3.7.14.1, SR 3.7.15.1, and SR 3.7.16.1 are the only remaining surveillance requirements in the POTS. These surveillances verify spent fuel pit water level, spent fuel pit boron concentration, and the initial enrichment and burnup of each fuel assembly and that the storage location meets LCO 3.7.16 requirements. These activities are not complex.

Thus, there is a reasonable expectation that they will be met when performed.

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 ,...,ith 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 1Nith ACTIONS or that Applicability that are required to comply with are part of a shutdovm of the unit.

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

SR 3.0.4 is modified by deleting the reference to MODES. Pursuant to 10 CFR 50.82(a)(2), the facility license for IP3 will no longer authorize operation of the reactor or placement or retention of fuel in the reactor.

MODES are not used in any POTS 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. The only remaining TSs with Required Actions are TS 3.7.14, TS 3.7.15, and TS 3.7.16, 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 IP3 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-20-033 Page 32 of 86 TECHNCIAL SPECIFICATION SECTION 3.1, REACTIVITY CONTROL SYSTEMS Technical Specification 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 IP3, 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.

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

Current IP3 TS Basis for Chanae TS 3.1.1, SHUTDOWN MARGIN (SOM) Technical Specification 3.1.1 is proposed for deletion.

Technical Specification 3.1.1 is applicable in MODE 2 with keff < 1.0, and MODES 3 through 5.

It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP3. 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 Technical Specification 3.1.2 is proposed for deletion.

Technical Specification 3.1.2 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 IP3. 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 aoolv in the permanently defueled condition.

Enclosure NL-20-033 Page 33 of 86 TS 3.1.3, Moderator Temperature Technical Specification 3.1.3 is proposed for Coefficient (MTC) deletion.

Technical Specification 3.1.3 is applicable in MODE 1 and MODE 2 with kett ~ 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 IP3. 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 3.1.4, Rod Group A,lignment Limits Technical Specification 3.1.4, including Table 3.1.4-1, is proposed for deletion.

Technical Specification 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 IP3. 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 Technical Specification 3.1.5 is proposed for deletion.

Technical Specification 3.1.5 is applicable in MODE 1 and MODE 2 with any control bank not fully inserted. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP3. 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.

Enclosure NL-20-033 Page 34 of 86 TS 3.1.6, Control Bank Insertion Limits Technical Specification 3.1.6 is proposed for deletion.

Technical Specification 3.1.6 is applicable in MODE 1 and MODE 2 with kett 2: 1.0. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP3.

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 Technical Specification 3.1.7 is proposed for deletion.

Technical Specification 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 IP3. 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 3.1.8, PHYSICS TESTS Exceptions - Technical Specification 3.1.8 is proposed for MODE2 deletion.

Technical Specification 3.1.8 is applicable in MODE 2 during PHYSICS TESTS. It will not be required after the certifications required under 10 CFR 50.82(a)(1) have been docketed for IP3.

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.

Enclosure NL-20-033 Page 35 of 86 TECHNICAL SPECIFICATION SECTION 3.2, POWER DISTRIBUTION LIMITS Technical Specification 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 IP3, 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.

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

Current IP3 TS Basis for Chanae TS 3.2.1, Heat Flux Hot Channel Factor Technical Specification 3.2.1 is proposed for (Fa(Z)) deletion.

Technical Specification 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 IP3. 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 Technical Specification 3.2.2 is proposed for Channel Factor (FNLiH) deletion.

Technical Specification 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 IP3. 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 permanentlv defueled condition.

Enclosure NL-20-033 Page 36 of 86 TS 3.2.3, AXIAL FLUX DIFFERENCE Technical Specification 3.2.3 is proposed for (AFD) (Constant Axial Offset Control deletion.

(CAOC) Methodology)

Technical Specification 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 IP3. 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 Technical Specification 3.2.4 is proposed for RATIO (QPTR) deletion.

Technical Specification 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 IP3. At that time, the 10 CFR Part SO. 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.

TECHNICAL SPECIFICATION SECTION 3.3, INSTRUMENTATION Technical Specification 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 IP3, 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.

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

Current IP3 TS Basis for Change TS 3.3.1, Reactor Protection System (RPS) Technical Specification 3.3.1, including Table Instrumentation 3.3.1-1, is proposed for 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 in those MODES.

Enclosure NL-20-033 Page 37 of 86 After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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). Operation in the applicable 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 Technical Specification 3.3.2, including Table Actuation System (ESFAS) Instrumentation 3.3.2-1, is proposed for 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 in those MODES.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.~2(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.

Enclosure NL-20-033 Page 38 of 86 TS 3.3.3, Post Accident Monitoring (PAM) Technical Specification 3.3.3, including Table Instrumentation 3.3.3-1, is proposed for deletion.

Technical Specification 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 IP3, 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 Technical Specification 3.3.4 is proposed for deletion.

Technical Specification 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 IP3, 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 Technical Specification 3.3.5 is proposed for Generator (DG) Start Instrumentation deletion Technical Specification 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."

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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

Enclosure NL-20-033 Page 39 of 86 Chapter 14 are no longer applicable in the permanently defueled condition. The analyses of the remaining DBAs (i.e., the FHA and the accidental 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 IP3 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 Technical Specification 3.3.6, including Table Pressure Relief Line Isolation 3.3.6-1, is proposed for deletion.

Instrumentation Technical Specification 3.3.6 is applicable in MODES 1 through 4 and during CORE ALTERATIONS, and during movement of irradiated fuel assemblies within containment.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 and CORE ALTERATIONS will no longer occur.

In addition, the POTS will not be implemented until after all of the fuel has been transferred from the reactor to the spent fuel pit. Thus, movement of irradiated fuel assemblies within the containment will no longer occur.

As a result, the containment purge system and pressure relief line isolation instrumentation will not be required in the permanently defueled confiquration.

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

Technical Specification 3.3.7 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR

Enclosure NL-20-033 Page 40 of 86 50.82(a)(1) are docketed for IP3, 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, CRVS actuation instrumentation will not be required in the permanently defueled configuration.

TS 3.3.8, Fuel Storage Building Emergency Technical Specification 3.3.8 is proposed for Ventilation System (FSBEVS) Actuation deletion.

Instrumentation Technical Specification 3.3.8 is applicable during movement of recently irradiated fuel in the fuel storage building. The Bases for TS 3.3.8 defines "recently irradiated" fuel as fuel that has occupied part of a critical reactor core within the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The POTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

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

TECHNICAL SPECIFICATION SECTION 3.4, REACTOR COOLANT SYSTEM (RCS)

Technical Specification 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 IP3, 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.

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

Current IP3 TS Basis for Change TS 3.4.1, RCS Pressure, *Temperature, and Technical Specification 3.4.1 is proposed for Flow Departure from Nucleate Boiling deletion.

(DNB) Limits Technical Specification 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 IP3, the 10 CFR Part 50 license will no longer authorize operation of the

Enclosure NL-20-033 Page 41 of 86 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 Technical Specification 3.4.2 is proposed for Criticality deletion.

Technical Specification 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 IP3, 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.

Enclosure NL-20-033 Page 42 of 86 TS 3.4.3, RCS Pressure and Temperature Technical Specification 3.4.3, including Figures (Prr) Limits 3.4.3-1 and 3.4.3-2, is proposed for deletion.

Technical Specification 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 restore the parameter(s) to within limit in accordance with Required Action C.1 and 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 IP3, 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 Prr limits are no longer applicable in the permanently defueled condition.

TS 3.4.4, RCS Loops - MODES 1 and 2 Technical Specification 3.4.4 is proposed for deletion.

Technical Specification 3.4.4 is applicable in MODES 1 and 2.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.

Enclosure NL-20-033 Page 43 of 86 TS 3.4.5, RCS Loops - MODE 3 Technical Specification 3.4.5 is proposed for deletion.

Technical Specification 3.4.5 is applicable in MODE 3.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.4.6 is proposed for deletion.

Technical Specification 3.4.6 is applicable in MODE4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.4.7 is proposed for Filled deletion.

Technical Specification 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 IP3, 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-20-033 Page 44 of 86 TS 3.4.8, RCS Loops - MODE 5, Loops Technical Specification 3.4.8 is proposed for Not Filled deletion.

Technical Specification 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 IP3, 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 Technical Specification 3.4.9 is proposed for deletion.

Technical Specification 3.4.9 is applicable in MODES 1 through 3.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.4.1 O is proposed for deletion.

Technical Specification 3.4.10 is applicable in MODES 1 through 3 and MODE 4 with all RCS cold leg temperatures > 330°F.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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-20-033 Page 45 of 86 TS 3.4.11, Pressurizer Power Operated Technical Specification 3.4.11 is proposed for Relief Valves (PO RVs) deletion.

Technical Specification 3.4.11 is applicable in MODES 1 through 3.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 3.4.12, Low Temperature Overpressure Technical Specification 3.4.12, including Figures Protection (LTOP) 3.4.12-1 through 3.4.12-3, is proposed for deletion.

Technical Specification 3.4.12 is applicable whenever the Residual Heat Removal System (RHR) System is not isolated from the RCS, in MODE 4 when any RCS cold leg temperature is

330°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 IP3, 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 4 through 6 will no longer occur.

In addition, LTOP was provided to protect the RCS from over-pressurization transients during shut down, in part, by providing a sufficient size RCS vent. In permanent shut down, the RCS is partially drained and adequately vented to prevent over-pressurization.

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

Enclosure NL-20-033 Page 46 of 86 TS 3.4.13, RCS Operational LEAKAGE Technical Specification 3.4.13 is proposed for deletion.

Technical Specification 3.4.13 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.4.14 is proposed for (PIV) Leakage deletion.

Technical Specification 3.4.14 is applicable in MODES 1 through 3 and MODE 4 with an exception for leakage limits for valves in the 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 IP3, 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 3.4.15, RCS Leakage Detection Technical Specification 3.4.15 is proposed for Instrumentation deletion.

TechnicalSpecification 3.4.15 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 aoolicable in a permanently defueled condition.

Enclosure NL-20-033 Page 47 of 86 TS 3.4.16, RCS Specific Activity Technical Specification 3.4.16 is proposed for deletion.

Technical Specification 3.4.16 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.4.17 is proposed for Integrity deletion.

Technical Specification 3.4.17 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 aoolicable in a oermanentlv defueled condition.

TECHNICAL SPECIFICATION SECTION 3.5, EMERGENCY CORE COOLING SYSTEMS (ECCS)

Technical Specification 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 IP3, 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.

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

Enclosure NL-20-033 Page 48 of 86 Current IP3 TS Basis for Change TS 3.5.1, Accumulators Technical Specification 3.5.1 is proposed for deletion.

Technical Specification 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 IP3, 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 Technical Specification 3.5.2 is proposed for deletion.

Technical Specification 3.5.2 is applicable in MODES 1 through 3.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Ln a permanently defueled condition.

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

Technical Specification 3.5.3 is applicable in MODE4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 aoolicable in a permanently defueled condition.

Enclosure NL-20-033 Page 49 of 86 TS 3.5.4, Refueling Water Storage Tank Technical Specification 3.5.4 is proposed for (RWST) deletion.

Technical Specification 3.5.4 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.

TECHNICAL SPECIFICATION SECTION 3.6, CONTAINMENT SYSTEMS Technical Specification 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 IP3, 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.

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

Current IP3 TS Basis for Change TS 3.6.1, Containment Technical Specification 3.6.1 is proposed for deletion.

Technical Specification 3.6.1 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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-20-033 Page 50 of 86 TS 3.6.2, Containment Air Locks Technical Specification 3.6.2 is proposed for deletion.

Technical Specification 3.6.2 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 aoolicable in a permanently defueled condition.

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

Technical Specification 3.6.3 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are d.ocketed for IP3, 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 3.6.4, Containment Pressure Technical Specification 3.6.4 is proposed for deletion.

Technical Specification 3.6.4 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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-20-033 Page 51 of 86 TS 3.6.5, Containment Air Temperature Technical Specification 3.6.5 is proposed for deletion.

Technical Specification 3.6.5 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.6.6 is proposed for Containment Fan Cooler System deletion.

Technical Specification 3.6.6 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.6.7 is proposed for deletion.

Technical Specification 3.6.7 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.6.8 is proposed for deletion. The deletion of this placeholder is an administrative chani:::ie.

Enclosure NL-20-033 Page 52 of 86 TS 3.6.9, Isolation Valve Seal Water Technical Specification 3.6.9 is proposed for (IVSW) System deletion.

Technical Specification 3.6.9 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.6.10 is proposed for Pressurization System (WC&PPS) deletion.

Technical Specification 3.6.1 O is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.

TECHNICAL SPECIFICATION SECTION 3.7, PLANT SYSTEMSSPENT FUEL PIT REQUIREMENTS Technical Specification 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 IP3, 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.

Technical Specification 3.7 is retitled to reflect that the remaining TSs address SFP requirements.

Technical Specification 3.7.1 through TS 3.7.13 and TS 3.7.17 are proposed for deletion in their entirety. Thus, mark-ups of these TSs are not provided.

Technical Specification 3.7.14 provides the limit regarding the SFP water level. It will be retained in the POTS, and modified to eliminate the reference to LCO 3.0.3.

Enclosure NL-20-033 Page 53 of 86 Technical Specification 3.7.15 provides requirements regarding the SFP boron concentration. It will be retained in the POTS, and modified to eliminate the reference to LCO 3.0.3.

Technical Specification 3. 7 .16 provides the limits for storing fuel assemblies in the SFP. It will be retained in the POTS, and modified to eliminate the reference to LCO 3.0.3.

Mark-ups of TS 3.7.14, TS 3.7.15, and TS 3.7.16 are provided in Attachment 1 to this enclosure.

Current IP3 TS Basis for Chanae 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 Technical Specification 3.7.1, including Tables (MSSVs) 3.7.1-1 and 3.7.1-2, is proposed for deletion.

Technical Specification 3. 7 .1 is applicable in MODES 1 through 3.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.7.2 is proposed for (MSIVs) and Main Steam Check Valves deletion.

(MSCVs)

Technical Specification 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 IP3, 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 Boiler Feedpump Discharge Technical Specification 3.7.3 is proposed for Valves (MBFPDVs), Main Feedwater deletion.

Regulation Valves (MFRVs), Main Feedwater Inlet Isolation Valves (MFIIVs)

Enclosure NL-20-033 Page 54 of 86 and Main Feedwater (MF) Low Flow Technical Specification 3.7.3 is applicable in Bypass Valves MODES 1 through 3 except when each main feedwater and bypass line is isolated by a closed and de-activated motor/air operated valve or isolated by a closed manual valve.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 3.7.4, Atmospheric Dump Valves Technical Specification 3.7.4 is proposed for (ADVs) deletion.

Technical.Specification 3.7.4 is applicable in MODES 1 through 3 and MODE 4 when the steam generator is relied upon for heat removal.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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) Technical Specification 3.7.5 is proposed for System deletion.

Technical Specification 3.7.5 is applicable in MODES 1 through 3 and MODE 4 when the steam generator is relied upon for heat removal.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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) Technical Specification 3.7.6 is proposed for deletion.

Enclosure NL-20-033 Page 55 of 86 Technical Specification 3.7.6 is applicable in MODES 1 through 3 and MODE 4 when the steam generator is relied upon for heat removal.

After the certifications required by 10 CFR .

50.82(a)(1) are docketed for IP3, 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 City Water (CW) Technical Specification 3.7.7 is proposed for deletion.

Technical Specification 3.7.7 is applicable in MODES 1 through 3 and MODE 4 when the steam generator is relied upon for heat removal.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 3.7.8, Component Cooling Water (CCW) Technical Specification 3.7.8 is proposed for System deletion.

Technical Specification 3.7.8 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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, Service Water System (SWS) Technical Specification 3.7.9 is proposed for deletion.

Technical Specification 3.7.9 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR

Enclosure NL-20-033 Page 56 of 86 50.82(a)(1) are docketed for IP3, 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, Ultimate Heat Sink (UHS) Technical Specification 3.7.10 is proposed for deletion.

Technical Specification 3. 7 .10 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 3.7.11, Control Room Ventilation Technical Specification 3.7.11 is proposed for System (CRVS) deletion.

Technical Specification 3. 7.11 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 IP3, 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 3.7.11 defines "recently irradiated" fuel as fuel that has occupied part of a critical reactor core within the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The POTS 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.

TS 3.7.12, Control Room Air Conditioning Technical Specification 3.7.12 is proposed for System (CRACS) deletion.

Technical Specification 3.7.12 is applicable in MODES 1 through 4.

Enclosure NL-20-033 Page 57 of 86 After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.

TS 3. 7 .13, Fuel Storage Building Technical Specification 3.7.13 is proposed for Emergency Ventilation System (FSBEVS) deletion.

Technical Specification 3.7.13 is applicable during movement of recently irradiated fuel assemblies in the fuel storage building.

The Bases for TS 3. 7.13 defines "recently irradiated" fuel as fuel that has occupied part of a critical reactor core within the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months .

The POTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

As a result, the FSBEVS will not be required in the permanently defueled confiQuration.

TS 3. 7 .14, Spent Fuel Pit Water Level Technical Specification 3. 7 .14 is retained in the POTS. The title of TS Section 3. 7 is administratively changed from addressing "plant systems" to addressing "spent fuel pit requirements" to comport with the remaining POTS 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 3.7.15, Spent Fuel Pit Boron Technical Specification 3.7.15 is retained in the Concentration POTS. The title of TS Section 3. 7 is administratively changed from addressing "plant systems" to addressing "spent fuel pit requirements" to comport with the remaining POTS 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.

Enclosure NL-20-033 Page 58 of 86 TS 3.7.16, Spent Fuel Assembly Storage Technical Specification 3. 7 .16, including Figure 3.7.16-1, is retained in the POTS. The title of TS Section 3.7 is administratively changed from addressing "plant systems" to addressing "spent fuel pit requirements" to comport with the remaining POTS 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 3.7.17, Secondary Specific Activity Technical Specification 3. 7.17 is proposed for deletion.

Technical Specification 3. 7 .17 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.

TECHNICAL SPECIFICATION SECTION 3.8, ELECTRICAL POWER SYSTEMS Technical Specification 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 IP3, 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 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 I P3 permanently shut down and defueled.

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

Current IP3 TS Basis for Change TS 3.8.1, AC Sources - Operating Technical Specification 3.8.1 is proposed for deletion.

Technical Specification 3.8.1 is applicable in MODES 1 through 4.

Enclosure NL-20-033 Page 59 of 86 After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.8.2 is proposed for deletion.

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

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 DBAs and transients analyzed in UFSAR Chapter 14 will no longer be applicable in the permanently defueled condition, with the exception of the FHA 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 IP3 permanently shut down and defueled.

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

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

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

Technical Specification 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 POTS because the TSs that it supports are no longer required after IP3 is permanently shut down and defueled.

Enclosure NL-20-033 Page 60 of 86 TS 3.8.4, DC Sources - Operating Technical Specification 3.8.4 is proposed for deletion.

Technical Specification 3.8.4 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.8.5 is proposed for deletion.

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

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 DBAs and transients analyzed in UFSAR Chapter 14 will no longer be applicable in the permanently defueled condition, with the exception of the FHA 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 IP3 permanently shut down and defueled.

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

TS 3.8.6, Battery Cell Parameters Technical Specification 3.8.6, including Table 3.8.6-1, is proposed for deletion.

Technical Specification 3.8.6 is applicable when the associated Direct Current (DC) electrical power subsystems are required to be OPERABLE.

