ML21083A000

Issuance of Amendment No. 63 Permanently Defueled Technical Specifications

ML21083A000
Person / Time
Site:	Indian Point
Issue date:	04/14/2021
From:	Richard Guzman NRC/NRR/DORL/LPL1
To:	Entergy Nuclear Operations
Guzman R
References
EPID L-2020-LLA-0146
Download: ML21083A000 (43)

Text

April 14, 2021 Vice President, Operations Entergy Nuclear Operations, Inc.

Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249

SUBJECT:

INDIAN POINT NUCLEAR GENERATING STATION, UNIT NO. 1 ISSUANCE OF AMENDMENT NO. 63 RE: PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS (EPID L-2020-LLA-0146)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 63 to Provisional Operating License No. DPR-5 for Indian Point Nuclear Generating Station, Unit No. 1 (Indian Point, Unit 1). The amendment consists of changes to the Provisional Operating License and the associated Technical Specifications (TSs) in response to your application dated June 30, 2020.

The amendment revises the Indian Point, Unit 1 Provisional Operating License and the TSs in Appendix A to reflect the current conditions at Indian Point, Unit 1 and the permanent cessation of power operations at Indian Point Nuclear Generating Unit No. 2, and to denote that certain Indian Point, Unit 1 systems also support Indian Point Nuclear Generating Unit No. 3.

A copy of the related Safety Evaluation is enclosed. Notice of Issuance will be included in the Commissions monthly Federal Register notice.

Sincerely,

/RA/

Richard V. Guzman, Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No.50-003

Enclosures:

1. Amendment No. 63 to DPR-5

2. Safety Evaluation cc Listserv

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

DOCKET NO.50-003 INDIAN POINT NUCLEAR GENERATING UNIT NO. 1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 63 License No. DPR-5

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

A. The application for amendment by Entergy Nuclear Operations, Inc. (the licensee) dated June 30, 2020, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate 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 amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Provisional Operating License No. DPR-5 is hereby amended to read as follows:

The title PROVISIONAL OPERATING LICENSE is to read PROVISIONAL LICENSE Paragraphs A and B are to read as follows:

A. The application for amendment by Entergy Nuclear Operations, Inc. (the licensee), dated May 30, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter 1; B. The facility will be maintained in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; Paragraphs 1 and 2 are to read as follows:

Accordingly, License No. DPR-5 is hereby amended as indicated in the attachment to this license amendment, and Provisional License No. DPR-5 is hereby amended to read as follows:

1. This license applies to the utilization facility consisting of a pressurized water reactor (the facility) and associated components and equipment which is owned by ENIP2. The facility is located in Westchester County, New York, and described in the Defueled Safety Analysis Report, as supplemented and amended, and is designated by ENIP2 as the Indian Point Station Unit No. 1.

2. Subject to the conditions and requirements incorporated herein, the U.S. Nuclear Regulatory Commission (hereinafter referred to as the Commission) hereby licenses:

a) ENIP2 and ENO, pursuant to Section 104b. of the Act and Title 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess but not operate the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations described in the application and this license; b) ENO, pursuant to the Act and 10 CFR Part 70, to possess up to 1918 kilograms of contained uranium-235 previously received for reactor operation; e) ENO, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special

materials as were produced by the prior operation of the facility; Paragraphs 3.b) and 3.c) are to read as follows:

b) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 63, are hereby incorporated in the license. ENO shall maintain the facility in accordance with the Technical Specifications.

ENIP2 and ENO, pursuant to Section 104b. of the Act and Title 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess but not operate the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations described in the application and this license; c) Deleted Paragraphs 4 and 5 are to read as follows:

4. Deleted

5. Deleted Paragraphs 9 and 10 are to read as follows:

9. The approved Decommissioning Plan supplements the Defueled Safety Analysis Report (DSAR) and the licensee may (i) make changes in the facility or procedures as described in the DSAR or the Decommissioning Plan and (ii) conduct tests, or experiments not described in the DSAR or Decommissioning Plan, without prior Commission approval, provided the requirements of 10 CFR 50.59 and 10 CFR 50.82(a)(6) and (7) are satisfied.

10. The amended license is effective as of the date of issuance and until the Commission notifies the licensee in writing that the license is terminated.