Technical Specification 3.8.4 and TS 3.8.5 provide the OPERABILITY requirements regarding the DC

Enclosure NL-20-033 Page 61 of 86 sources. As previously discussed, these TSs are proposed for deletion. Thus, TS 3.8.6 is not included in the POTS because the TSs that it supports are no longer required after IP3 is permanently shut down and defueled.

TS 3.8.7, Inverters - Operating Technical Specification 3.8. 7 is proposed for deletion.

Technical Specification 3.8. 7 is applicable in MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 aoolicable in a permanently defueled condition.

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

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

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 DBAs and transients analyzed in UFSAR Chapter 14 will no longer be applicable in the permanently defueled condition, with the exception of the FHA 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 I P3 permanently shut down and defueled.

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

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

Technical Specification 3.8.9 is applicable in

Enclosure NL-20-033 Page 62 of 86 MODES 1 through 4.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 aoolicable in a permanently defueled condition.

TS 3.8.10, Distribution Systems - Technical Specification 3.8.1 O is proposed for Shutdown deletion.

Technical Specification 3.8.10 is applicable in MODES 5 and 6, and during movement of irradiated fuel assemblies.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 DBAs and transients analyzed in UFSAR Chapter 14 will no longer be applicable in the permanently defueled condition, with the exception of the FHA 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 IP3 permanently shut down and defueled.

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

TECHNICAL SPECIFICATION SECTION 3.9, REFUELING OPERATIONS Technical Specification 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 IP3, 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 and the accidental releases of

Enclosure NL-20-033 Page 63 of 86 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 I P3 permanently shut down and defueled.

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

Current IP3 TS Basis for Change TS 3.9.1, Boron Concentration Technical Specification 3.9.1 is proposed for deletion.

Technical Specification 3.9.1 is applicable in MODE 6.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 3.9.2, Nuclear Instrumentation Technical Specification 3.9.2 is proposed for deletion.

Technical Specification 3.9.2 is applicable in M0DE6.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Technical Specification 3.9.3 is proposed for deletion.

Technical Specification 3.9.3 is applicable during movement of recently irradiated fuel assemblies within containment. As stated in the Bases for TS 3.9.3, the release of radioactivity from the containment following an FHA is limited by several conditions, including a minimum decay time of 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months prior to moving irradiated fuel. "Recently irradiated" fuel is fuel that has occupied part of a critical reactor core within the previous 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months .

The POTS will not be implemented until after that time period, so that the specific condition of Applicability will no longer occur.

Enclosure NL-20-033 Page 64 of 86 As a result, this TS will not be applicable in a permanently defueled condition.

TS 3.9.4, Residual Heat Removal (RHR) Technical Specification 3.9.4 is proposed for and Coolant Circulation - High Water Level deletion.

Technical Specification 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 IP3, 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) Technical Specification 3.9.5 is proposed for and Coolant Circulation - Low Water Level deletion.

Technical Specification 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 IP3, 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 3.9.6, Refueling Cavity Water Level Technical Specification 3.9.6 is proposed for deletion.

Technical Specification 3.9.6 is applicable during CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, and during movement of irradiated fuel assemblies within containment.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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, CORE AL TE RATIONS with the

Enclosure NL-20-033 Page 65 of 86 specified conditions will no longer occur.

In addition, the POTS will not be implemented until after all of the fuel has been transferred from the reactor to the spent fuel pit. Thus, movement of irradiated fuel assemblies within the containment will no longer occur.

As a result, this TS will no longer be applicable in the permanently defueled condition.

TECHNICAL SPECIFICATION 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 IP3, 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.

Technical Specification 4.3.1.1 is modified by making grammatical corrections regarding punctuation.

Technical Specification 4.3.1.2 is proposed for deletion.

Current IP3 TS Basis for Change Current TS 4.3.1.1 Proposed TS 4.3.1.1 4.3.1.1 The spent fuel storage racks are 4.3.1.1 The spent fuel storage racks are designed designed and shall be maintained and shall be maintained with:

with:

a. Fuel assemblies having a maximum

a. Fuel assemblies having a U-235 enrichment of 5.0 weight maximum U-235 enrichment percent; of 5.0 weight percent; b. kett :s; 0.95 if assemblies are inserted in

b. kett :s; 0.95 if assemblies are accordance with Technical inserted in accordance with Specification 3.7.16, Spent Fuel Technical Specification Assembly Storage.,.;

3.7.16, Spent Fuel Assembly c. A nominal 9.075 inch center to center Storage. distance between fuel assemblies

c. A nominal 9.075 inch center placed in the high density fuel storage to center distance between racks (Region II);

fuel assemblies placed in the d. A nominal 10. 76 inch center to center high density fuel storage distance between fuel assemblies racks (Region II); placed in low density fuel storage

d. A nominal 10. 76 inch center racks (Region I};-.

to center distance between fuel assemblies placed in low density fuel storage racks (Region I);

Enclosure NL-20-033 Page 66 of 86 Basis Technical Specification 4.3.1.1 is modified to correct qrammatical errors reqarding punctuation.

Current TS 4.3.1.2 Proposed TS 4.3.1.2 The new fuel storage racks are designed None and shall be maintained with ...

Basis Technical Specification 4.3.1.2 is proposed for deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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). Indian Point Nuclear Generating Station Unit 3 will never acquire new fuel again. Thus, this TS is not applicable in the permanently shut down and defueled condition.

TECHNICAL SPECIFICATION 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 IP3 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 UFSARDSAR and Quality Assurance Plan, the UFSAR and Quality Assurance Plan, as appropriate; as appropriate; Basis

Enclosure NL-20-033 Page 67 of 86 Technical Specification 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 IP3, the IP3 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 FSAR.DSAR; Updated FSAR.

b. Deleted;

b. Deleted C. Quality assurance for effluent and C. Quality assurance for effluent and environmental monitoring; environmental monitoring;

d. Fire Protection Program implementation

d. Fire Protection Program implementation; Deleted; and ...

and ...

Basis Following the permanent shut down and defueling of IP3, the IP3 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.

Technical Specification 5.4.1.d is proposed for deletion. This is consistent with the proposed deletion of License Condition 2.H. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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. Indian Point Nuclear Generating Station Unit 3 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.

Enclosure NL-20-033 Page 68 of 86 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 5.5.4, Radioactive Effluent Controls Technical Specification 5.5.4 will be retained, but Program TS 5.5.4.d, TS 5.5.4.h, and TS 5.5.4.i are proposed to be modified to replace "unit" with "uniUfacility." These are administrative changes to the IP3 TSs. It establishes consistency with the IP2 POTS.

TS 5.5.5, Component Cyclic or Transient Technical Specification Q.5.5 is proposed for Limit deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.

TS 5.5.6, Reactor Coolant Pump Flywheel Technical Specification 5.5.6 is proposed for Inspection Program deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.

Technical Specification 5.5.6 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 proQram.

TS 5.5.7, lnservice Testing Program Technical Specification 5.5.7 is proposed for deletion.

Technical Specification 5.5.7 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 OBA.

Enclosure NL-20-033 Page 69 of 86 As such, the inservice testing program will no longer be relevant in the permanently shut down and defueled condition.

TS 5.5.8, Steam Generator (SG) Program Technical Specification 5.5.8 is proposed for deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.

Technical Specification 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 proQram is aooropriate.

TS 5.5.9, Secondary Water Chemistry Technical Specification 5.5.9 is proposed for Program deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 5.5.10, Ventilation Filter Testing Program Technical Specification 5.5.1 O is proposed for (VFTP) deletion.

As previously discussed, TS 3.6.6, Containment Spray System and Containment Fan Cooler System, and TS 3.7.11, Control Room Ventilation System (CRVS), are proposed for deletion. Thus, this support program is not required in the permanently shut down and defueled condition.

Enclosure NL-20-033 Page 70 of 86 TS 5.5.12, Diesel Fuel Oil Testing Program Technical Specification 5.5.12 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 diesel generators. Thus, this support program is not required in the permanently shut down and defueled condition.

TS 5.5.13, Technical Specifications (TS) Technical Specification 5.5.13 will be retained, but Bases Control Program modified by replacing the references to the "updated UFSAR" and UFSAR in TS 5.5.13.b.2 and TS 5.5.13.c with references to the DSAR.

Following the permanent shut down and defueling of IP3, the IP3 UFSAR will be updated to reflect this condition. The document will be retitled as the DSAR. Thus, these proposed changes are administrative changes.

TS 5.5.14, Safety Function Determination Technical Specification 5.5.14 is proposed for Program (SFDP) deletion.

This program was established to ensure loss of safety function is detected and appropriate actions taken. The LCOs remaining in the POTS 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.

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 5.5.15, Containment Leakage Rate Technical Specification 5.5.15 is proposed for Testing Program deletion.

The IP3 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 3.6.1 O regarding the

Enclosure NL-20-033 Page 71 of 86 containment systems are proposed for deletion.

Thus, the deletion of this TS is aooropriate.

TS 5.5.16, Control Room Envelope Technical Specification 5.5.16 is proposed for Habitability Program deletion.

As previously discussed, TS 3.7.11, 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 Technical Specification 5.6.5 is proposed for (COLR) deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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, there will no longer be a need to establish core operating limits. As a result, this TS will not be applicable in the permanently defueled condition.

TS 5.6.6, Not Used The placeholder for TS 5.6.6 is proposed for deletion. This is an administrative change to reflect reorganization of the TS.

TS 5.6.7, Post Accident Monitoring Technical Specification 5.6.7 is proposed for Instrumentation (PAM) Report deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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).

Technical Specification 3.3.3 provides the operability requirements for the PAM instrumentation. It is 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. 7 reportinQ details is appropriate.

Enclosure NL-20-033 Page 72 of 86 TS 5.6.8, Steam Generator Tube Inspection Technical Specification 5.6.8 is proposed for Report deletion.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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.

Technical Specification 3.4.17 provides the requirements to ensure SG tube integrity in MODES 1 through 4. It is proposed for deletion.

In addition, TS 5.5.8, Steam Generator (SG)

Program, is proposed for deletion. Thus, the proposed deletion of this supporting TS program is appropriate.

APPENDIX B TO FACILITY QPeRA+ING LICENSE Current Cover Page for Part I Pro12osed Cover Page for Part I APPENDIX B TO FACILITY OPERATING APPENDIX B TO FACILITY OPERATING LICENSE ... LICENSE ...

Basis Appendix B, Part I, 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 IP3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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 Pro12osed 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:

(1) Verify that the plant is operated in an environmentally acceptable manner, as (1) Verify that the plant is operatedfaci/ity is established by the FES and other NRG maintained in an environmentally environmental impact assessments. acceptable manner, as established by the FES and other NRG environmental impact assessments.

Enclosure NL-20-033 Page 73 of 86 (2) Coordinate NRC requirements and maintain (2) Coordinate NRC requirements and maintain consistency with other Federal, State and consistency with other Federal, State and local requirements for environmental local requirements for environmental protection. protection.

(3) Keep NRC informed of the environmental (3) Keep NRC informed of the environmental effects of facility construction and operation effects of handling and storage of spent and of actions taken to control those effects. fuel and maintenance of the facility construction and operation and of actions taken to control those effects.

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

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 ENO may make changes in station design or ENO may make changes in stationfacility operations or perform tests or design or operations or perform tests or experiments ... Changes in plant design or experiments ... Changes in f}lamfacility design operation or performance of tests or 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-Bf power level; ... -- *-- ,_ -'*

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 IP3, 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-20-033 Page 74 of 86 Current Section 3.3 Proposed Section 3.3 Changes in plant design or operation and ... Changes in J*affifacility design or operation and ...

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 f*8ffi operation shall be recorded and ... operation the handling and storage of spent fuel and maintenance of the facility shall be recorded and ...

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.

The proposed change to Section 4.2 concludes that the Biological Opinion rendered during the evaluation of the continued operation of IP2 and I P3 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 IP3 will not impact the Biological Opinion regarding the shortnose sturgeon and Atlantic sturgeon in an adverse manner when compared to the continued operation of IP2 and IP3.

Enclosure NL-20-033 Page 75 of 86 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 inspection. These records and logs shall be made available to the NRC on request.

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

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," and an editorial correction to replace "en\(ironmental" with "environment" is also made. 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

... and an assessment of the observed impacts ... and an assessment of the observed impacts of the plant operation on the environment. .. of the previous plant operation and the handling and storage of spent fuel and maintenance of the facility on the environment. ..

(b) A list of all changes in station design or (b) A list of all changes in stationfaci/ity 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 ...

The proposed changes to Section 5.4 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, a reference to "station" is replaced with a reference to "facility." These proposed changes reflect the revised mission of the facility in the permanently shut down and defueled condition.

Enclosure NL-20-033 Page 76 of 86 Current Section 5.4.2 Proposed Section 5.4.2

... The report shall (a) describe, analyze, and ... The report shall (a) 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, (b) ... sl:laFasteFistissfacilitv conditions, (b) ...

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.

APPENDIX B TO FACILITY OPERA+ING LICENSE Current Cover Page for Part II Proposed Cover Page for Part 11 APPENDIX B TO FACILITY OPERATING APPENDIX B TO FACILITY OPERATING LICENSE ... LICENSE ...

Basis Appendix B, Part II, is modified by replacing the references to "Facility Operating License" with "Facility License." This change reflects the upcoming change in status regarding IP3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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).

APPENDIX C TO FACILITY OPERA+ING LICENSE Current Cover Page for Part I Proposed Cover Page for Part I APPENDIX C TO FACILITY OPERATING APPENDIX C TO FACILITY OPERATING LICENSE ... LICENSE ...

Current Header for Part I Proposed Header for Part I Facility Operating License ... Facility OpeFating License ...

Current Cover Page for Part II Proposed Cover Page for Part II APPENDIX C TO FACILITY OPERATING APPENDIX C TO FACILITY OPERATING LICENSE ... LICENSE ...

Basis Appendix C, Parts I and II, are modified by replacing the references to "Facility Operating License" with "Facility License." These changes reflect the upcoming change in status regarding IP3. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, 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-20-033 Page 77 of 86 The mark-ups of the Appendix A TSs Bases and re-typed versions of the POTS 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 5.5.13, "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 IP3 would permanently cease operations no later than April 30, 2021 (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 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 fo 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 1O of the Code of Federal Regulations (1 O 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 IP3 in the permanently shut down and defueled condition, this criterion is not applicable.

Enclosure NL-20-033 Page 78 of 86 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 OBA 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 IP3 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 OBA 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 IP3, 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 IP3 is discussed in more detail within this license amendment request.

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 IP3 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 IP3 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.

Enclosure NL-20-033 Page 79 of 86 This proposed amendment deletes the portions of the previous IP3 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, 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."

Enclosure NL-20-033 Page 80 of 86 10 CFR 50, Appendix A. General Design Criteria (GDC) for Nuclear Power Plants Section 1.3 of the IP3 UFSAR states:

"The General Design Criteria presented and discussed in specific sections of the FSAR (which describe systems, structures, equipment and components important to safety) are those which were in effect at the time when Indian Point 3 was designed and constructed. The General Design Criteria which formed the bases for the Indian Point 3 design were published by the Atomic Energy Commission in the Federal Register of July 11, 1967 and subsequently made part of 10 CFR 50.

The Authority completed a study of the method by which the Indian Point 3 facility complied with the safety rules and regulations, in particular those contained in 10 CFR Parts 20 and 50, that were in effect at the time of the study. The study was conducted in accordance with the provisions of NRC Confirmatory Order of February 11, 1980 and were submitted to the NRC on August 11, 1980. The NRC audit of submittal indicated that the Indian Point 3 design and operation meet the applicable regulations. The following sections provide the results of the compliance study, updated to reflect changes made to the configuration since the study was completed."

The Indian Point Nuclear Generating Station Unit 3 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-20-033 Page 81 of 86 Design Basis Accidents (DBAs)

Section 14 of the IP3 UFSAR describes the OBA 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 IP3, 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 IP3 is in the permanently defueled condition. The only remaining DBAs will be the FHA 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 In accordance with Title 10 of the Code of Federal Regulations (1 O CFR) 50.92, Entergy Nuclear Operations, Inc. (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 Nuclear Generating Station Unit No. 3 (IP3) no later than April 30, 2021 (Reference 1). After the certifications for permanent cessation of operations and permanent fuel removal from the reactor vessel are docketed for IP3, the 10 CFR Part 50 license for IP3 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 IP3 Facility Operating License (FOL); revise the Technical Specifications (TSs) in Appendix A of the FOL to Permanently Defueled Technical Specifications (POTS); revise the Environmental Technical Specification Requirements in Appendix B of the FOL; and revise 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 IP3 FOL and TSs and remove the requirements that would no longer be applicable after IP3 is permanently shut down and defueled.

The existing IP3 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. The LCOs and associated Surveillance Requirements (SRs) that will not apply in the permanently defueled condition are being proposed for deletion. The remaining portions of the

Enclosure NL-20-033 Page 82 of 86 Appendix A TSs are being proposed for revision and incorporation as the POTS 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 IP3 has permanently ceased operation, entered a permanently defueled condition, met the decay requirements established in the analysis of the Fuel Handling Accident (FHA), and implemented the NRC approved license amendment regarding administrative controls for the permanently defueled condition. The proposed amendment would modify the IP3 FOL and TSs in Appendices A through C by deleting the portions of the FOL 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 IP3 Updated Final Safety Analysis Report (UFSAR) describes the OBA and transient scenarios applicable to IP3 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 IP3 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 IP3 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, reactor heat removal, reactor coolant system (RCS) inventory control, and containment integrity are no longer applicable at IP3 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 IP3.

Enclosure NL-20-033 Page 83 of 86 After IP3 permanently ceases operation, the future generation of fission products will cease and the remaining source term will decay. The radipactive 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. The 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 condition 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 IP3 FOL 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 IP3 will no longer be authorized to operate the reactor.

The proposed deletion and modification of requirements of the IP3 FOL and Appendices A through C TSs do not affect systems credited in the accidents that remain applicable at IP3 in the permanently defueled condition. The proposed FOL and TSs will continue to require proper control and monitoring of safety significant parameters and activities.

Enclosure NL-20-033 Page 84 of 86 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 OBA.

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 condition 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 IP3 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 IP3 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 IP3 FOL and TSs in Appendices A through C by deleting the portions of the FOL 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 IP3 FOL 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 IP3 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.

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.

Enclosure NL-20-033 Page 85 of 86 3.3 PRECEDENT The proposed changes to the IP3 FOL and Appendices A through C TSs are consistent with the intent of the license and accompanying POTS 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 6); (2) Oyster Creek Nuclear Generating Station, for which an amendment was issued on October 26, 2018 (Reference 7); (3) San Onofre Nuclear Generating Station, Units 2 and 3, for which an amendment was issued on July 17, 2015 (Reference 8); (4) Crystal River Nuclear Plant, Unit 3, for which an amendment was issued on September 4, 2015 (Reference 9).

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 in 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-20-033 Page 86 of 86

5. REFERENCES

1. Entergy Nuclear Operations, Inc. (Entergy) letter to U.S. Nuclear Regulatory Commission

(NRC), "Notification of Permanent Cessation of Power Operations," (Letter No. NL-17-021)

(ADAMS Accession No. ML17044A004), dated February 8, 2017

2. NRC letter to Entergy, "Indian Point Nuclear Generating Unit Nos. 2 and 3 - Issuance of Amendment Nos. 292 and 267 RE: Changes to Technical Specification Sections 1.1,

'Definitions'; 4.0, 'Design Features'; and 5.0, 'Administrative Controls,' for a Permanently Defueled Condition (EPID L-2019-LLA-0081)," (ADAMS Accession No. ML20071Q717),

dated April 10, 2020

3. Entergy letter to NRC, "Proposed Technical Specifications (TS) Changes - Indian Point Nuclear Generating Unit 3 TS SR 3.7.7.2 and TS 3.7.6, Required Action A.1," (Letter No.

NL-19-093) (ADAMS Accession No. ML19325E913), dated November 21, 2019

4. Holtec Decommissioning International, LLC (HDI) letter to NRC, "Post Shutdown Decommissioning Activities Report including Site-Specific Decommissioning Cost Estimate for Indian Point Nuclear Generating Units 1, 2, and 3," (ADAMS Accession No. ML19354A698), dated December 19, 2019

5. NRC letter to Entergy Services, LLC, "Arkansas Nuclear One, Units 1 and 2; Grand Gulf Nuclear Station, Unit 1; Indian Point Nuclear Generating Unit Nos. 2 and 3; Palisades Nuclear Plant; River Bend Statiop, Unit 1; Waterford Steam Electric Station, Unit 3 - Re:

Issuance of Amendments to Adopt TSTF-529, 'Clarify Use And Application Rules' (EPID L-2019-LLA-0013)," (ADAMS Accession No. ML19175A042), dated September 11, 2019

6. NRC letter 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, EPID L-2017-LLA-0192)," (ADAMS Accession No. ML18010A087), dated March 6, 2018

7. NRC letter 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)," (ADAMS Accession No. ML18227A338), dated October 26, 2018

8. NRC letter 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),"

(ADAMS Accession No. ML15139A390), dated July 17, 2015

9. NRC letter 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)," (ADAMS Accession No. ML152248286),

dated September 4, 2019

Enclosure, Attachment 1 NL-20-033 Indian Point Nuclear Generating Station Unit 3 Mark-up of the Current Facility Operating License, Appendices A through C Technical Specifications, and Appendix A Technical Specifications Bases

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 RENEWED FACILITY OPERATING LICENSE Renewed License No. DPR-64

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

A. The application for a renewed license filed by Entergy Nuclear Indian Point 3, LLC (ENIP3) (the licensee) and Entergy Nuclear Operations, Inc. (ENO) (operator) for Indian Point Nuclear Generating Unit No. 3 (IP3 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; be maintained B. The facility will in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. 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 Commission's regulations; D. ENIP3 and ENO are financially and technically qualified to engage in Arndt. 203 the activities authorized by this amendment; 11/27/00 E. ENIP3 and ENO have satisfied the applicable provisions of 10 CFR Arndt. 203 Part 140, "Financial Protection Requirements and Indemnity 11/27/00 Agreements" of the Commission's regulations; F. 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; G. 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;

H. The issuance of this renewed license is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. 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 facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commission's regulations.

2. Accordingly, Renewed Facility Operating License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows: maintained A. This renewed license applies to the Indian Pain Arndt. 203 Unit No. 3, a pressurized water nuclear reacto nd associated 11/27/00 equipment (the facility), owned by ENIP3 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 "F"inal F"acility Description and afety Analysis Report" as supplemented and amended, an t e Environmental Report, as amended. Defueled B. Subject to the conditions and requirements incorporated herein, the Commission licenses:

and (1) Pursuant to Section 104b of the Act and 1O CF Part 50, Arndt. 203 "Licensing of Production and Utilization Faciliti s," (a) ENIP3 11/27/00 to possess and use, and (b) ENO to possess, se afla 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; that was used (2) ENO pursuant to the Act and 10 CFR Part 70, to fe6eM3::- Arndt. 203 possess, and use, at any time, special nuclear material s 11/27/00 reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final F"acility Description and fety Analysis Report, as supplemented and amended; Defueled that were used ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, Arndt. 203 t eceive, possess, and use, at any time, any byproduct 11/-27.'00 sou e and special nuclear material as sealed neutron sourc for reactor startup, sealed sources for ~ r ~

instrumentation an adiation monitoring e ipment ~

calibration, a as fi sion detectors in amoun as required; are used in the that were used that were used calibration of

(4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, Arndt. 203 to receive, possess, and use in amounts as required any 11/27/00 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 Arndt. 203 possess, but not separate, such byproduct and special 11/27/00 nuclear materials 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; and 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 below:

Deleted per (1) Mmcimum Pov,er Level ~----iAmendment [###]

E:~JO is authorized to operate the facility at steady state reactor core po'a'Ver levels not in mccess of 3216 megavvatts thermal (100% of rated power).

(2) Technical Specifications ~

The Technical Specificati<yPs contained in Appendices A, B, and C, as revised through Amendment No. ~ . are hereby incorporated in the renewed license.

ENO shall the facility in accordance with the Technical Specifications.

f3t (DE:LE:TE:D) Arndt. 205 2 27 01 f41 (DE:LE:TE:D) Arndt. 205 2 27 01 D. (DELETED) Arndt. 46 2 16 g3 E. (DELETED) Arndt. 37 5 14 g1 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.

Amendment No -2e8-

G. 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 CFR 50.54(p). The combined set of plans 1 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 O," and was submitted by letter dated October 14, 2004, as supplemented by letter dated May 18, 2006.

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. 243, as supplemented by changes approved by License Amendment Nos. 254, 260, and 263.

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 Deleted per letters dated April 27, 2011, and January 4, 2012. ENO shall fully implement and Amendment maintain in effect the provisions of the Commission-approved authorization.

[###]

Protection Program as described in tt:ie Final Safety Analysis Report for Indian Point

~Juclear Generating Unit No. a and as approved in NRG fire protection safety evaluations (Ses) dated September 21, 197a, Marsh 6, 1979, May 2, 1980,

~Jovember 18, 1982, December ao, 1982, February 2, 1984, April 16, 1984, January 7, 1987, September 9, 1988, October 21, 1991, April 20, 1994, January a, 1995, and supplements thereto, subject to the follo*1.ring provision:

ENO may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affeet the ability to achieve and maintain safe shutdown in the event of a fire.

I. DELETED Amelt. 20§ 2 27 01 J. DELETED Amelt. 20§ 2 27 01 K. DELETED Amelt. 49

§ 2§ 84 1

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

L. DELETED Arndt. 20§ 2 27 01 M. DELETED Arndt. 2 05 2 27 01 N. DELETED Arndt. 49 Deleted per

§ 2§ 84 Amendment

[###]

0. Evaluation, status and sehedule for eoR'lpletion of balanee of plant Arndt. 47 R'lodifieations as outlined in letter dated February 12, 1983, shall be 5 27 83 foRNarded to the ~JRG by January 1, 1984.

P. ENIP3 and ENO shall take no action to cause Entergy Global Arndt. 203 Investments, Inc. or Entergy International Ltd. LLC, or their parent 11121/00 companies to void, cancel, or modify the $70 million contingency commitment to provide funding for the facility as represented in the application for approval of the transfer of the license from PASNY to ENIP3 and ENO, without the prior written consent of the Director, Office of Nuclear Reactor Regulation.

Q. DELETED R. DELETED S. DELETED T. DELETED U. DELETED V. DELETED

W. For purposes of ensuring public health and safety, ENIP3, upon the transfer of this license to it, and upon transfer of decommissioning funds from PASNY to ENO, shall provide decommissioning funding assurance for the facility by the prepayment or equivalent method, to be held in a decommissioning trust fund for the facility, of no less than the amount required under NRC regulations at 10 CFR 50.75. Any amount held in any decommissioning trust maintained by ENO for the facility after the transfer of the facility license to ENIP3 may be credited towards the amount required under this paragraph.

X. ENIP3 shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the application for the transfer of this license to ENIP3 and ENO, as modified by the request to transfer decommissioning funds from PASNY, and the requirements of the order approving the transfer and order Deleted per approving the transfer of decommissioning funds from PASNY to ENO, and Amendment consistent with the safety evaluations supporting such orders.

[###]

Arndt. 205 2/27,101

+.- 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 GTS,"

and Table LA, "Removed Details and Less Restrictive Administrative Changes to the GTS" attached to the NRG 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.

&. The follo*Ning 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 surveillanoe interval that begins on the date of implementation of this amendment.

For SRs that mcisted prior to this amendment 1Nhose intervals of performance are being reduced, the first reduoed surveillance interval begins upon completion of the first surveillance performed after the date of implementation of this amendment.

For SRs that e3Eisted prior to this amendment that hmm 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 m<isted prior to this amendment i.vhose intervals of performance are being extended, the first Deleted per extended surveillance interval begins upon completion of the Amendment last SUF\'eillanee performed prior to the date of implementation of this amendment.

[###]

With the reactor critical, entergy shall maintain the reactor coolant system cold leg at a temperature (Teo1a) greater than or equal to §26 °F. entergy shall maintain a record of the cumulative time that the plant is operated 1Nith the reactor critical 1,1vhile Teo1a is below §2§ °F. Upon determination by entergy that the cumulative time of plant operation with the reactor critical while Tea1a is below §2§ °F has exceeded one (1) year, entergy must:

(a) within one (1) month, inform the NRG, in 1.Nriting, and (b) within si3< (e) months submit the results of an analysis of the impact of the operation with Tca1a below 525 °F on the pressuri:zed thermal shook reference temperature (RTpts).

AC. 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 o_f readily-available pre-staged equipment

6. Training on integrated fire response strategy

7. Spent fuel pool mitigation measures Deleted per Amendment (c) Actions to minimize release to include consideration of:

[###] 1. Water spray scrubbing

2. Dose to onsite responders AD.

Upon implementation of Amendment No. 239 adopting TSTF 4 48, Revision 3 (as supplemented), the determination of control room envelope (GRe) unfiltered air inleakage as required by Technical Specification (TS) Surveillance Requirement (SR) 3. 7.11.4, in aeeordanee 1.vith TS e.e.1e.e.(i), the assessment of GRe habitability as required by TS e.e.1e.e.(ii), and the measurement of GR.e pressure as required by TS 6.6.1e.d, shall be considered met. Follo11.:ing implementation:

~ The first performanoe of SR 3.7.11.4, in aooordanoe with TS 6.6.16.o.(i),

shall be v,ithin the speoified Frequenoy of 6 years, plus the 18 month allowanee of SR 3.0.2, as measured from February 1, 2006, the date of the ffiost resent sueeessful traoer gas test, as stated in the dune 28, 2005, letter response to Generie Letter 2003 01.

~ The first performanee of the periodie assessment of GRE: habitability, TS 5.6.16.o.(ii), shall be within the nmct 9 ffionths sinoe the time period sinoe the most resent suooessful traeer gas test is greater than 3 years.

(a), The first performanoe of the periodio ffieasurement of GRE: pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from dune 18, 2007, the date of the most resent suooessful pressure measurement test.

AE. 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 ENIP3.

AF. 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 AF(1) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).

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.

o. The lioensee shall notify the NRG in writing 1Nithin 30 days after having aeeoffiplished item (2)a above and inelude the status of those aotivities that have been or remain to be eoffipleted in item (2)b above.

3. This renewed license is effective as of the date of issuance, antisllall-exeiFEHM-miafliwi-t April 30, 2025.

until the Commission notifies the licensee in writing that the license is terminated FOR THE NUCLEAR REGULATORY COMMISSION PERMANENTLY DEFUELED Ho K. Nieh, Director Office of Nuclear Reactor Regulation Attachments:

Appendix A- echnical Specifications Appendix 8- Environmental Technical Specification Requirements Appendix C - Inter-Unit Fuel Transfer Technical Specifications Date of Issuance: September 17, 2018

~ T o Be Determined

APPENDIX A PERMANENTLY DEFUELED

.FACILITY OPBR:ATINC LICENSE DPR-64, TECHNICAL SPECIFICATIONS AND BASES FOR THE

,INDIAN 20INT 3 NUCLEAR GENERATING STATION UNIT NO. 3, WESTCHESTER COUNTY. NEW YORK ENTERGY NUCLEAR INDIAN POINT *3. LLC (ENIP3).

AND ENTERGY NUCLEAR OPERATIONS I INC . (ENO)_

DOCKET NO. 50-286.

Date of Issuance:

April .15, 1976 Amendment No. ~

Facility Operating License No. DPR-64 Appendix A - echnical Specifications TABLE OF CONTENTS PERMANENTLY

~~~~~~~~~~~~~ -===iDEFUELE D 1.0 USE AND APPLICATION 1.1 Definitions 1.2 Logical Connectors 1.3 Completion Times 1.4

~DELETED I 2.0 b-+/- Safety Limits

&.-2, Safety Limit Violations 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY SURVEILLANCE REQUIREMENT (SR) APPLICABILITY J-. REACTIVITY CONTROL SYSHMS J.-....l--.-.l. Shutdown Margin (SDM)

~ Core Reactivity

~ Moderator Temperature Coefficient (MTG)

J-.h4 Rod Group Alignment Limits .-l--.-& Shutdown Bank Insertion Limits

~ Control Bank Insertion Limits

-hh Rod Position Indication J-....l--.S PHYSICS TtSTS txceptions MODt 2

-POWER DISTRIBUTION LIMITS Meat flux Mot Channel Factor ( FQ (Z))

Nuclear tnthalpy Rise Mot Channel Factor (~J AXIAL FLUX DIFFtRtNCt (AFD)

QUADRANT POWtR TI LT RATIO (QPTR)

~ INSTRUMENTATION

~ Reactor Protection System (RPS) Instrumentation

~ tngineered Safety Feature Actuation System (tSFAS) Instrumentation

~ Post Accident Monitoring (PAM) Instrumentation J-..-J..-4 Remote Shutdown 1

~ Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation

~ Containment Purge System and Pressure Relief Line Isolation Instrumentation J...-J-..-7. Control Room '.'enti 1ati on (CRVS) Actuation Instrumentation J......J-....g Fuel Storage Building tmergency Ventilation System (FSBtVS)

Actuation Instrumentation (continued)

INDIAN POINT 3 Amendment~

Facility Operating License No. DPR-64 Appendix A - chnical Specifications PERMANENTLY TABLE OF CONTENTS DEFUELED REACTOR COOLANT SYSTEM (RCS)

RCS PressHre, TemperatHre, and Flow DepartHre from NHcleate Boiling (DNB) Limits J...4-...i RCS MinimHm TemperatHre for Criticality

~ RCS PressHre and TemperatHre (P/T) Limits J...4-.-4 RCS Loops MODES 1 and 2

~ RCS Loops MODE 3 J...4-.-e. RCS Loops MODE q J....4-..-7. RCS Loops MODE a, Loops Filled J-.-4-..8 RCS Loops MODE a, Loops Not Filled

-h4-. PressHrizer 3 .4 .10 PressHrizer Safety Valves 3 .4 .11 PressHrizer Power Operated Relief Valves (PORVs) 3.4.12 Low TemperatHre OverpressHre Protection (LTOP) 3 .4 .13 RCS Operational LEAKAGE

3. 4, lq RCS PressHre Isolation Valve (PIV) Leakage 3.4.ls RCS Leakage Detection InstrHmentation 3,q,Hi RCS Specific Activity 3.4.17 Steam Generator (SG) THbe Integrity EMERGENCY CORE COOLING SYSTEMS (ECCS)

AccHmHlators EGGS Operating EGGS ShHtdown RefHeling Water Storage Tank (RWST)

~ CONT.o.INMENT SYSTEMS .-e-.-l- Containment

~ Containment Air Locks

~ Containment Isolation Valves J.....e...4 Containment PressHre J......e-.s. Containment Air TemperatHre J......a-.:-e. Containment Spray System and Containment Fan Cooler System J-.-&-.-7. RecircHlation pH Control System

~ Not Used

~ Isolation Valve Seal Water (IVSW) System 3 .e.10 Weld Channel and Penetration PressHrization System (WC & PPS)

(continued)

INDIAN POINT 3 ii Amendment~

Facility Operating License No. DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS

1SPEN T FUEL PIT PERMANENTLY i--~~~~~~-

REQUIREMENTS DEFUELED Main Steam Safety Valves (MSSVs)

Main Steam Isolation Valves (MSIVs) and Main Steam Check Valves (MSCVs)

Main Boiler FeedpYmp Discharge Valves (MBFPDVs), Main Feedwater Regylation Valves (MFRVs), Main Feedwater Inlet Isolation Valves (MFIIVs) and Main Feedwater (MF) Low Flow Bypass Valves J....+-.-4 Atmospheric DYmp Valves (ADVs)

~ AYxi 1i ary Feed1*1ater (AFW) System J-.-7-....& Condensate Storage Tank (CST)

~ City Water (C 1.~)

J...-+...S Component Cooling Water (CCW) System J....-7-.-9. Service Water (SW) System 3 .7.10 Ultimate Heat Sink (UHS) 3 .7 .11 Control Room Ventilation System (CRVS) 3.7.12 Control Room Air Condi ti oni ng System (CRAGS) 3.7.13 FYel Storage BYilding Emergency Ventilation System (FSBEVS) 3.7.14 Spent Fuel Pit Water Level 3.7.15 Spent Fuel Pit Boron Concentration 3.7.16 Spent Fuel Assembly Storage 3.7.17 Secondary Specific Activity 3-rS fUGTRIC.O.b POWfR SYSHMS ..&-.-l- AC SoYrces Operating J....S..-2. AC SoYrces ShYtdown

~ Diesel FYel Oil and Starting Air

~ DC SoYrces Operating

~ DC SoYrces ShYtdown J......8-.-& Battery Cell Parameters J-..-8-..-7. Inverters Operating

~ Inverters ShYtdown J-..&..-9. DistribYtion Systems Operating 3.8.10 DistribYtion Systems ShYtdown i.....9- RHUHING OPfR,O.TIONS J...-9-..-1. Boron Concentrat1on

~ NYclear InstrYmentation J.....9...-J. Containment Penetrations J-.-9-.-4 ResidYal Heat Removal (RHR) and Coolant CircYlation High Water Level J-.-9...-s. ResidYal Heat Removal (RHR) and Coolant CircYlation Low Water Level J....-9-.-e. Refyeling Cavity W~ter Level (continued)

INDIAN POINT 3 iii Amendment 2-JJ

Facility Operating License No. DPR-64 Appendix A - Technical Specifications TABLE OF CONTENTS PERMANENTLy

~~~~~~~~~~~~~~~~----1DEFUELED 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 5.3 Facility Staff Qualifications 5.4 Procedures 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) ~

5.5.2 Primary Coolant Sourees Outside Containment 5.5.3 NOT USED 5.5.4 Radioactive Effluent Controls Program DELETED 5.5.5 .... ~

5.5.6 Reaetor Coolant Pump Fly,Nheel lnspeetion Program I I 5.5.7 lnserviee Testing Program( DELETED 5.5.8 Steam Generator (SG) ProgramnDELETED I !DELETED 5.5.9 Seeondary Water Chemistry Program<

5.5.10 Ventilation Filter Testing Program ( V F T P ) ~

5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5.12 Diesel Fuel Oil Testing Program( !DELETED I 5.5.13 Technical Specification (TS) Bases Control Program

~ Safety Funetion Determination Program (SFDP)

~ Containment Leakage Rate Testing Program

~ 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

~ NOT USED

&:&-e GORE OPERATING LIMITS REPORT (GOU~)

~ ~JOT USED

&:&.-7 Post Aeeident Monitoring Instrumentation (PAM) Report

&:&:-8 Steam Generator Tube lnspeetion Report 5.7 High Radiation Area INDIAN POINT 3 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.

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 .....ith each possible interlock logic state and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall inolude a oontinuity oheck of output devices.