3. This license amendment is effective as of the date of its issuance and shall be implemented within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Bruce A. Bruce A. Watson Date: 2021.04.14 Watson 11:00:33 -04'00' Bruce A. Watson, CHP, Chief Reactor Decommissioning Branch Division of Decommissioning, Uranium Recovery and Waste Program Office of Nuclear Material Safety and Safeguards

Attachment:

Changes to the Provisional Operating License and Technical Specifications Date of Issuance: April 14, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 63 INDIAN POINT NUCLEAR GENERATING UNIT NO. 1 PROVISIONAL OPERATING LICENSE NO. DPR-5 DOCKET NO.50-003 Replace the following pages of the Provisional Operating License and Appendix A to the Provisional Operating License with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Provisional Operating License No. DPR-05 REMOVE INSERT through through Appendix A to Provisional Operating License No. DPR-05 REMOVE INSERT Title Page Title Page Table of Contents Table of Contents Page 1 through Page 12 Page 1 through Page 3

ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO.50-003 INDIAN POINT NUCLEAR GENERATING STATION, UNIT NO. 1 AMENDMENT TO PROVISIONAL LICENSE Amendment No. 63 License No. DPR-5 The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Entergy Nuclear Operations, Inc. (the licensee), dated May 30, 2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions rules and regulations set forth in 10 CFR Chapter 1; 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 amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commissions regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commissions regulations, and all applicable requirements have been satisfied.

Accordingly, License No. DPR-5 is hereby amended as indicated in the attachment to this license amendment, and Provisional License No. DPR-5 is hereby amended to read as follows:

1. This license applies to the utilization facility consisting of a pressurized water reactor (the facility) and associated components and equipment, which is owned by ENIP2. The facility is located in Westchester County, New York, and described in the Defueled Safety Analysis Report, as supplemented and amended, and which is designated by ENIP2 as the Indian Point Station Unit No. 1.

2. Subject to the conditions and requirements incorporated herein, the U.S. Nuclear Regulatory Commission (hereinafter referred to as "the Commission") hereby licenses:

a) ENIP2 and ENO, pursuant to Section 104b. of the Act and Title 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess but not operate the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations described in the application and this license; Amendment No. 63

b) ENO, pursuant to the Act and 10 CFR Part 70, to possess up to 1918 kilograms of contained uranium-235 previously received for reactor operation; c) Deleted; d) Deleted; e) ENO, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special materials as were produced by the prior operation of the facility; f) Deleted.

3. This license shall be deemed to contain and is subject to the conditions specified in Sections 50.54 and 50.59 of Part 50, Section 70.32 of Part 70, Section 40.41 of Part 40, and Section 30.32 of Part 30 of the Commission's regulations; is subject to all applicable provisions of the Act and rules, regulations and orders of the Commission now and hereafter in effect; and is subject to the additional conditions specified below:

a) Maximum Power Level ENO is prohibited from taking the reactor to criticality, and the facility shall not be operated at any power level.

b) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 63, are hereby incorporated in the license. ENO shall maintain the facility in accordance with the Technical Specifications.

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

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. 55, as supplemented by changes approved by License Amendment Nos. 57, 59, and 60.

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

Amendment No. 63

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

4. Deleted

5. Deleted

6. On the closing date of the transfer of the license, Con Edison shall transfer to ENIP2 all of the accumulated decommissioning trust funds for Indian Point Nuclear Generating Unit No.1 (IP1) and such additional funds to be deposited in the decommissioning trusts for IP1 such that the total amount transferred for IP1 and Indian Point Nuclear Generating Unit No. 2 (IP2) is no less than $430,000,000. Furthermore, ENIP2 shall either (a) establish a provisional trust for decommissioning funding assurance for IP1 and IP2 in an amount no less than $25,000,000 (to be updated as required under applicable NRC regulations, unless otherwise approved by the NRC) or (b) obtain a surety bond for an amount no less than $25,000,000 (to be updated as required under applicable NRC regulations, unless otherwise approved by the NRC). The total decommissioning funding assurance provided for IP1 by the combination of the decommissioning trust and the provisional trust or surety bond at the time of transfer of the licenses shall be at a level no less than the amounts calculated pursuant to, and required under, 10 CFR 50.75. The provisional trust and surety bond shall be subject to or be consistent with the following requirements, as applicable:

a) Deleted b) Provisional Trust (i) The provisional trust agreement must be in a form acceptable to the NRC.

(ii) Investments in the securities or other obligations of Entergy Corporation or its affiliates, subsidiaries, successors, or assigns are and shall be prohibited.

Except for investments tied to market indexes or other non-nuclear-sector mutual funds, investments in any entity owning one or more nuclear power plants are and shall be prohibited.