AXIAL FLUX DIFFERENCE /\FD shall be the difference in normalized flux fAp-9t signals beti.veen the top and bottom halves of a t\vo section excore neutron detector.

CERTIFIED FUEL HANDLER A CERTIFIED FUEL HANDLER is an individual who (CFH) 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 so that it responds v1ithin the required range and acouraoy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, display, and trip funotions. Calibration of instrument channels *.vith resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devises in the channel.

Whenever a sensing element is replaced, the next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements ,...,ith the (continued)

INDIAN POINT 3 1.1-1 Amendment No. ~

r

l

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_J; J; Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions MODE (continued) vessel head closure bolt tensioning specified in Table 1.1 1 with fuel in the reactor vessel.

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 'Nhen all necessary attendant instrumentation, controls, normal or emergency electrical po\ver, sealing and seal v,ater, lubrisation, and other auxiliary equipment that are required for the system, subsystem, train, component, or devise to perform its specified safety funstion(s) are also sapable of performing their related support function(s).

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

a-: Described in FSAR Chapter 1a, Initial Tests and Operations;

&.- l\uthorized under the provisions of 10 CFR 60.69; OF e-: OtheANise approved by the Nuclear Regulatory Commission.

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

RATED THERMAL POVVER RTP shall be a total reastor sore heat transfer

~ rate to the reactor sealant of 321e MWt.

(continued)

INDIAN POINT 3 1.1-5 Amendment No. ~

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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 n ec n,cal Specifications (TS) to discriminate between, and t 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 T=iffle.;--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.

(continued)

INDIAN POINT 3 1.2-1 Amendment 2GS

Logical Connectors 1.2 1.2 Logical Connectors (continued)

EXAMPLES The following examples i l l ~ the use of logical connectors.

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

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.

(continued)

INDIAN POINT 3 1.2-2 Amendment 29§.

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 Ope tion (LCOs) specify requirements for ensuring saf 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 completing a Required Action. It facility I discovery of a situation (e.g., 7°-Ae-aeiF-iH:rtt~~+emeA-c--&1~

~ v a r i a b l e not within limits) that requires entering a Condition unless otherwise specified, providing the is in a MODE or specified condition stated in the Applicability of the LCO.

URless otherwise specified, the CompletioR Time begiRs wheR a seRior liceRsed operator OR the operatiRg shift crew with respoRsibility for plaRt operatioRs makes the determiRatioR that aR LCO is Rot met aRd aR ACTIONS CoRditioR is eRtered.

The "otherwise specified" excepti ORS are *varied, such as a Required ActioR Note or SurveillaRce RequiremeRt Note that provides aR alterRative time to perform specific tasks, such as testiRg, without startiRg the CompletioR Time. While utiliziRg the Note, should a CoRditioR be applicable for aRy reasoR Rot addressed by the Note, the CompletioR Time begiRs.

Should the time allowaRce iR the Note be exceeded, the CompletioR Time begiRs at that poiRt. The exceptioRs may also be iRcorporated iRto the CompletioR Time. For example, LCO 3.8.1, "AC Sources OperatiRg," Required ActioR B.2, requires declariRg required features supported by aR iRoperable diesel geRerator, iRoperable wheR the reduRdaRt required feature is iRoperable. The CompletioR Time states, "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from discovery of CoRditioR B coRcurrciRt with iRoperability of reduRdaRt required feature." IR this case the CompletioR Time does Rot begiR uRtil the coRditioRs iR the CompletioR Time are satisfied.

(continued)

INDIAN POINT 3 1. 3-1 Amendment 2 Completion Times

1. 3 1.3 Completion Times DESCRIPTION Required Actions must be completed prior to the expiration (continued) of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the

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 discovery of the situation that required entry into the Condition, unless otherwise specified.

Once a Condition has been entered, subsequent trains, subsystems, components, or variables expressed in the Condition, discovered to be inoperable or not within limits, wil 1 DQ1 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, unless otherwise specified.

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 .fiE.tl inoperability; and

a. Must remain i nope rabl e or not within 1 i mi ts 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 ;

&F-

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

(continued)

INDIAN POINT 3 1. 3-2 Amendment~

.l l

Completion Times Required Actions 1. 3 Initiate action to restore

1. 3 Com l eti on Ti mes continued spent fuel pit boron

=-=--=------===~..:....,_c=..,_-'---'-'-'==-----"--'==..,..'-=--'--'-'-=-='---+-------.-..----+----1conce ntratio n to with in limit.

EXAMPLE£ The following example& "llustrate wi th di ff e rent types of t0-1'!-ffil-8-e-R-5-iaftt!l---i:=fl-a:Ha-'l-ffl~::&Flf&=l-'E-'HEtfl Suspend movement of fuel assemblies in the 1. 3-1 spent fuel pit.

CONDITIO COMPLETION TIME

~ A Spent fuel pit boron Required ActioA aAd ~hoursjlmmediately I concentration not associated within limit. CompletioA ~~mmediately I Time Rot met.

is immediately suspend movement of fuel assemblies in the spent fuel pit and initiate action to restore spent fuel pit boron concentration to within limit.

(continued)

INDIAN POINT 3 1.3-4 Amendment 2-66

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Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMPLE 1.3 7 (CORtiRued)

If after CoRditioR A is eRtered, Required ActioR A.1 is Rot met withiR either the iRitial 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or aRy subsequeRt 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months iRterval from the previous performaRce (plus the exteRsioR allowed by SR 3.0.2), CoRditioR Bis eRtered. The CompletioR Time clock for CoRditioR A does Rot stop after CoRditioR Bis eRtered, but coRtiRues from the time CoRditioR A was iRitially eRtered. If Required ActioR A.1 is met after CoRditioR Bis eRtered, CoRditioR Bis exited aRd operatioR may coRtiRue iR accordaRce with CoRditioR A, provided the CompletioR Time for Required ActioR A.2 has Rot expired.

IMMEDIATE COMPLETION TIME When "Immediately" is used as a Completion Time, the Required Action should be pursued without delay and in a controlled manner.

INDIAN POINT 3 1. 3-14 Amendment~

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, Surveillance Requirement (SR) Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR as well as certai A Notes iA the SurveillaAce columA that modify performaAce requiremeAts.

Si tuati OAS ~.*here a Survei 11 aAce could be required ( i . e. , its FrequeAcy could mcpi re) , but where it is Rot possible or Rot desired that it be performed uAtil sometime a~er the associated LGO is wi thi A its Appli cabi l i ty, represeAt poteAti al SR 3. O.4 coAfl i cts.

To avoid these coAflicts, the SR (i.e., the SurveillaAce or the FrequeAcy) is stated such that it is oAl y "required" ~.*heA it caA be aAd should be performed. With aA SR satisfied, SR 3.0.4 imposes AO restrictioA.

EXAMPLE£ illustrates the type of Frequency statement that appears in the Technical Specifications (TS).

(continued)

INDIAN POINT 3 1.4-1 Amendment 2-%

Frequency 1.4 1.4 Frequency EXAMPLE& EXAMPLE 1.4-1 (continued)

SURVEILLANCE REQUIREMENTS Verify level is within limits. SURVEILLANCE FREQUENCY Perform CHANN[L CH[CK. 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 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 times, even when the SR is not required to be met per SR 3.0.1 (such as when th- facility equipment is inoperable, a variable is outside specified s, or

~ ~llfl-i-.t. is outside the Applicability of the CO the interval

~ s p e c i f i e d by SR 3.0.2 is exceeded while the

is in a MOD[ or other specified condition in the Applicability of the LCD, and the performance of the Surveillance is not otherwise modified (refer to

[xample 1.4 3), then SR 3.0.3 becomes applicable. -~

If the interval as specified by SR 3.0.2 is exceeded while t h e ~ ~

is not in a MOD[ or other specified condition in the Applicability of the LCD 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 MOD[ or other specified condition or the LCD is considered not met (in accordance with SR 3.0.1) and LCO 3.0.4 becomes applicable.

(continued)

INDIAN POINT 3 1.4 - 2 Amendment~

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2.1 L 1 2.1.1 Reactor Core gLs In ~mm;g 1 and 2 1 the combination of THE:RH-Z'.L POWER, Reactor Vessel inlet temperature, and pressuri2er pressure shall not exceed the limits specified in the COLR; and the follm.ring gLs shall not be exceeded:

2.1.1.1 The departure from. nucleate boiling ratio (DNBR) shall be maintained> 1.17 for the WRB 1 DNB correlations.

2.1.1.2 The peak fuel centerline temperature shall be maintained< 5080°F, decreasing by .§-82.F per 10,000 HWD/HTU of burnup.

2.1.2 Reg Pressure gL In ~40Df:g 1, 2, 3, 4 1 5, and in MODE: 6 when the reactor vessel head is on, the Reg pressure shall be maintained< 2735 psig.

2.2 gL Violations 2.2.1 If gL 2.1.1 is violated, restore compliance and be in HODE: 3 r.:ithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If gL 2.1.2 is violated:

2. 2. 2 .1 In ~mm:; 1 or 2 1 restore compliance and be in NODE 3 r,.rithin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE: 3 1 4, 5, or 'Ei, restore compliance within 5 minutes, INDIAN POINT 3 2.0-1 Amendment 2 LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the MODES or otRer 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.S 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.

'1!Ren an LCO is not met and tRe associated ACTIONS are not met, an associated ACTION is not provided, or if directed by tRe associated ACTIONS, tRe linit sRall be placed in a MODE or otRer specified condition in wRicR tRe LCO is not applicable. Action sRall be initiated witRin 1 ROlir to place the linit, as applicable, in:

a-,- MODE 3 witRin 7 ROlirs;

b. MODE 4 witRin 13 ROlirs; and

. MODE s witRin 37 ROlirs.

Exceptions to tRis Specification are stated in tRe individlial Speci fi cati ens.

WRere corrective measblres are completed that permit operation in accordance witR tRe LCO or ACTIONS, completion of tRe actions reqliired by LCO 3.0.3 is not reqliired.

LCO 3.0.3 is only applicable in MODES 1, 2, 3, and 4.

(continued)

INDIAN POINT 3 3.0 - 1 Amendment 2-9

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SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MODES er ether 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 e-r--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 Speci fi cati ens.

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. The delay period is only {

applicable when there is a reasonable expectation the surveillance will be met when performed. 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.

(continued)

INDIAN POINT 3 3.0 - 4 Amendment~

SR Applicability 3.0 3.0 SR APPLICABILITY (continued)

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. WheR aR LCD is Rot met due to SurveillaRces Rot haviRg beeR met, eRtry iAto a MODE or other specified coRditioR iA the Applicability shall oRly be made iA accordaAce with LCD 3.0.4.

This provisioR shall Rot preveRt eAtry iAto MODES or other specified coRditioRs iA the Applicability that are required to comply with ACTIONS or that are part of a shutdowR of the tm4-t-,-

INDIAN POINT 3 3.0 - 5 Amendment~

Spent Fuel Pit Water Level SPENT FUEL PIT 3.7.14 REQUIREMENTS 3.7.14 Spent Fuel Pit Water Level LCO 3.7.14 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 f ue 1 pit water A.1 NOTE level not within limit. LCO 3.0.3 is Aot applicable.

Suspend movement of Immediately irradiated f ue 1 assemblies in the spent fuel pit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pit water level is~ 23 ft 7 days above the top of the irradiated fuel assemblies seated in the storage racks.

INDIAN POINT 3 3.7.14-1 Amendment~

SPENT FUEL PIT Spent Fuel Pit Boron Concentration REQUIREMENTS 3.7.15 3.7.15 Spent Fuel Pit Boron Concentration LCD 3.7.15 The Spent Fuel Pit boron concentration shall be~ 1000 ppm.

NOTE----------------------- --

During inter-unit transfer of fuel the spent fuel pit boron concentration must also meet Appendix C LCD 3.1.1, "Boron Concentration".

APPLICABILITY: When fuel assemblies are stored in the spent fuel pit and a spent fuel pit verification has not been performed since the last.

movement of fuel assemblies in the spent fuel pit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pit boron Non:

concentration not within LCO 3.0.3 is Rot a~~licable.

limit.

A.1 Suspend movement of Immediately fuel assemblies in the spent fuel pit.

A.2.1 Initiate action to Immediately restore spent fuel pit boron concentration to with i n l i mi t.

A.2.2 Initiate action to Immediately perform a spent fuel pit verification.

INDIAN POINT 3 3.7.15-1 Amendment~

Spent Fuel Assembly Storage SPENT FUEL PIT 3.7.16 REQUIREMENTS 3.7 3.7.16 Spent Fuel Assembly Storage LCO 3.7 .16 Fuel assemblies stored in the spent fuel pit shall be classified in accordance with Figure 3.7.16-1 based on initial enrichment and burnup; and, Fuel assembly storage location within the spent fuel pit shall be restricted based on the Figure 3.7.16-1 classification as follows:

a. Fuel assemblies classified as Type 2 may be stored in any location in either Region 1 or Region 2:

b. Fuel assemblies classified as Type lA, 18 or lC shall be stored in Region 1:

c. Fuel assembly storage location within Region 1 shall be restricted as follows:

1. Type 1A assemblies may be stored anywhere in Region 1:

2. Type 18 assemblies may be stored anywhere in Region 1, except a Type 18 assembly shall not be stored face-adjacent to a Type lC assembly;

3. Type lC assemblies shall not be stored in Row 64 or in Column ZZ: and

4. Type lC assemblies shall be stored in Region 1 locations where all face-adjacent locations are as follows:

a) occupied by Type 2 or Type 1A assemblies, or b) occupied by non-fuel components. or c) empty.

APPLICABILITY: Whenever any fuel assembly is stored in the spent fuel pit.

INDIAN POINT 3 3.7.16-1 Amendment~

Spent Fuel Assembly Storage 3.7.16 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 NOTE not met. LGO 3.0.3 is AOt applicable.

Initiate action to move Immediately fuel to restore compliance with LCO 3.7 .16.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means the initial Prior to storing enrichment and burnup of each fuel assembly and the fuel assembly that the storage location meets LCO 3.7.16 in the spent fuel requirements. pit INDIAN POINT 3 3.7.16-2 Amendment~

Design Features 4.0

4. 0 DESIGN FEATURES (continued) 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. ~ff :s O. 95 if assemb1i es are inserted in accordance with Technical Specification 3.7.16, Spent Fuel Assembly Storage-;-~

c. A nominal 9.075 inch center to center distance between fuel assemblies placed in the high density fuel storage racks (Region II) ;

d. A nominal 10.76 inch center to center distance between fuel assemblies placed in low density fuel storage racks (Region I)~

4. 3.1. 2 The nm, fuel storage racks are designed and shal 1 be maintained

~

a-:- Fuel assembl i cs having a mm<i mum U 235 enri chmcnt of

5. 0 *.,ci ght percent; a-:- *e# :S 0.95 under all possible moderation conditions (Credit may be taken for burnable integral neutron absorbers ) :

e-;- A nominal 20.5 inch center to center distance bet'n'een fuel assemblies placed in the storage racl<s.

4.3.2 Drainage The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pool below a nominal elevation of 88 ft.

(continued)

INDIAN POINT 3 4.0-2 Amendment~

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

~ and Quality Assurance Plan, as appropriate;

b. The plant manager shall be responsible for overall safe maintenance of the facility and shall have control over those onsite activities necessary for safe storage and maintenance of nuclear fuel;

c. The corporate officer with direct responsibility for IP3 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.

(continued)

INDIAN POINT 3 5.0-2 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 FSAR1"- ~

.-~

b. Deleted~

c. Quality assurance for effluent and environmental monitoring;

!Deleted k 1

d~Fire Protection Program implementation; and

e. All programs specified in Specification 5.5.

INDIAN POINT 3 5.0-6 Amendment No. ~

i Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) (continued) 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.

5.5.2 , ; Geela;t GeYffies Oetsi~e GeRtaiRRlORt Deleted (continued)

INDIAN POINT 3 5.0-8 Amendment No. ~

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

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 uni 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;

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 limited to the following:

a. For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and

b. For iodine-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to dose rate of 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 uni to areas beyond the site boundary, conforming to 10 CFR 50, Appe I;

/facility (continued)

INDIAN POINT 3 5.0 - 10 Amendment~

Programs-and Manuals 5.5 5.5 Programs and Manuals 5.5.4 Radioactive Effluent Controls Program (continued)

i. Limitations on the annual and quarterly doses to a member of the

.. public from'iodine-131, tritium, and all radionuclides in particulate

~ ~ ~ ; ~ w i t h half lives> 8 days in gaseous effluents released from each

~ ~ 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 Effluents Controls Program surveillance frequency.

!Deleted 5.5.5 1,mio*e*t Cyelie,. T*a*sieRt Limit This program provides coRtrols to track the FSAR, SectioR q.1.§, cyclic aRd traRsieRt occurreRces to eRsure that compoReRts are maiRtaiRed withiR the

~ID_e_le_t_e_d_ ___.!-i des i gR l i mit s .

5.5.6 ~eactor CoolaRt Pump Flywheel IRspectioR Program This program shall provide for the iRspectioR of each reactor coolaRt pump flywheel. The program shall iRclude iRspectioR frequeRcies aRd acceptaRce criteria. The iRspectioR frequeRcy will ensure that each reactor coolant pump flywheel is surface and volumetrically inspected at 20 year intervals.

(cont4nued)

INDIAN POINT 3 5.0 - 11 Amendment~

Programs and Manuals 5.5 5.5 Programs and Manuals

~

5.5.7 lnservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components including applicable supports. The program shal I include the fol lowing:

a.- Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as fo I Iows:

ASME OM Code and applicable Addenda terminology for Required Frequencies inservice testing for performing inservice activities testing activities WeeM-y At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semi annua I ly 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

&,- The provisions of SR 3 . 0. 2 are app I icab Ie to the above required Frequencies and to other normal and accelerated frequencies specified as 2 years or less in the lnservice Testing Program for performing inservice testing activities;

&,- The provisions of SR 3.0.3 are applicable to inservice testing activities; and a.- Nothing in the ASME OM Code shal I be construed to supersede the requirements of any TS.

(continued)

INDIAN POINT 3 5.0 -12 Amendment~

Programs and Manuals 5.5 5.5 Programs and Manuals A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall inslude the f:ollowing:

a:- Provisions f:or sondition monitoring assessments. Condition monitoring assessment means an evaluation of the "as f:ound" sondition of the tubing with respest to the performanse sriteria for strustural integrity and assident indused leakage. The "as f:ound" sondition refers to the sondition of the tubing during an SG inspestion outage, as determined from the inservioe inspestion results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be sondusted during eash outage during 1.vhish the SG tubes are inspested or plugged, to sonfirm that the performanse sriteria are being met.

&. Performanse sriteria for SG tube integrity. SG tube integrity shall be maintained by meeting the performanse sriteria f:or tube strustural integrity, assident indused leakage, and operational LEAKAGE.

4.,. Strustural integrity performanse sriterion: All in servise steam generator tubes shall retain strustural integrity over the full range of normal operating sonditions (insluding startup, operation in the pmver range, hot standby, and sool down), all antisipated transients insluded in the design spesif.isation and design basis assidents. This insludes retaining a safetyfastor of 3.0 against burst under normal steady state full power operation primary to sesondary pressure differential and a safety fastor of 1.4 against burst applied to the design basis assident primary to sesondary pressure differentials. Apart from the above requirements, additional loading sonditions assoeiated v:ith the design basis assidents, or eombination of aeeidents in aseordanse with the design and lisensing basis, shall also be evaluated to determine if the assosiated loads sontribute signifieantly to burst or sollapse. In the assessment of tube integrity, those loads that do signifisantly affest burst or sollapse shall be determined and assessed in sombination with the loads due to pressure with a safety faster of 1.2 on the eombined primary loads and 1.0 on axial sesondary loads.

2. Assident indused leakage performanse eriterion: The primary to sesondary assident indused leakage rate f:or any design basis assident, other than a SG tube rupture, shall not exseed the (continued)

INDIAN POINT 3 5.0 - 13 Amendment 2-a-7

Programs and M arniaJ s 5......5.

5 5 Prag rams and JNDJAN POINT 3 5 -O - 14

Programs and Manuals 5.5 5.5 Programs and Manuals

5. 5. 9 SecoAdary Water Gt:1emi stry Program

~ T h i s program provides coAtrols for moAitoriAg secoAdary water chemistry to iAhibit SG tube degradatioA. The program shall iAclude:

a-:- IdeAtificatioA of a sampliAg schedule for the critical variables aAd coAtrol poiAts for these variables:

fr=- IdeAtificatioA of the procedures used to measure the values of the critical variables:

e-;- IdeAtificatioA of process sampliAg poiAts, ~ihich shall iAclude moAitoriAg the COAdeAser hot wells for evideAce of COAdeAser iA leakage;

&;- Procedures for the recordiAg aAd maAagemeAt of data; e-:- Procedures defi Ai Ag corrective acti OAS for all off coAtrol poi At chemistry COAditiOAS; aAd

.f.:- A procedure ideAtifyiAg the authority respoAsible for the iAterpretatioA of the data aAd the sequeAce aAd timiAg of admiAistrative eveAts, \Jhich is required to iAitiate corrective actioA.

(continued)

INDIAN POINT 3 5.0-20 Amendment~

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Ventilation Filter Testing Program (VFTP)

~ T h i s program pro:sri des controls for imp] emeutati on of reqJJi red testing of the :ventilation filter fnoctioo for the Control Boom :vent i 1 at i an System and Canta i oment Fan Cool er Ilni ts Applicable tests described in Specifications 5 5 JO a, 5 5 JO b, 5 5 JO c aud 5 5 JO d shaJJ be performed*

J) After 720 honrs of charcoal adsorber 11se si nee the J ast test* and ;

2) Every 24 months for tbe Control Boom Veuti J ati on System, and Containment Eau Cooler Units* aud

3) After eacb complete or partial replacement of the HEPA filter train or charcoal adsorber filter* aud ;

4) After any stnictnraJ mai uteuauce on the system ho11si ug tbat co11J d aJ ter system i utegd ty* aud,

5) After significant painting, fire ; or chemical release in any 3Tentilation zone cororoJJnicating ,with the system while it is in operation SB 3 0 2 is appJ i cab] e to the Veuti J ati on Fi J ter Testi ug Program (continued)

INDIAN POINT 3 5.0 - 21 Amendment 22..4.

Jii ~Hi Hli

. . . ]i !. !,

1~ Jh i

I ti ti Ji !, ;,

A ); lh j

. * . j1{1 ~' ;,

0 J;

Programs and Manuals 5.5 5.5 Programs and Manuals IDelete~. 5.12 J1.:.i ::eesle~uel Oil Testi Ag Program

_ ~ ~-- a,---, fuel oil testiAg program to implemeAt requires testiAg of both Ae~J fuel oil aAs stores fuel oil shall be establishes for the 06 fuel oil oAsite storage taAks aAs the 06 reserve fuel oil storage taAl<s. The program shall iAcluse sampliAg aAs testiAg requiremeAts. aAs acceptaAce criteria, all iA accorsaAce with applicable AS"FM StaAsarss. The purpose of the program is to establish the followiAg:

-a-:- Verifi cati oA of the acceptability of A6~ fuel oil for use prior to assitioA to the 06 fuel oil oAsite storage taAl<s by setermiAiAg that the fuel oil has:

h Relative seAsity \iithiA the limits of 0.83 to 0.89, 2-:- kiAematic viscosity withiA the limits of 1.8 to 5.8, aAs 3-:- a clear aAs bright appearaAce with proper color eh VerificatioA of the acceptability of the fuel oil iA the oAsite storage taAks aAs the reserve storage tanl<s every 92 says by veri fyi ng that the properties of the fuel oil iA the taAks, other thaR those assresses iR item a .. are withiA limits for AS"FM20 fuel oil. The sampliRg techAique for the reserve storage taRks may seviate from AS"FM 0270 1975 iR that oRly a bottom sample is requires.

6f' 62-:- Verification of the acceptability of each Rew fuel assitioR masc subsequeRt to the last verificatioA mase iR accorsance *,,ith item bl. by verifyi Ag ~Ji thi A 31 says foll owi Ag the assi ti oA that the properties of the Re,, fuel oil, other thaA those properties assresses iR item a. are

~,ithiA limits for AS"FM 20 fuel oil.

e-;- Veri fi cati OR e*,ery 92 says that total particulate coRceRtrati oA of the fuel oil iA the oRsite aAs reserve storage taAks is less thaR or equal to 10 mg/1 wheR testes iR accorsaAce *.~ith AS"FM O 2276, Methos A 2 or A 3-;- The sampl i Ag techAi que for the reserve storage tanks may se*;*i ate from AS"FM 0270 1975 in that oAly a bottom sample is requires.

(continued)

INDIAN POINT 3 5.0-27 Amendment~

Programs and Manuals 5.5 5.5 Programs and Manuals Diesel Fuel Oil TestiAg Program (coAtiAued)

TAC provisioAs of SR 3.0.2 aAd SR 3.0.3 are applicable to tAe Diesel Fuel Oil TestiAg Program tcstiAg frequeAcies.

5.5.13 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 involve either of the following:

1. a change in the TS incorporated in the license; or

2. a change to the HBEiaB'!B-~itt or Bases that requires NRC approva 1 pursuant to 10 CFR 50. 59. DSAR DSAR

c. The Bases Control Program shall contain visions to ensure that the Bases are maintained consistent with the

d. Proposed changes that do not meet the criteria of Specification 5.5.13.b 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).

Safety FuActioA DetermiAatioA Program (SrDP)

TAis program eAsures loss of safety fuActioA is detected aAd appropriate acti OAS tal<eA. UpoA cAtry i Ate LCO 3. O. 6, aA cval uati oA sAall be made to dctcrmiAe if loss of safety fuActioA exists. AdditioAally, otAer appropriate actioAs may be takeA as a result of tAe support system iAoperability aAd correspoAdiAg exceptioA to eAteriAg supported system CoAditioA aAd Required ActioAs. TAis program implemcAts tAc rcquircmcAts of LCO 3.0.6. TAC SFDP SAall COAtaiA tAe folla,,iAg:

(continued)

INDIAN POINT 3 5.0-28 Amendment GS

I

Pro§rams and ManHals

.-5

-s-.-s Pro§rams and ManHals a.a.le Control Room Envelope Habitability Program (continHed)

a. The definition of the CRE and the CRE bmmdary.

b. ReqHi rements for mai ntai ni n§ the CRE boHndary in its desi §R condition inclHdin§ confi§Hration control and preventive maintenance.

c. ReqHi rements for (i) determiRi R§ the Hnfi ltered air i nl eaka§e past the CRE boHndary into the CRE in accordance with the testi R§ methods and at the FreqHencies specified in Sections C.1 and C.2 of Re§Hlatory GHide 1.197, "Demon strati R§ Control Room Envelope IRte§rity at NHcl ear Power Reactors, 11 Revision Q., May- 2-0W, -affG fi-4+ assessin§ ~ habitability -a-t- th-e-FreqHenci es specified in Sec ti ens C.1 and C. 2 of Re§Hl atory GHi de 1.197, Revision 0.

d. MeasHrement, at desi§nated locations, of the CRE pressHre relative to all external areas adjacent to the CRE boHndary dHrin§ the pressHrization mode of operation by one train of the CRVS, operatin§ at the flow rate reqHired by the VFTP, at a FreqHency of 24 months on a STAGGERED TEST BASIS. The resHlts shall be trended and Hsed as part of the 24 month assessment of the CRE boHndary.

c. The qHantitati ve limits on Hnfi ltcrcd air i nl eaka§e iRto the CRE. These limits shall be stated in a manner to allow direct comparison to the Hnfiltered air inleaka§e measHred by the testin§ described in paragraph c.

The i,mfi ltered air i nl eaka§e limit for radi el ogi cal chall en§es is the inleaka§e flow rate assHmed in the licensin§ basis analysis of OBA conseqHences. Unfiltered air inleaka§e limits for hazardoHs chemicals mHst ensHre that exposHre of CRE occHpants to these hazards will be within the assHmptions in the licensin§ basis.

f. The provisions of SR 3.0.2 are applicable to the FreqHencies for assessin§ CRE habitability, determinin§ CRE Hnfiltered inleaka§e, and measHrin§ CRE pressHre and assessing the CRE boHndary as reqHired by paragraphs c and d, respectively.

INDIAN POINT 3 5.0 31a Amendment 239

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued)

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 Part 50.36a and 10 CFR 50, Appendix I, Section IV.8.1.

e-,e:4 Not Used

~ GORE OPERATING LIMITS REPORT (GOLR) a-:- Gore 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 GOLR for the follm'.'ing:

(continued)

INDIAN POINT 3 5.0-33 Amendment No. ~

. . . .j . ..

di

\.

\

APPENDIX B TO FACILITY OPER/\Tl~JG LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT 3 NUCLEAR POWER PLANT ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART I: NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION PLAN FACILITY LICENSE NO. DPR-64 DOCKET NUMBER 50-286 Renewed License No. DPR-64

handling and storage of spent fuel and maintenance 1.0 The Enviro mental Protection Plan (EPP) is to provide for protection of environmental values during eonstruetion 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 eonstruetion 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 No. DPR-64

3.0 Consistency Requirements 3.1 Plant Design and Operation facility facility ENO may make changes in operations or perform tests or experiments affecting the environment provided sue changes, tests or experiments do not involve an unreviewed environmental question and do not involve a change in the Environmental Protection Plan.* Changes in the f)laRt design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of this 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, ENO shall prepare and record an environmental evaluation of such activity. When the evaluation indicates that such activity involves an unreviewed environmental question, ENO 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 (FES) or final supplemental environmental impact statement (FSEIS), as modified by the staffs testimony to the Atomic Safety and Licensing Boards, supplements to the FES or FSEIS, environmental impact appraisals, or in any decision of the Atomic Safety and Licensing Board;

This provision does not relieve the ENO of the requirements of 10 CFR 50.59.

3-1 Renewed License No. DPR-64

or (2) a significant change in effluents or pm.var 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.

ENO 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 this EPP to meet the objectives specified in Section 1.0. ENO 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 NPDES Permits and State Certifications Violations of the NPDES Permit or the State certification (pursuant to Section 4.1 of the Clean Water Act) shall be reported to the NRC by submittal of copies of the reports required by the NPDES Permit or certification.

Changes and additions to the NPDES Permit or the State certification 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.

The NRC shall be notified of changes to the effective NPDES Permit proposed by ENIP3 and ENO 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. ENO shall provide the NRC a copy of 3-2 Renewed License No. DPR-64

l the application for renewal of the NPDES permit at the same time the application is submitted to the permitting agency.

3.3 Changes Required for Compliance with Other Environmental Regulations facility 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 No. DPR-64

i 4.0 Environmental Conditions 4.1 Unusual or Important Environmental Events the handling and storage of spent fuel and maintenance of the facility Any occurrence of an unusual or important ev t that indicates or could result in significant 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 No. DPR-64

5.0 Administrative Procedures 5.1 Review and Audit ENO shall provide a 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 is utilized to achieve the independent review and audit function and results of the audits 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 lant operation hall 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 1*aAt structures, systems and components determined to p tentially affect the continued protection of the environmentffi shall be retained for the life of the f*Sffi. 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 in the form of a license amendment incorporating the appropriate revision to the Environmental Protection Plan. This EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.

5-1 Renewed License No. DPR-64

i 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 report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this Environmental Protection Plan for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous non-radiological environmental monitoring reports, and an assessment of the observed on the environment. If harmful effects or evidence of trends the environment are observed, ENO shall provide a detailed osed course of action to alleviate the problem.

previous and the handling and storage of spent fuel and maintenance of the facility The Annual Environmental Protection Plan Report shall also include:

(a) A list of EPP noncompliances and the corrective actions taken to remedy them.

(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.

5-2 Renewed License No. DPR-64

5.4.2 Nonroutine Reports A written report shall be submitte o the NRC within 30 days of occurrence of a nonroutine event. The report shall (a) des ibe, analyze, and evaluate the event, including extent and magnitude of the impact and * * , (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) 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 No. DPR-64

APPENDIX B TO FACILITY OPERATING LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT 3 NUCLEAR POWER PLANT ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART II: RADIOLOGICAL ENVIRONMENTAL FACILITY LICENSE NO. DPR-64 DOCKET NUMBER 50-286 Renewed License No. DPR-64

APPENDIXC TO FACILITY OPERATING LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 3 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART I: SPENT FUEL TRANSFER CANISTER AND TRANSFER CASK SYSTEM FACILITY LICENSE NO. DPR-64 DOCKET NO. 50-286 Amendment No. ~

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 1000) (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% inch steel, 2 % inch lead and% inch steel, respectively. The canister closure incorporates two 0-ring seals to ensure its confinement function. The confinement system consists of the canister inner shell, bottom plate, top flange, top lid, top lid 0-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 % inch steel, 2 Ye inch lead and 1 inch steel, respectively. An outside shell called the "water jacket" a

contains water for neutron shielding, with 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 3 1 Amendment~

Facility Operating License Appendix C - Inter-Unit Fuel Transfer Technical Specifications 2.0 CONDITIONS 2.1 OPERATING PROCEDUR,ES 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 Hl-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 Hl-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.

I) 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 3 2 Amendment~

APPENDIXC TO FACILITY OPE:R/\TING LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 3 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART II: TECHNICAL SPECIFICATIONS FACILITY LICENSE NO. DPR-64 DOCKET NO. 50-286 Amendment No. ~

Facility 013eratiAg License No DPR-64 Appendix A - Technical Specification Bases PERMANENTLY TABLE OF CONTENTS DEFUELED B.2.0 8+/--hl

~

c.:~~~~s::

Reaeter Care SLs Reaeter CeelaAt SysteFA Pressttre SL B.3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY SURVEILLANCE REQUIREMENT (SR) APPLICABILITY

&H REACTIVITY CONTROL SYSTEMS 83+/- SI IUTDOVVN MARGIN

~ Core Reaetivity

~ Moderator TeFA13eratttre CoeffieieAt (MTC)

B-3+/-4 Rod Grettf3 AligAFAeAt LiFAits

&-3:-M Shtttdovm BaAI< IAsertioA LiFAits

~ CoAtrol BaAk IAsertioA LiFAits 8:-3-+/-:-7 Rod PositioA IAdieatioA

~ PIIYSICS TESTS E>Eee13ti0As MODE 2

~ POVJER DISTRIBUTION LIMITS

~ Heat Fltt>E Hot ChaAAel Faetor (FQ(Z))

~ Nttelear EAthal13y Rise Hat ChaAAel Faetor (FN M)

~ AXIAL FLUX DIFFERENCE (AFD)

~ QUADRANT POWER TILT RATIO (QPTR)

INSTRUMENTATION Reaetor ProteetioA SysteFA (RPS) IAStFttFAeAtatioA EAgiAeered Safety Feature AetuatioA System (ESFAS)

IAStFttFAeAtatioA

~ Post AeeideAt MoAitoriAg (PAM) IAStFttFAeAtatioA

~ ReFAote ShtttdowA 8-:-3-3:-§. Less of Po*.ver (LOP) Diesel Ge Aerator (DG) Start IAStFttFAeAtatioA

~ CoAtaiRFAeAt Purge SysteFA aAd Pressttre Relief LiAe lsolatioA IAstrnmeAtatioA CoAtrol Room VeAtilatioA (CRVS) AetttatioA IAstrnmeAtatioA Fttel Storage BttildiAg EmergeAEY VeAtilatioA SvsteFA (FSBEVS) AetttatioA IAstrttmeAtatioA (continued)

INDIAN POINT 3 Bi Revision 4

I j

J*

I

i *: 1 ii jjjjj~iiiiiiiiii ~ill~~~ l!ill~JJJ!J

Facility OperntiAg License No DPR-64 Appendix A -Technical Specification Bases PERMANENTLY TABLE OF CONTENTS (continued)

DEFUELED

~SPENT FUEL PIT REQUIREMENTS B.3.7 PLANT SYSTEMS

&-3-;-7-d- MaiA Steam Safety Valves (MSSVs) 8-:-3-,H MaiA Steam lsolatioA Valves (MSIVs) aAel Mai A Steaffi Check Valves (MSCVs)

&-3-;-7-:-3 MaiA Boiler Feeelpump Discharge Valves (MBFPDVs), MaiA Feeelwater RegulatioA Valves (MFRVs), MaiA Feeelv,ater IAlet lsolatioA Valves (MFIIVs) aAel MaiA Feeelwater (MF) Lo*.v Flow Bypass Valves

&3-,-7-:4 Atmospheric Dump Valves (ADVs)

~ Auxiliary Feeelwater (AP.A') System

~ CoAeleAsate Storage TaAIE (CST)

~ City '.Vater (C'.:'.')

B-:-3-:-+.-8 CompoAeAt CooliAg Water (CC'.:'/) System

&-3-,-7-:-9 Service '.:'later (SW) System B.3.7.10 Ultimate Heat SiAk (UI-IS)

B.3.7.11 CoAtFOI Room VeAtilatioA Systeffi (CRVS)

B.3.7.12 CoAtrol Room Air CoAelitioAiAg System (CRAGS)

B.3.7.13 Fuel Storage BuileliAg EmergeAcy VeAtilatioA System (FSBEVS)

B.3.7.14 Spent Fuel Pit Water Level B.3.7.15 Spent Fuel Pit Boron Concentration B.3.7.16 Spent Fuel Assembly Storage B.3.7.17 SecoAelary Specific Activity

&3-:8 ELECTRICAL POWER SYSTEMS B-:-3-:-8-:-1 AC Sources OperatiAg

&-3-;-8-; AC Sources ShutelowA 8-:-3-:-8-3 Diesel Fuel Oil aAel StartiAg Air 8-:-3-:-&4 DC Sources OperatiAg

~ DC Sources ShutelowA

~ Battery Cell Parameters

&-3.,.&-+ IAverters OperatiAg B-3:-8-:-8 IA*verters ShutelowA

&.-3-:&9 DistributioA Systems OperatiAg B.3.8.10 DistributioA Systems ShutelowA (continued)

INDIAN POINT 3 B iii Revision 4

.. ]

~ jJ!!! !

!Deleted B 2.0 B 2.1.1

~:~M~::~L Reactor Core Sls BACKGROUND GOG 10 (Ref. 1) requires thats e . .

are n~t exceeded during steady ~t:~:1ed acc:ptable fuel design limits transients, and anticipated . operation, normal operational accomplished by having a de op:rat,onal occurrences (AOOs). This is basis' *,Jhi ch corresponds t/:r 9~~e fr~m b~u~l eate boiling CDNB) design leve~ ~the 95/95 DNB criterion) thp;oD~B,.~.:ty at a 95?& confidence requ1r1ng that fuel centerline t a "111 not occur and by temperature. emperature stays bel o*,., the melting The restrictions of this SL

~l adding, as \*,*ell as possi bl :r:~:::. overheating. of the fuel and

,~ the r:lease of fission products ~ngt:erforat,on, that Hould result 0

011 erheat, ng of the fuel . e reactor cool ant peak: 1 .