(iii) The provisional trust agreement must provide that no disbursements or payments from the trust, other than for ordinary administrative expenses, shall be made by the trustee unless the trustee has first given the Director of the Office of Nuclear Reactor Regulation 30 days prior written notice of payment.

The provisional trust agreement shall further contain a provision that no disbursements or payments from the trust shall be made if the trustee receives prior written notice of objection from the NRC.

(iv) The provisional trust agreement must provide that the agreement cannot be amended in any material respect, or terminated, without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

Amendment No. 63

(v) The appropriate section of the provisional trust agreement shall state that the trustee, investment advisor, or anyone else directing the investments made in the trust shall adhere to a "prudent investor" standard, as specified in 18 CFR 35.32(a)(3) of the Federal Energy Regulatory Commission's regulations.

(vi) Use of assets in the provisional trust, in the first instance, shall be limited to the expenses related to decommissioning IP1 or IP2 as defined by the NRC in its regulations and issuances, and as provided in this license and any amendments thereto.

(c) Surety Bond (i) The surety bond agreement must be in a form acceptable to the NRC and be in accordance with all applicable NRC regulations.

(ii) The surety company providing any surety bond obtained to comply with the requirements of the Order approving the transfer shall be one of those listed by the U.S. Department of the Treasury in the most recent edition of Circular 570 and shall have a coverage limit sufficient to cover the amount of the surety bond.

(iii) ENIP2 shall establish a standby trust to receive funds from the surety bond, if a surety bond is obtained, in the event that ENIP2 defaults on its funding obligations for the decommissioning of IP1. The standby trust agreement must be in a form acceptable to the NRC, and shall conform with all conditions otherwise applicable to the decommissioning trust agreement, and with all conditions that would be applicable to the provisional trust above, if established.

(iv) The surety agreement must provide that the agreement cannot be amended in any material respect, or terminated, without 30 days prior written notification to the Director of the Office of Nuclear Reactor Regulation.

7. Deleted

8. ENIP2 and ENO shall take no action to cause Entergy Global Investments, Inc., or Entergy International Ltd. LLC or their parent companies to void, cancel, or modify the

$55 million contingency commitment to provide funding for the IP1 and IP2 plants as represented in the application without the prior written consent of the Director of the Office of Nuclear Reactor Regulation.

9. The approved Decommissioning Plan supplements the Defueled Safety Analysis Report (DSAR) and the licensee may (i) make changes in the facility or procedures as described in the DSAR or the Decommissioning Plan and (ii) conduct tests, or experiments not described in the DSAR or Decommissioning Plan, without prior Commission approval, provided the requirements of 10 CFR 50.59 and 10 CFR 50.82(a)(6) and (7) are satisfied.

Amendment No. 63

10. The amended 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 ATOMIC ENERGY COMMISSION Original signed by E. G. Case R. L. Doan, Director Division of Reactor Licensing Date of Issuance: October 29, 1965 Amendment No. 63

Appendix A to Provisional License DPR-5 for Entergy Nuclear Indian Point 2, LLC and Entergy Nuclear Operations, Inc.

Indian Point Station Unit No. 1 Docket No. 50-3 Amendment No. 63

PERMANENTLY DEFUELED TECHNICAL SPECFICATIONS TABLE OF CONTENTS Section Title Page 1.0 General Information 1 1.1 Definitions 1 1.2 Exclusion Distance 2 2.0 Reactor Facility Design Performance Requirements 2 2.1 Deleted 2 2.2 Fuel Storage 2 3.0 Administrative and Procedural Safeguards 2 3.1 Responsibility 2 3.2 Organization 2 3.3 Operating Instructions and Procedures 2 4.0 Operating Limitations 2 4.1 Deleted 2 4.2 Release of Radioactive Liquids and Gases 2 4.3 Deleted 3 4.4 Deleted 3 4.5 Radiological Environmental Monitoring 3 5.0 Maintenance 3 5.1 General 3 5.2 Testing 3 6.0 Plant Reporting Requirements 3 Amendment No. 63

Appendix A to Provisional License DPR-5 For the Entergy Nuclear Indian Point 2, LLC and Entergy Nuclear Operations, Inc.

1.0 GENERAL INFORMATION The facility, known as the Indian Point Station Unit No. 1, is located on a site in the Village of Buchanan, Westchester County, New York. The Indian Point Station Unit No. 2 and the Indian Point Station Unit No. 3 share this site.