. rnear heat rate (Ll=IR) belo"' ~ a,n,ng the steady

, s prevented by mai nt . . .

state melting occurs. Overheating of ~h the 1e *el at ~\/hi ch fuel centerline restricting fuel operation to ,.,i th 7 f~:l cladding is prevented by w*here the heat transfer coeff1'". t,n e nucleate boiling regime temperat ure 1s* slightly abo11e th c, en isl largeeand th c l adding surface

e coo ant saturation temperature.

~uel centerline melting occurs ***hen th

,n a region of the fuel is high" e local Ll=IR, or pm*mr peak:i ng temperature to reach the melt" eno~gh to cause the fuel centerline' pellet:penu centerline melting,ngma,point of the fuel .xpansion ~ of the cladding to the point off *1 } caus: the pellet to stress the acti v,*tY to the reactor cool a, ant.

ure

all" e***rng an uncontrolled rel ease of Operati~n above the boundary of the ..

result ,n excessi"e cladding t nucleate bo1l1ng regime could and . th e resultant " emperature sharp reduction . h becau s e of t he onset of DNB Ins1d7 the steam film, high cladd',n eat transfer coefficient.

cl ad~,ng ~*Jater (zirconium water) ~:g !:mperatures are reached, and a chemical reaction results in oxidat~c ,on may tal~e place. This structurally ',veak:er form. Th. ,., ,on of the fuel cl adding to a resultin g in* an uncontrolled 15 rel .. eak:erOf form .ma:Y

l ose 1*ts integrity coolant. ease act1v1ty to the reactor '

The proper functioning of the React steam generator safety ,, l" or Protect,. on System ( RPS) and

-St:&. va .es prevents violation of the reactor core (continued)

INDIAN POINT 3 ~-1 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.~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 Applicabili y statement within each individual Specification as the requi ement for when the LCO is required to be met (i.e., when the is in the MODEa 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, unless othonlise specified. 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.

Thero are t110 basic types of Required 1\ctions. The first type of Required Action speeifies a time limit in 11hieh the LCO must be mot. This time limit is the Completion Time to restore an inoperable system or component to OPER..~gLg status or to restore variables to Hithin speeified limits. If this type of Required AetioR is not eompleted 1cithin the speeified Completion Time, a shutdo1m may be required to plaeo the unit in a MODE or conditioR in 11hieh the apecifieation is not applieable. (Whether stated as a Required Action or not, eorrection of the entered Condition is an action that may al11ays be considered upon entering ACTIONa.)

The second type of Required Action specifies the remedial measures that permit eontinued operation of tho unit that is not further restricted by the Completion Time. In this case, complianee 11ith the Required Actions provides an aeceptable level of safety for continued operation.

(continued)

INDIAN POINT 3 B 3.0 - 1 Revision -a-

LCO Applicability B 3.0 BASES LCO 3.0.2 Completing the Required Actions is not required when an LCO is (continued) met or is no longer applicable, unless otherwise stated in the individual Specifications.

The nature of some Required Actions of some Conditions neeessitatea that, onee the Condition is entered, tho Required Actions must be completed even though the associated Conditions no longer mdst. The individual LCO' s ACTION£ specify the Required .'\et ions 11here this is the ease. An-mrnFRple of this is in LCO 3. 4. 3 1 "RC£ Pressure and TeFRperature (P/T) Limits,"

The Completion Times of the Required Actions are also applicable 11hen a syatem or eoFRponent ia removed from service intentionally. Tho ACTION£ for not meeting a single LCO adequately manage any inereaao in plant risk, provided any unusual m,tornal conditions ( o. g., severe 11oathor, off site pm1or instability) are considered. In addition, the increased risk associated 11ith simultaneous removal of multiple structures, syatoms, trains or eoFRpononts from aervieo is assessed and managed in aeeordanee Hith 10 CFR 50. e5 (a) ( 4).

Individual epoeifieations may specify a time limit for performing an eR Hhon equipment is removed from service or bypassed for testing, In this ease, tho CoFRplotion Times of tho Required Actions are applicable 1ilion this time limit oEpires, if the equipment remains removed from service or bypassed.

When a change in MODE or other specified condition is required to eoFRply with Required Aotions, tho unit may enter a ~40DE or other specified condition in ,iliieh another epeeifieation boeomos applicable, In this ease, tho Completion Times of tho associated Required Actions './Ould apply from tho point in time that tho nor,;

epoeifieation becomes applicable, and tho ACTimle Condition ( s) are entered.

(continued)

INDIAN POINT 3 B 3.0 - 2 Revision.§.

LCO ~pplicability

~

LCO 3.0.3 LCO 3.0.3 establishes met the and:actions that HlUSt be implemented 11hen an LCO.

An associated met an Re qtured no d .

other Condit'Action and Completion~*

ion applies; or zime is not The b1 condition r th of tttO~h .

0 associated n ~CTIunit is' not specifically addre combination of c*~ ~m,rn. This means that no ssed made that exactl:nditions stated in the ACTIONS the .

unit.

are such th t

'.! corresponds t Sometimes . o the actual c

. , possible coHlbi t.

. ~

can be a ~~

ondition of a entering LCO na ions of c d' cases, the ACTIONS 3.0.3 is r.,arranted~ ~on itions correspond' pacifically stat ' in such 3 0 3 ing to such combinat' ea Condition

. . be entered ifflfFlediately. ions and also that LCO This s *

.pacification unit in a safe MODE or d 1*

operat. e ineates the . time 1* i t s for pla .

ion cannot be . other

_e spo i f ied

.im condit . c i n g the

~

operation as defin dHl:~ntained llithin the 1* ~on Hhen ea<ry. ia<o LCO '. 0 ~' ~ho LCO *** He ACT;::' <or ea<o practicable to F

s ould be avoided . .. Planned should be a?oid planned entry i t . If it is 3 not assessed and 50.i5(a) (4), aEd no LCO 3 o managed in aooorda*;* , plant risk less effect on. the planned entry into Lnce Hlth 10 CFR alternatiues *

plant safety than oth~r practicable CO ~. 0. 3 should have Upon entering LCO 3 oreerly Th. shute .0.3, . 1 . hour is allm1ed to pr is includes mm t* before initiating .

a ch . e p a r e for an reduction in el:::r~o perHlit the operat::g: in uni~ operation.

ensure the stabil. i.cal generation 11ith th o ooord~nate the The time limits sit:i _a~d availability oft: load dispatcher to permit the shutdo~:c:fied to enter lor.1er HO~E:le:trical O grid.

manner that i O proceed in a cot operation s 11011 "ith

n rolled a e an ,lithin th ': in the specifi d . n orderly thed minimum r:q~~pabilities of the uni: maiamum oooldor.m rate

'* uired equi pment is OPEPu.".BLE. , assuming .a tht only (continued)

INDL".~l POINT 3 B 3.0 3 Revision 5

t

~ \ \ '\... \ \

~ \ \

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- ~ ~ - - - - - - - - - - - - - - - - -

\ \

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- --- - -- - -- - - -- ------ ~ ~-~-----,

i J

\ \ \ \ ~

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~ ------**--------- --- - - -- ----- --- _________ - .. ~ ~-------,

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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 ~40DEe 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 OPER.~BILITY of systems and 00H1.ponents 1 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.

eystems and 00H1.ponents are assumed to be OPERABLE 11hen the assooiated eRs have been mot. Nothing in this epeoifioation, hm10ver 1 is to be oonstrued as iH1.plying that systems or GOHl.pOnents are OPER.~BLE 11hon:

-. The systems or 00H1.pononts are knmm to be inoperable, although still meeting the eRs; or

,e..,. The requirements of the eurveillanoe ( s) are lmmm not to be met bot'.leen required eurveillanoe p e r f o r m a n o e s . ~

Surveillances do not have to be performed when the u-R-i-t! is in a

~40DE or other specified condition for which the requirements of the associated LCO are not applicable, unless othen1ise speoifiod. Tho £Rs assooiated 11ith a test mweption are only applioable when the test oiroeption is used as an allm1able eiweption to the requirements of a epeoifioation.

Unplanned events may satisfy the requirements (inoluding applioable aooeptanoe oriteria) for a given eR. In this ease, the unplanned event may be oredited as fulfilling the performanoe of the eR.

This allmianoe inoludes those eRs Hhose performanoe is normally preoluded in a given NODE or other speoified oondition.

(continued)

INDIAN POINT 3 B 3.0 - 15 Revision -e-

SR Applicability B 3.0 BASES __, to restore variables within their variables that are outside specified limits. their specified limits, SR 3.0.1 (continued) Survei lances,--Hlc&.8*a-3c~f--1*'"l?.'ce.;ic+/-1,,a.R-,e,,,-:>,;;~'o<e.JH.G--B-V-~~tt+/--£-E~

do not have to be the ACTIONS define Upon completion of maintenance, appropriate post maintenance testing is required to declare equipment OPER.~BLE. ~

includes ensuring applicable Surveillances are not failed and their most recent performance is in accordance Hith 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. ffi these situations, the equipment may be considered OPER.7\.BLE provided testing has been satisfactorily completed to the 01,tent possible and the equipment is not other11ise believed to be incapable of performing its function. This 11ill allo11 operation to proceed to a MODi or other specified condition Hhere other necessary post maintenance tests can be completed.

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action 14th a Completion Time that requires the periodic performance of the Required Action on a "once per .

facility SR 3.0.2 permits a 25% extension of the inte val specified in the Frequency. This extension facilitate urveillance scheduling and considers plant operating onditions 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 01weptiens to SR 3. 0. 2 are these Surveillances for Hhich the 25% 01,tension of the interval specified in the Frequency does not apply, These 01rneptions are stated in the individual Specifications. An euample of 11here SR 3.0.2 does not apply is a Surveillance 11ith a Frequency of "in accordance 11ith 10 CFR 50 1 Appendi1f J, as modified by approved 01rnmptions." -T-h-e requirements of regulations take precedence over the TS. -T-h-e TS cannot in and of themselves eutend a test interval specified in the regulations. Therefore, there is a Wote in the Frequency stating, "SR 3.0.2 is not applicable."

(continued)

INDIAN POINT 3 B 3.0 - 16 Revision.§.

SR Applicability B 3.0 BASES SR 3.0.2 As stated in SR 3.0.2 1 tho 25% extension also does not apply to (continued) the initial portion of a poriodio Completion Time that requires performanoo on a "onoo per ... " basis. Tho 25% extension applies to eaoh porformanoo after tho initial porformanoo. ~

initial porformanoo of tho Required *11,otion, ,.'bother it is a partioular Survoillanoe or some other remedial aotion, is oonsidorod a single aotion .:ith a single Completion Time. Gn-e reason for not allm1ing tho 25% ontonsion to this Completion Time is that suoh an aotion usually verifies that no loss of funotion has eoourrod by ohooking tho status of redundant or merely as a diverse oompononts or aooomplishos tho funotion of tho convenience inoperable equipment in an alternative manner.

The p isions of SR 3.0.2 are not intended to be used repeatedl to extend Surveillance intervals (other than those oonsistent r.1ith refueling intervals) or periodio Completion Time intervals beyond those specified.

completed SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declarin affeoted equipment inoperable or an affected variable utside the specified limits when a Surveillance has not been performed 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.

0complete I This delay period provides adequate time to perform Surveillances that have been missed. ~

y This delay period permits the porformanoe of a Surveillance before complying with Required Actions or other remedial measures that might preclude f the Surveillance.

completion (continued)

INDIAN POINT 3 B 3.0 - 17 Revision.§.

SR Applicability B 3.0 BASES D!acility SR 3.0.3 The basis for this delay period includes consideration o f ~

(continued) 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 tiHlo intervals, but upon specified unit conditions, operating situations, er requirements of regulations (e.g., prior to entering MODE 1 after each fuel loading, or in accordance 11ith 10 CFR 50 1 Appendi1, J, as modified by approved mrnmptions, etc.) is discovered to not have been performed 11hen specified, 8R 3.0.3 allo11s for the full delay period of up to the specified Frequency to perform. the aurveillance. '!l:01101.rer, since there is not a tim.e interval specified, the Hlissed 8urveillance should be perform.ad at the first reasonable opportunity.

8R 3. 0. 3 provides a time limit for, and all011ances for the perform.ance of, 8urveillances that becom.o applicable as a consequence of MODE changes imposed by Required Actions.

8R 3.0.3 is only applicable if there is a reasonable m,pectation the associated equipment is OPER.7illLE or that variables are Hithin lim.its, and it is m,pected that the 8urveillance Hill be HJ.et 11hen performed. Many factors should be considered, such as the period of time since the 8urveillance

\las last performed, or 11hether the aurvoillance, or a portion thereof, has ever boon perform.ad, and any other indications, tests, or activities that m.ight support the expectation that the aurveillance 11ill be met 11hen performed. An m.ample of the use of 8R 3. 0. 3 Hould be a relay contact that 11as not tooted as required in accordance Hith a particular aR, but previous successful perforHlances of the 8R included the relay contact; the adjacent, physically connected relay contacts ;1ore tested during the aR perforHJ.ance; tho subject relay contact has been tested by another 8R; or historical operation of the subject relay contact has boon suooessful. It is not sufficient to infer the behavior of tho associated equipment from tho porform.ance of similar equipm.ont. The rigor of determining 11hethor there is a reasonable m,pectation a 8urvoillanoe 1lill be met 11hon performed should increase based on tho length of time since the last perform.ance of the 8urvoillance. If tho 8urvoillance has been performed recently, a revieH of the 8urveillance history and equipment perfonaance Hlay be sufficient to support a reasonable expectation that the 8urveillance 1lill be HJ.et 1~en performed.

(continued)

INDIAN POINT 3 B 3.0 - 18 Revision.§.

SR Applicability B 3.0 BASES SR 3.0.3 For .Surveillances that have not been performed for a long (continued) period or that have never been performed, a rigorous evaluation based on objective evidence should provide a high degree of confidence that the equipment is OPER.",BLE. The evaluation should be documented in sufficient detail to all011 a lm011ledgeable individual to understand the basis for the determination.

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 repeatedly to extend Surveillance facility facility ~as a convenience to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months limit of the specified Frequency is to perform the issed Surveillance, it is expected facility missed Surveilla ce will be performed at the first reason ble opportunity. he determination of the first opportunity sho ld include consideration of the

~ risk (from delaying the Surveillance as well as a n y ~ configuration cha ges required or shutting the plant 4ew-n-to perform the Surveil ance) and impact on any analysis assumptions, in addition to 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. 5 ( a) ( 4) and its implementation guidance, NRG Regulatory Guide 1.1@2, 'Assessing and Managing Risk Before Maintenance Activities at Nuclear P011er Plants.' This Regulatory Guide addresses consideration of temporary and aggregate risk impacts, determination of risk management action thresholds, and risk management action up to and including plant shutd01m. 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 Hith the importance of the component. Missed .Surveillances for important components (continued)

INDIAN POINT 3 B 3.0 - 19 Revision.§.

SR Applicability B 3.0 BASES SR 3.0.3 should be analy3ed quantitatively. If the results of the risk (continued) 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 licensee's 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.

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. rV-a-r-ia_b_l_e_S_W~it-h-in~s-p_e_c_if-ie-d~li_m_i-ts-,

This Specification ensures that system and component OPER.~BILITY requirements and variable limits are met bef entry into MODES or other specified conditions in the Applicability for which these,,,.:S-'v<&1==-el'a&-aB1i--,;:l-f*l$<&n-eBt-E;......-~&u-r-e-The provisions of variables ensure safe Specificat. ould not be interpreted as endorsi handling and storage of L__~~ e to exercise the good practice of restoring systems or spentfuel components to OPER.~BLE status before entering an associated MODE or other specified condition in the Applicability.

A pror.rision is included to alloH entry into a MODE or other specified condition in the Applicability Hhen an LCO is not met due to Surveillance not being met in accordance Hith LCO 3.0.4.

variables that are outside their specified limits Howe in certain circumstances, failing to meet an SR will not result

R 3.0.4 restricting a MODE change or other specified conditio When a system, subsystem, t8c~c&.~&r---e.eH\l*ffi<~br--l,e,,;;'+/-t"!-9'ac::-,~~variable is inoperable or a variable is outside its specified limits, t ssociated SR(s) are not specified limit be performed, per SR 3.0. , ic do not have to be performed on-~.e,i:;;.e.;i:;.a-e-+/-~

equipment. Whe 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 HODES or (continued)

INDIAN POINT 3 B 3.0 - 20 Revision -e-

SR Applicability B 3.0 BASES SR 3.0.4 other specified conditions of the Applicability. H0\10vor, (continued) sinoo tho LCO is not met in this instanoo, LCO 3.0.4 11ill govern any restriotions that my (or may not) apply to MODE or other spooifiod oondition ohangos. SR 3.0.4 does not restrict changing MODEa 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 aR 3.0.4 shall not prevent entry into MODEa or other speoified oonditions in the Applioability that are required to oomply 11ith ACTIOl'la. In addition, the provisions of aR 3.0.4 shall not prevent ohanges in MODEa or other speoified oonditions in the .".pplioability that result from any unit shutdO\m. In this oontmrt a unit shutdmm is defined as a ohange in MODE or other speoifiod oondition in the Applioability assooiated Hith transitioning from MODE 1 to MODE 2, MODE 2 to MODE 3 1 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 in 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 3 B 3.0 - 21 Revision.§.

Spend Fuel Pit Water Level SPENT FUEL PIT B 3.7.14 REQUIREMENTS B 3.7 B 3.7.14 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 design and the Spent Fuel Cooling and Cleanup System is given in the FSAR, Section 9.5 (Ref. 1).

The assumptions of the fuel handling accident are given in the FSAR, Section 14.2 (Ref. 2).

APPLICABLE SAFETY ANALYSES The minimum water level in the spent fuel pit meets the assumptions of the fuel handling accident described in FSAR, Section 14.2 (Ref. 2).

The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose per person at the exclusion area L/

boundary satisfies the 10 CFR 50.67 (Ref. 3) limits. /1 According to Reference 2, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference v' 4 can be used directly. In practice, this LCD preserves this /1 assumption for the bulk of the fuel in the storage racks.

The Spent Fuel Pit water level satisfies Criteria 2 and 3 of 10 CFR 50.36.

LCD 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.

(continued)

INDIAN POINT 3 B 3.7.14 - 1 Revision J.

---~~------

Spend Fuel Pit Water Level B 3.7.14 BASES APPLICABILITY This LCD applies during movement of irradiated fuel assemblies in the spent fuel pit, since the potential for a release of fission products exists.

ACTIONS A.,.l 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 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 'IJhile in MODE 5 or i, LCO 3.0.3

',*mul d not specify any action. If moving irradiated fuel assemblies

~\lhi 1e 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 shutdO\~n.

SURVEILLANCE REQUIREMENTS SR 3.7.14.1 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 normally in equilibrium *,Jith the refueling canal and reactor cavity, and the level in the refueling reactor cavity is checked daily in accordance *.:ith SR 3.9.i.l.