Indian Point Unit No. 1 includes a pressurized water reactor, which operated with an authorized maximum steady state power level of 615 thermal megawatts until October 31, 1974. Pursuant to the June 19, 1980 Commission Order Revoking Authority to Operate Facility and the Decommissioning Plan for Indian Point Unit No. 1 approved by the NRC in an Order dated January 31, 1996, the reactor remains in a defueled status and some of the facilities structures, systems, and components support activities at Indian Point Units 1, 2, and 3. Unit No. 1 and Unit No. 2 are physically contiguous and share a number of systems and facilities as well as a common organization. The technical specifications contained herein recognize this commonality as well as the intended use of the Unit No. 1 facilities to support Unit Nos. 2 and 3 until the fuel for these units is transferred from the spent fuel pits to an offsite storage facility, and contain specific references to Appendix A to the Indian Point Unit No. 2 Facility License No. DPR-26. Unit No. 1 contains radioactive waste processing facilities, which provide waste processing services for Unit Nos. 1, 2, and 3. Radiological effluent limits are met on an overall site basis and specific operating limits and surveillance requirements for effluent monitoring instrumentation, including stack noble gas monitoring, are discussed in the Offsite Dose Calculation Manual.

1.1 Definitions Offsite Dose Calculation Manual (ODCM)

The Offsite Dose Calculation Manual contains the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the environmental radiological monitoring program. Requirements for the ODCM are specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

Amendment No. 63 Page 1

1.2 Exclusion Distance 1.2.1 The minimum distance from the reactor facility to the nearest land boundary of the exclusion area, as defined in 10 CFR 100 shall be 1400 feet.

1.2.2 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 460 meters and 1100 meters, respectively.

2.0 REACTOR FACILITY DESIGN PERFORMANCE REQUIREMENTS 2.1 Deleted 2.2 Fuel Storage No fuel shall be stored in the Unit No. 1 spent fuel storage area.

3.0 ADMINISTRATIVE AND PROCEDURAL SAFEGUARDS 3.1 Responsibility Responsibilities are as specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

3.2 Organization The organization requirements are as specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

The Shift Manager is responsible for the Unit No. 1 facility.

3.3 Operating Instructions and Procedures 3.3.1 No fuel will be loaded into the reactor core or moved into the reactor containment building without prior review and authorization by the Nuclear Regulatory Commission.

3.3.2 Detailed written instruction setting forth procedures used in connection with the facility shall conform to the requirements specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

4.0 OPERATING LIMITATIONS 4.1 Deleted 4.2 Release of Radioactive Liquids and Gases The concentration of radioactive materials released in liquid or gaseous form to unrestricted areas shall not exceed the limits specified in 10 CFR Part 20. Release of radioactive liquids and gases shall also be consistent with the requirements of 10 CFR Part 50, Appendix I, as specified in the ODCM.

Amendment No. 63 Page 2

4.3 Deleted 4.4 Deleted 4.5 Radiological Environmental Monitoring The Indian Point site Radiological Environmental Monitoring Program shall be conducted as specified in the ODCM.

5.0 MAINTENANCE 5.1 General Components addressed in these technical specification requirements, which have been repaired, replaced, or otherwise subjected to temporary or permanent modification, shall be tested in accordance with procedures, which are appropriate in view of the nature of the repair, replacement, or modification, and the condition of the system.

5.2 Testing Unit 1 radioactive effluent monitoring instrumentation shall satisfy the surveillance requirements as specified in the ODCM.

6.0 PLANT REPORTING REQUIREMENTS Reporting Requirements are as specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

Amendment No. 63 Page 3

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 63 PROVISIONAL OPERATING LICENSE NO. DPR-5 ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING STATION, UNIT NO. 1 DOCKET NO.50-005

1.0 INTRODUCTION

By application dated June 30, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20182A679), Entergy Nuclear Operations, Inc. (Entergy or the licensee) requested changes to Provisional Operating License (POL) No. DPR-5 and the associated Technical Specifications (TSs) for Indian Point Nuclear Generating Station, Unit No. 1 (Indian Point Unit 1 or IP1). Entergy requested an amendment to revise the IP1 POL and the TSs in Appendix A to reflect the current conditions at IP1 and the permanent cessation of power operations at Indian Point Nuclear Generating Unit No. 2 (Indian Point Unit 2 or IP2),

and to denote that certain IP1 systems also support Indian Point Nuclear Generating Unit No. 3 (Indian Point Unit 3 or IP3).