(continued)

INDIAN POINT 3 B 3 .7 .14 - 2 Revision -+/--

SPENT FUEL PIT Spent Fuel Pit Boron Concentration 83.7.15 REQUIREMENTS B 3.7.15 Spent Fuel Pit Boron Concentration BASES BACKGROUND In the Maximum Density Rack (MOR) design, the spent fuel storage pool is divided into two separate and distinct regions. The layout of the IP3 MDR is shown in Figure B 3.7.16-1. As shown in Figure B 3.7.16-1, Region 1 (Columns SS-ZZ, Rows 35-64) includes 240 storage positions and Region 2 (Columns A-RR, Rows 1-34) includes 1105 storage positions. Region 1 is analyzed for storage of high-enrichment and low-burnup fuel. Region 2 is analyzed for storage of fuel with either higher burnup or lower enrichment. Each region has been separately analyzed for close packed storage when all cells in that region contain fuel of the highest reactivity stored in accordance with LCO 3. 7.16, Spent Fuel Assembly Storage. This analysis is the basis for the restrictions on fuel storage locations established by LCO 3.7.16.

Limits, based on a combination of initial enrichment and burnup, are used to determine if a fuel assembly must be stored in region 1 or if the fuel assembly may be stored in either region 1 or region 2. Fuel with the highest initial enrichments are subject to additional restrictions even when stored in region 1. Fuel assemblies with an initial enrichment> 5.0 wt%

U-235 cannot be stored in the spent fuel pit in accordance with paragraph 4.3.1.1 in Section 4.3, Fuel Storage.

The water in the spent fuel pit normally contains soluble boron, which results in large subcriticality margins under actual operating conditions.

However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting kett of 0.95 be evaluated in the absence of soluble boron. Hence, the design of both regions is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the regions fully loaded when fuel storage locations, enrichment and burnup are in conformance with analysis assumptions as specified in LCO 3.7.16. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 1) allows credit for soluble boron under other abnormal or accident conditions, because only a single accident need be considered at one time. For example, the accident (continued)

INDIAN POINT 3 B 3.7.15-1 Revision 4

Spent Fuel Pit Boron Concentration 83.7.15 BASES LCO (continued) confirms that there are no misloaded fuel assemblies. With no misloaded fuel assemblies and unborated water, the spent fuel pit design is sufficient to maintain the core at Kett~ 0.95.

The LCO is modified by a Note that states that during inter-unit transfer of fuel the spent fuel pit boron concentration must also meet Appendix C LCO 3.1.1, "Boron Concentration". This requirement ensures that the criticality analysis of the fuel within the Shielded Transfer Canister remains bounding.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pit, until a complete spent fuel pit verification has been performed following the last movement of fuel assemblies in the spent fuel pit. This LCO does not apply following the verification, since the verification would confirm that there are no misloaded fuel assemblies. With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.

ACTIONS A.1, A.2.1 and A.2.2 The Required /\etions are modified by a Note indicating that LGO 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. Alternatively, beginning a verification of the Spent Fuel Pit fuel locations, to ensure proper locations of the fuel, can be performed. However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.

(continued)

INDIAN POINT 3 B 3.7.15-3 Revision 4

l Spent Fuel Pit Boron Concentration B 3.7.15 BASES ACTIONS /\.1, /\.2.1 and A.2.2 (contin':Jed)

If the LGO is not met *while moving irradiated fuel assemblies in MODE 5 or 6, LGO 3.0.3 *.vould not be applicable. If moving irradiated fuel assemblies 'Nhile 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 REQUIREMENTS SR 3.7.15.1 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 31 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. This SR is not required to be met or performed if a spent fuel pit verification for conformance with LCO 3.7.16, Figures 3.7.16-1 and B 3.7.16-1, has been performed on all fuel assemblies since the last verification following the last movement of fuel assemblies in the spent fuel pit.

REFERENCES 1. Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

2. SER related to Amendment 173 to Facility Operating License No. DPR-64, Indian Point Nuclear Generating Unit No. 3, April 15, 1997.

3. Criticality Analysis of the Indian Point 3 Fresh and Spent Fuel Racks, Westinghouse Commercial Nuclear Fuel Division, October, 1996.

INDIAN POINT 3 83.7.15-4 Revision4

Spent Fuel Assembly Storage SPENT FUEL PIT B 3.7.16 REQUIREMENTS B 3.7 B 3.7.16 Spent Fuel Assembly Storage BASES BACKGROUND In the Maximum Density Rack (MOR) design, the spent fuel pit (SFP) is divided into two separate and distinct regions. The layout of the IP3 MOR is shown in Figure B 3.7.16-1, IP3 Maximum Density Spent Fuel Pit Racks, Regions and Indexing. As shown in Figure B 3.7.16-1, Region 1 (i.e., Columns SS-ZZ, Rows 35-64) includes 240 storage positions and Region 2 (i.e., Columns A-RR. Rows 1-34) includes 1105 storage positions. Region 1 is analyzed for storage of high-enrichment and low-burnup fuel. Region 2 is analyzed for storage of fuel with either higher burnup or lower enrichment. Each region has been separately analyzed for close packed storage when all cells in that region contain fuel of the highest reactivity that is allowed by this LCO. This analysis is the basis for the restrictions on fuel storage locations established by this LCO.

Prior to storage in the spent fuel pit, fuel assemblies are classified as to the level of reactivity based on the initial enrichment and burnup. This classification is made using Figure

3. 7.16-1. "Fuel Assembly Cl assifi cation for Storage in the Spent Fuel Pit". This classification is used to determine in which region a particular fuel assembly may be stored and if additional restrictions must be applied to the assemblies in adjacent locations. Figure 3.7.16-1, "Fuel Assembly Classification for Storage in the Spent Fuel Pit", is used to classify each assembly into one of the following categories based on initial U-235 enrichment and burnup:

Type 2 assemblies are the least reactive assemblies and include any assembly for which the combination of initial enrichment and burnup places the assembly in the domain labeled Type 2 in Figure 3.7.16-1.

Type 2 assemblies may be stored in any location in Region 1 or Region 2 of Figure B 3.7.16-1.

Type 1A assemblies are more reactive than Type 2 assemblies and include any assembly for which the combination of initial enrichment and burnup places the assembly in the domain labeled (continued)

INDIAN POINT 3 B 3. 7.16-1 Revision G

Spent Fuel Assembly Storage B 3.7.16 BASES LCD Fuel assemblies stored in the spent fuel pit are classified in accordance with Figure 3.7.16-1 based on initial enrichment and burnup which is indicative of fuel assembly reactivity. Based on this classification, fuel assembly storage location within the spent fuel pit and storage location relative to other assemblies is restricted in accordance with the rules established by this LCD.

Fuel assemblies with an initial enrichment> 5.0 wt% U-235 are not shown on Figure 3.7.16-1 because fuel assemblies with this enrichment cannot be stored in the spent fuel pit in accordance with limits established in Technical Specification Section 4.3.

APPLICABILITY This LCD applies whenever any fuel assembly is stored in the spent fuel pit.

ACTIONS Required ActioA A.l is modified by a Note iAdicatiAg that LCO 3.0.3 does Aot apply.

When the configuration of fuel assemblies stored in the spent fuel pit is not in accordance with this LCD, the immediate action is to initiate action to make the necessary fuel assembly movement(s) to bring the configuration into compliance with this LCO.

If uAable to move irradiated fuel assemblies ~:hile iA MODE 5 or 6, LGO 3.0.3 would Aot be applicable. If uAable to move irradiated fuel assemblies ~:hile in MODE 1, 2, 3, or 4, the action is independent of reactor operatioA. Therefore, iAability to move fuel assemblies is Aot sufficient reasoA to require a reactor shutdo~m.

(continued)

INDIAN POINT 3 B 3. 7.16-4 Revision G

Enclosure, Attachment 2 NL-20-033 Indian Point Nuclear Generating Station Unit 3 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 I -

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 RENEWED FACILITY LICENSE Renewed License No. DPR-64

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

A. The application for a renewed license filed by Entergy Nuclear Indian Point 3, LLC (ENIP3} (the licensee) and Entergy Nuclear Operations, Inc. (ENO)

(operator) for Indian Point Nuclear Generating Unit No. 3 (IP3 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. The facility will be maintained in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. 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 Commission's regulations; D. ENIP3 and ENO are financially and technically qualified to engage in the activities authorized by this amendment; E. ENIP3 and ENO have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements" of the Commission's regulations; F. 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; G. 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; Amendment [###]

H. The issuance of this renewed license is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. 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 facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commission's regulations.

2. Accordingly, Renewed Facility License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows:

A. This renewed license applies to the Indian Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and associated equipment (the facility),

owned by ENIP3 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 licenses:

(1) Pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," (a) ENIP3 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; (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 calibration of reactor instrumentation and are used in the calibration of radiation monitoring equipment, and that were used as fission detectors in amounts as required; Amendment[# ##]

(4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to 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; and 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 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.

D. (DELETED)

E. (DELETED)

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. 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 CFR 50.54(p). The combined set of plans 1 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 O," and was submitted by letter dated October 14, 2004, as supplemented by letter dated May 18, 2006.

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

Amendment[# ##]

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. 243, as supplemented by changes approved by License Amendment Nos. 254, 260, and 263.

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 supplemen ted 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 Commissio n-approved authorization.

H. Deleted per Amendmen t [###]

I. DELETED J. DELETED K. DELETED L. DELETED M. DELETED N. DELETED

0. Deleted per Amendmen t [###]

P. ENIP3 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 $70 million contingency commitmen t to provide funding for the facility as represented in the application for approval of the transfer of the license from PASNY to ENIP3 and ENO, without the prior written consent of the Director, Office of Nuclear Reactor Regulation.

Q. DELETED R. DELETED S. DELETED T. DELETED U. DELETED V. DELETED Amendmen t[###]

W. For purposes of ensuring public health and safety, ENIP3, upon the transfer of this license to it, and upon transfer of decommissioning funds from PASNY to ENO, shall provide decommissioning funding assurance for the facility by the prepayment or equivalent method, to be held in a decommissioning trust fund for the facility, of no less than the amount required under NRC regulations at 10 CFR 50. 75. Any amount held in any decommissioning trust maintained by ENO for the facility after the transfer of the facility license to ENIP3 may be credited towards the amount required under this paragraph.

X. ENIP3 shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the application for the transfer of this license to ENIP3 and ENO, as modified by the request to transfer decommissioning funds from PASNY, and the requirements of the order approving the transfer and order approving the transfer of decommissioning funds from PASNY to ENO, and consistent with the safety evaluations supporting such orders.

AA. Deleted per Amendment [###]

AB. Deleted per Amendment[###]

AC. 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

7. Spent fuel pool mitigation measures (c} Actions to minimize release to include consideration of:

1. Water spray scrubbing

2. Dose to onsite responders AD. Deleted per Amendment[###]

AE. 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 ENIP3.

Amendment[###]

AF. License Renewal License Conditions (1) The information in the UFSAR supplement, submitted pursuant to 1o 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 AF(1) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).

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.

3. 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 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[###]

APPENDIX A TO FACILITY LICENSE DPR-64 PERMANENT LY DEFUELED TECHNICAL SPECIFICATIONS AND BASES FOR THE INDIAN POINT 3 NUCLEAR GENERATING STATION UNIT NO. 3 WESTCHESTER COUNTY, NEW YORK ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

DOCKET NO. 50-286 Date of Issuance:

April 15, 1976 Amendment No. ###

Facility Operating License No. DPR-64 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 3.7.14 Spent Fuel Pit Water Level 3.7.15 Spent Fuel Pit Boron Concentration 3.7.16 Spent Fuel Assembly 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 5.3 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 NOT USED 5.5.4 Radioactive Effluent Controls Program 5.5.5 DELETED 5.5.6 DELETED 5.5.7 DELETED 5.5.8 DELETED 5.5.9 DELETED 5.5.10 DELETED 5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5.12 DELETED 5.5.13 Technical Specification (TS) Bases Control 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 5.7 High Radiation Area Indian Point 3 Amendment No.

Definitio ns 1.1 1.0 USE AND APPLICATION 1.1 Definitions

NOTE--------------------------------------------

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

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

CERTIFIED FUEL HANDLER A CERTIFIED FUEL HANDLER is an individual who (CFH) 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 3 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 ...

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 3 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.

Indian Point 3 1.3-1 Amendment No.

Completion Times 1.3 1.3 Completion Time EXAMPLE (continued)

EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pit A.1 Suspend Immediately boron movement of fuel concentration not assemblies in the within limit. spent fuel pit.

A.2 Initiate action to Immediately restore spent fuel pit boron concentration to within limit.

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 within limit.

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 3 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, Surveillance Requirement (SR)

Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR.

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 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 3 1.4-1 Amendment No.

Deleted 2.0 2.0 DELETED Indian Point 3 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 3 3.0-1 Amendment No.

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.i 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 3 3.0-2 Amendment No.

Spent Fuel Pit Water Level 3.7.14 3.7 SPENT FUEL PIT REQUIREMENTS 3.7.14 Spent Fuel Pit Water Level LCO 3.7.14 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 .14.1 Verify the spent fuel pit water level is ~ 23 ft 7days above the top of the irradiated fuel assemblies seated in the storage racks.

Indian Point 3 3.7.14-1 Amendment No.

Spent Fuel Pit Boron Concentration 3.7.15 3.7 SPENT FUEL PIT REQUIREMENTS 3.7.15 Spent Fuel Pit Boron Concentration LCO 3.7.15 The spent fuel pit boron concentration shall be ~ 1000 ppm.

NOTE--------------------------------------------

During inter-unit transfer of fuel the spent fuel pit boron concentration must also meet Appendix C LCO 3.1.1, "Boron Concentration."

APPLICABILITY: When fuel assemblies are stored in the spent fuel pit and a spent fuel pit verification has not been performed since the last movement of fuel assemblies in the spent fuel pit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pit boron A.1 Suspend movement of Immediately concentration not within fuel assemblies in the limit. spent fuel pit.

A.2.1 Initiate action to restore Immediately spent fuel pit boron concentration to within limit.

OR A.2.2 Initiate action to perform a Immediately spent fuel pit verification.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the spent fuel pit boron concentration is 31 days within limit.

Indian Point 3 3.7.15-1 Amendment No.

Spent Fuel Assembly Storage 3.7.16 3.7 SPENT FUEL PIT REQUIREMENTS

3. 7 .16 Spent Fuel Assembly Storage LCO 3.7.16 Fuel assemblies stored in the spent fuel pit shall be classified in accordance with Figure 3. 7 .16-1 based on initial enrichment and burn up; and, Fuel assembly storage location within the spent fuel pit shall be restricted based on the Figure 3.7.16-1 classification as follows:

a. Fuel assemblies classified as Type 2 may be stored in any location in either Region 1 or Region 2;

b. Fuel assemblies classified as Type 1A, 1B or 1C shall be stored in Region 1;

c. Fuel assembly storage location within Region 1 shall be restricted as follows:

1. Type 1A assemblies may be stored anywhere in Region 1;

2. Type 18 assemblies may be stored anywhere in Region 1, except a Type 1B assembly shall not be stored face-adjacent to a Type 1C assembly;

3. Type 1C assemblies shall not be stored in Row 64 or in Column ZZ; and

4. Type 1C assemblies shall be stored in Region 1 locations where all face-adjacent locations are as follows:

a) occupied by Type 2 or Type 1A assemblies, or b) occupied by non-fuel components, or c) empty.

APPLICABILITY: Whenever any fuel assembly is stored in the spent fuel pit.

Indian Point 3 3.7.16-1 Amendment No.

Spent Fuel Assembly Storage 3.7.16 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Initiate action to move fuel Immediately LCO not met. to restore compliance with LCO 3.7.16.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3. 7.16.1 Verify by administrative means the initial Prior to storing the fuel enrichment and burnup of each fuel assembly assembly in the spent fuel and that the storage location meets LCO 3. 7 .16 pit requirements.

Indian Point 3 3.7.16-2 Amendment No.

Spent Fuel Assembly Storage 3.7.16 40000 If

/

J 35000 /

J "

'/

,,, J 30000 ,

/

I/

, J 25000 I' ,

sI-c== ,

-I a.

I E

20000

I m

, /

* ..,.._-. n 15000 ,

10000 ,

Figure 3.7.16-1 (Page 1 of 1)

Fuel Assembly Classification for Storage in the Spent Fuel Pit Indian Point 3 3.7.16-3 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location Indian Point 3 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 350 meters and 1100 meters, respectively.

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 s 0.95 if assemblies are inserted in accordance with Technical Specification 3.7.16, Spent Fuel Assembly Storage; C. A nominal 9.075 inch center to center distance between fuel assemblies placed in the high density fuel storage racks (Region II);

d. A nominal 10.76 inch center to center distance between fuel assemblies placed in low density fuel storage racks (Region I);

4.3.2 Drainage The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pool below a nominal elevation of 88 ft.

4.3.3 Capacity The spent fuel pit is designed and shall be maintained with a storage capacity limited to no more than 1345 fuel assemblies.

Indian Point 3 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 3 5.0-1 Amendment No.

i I

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 and Quality Assurance Plan, as appropriate;

b. The plant manager shall be responsible for overall safe maintenance of the facility and shall have control over those onsite activities necessary for safe storage and maintenance of nuclear fuel;

c. The corporate officer with direct responsibility for IP3 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.

Indian Point 3 5.0-2 Amendment No.

Organization 5.2 5.2 Organization 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.

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 CERTIFIED FUEL HANDLER.

f. Deleted.

Indian Point 3 5.0-3 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 3 5.0-4 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; and

e. All programs specified in Specification 5.5.

Indian Point 3 5.0-5 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 Specification 5.6.2 and 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 1o 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.

5.5.2 Deleted Indian Point 3 5.0-6 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Not Used 5.5.4 Radioactive Effluent Controls Program This program conforms to 1o 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 1O times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2;

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 1n 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 3 5.0-7 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.4 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 limited to the following:

a. For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and

b. For iodine-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to dose rate of 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 Effluents Controls Program surveillance frequency.

5.5.5 through Deleted 5.5.10 Indian Point 3 5.0-8 Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 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."

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 shall be limited to 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.

5.5.12 Deleted Indian Point 3 5.0-9 Amendment No.

l Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 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 involve 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.

d. Proposed changes that do not meet the criteria of Specification 5.5.13.b 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 3 5.0-10 Amendment No.

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 1 O 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 un_it/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.8.2, IV.8.3, 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 Part 50.36a and 10 CFR 50, Appendix I, Section IV.B.1.

Indian Point 3 5.0-11 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 3 5.0-12 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.

Indian Point 3 5.0-13 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

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.

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 Indian Point 3 5.0-14 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)

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.

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 3 5.0-15 Amendment No.

APPENDIX B TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3) AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT 3 NUCLEAR POWER PLANT ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART I: NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION PLAN FACILITY LICENSE NO. DPR-64 DOCKET NUMBER 50-286 Renewed License No. DPR-64

INDIAN POINT NUCLEAR GENERATING PLANT UNIT3 ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART I: 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 the NPDES Permits and State Certifications ........................ 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 .............................................................................. 5-2 Renewed License No. DPR-64

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 maintenance 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 No. DPR-64

3.0 Consistency Requirements 3.1 Plant Design and Operation ENO may make changes in facility design or operations 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: Changes in the facility design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of this 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, ENO shall prepare and record an environmental evaluation of such activity.

When the evaluation indicates that such activity involves an unreviewed environmental question, ENO 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 (FES) or final supplemental environmental impact statement (FSEIS), as modified by the staff's testimony to the Atomic Safety and Licensing Boards, supplements to the FES or FSEIS, environmental impact appraisals, or in any decision of the Atomic Safety and Licensing Board;

This provision does not relieve the ENO of the requirements of 10 CFR 50.59.

3-1 Renewed License No. DPR-64

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.

ENO 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 this EPP to meet the objectives specified in Section 1.0. ENO 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 NPDES Permits and State Certifications Violations of the NPDES Permit or the State certification (pursuant to Section 4.1 of the Clean Water Act} shall be reported to the NRC by submittal of copies of the reports required by the NPDES Permit or certification.

Changes and additions to the NPDES Permit or the State certification 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.

The NRC shall be notified of changes to the effective NPDES Permit proposed by ENIP3 and ENO 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. ENO shall provide the NRC a copy of 3-2 Renewed License No. DPR-64

the application for renewal of the NPDES 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 No. DPR-64

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 I P2 and I P3 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 4-1 Renewed License No. DPR-64

conservatively bounds the conditions that will occur in the permanently shut down and defueled condition.

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 NRG, as the authorizing agency, and the NMFS.

4-2 Renewed License No. DPR-64

5.0 Administrative Procedures 5.1 Review and Audit ENO shall provide a 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 is utilized to achieve the independent review and audit function and results of the audits activities shall be maintained and made available for inspection.

5.2 Records Retention Records and logs relative to the environmental aspects of previous 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 in the form of a license amendment incorporating the appropriate revision to the Environmental Protection Plan. This EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.

5-1 Renewed License No. DPR-64

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 report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this Environmental Protection Plan for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous non-radiological environmental monitoring reports, and an assessment of the observed impacts of the previous plant operation and the handling and storage of spent fuel and maintenance of the facility on the environment. If harmful effects or evidence of trends towards irreversible damage to the environment are observed, ENO shall provide a detailed analysis of the data and a proposed course of action to alleviate the problem.

The Annual Environmental Protection Plan Report shall also 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.

5-2 Renewed License No. DPR-64

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 (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact and facility conditions, (b) describe the probable cause of the event, I (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) 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 No. DPR-64

APPENDIX B TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT 3 NUCLEAR POWER PLANT ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART II: RADIOLOGICAL ENVIRONMENTAL FACILITY LICENSE NO. DPR-64 DOCKET NUMBER 50-286 Renewed License No. DPR-64

APPENDIX C TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 3 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART I: SPENT FUEL TRANSFER CANISTER AND TRANSFER CASK SYSTEM FACILITY LICENSE NO. DPR-64 DOCKET NO. 50-286 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 1000) (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%

inch steel, 2 3A inch lead and 3A inch steel, respectively. The canister closure incorporates two 0-ring seals to ensure its confinement function. The confinement system consists of the canister inner shell, bottom plate, top flange, top lid, top lid 0-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 % inch steel, 2 Y:! 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 tipoveraccident.

INDIAN POINT 3 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 1O of the Licensing Report (Holtec International Report Hl-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 Hl-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.

I) 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 3 2 Amendment ###

APPENDIXC TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 3 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART II: TECHNICAL SPECIFICATIONS FACILITY LICENSE NO. DPR-64 DOCKET NO. 50-286 Amendment No. ###

Facility License No. DPR-64 Appendix A - Permanently Defueled Technical Specifications Bases TABLE OF CONTENTS B.2.0 DELETED 8.3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY B.3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY B.3.7 SPENT FUEL PIT REQUIREMENTS B.3.7.14 Spent Fuel Pit Water Level B.3.7.15 Spent Fuel Pit Boron Concentration B.3.7.16 Spent Fuel Assembly Storage Indian Point 3 Bi Revision

LCO Applicability 83.0 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 Re.quired 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 3 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 3 B 3.0-2 Revision

SR Applicability 83.0 BASES SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring 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 licensee's 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.

Indian Point 3 83.0-3 Revision

SR Applicability 83.0 BASES SR 3.0.3 (continued)

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.

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 in 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 3 B 3.0-4 Revision

Spent Fuel Pit Water Level B 3.7.14 B 3.7 SPENT FUEL PIT REQUIREMENTS B 3.7.14 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 design and the Spent Fuel Cooling and Cleanup System is given in the FSAR, Section 9.5 (Ref. 1).

The assumptions of the fuel handling accident are given in the FSAR, Section 14.2 (Ref. 2).

APPLICABLE The minimum water level in the spent fuel pit meets the assumptions of SAFETY the fuel handling accident described in FSAR, Section 14.2 (Ref. 2).

ANALYSES The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose per person at the exclusion area boundary satisfies the 10 CFR 50.67 {Ref. 3) limits.

According to Reference 2, there is 23 ft of water between the top of the damaged fuel bundle and the fuel pool surface during a fuel handling accident. With 23 ft of water, the assumptions of Reference 4 can be used directly. In practice, this LCO preserves this assumption for the bulk of the fuel in the storage racks.

The spent fuel pit water level satisfies Criteria 2 and 3 of 10 CFR 50.36.

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.

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel pit, since the potential for a release of fission products exists.

Indian Point 3 B 3.7.14-1 Revision

Spent Fuel Pit Water Level B 3.7.14 BASES ACTIONS 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.

SURVEILLANCE SR 3.7.14.1 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 procedures and are acceptable based on operating experience.

REFERENCES 1. FSAR, Section 9.5.

2. FSAR, Section 14.2.

3. 10 CFR 50.67.

4. Safety Evaluation Report (SER) for IP3 Amendment.

Indian Point 3 B 3.7.14-2 Revision

Spent Fuel Pit Boron Concentration B 3.7.15 B 3.7 SPENT FUEL PIT REQUIREMENTS B 3.7.15 Spent Fuel Pit Boron Concentration BASES BACKGROUND In the Maximum Density Rack (MDR) design, the spent fuel storage pool is divided into two separate and distinct regions. The layout of the IP3 MDR is shown in Figure B 3.7.16-1. As shown in Figure B 3.7.16-1, Region 1 (Columns SS-ZZ, Rows 35-64) includes 240 storage positions and Region 2 (Columns A-RR, Rows 1-34) includes 1105 storage positions. Region 1 is analyzed for storage of high-enrichment and low-burnup fuel. Region 2 is analyzed for storage of fuel with either higher burnup or lower enrichment. Each region has been separately analyzed for close packed storage when all cells in that region contain fuel of the highest reactivity stored in accordance with LCO 3.7.16, Spent Fuel Assembly Storage. This analysis is the basis for the restrictions on fuel storage locations established by LCO 3.7.16.

Limits, based on a combination of initial enrichment and burnup, are used to determine if a fuel assembly must be stored in Region 1 or if the fuel assembly may be stored in either Region 1 or Region 2. Fuel with the highest initial enrichments are subject to additional restrictions even when stored in Region 1. Fuel assemblies with an initial enrichment > 5.0 wt%

U-235 cannot be stored in the spent fuel pit in accordance with paragraph 4.3.1.1 in Section 4.3, Fuel Storage.

The water in the spent fuel pit normally contains soluble boron, which results in large subcriticality margins under actual operating conditions.

However, the NRC guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting ken of 0.95 be evaluated in the absence of soluble boron. Hence, the design of both regions is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the regions fully loaded when fuel storage locations, enrichment and burnup are in conformance with analysis assumptions as specified in LCO 3.7.16. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRC letter (Ref. 1) allows credit for soluble boron under other abnormal or accident conditions, because only a single accident need be considered at one time. For example, the accident Indian Point 3 B 3.7.15-1 Revision

Spent Fuel Pit Boron Concentration B3.7.15 BASES BACKGROUND (continued) scenarios include movement of fuel from Region 1 to Region 2, or accidental misleading of a fuel assembly in Region 1. This event could increase the potential for criticality of the spent fuel pit. To mitigate these postulated criticality related accidents, boron concentration is verified by SR 3.7.15.1 to be within the limits specified in this LCO prior to movement of fuel assemblies in the spent fuel pit. Safe operation of the MDR with no movement of assemblies is achieved by controlling the location of each assembly in accordance with LCO 3. 7 .16, "Spent Fuel Assembly Storage." Prior to movement of an assembly, it is necessary to perform SR 3.7.15.1.

APPLICABLE Most accident conditions do not result in an increase in the reactivity of SAFETY either of the two regions. Examples of these accident conditions are the ANALYSES loss of cooling (reactivity increase with decreasing water density) and the dropping of a fuel assembly on the top of the rack. However, accidents can be postulated that could increase the reactivity. This increase in reactivity is unacceptable with unborated water in the storage pool. Thus, for these accident occurrences, the presence of soluble boron in the storage pool prevents criticality in both regions. The postulated accidents are basically of two types. A fuel assembly could be incorrectly transferred from Region 1 to Region 2 (e.g., an unirradiated fuel assembly or an insufficiently depleted fuel assembly). The second type of postulated accidents is associated with a fuel assembly which is dropped adjacent to the fully loaded storage rack. This could have a small positive reactivity effect in the region. However, the negative reactivity effect of the soluble boron compensates for the increased reactivity caused by either one of the two postulated accident scenarios. The accident analyses is described in References 2 and 3.

The concentration of dissolved boron in the spent fuel pit satisfies Criterion 2 of 10 CFR 50.36.

LCO The spent fuel pit boron concentration is required to be i:: 1000 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 3. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the spent fuel pit until a spent fuel pit verification confirms that there are no misloaded fuel assemblies. With no misloaded fuel assemblies and unborated water, the spent fuel pit design is sufficient to maintain the core at Kett s 0.95.

Indian Point 3 B 3.7.15-2 Revision

Spent Fuel Pit Boron Concentration B 3.7.15 BASES LCO (continued)

The LCO is modified by a Note that states that during inter-unit transfer of fuel the spent fuel pit boron concentration must also meet Appendix C LCO 3.1.1, "Boron Concentration." This requirement ensures that the criticality analysis of the fuel within the Shielded Transfer Canister remains bounding.

APPLICABILITY This LCO applies whenever fuel assemblies are stored in the spent fuel pit, until a complete spent fuel pit verification has been performed following the last movement of fuel assemblies in the spent fuel pit. This LCO does not apply following the verification, since the verification would confirm that there are no misloaded fuel assemblies. With no further fuel assembly movements in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly.

ACTIONS A.1. A.2.1 and A.2.2 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. Alternatively, beginning a verification of the spent fuel pit fuel locations, to ensure proper locations of the fuel, can be performed. However, prior to resuming movement of fuel assemblies, the concentration of boron must be restored. This does not preclude movement of a fuel assembly to a safe position.

SURVEILLANCE SR 3.7.15.1 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 31 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. This SR is not required to be met or performed if a spent fuel pit verification for conformance with LCO 3.7.16, Figures 3.7.16-1 and B 3.7.16-1, has been performed on all fuel assemblies since the last verification following the last movement of fuel assemblies in the spent fuel pit.

Indian Point 3 83.7.15-3 Revision

Spent Fuel Pit Boron Concentration B 3.7.15 BASES REFERENCES 1. Double contingency principle of ANSI N 16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

2. SER related to Amendment 173 to Facility Operating License No.

DPR-64, Indian Point Nuclear Generating Unit No. 3, April 15, 1997.

3. Criticality Analysis of the Indian Point 3 Fresh .and Spent Fuel Racks, Westinghouse Commercial Nuclear Fuel Division, October, 1996.

Indian Point 3 B 3.7.15-4 Revision

Spent Fuel Assembly Storage B 3.7.16 B 3.7 SPENT FUEL PIT REQUIREMENTS 83.7.16 Spent Fuel Assembly Storage BASES BACKGROUND In the Maximum Density Rack (MDR) design, the spent fuel pit (SFP) is divided into two separate and distinct regions. The layout of the IP3 MOR is shown in Figure B 3.7.16-1, IP3 Maximum Density Spent Fuel Pit Racks, Regions and Indexing. As shown in Figure B 3.7.16-1, Region 1 (i.e., Columns SS-ZZ, Rows 35-64) includes 240 storage positions and Region 2 (i.e., Columns A-RR, Rows 1-34) includes 1105 storage positions. Region 1 is analyzed for storage of high-enrichment and low-burnup fuel. Region 2 is analyzed for storage of fuel with either higher burnup or lower enrichment. Each region has been separately analyzed tor close packed storage when all cells in that region contain fuel of the highest reactivity that is allowed by this LCO. This analysis is the basis for the restrictions on fuel storage locations established by this LCO.

Prior to storage in the spent fuel pit, fuel assemblies are classified as to the level of reactivity based on the initial enrichment and burnup. This classification is made using Figure 3.7.16-1, "Fuel Assembly Classification for Storage in the Spent Fuel Pit." This classification is used to determine in which region a particular fuel assembly may be stored and if additional restrictions must be applied to the assemblies in adjacent locations. Figure 3. 7.16-1, "Fuel Assembly Classification for Storage in the Spent Fuel Pit," is used to classify each assembly into one of the following categories based on initial U-235 enrichment and burnup:

Type 2 assemblies are the least reactive assemblies and include any assembly for which the combination of initial enrichment and burnup places the assembly in the domain labeled Type 2 in Figure 3.7.16-1.

Type 2 assemblies may be stored in any location in Region 1 or Region 2 of Figure B 3.7.16-1.

Type 1A assemblies are more reactive than Type 2 assemblies and include any assembly for which the combination of initial enrichment and burnup places the assembly in the domain labeled Indian Point 3 B 3.7.16-1 Revision

Spent Fuel Assembly Storage 8 3.7.16 BASES BACKGROUND (continued)

Type 1A in Figure 3.7.16-1. Type 1A assemblies must be stored in Region 1 of Figure B 3.7.16-1 but may be stored in any location in Region 1.

Type 1B assemblies are more reactive than Type 1A assemblies and include any assembly with an initial enrichment> 4.2 buts 4.6 wt%

U-235 with a burnup that places the assembly in the domain labeled Type 18 in Figure 3.7.16-1. Type 18 assemblies must be stored in Region 1 of Figure B 3.7.16-1 but may be stored in any location in Region 1 except in locations that are face-adjacent to a Type 1C assembly.

Type 1C assemblies are the most reactive bundles permitted in accordance with Specification 4.3, Fuel Storage. Type 1C assemblies include any assembly with an initial enrichment > 4.6 buts 5.0 wt%

U-235 with a burnup that places the assembly in the domain labeled Type 1C on Figure 3.7.16-1. Type 1C assemblies must be stored in Region 1 of Figure B 3.7.16-1. Type 1C assemblies cannot be stored in Row 64 or in Column ZZ. Additionally, Type 1C assemblies must be stored in a location where all face-adjacent locations are as follows:

a) occupied by Type 2 or Type 1A assemblies; b) occupied non-fuel components; or, c) empty.

Fuel assemblies with an initial enrichment > 5.0 wt% U-235 are not shown on Figure 3.7.16-1 and cannot be stored in the spent fuel pit in accordance with paragraph 4.3.1.1 in Section 4.3, Fuel Storage.

The water in the spent fuel pit normally contains soluble boron, which results in large subcriticality margins under actual operating conditions.

However, the NRG guidelines, based upon the accident condition in which all soluble poison is assumed to have been lost, specify that the limiting kett of 0.95 be evaluated in the absence of soluble boron. Hence, the design of both regions is based on the use of unborated water, which maintains each region in a subcritical condition during normal operation with the regions fully loaded and fuel storage locations, enrichment and burnup are in conformance with analysis assumptions and this LCO. The double contingency principle discussed in ANSI N-16.1-1975 and the April 1978 NRG letter (Ref. 1) allows credit for soluble boron under other abnormal or accident conditions because only a single accident need be considered at one time. For example, the accident scenarios include movement of a Type 1C fuel assembly from Region 1 to Region 2, or accidental misleading of a fuel assembly in Region 1. These events Indian Point 3 B 3.7.16-2 Revision

Spent Fuel Assembly Storage B 3.7.16 BASES BACKGROUND (continued) could increase the potential for criticality in the spent fuel pit. To mitigate these postulated criticality related accidents, boron concentration is verified to be within the limits specified in LCO 3.7.15, Spent Fuel Pit Boron Concentration, prior to movement of any fuel assembly. Safe operation of the SFP with no movement of assemblies is achieved by controlling the location of each assembly in accordance with the accompanying LCO. However, prior to movement of an assembly, it is necessary to perform SR 3.7.15.1 (i.e., verification that the spent fuel pit boron concentration is within limit).

APPLICABLE The restrictions on the placement of fuel assemblies within the spent fuel SAFETY pit are based on initial enrichment and burnup which is indicative of fuel ANALYSES assembly reactivity. Storage locations are then restricted to ensure the kett of the spent fuel pit will always remain< 0.95, assuming the pool to be flooded with unborated water. Fuel assemblies not meeting the criteria of Figure 3. 7 .16-1 may not be stored in accordance with Specification 4.3.1.1 in Section 4.3.

The hypothetical accidents can only take place during or as a result of the movement of an assembly (References 2 and 3). For these accident occurrences, the presence of soluble boron in the spent fuel storage pit (controlled by LCO 3.7.15, "Spent Fuel Pit Boron Concentration")

prevents criticality in both regions. By closely controlling the movement of each assembly and by checking the location of *each assembly after movement, the time period for potential accidents may be limited to a small fraction of the total operating time. During the remaining time period with no potential for accidents, the operation may be under the auspices of the accompanying LCO.

The configuration of fuel assemblies in the fuel storage pit satisfies Criterion 2 of 10 CFR 50.36.

LCO Fuel assemblies stored in the spent fuel pit are classified in accordance with Figure 3.7.16-1 based on initial enrichment and burnup which is indicative of fuel assembly reactivity. Based on this classification, fuel assembly storage location within the spent fuel pit and storage location relative to other assemblies is restricted in accordance with the rules established by this LCO.

Fuel assemblies with an initial enrichment > 5.0 wt% U-235 are not shown on Figure 3.7.16-1 because fuel assemblies with this enrichment cannot be stored in the spent fuel pit in accordance with limits established in Technical Specification Section 4.3.

Indian Point 3 83.7.16-3 Revision

Spent Fuel Assembly Storage B 3.7.16 BASES APPLICABILITY This LCO applies whenever any fuel assembly is stored in the spent fuel pit.

ACTIONS When the configuration of fuel assemblies stored in the spent fuel pit is not in accordance with this LCO, the immediate action is to initiate action to make the necessary fuel assembly movement(s) to bring the configuration into compliance with this LCO.

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies by administrative means that the initial enrichment and burnup of the fuel assembly in each location is in accordance with the accompanying LCO.

REFERENCES 1. Double contingency principle of ANSI N 16.1-1975, as specified in the April 14, 1978 NRG letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

2. SER related to Amendment 173 to Facility Operating License No.

DPR-64, Indian Point Nuclear Generating Unit No. 3, April 15, 1997.

3. Criticality Analysis of the Indian Point 3 Fresh and Spent Fuel Racks, Westinghouse Commercial Nuclear Fuel Division, October, 1996.

Indian Point 3 B 3.7.16-4 Revision

Spent Fuel Assembly Storage 83.7.16

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Hax11111 Density Spent Fuel Pit (SFP)

Racks, Regions and Indexing Indian Point 3 B 3.7.16-5 Revision

v t e Letter Sequence Request
EPID:L-2020-LLR-0045, TSTF-563, TSTF-448, TSTF-529, TSTF-4, TSTF-563 (Approved, Closed)
Initiation Request, Request, Request, Request, Request, Request Acceptance Administration Questions Answers 30 March 2020 Results Approval Other: ML20091M199, ML20344A419, ML20345A305
MONTHYEAR2020-03-31 [Table View]

NL-20-033, Technical Specifications Proposed Change - Permanently Defueled Technical Specifications

Text

Subject:

References:

Enclosure:

Enclosure:

SUMMARY

3.4 CONCLUSION

1.0 DESCRIPTION

1.0 DESCRIPTION

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