Specifically, the revised IP1 POL would be renamed the Provisional License and the IP1 TSs in Appendix A would be renamed the IP1 Permanently Defueled Technical Specifications (PDTS). The proposed changes would revise certain requirements contained within the IP1 POL and TSs to align with those required for the PDTS, to reflect decommissioning requirements, and remove the requirements that are no longer applicable.

2.0 BACKGROUND

Indian Point Unit 1 includes a pressurized-water reactor, which operated with an authorized maximum steady state power level of 615 thermal megawatts until October 31, 1974. Pursuant to the June 19, 1980 Commission Order Revoking Authority to Operate Facility and the Decommissioning Plan for Indian Point Unit No. 1 approved by the NRC in an Order dated January 31, 1996, the reactor remains in a SAFSTOR defueled status and some of the facilities structures, systems, and components (SSCs) support activities at Indian Point Units 1, 2, and 3.

The IP1 and IP2 facilities are physically connected with shared systems and facilities such as the integrated liquid waste system, the air system, and the chemistry and health physics laboratories. IP1 contains radioactive waste processing facilities, which provide waste processing services for Indian Point Units 1, 2, and 3. Radiological effluent limits are met on an overall site basis. The specific operating limits and surveillance requirements for effluent monitoring instrumentation are discussed in the Offsite Dose Calculation Manual (ODCM). The Enclosure 2

IP1 TSs in Appendix A recognize these commonalities as well as the intended use of the IP1 facilities to support IP2 and IP3 until the fuel for these units is transferred from the spent fuel pits to an offsite storage facility, and contain specific references to IP2 Facility License No. DPR-26.

License Condition 2.a) of the IP1 POL no longer authorizes operation of the reactor, and IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved to the reactor containment building without prior review and authorization by the NRC. All of the IP1 spent fuel1 has been transferred from the IP1 spent fuel storage pool to the Indian Point Energy Center (IPEC) Independent Spent Fuel Storage Installation (ISFSI).

By letter dated February 8, 2017 (ADAMS Accession No. ML17044A004), the licensee submitted a Notification of Permanent Cessation of Power Operations for IP2 and IP3. In this letter, Entergy provided notification to the NRC of its intent to permanently cease power operations at IP2 and IP3 no later than April 30, 2020, and April 30, 2021, respectively, subject to operating extensions through, but not beyond, April 2024 and April 2025, respectively.

By letter dated April 28, 2020 (ADAMS Accession No. ML20081J402), the NRC issued Amendment No. 294 to the IP2 operating license, which revised the IP2 Renewed Facility Operating License and converted the associated TSs to PDTS, consistent with the permanent cessation of operations and permanent removal of fuel from the IP2 reactor vessel.2 By letter dated May 12, 2020 (ADAMS Accession No. ML20133J902), Entergy certified in accordance with Sections 50.82(a)(1)(i) and (ii) of Title 10 of the Code of Federal Regulations (10 CFR) that power operations ceased at IP2 on April 30, 2020, and that the fuel was permanently removed from the IP2 reactor vessel and placed in the IP2 spent fuel pool on May 12, 2020. With the docketing of the IP2 certifications of permanent cessation of power operations and permanent removal of fuel from the IP2 reactor vessel, the 10 CFR Part 50 license for IP2 no longer permits operation of the reactor or emplacement or retention of fuel in the reactor vessel. The IP2 PDTS were implemented on June 1, 2020.

By application dated April 28, 2020 (ADAMS Accession No. ML20132A200), Entergy submitted a license amendment request to the NRC to revise the IP3 Renewed Facility Operating License No. DPR-64 and to convert the associated TSs to PDTS, consistent with the permanent cessation of operations and permanent removal of fuel from the IP3 reactor vessel. As indicated in the application, IP3 will continue to operate until its planned permanent cessation of operations on April 30, 2021. The licensee also indicated that while the IP1 Appendix A TSs will be revised to denote that IP1 systems are used to support IP3, the continued operation of IP3 and planned permanent cessation of operations on April 30, 2021, do not impact the implementation of the IP1 PDTS.

1 All of the IP1 spent fuel refers to onsite fuel since some spent fuel was shipped offsite.

2 Amendment No. 294 was corrected by letter dated May 4, 2020 (ADAMS Accession No. ML20122A262).

3.0 REGULATORY EVALUATION

3.1 Technical Specifications Section 182a of the Atomic Energy Act of 1954, as amended, requires applicants for nuclear power plant operating licenses to include TSs as part of the application. The NRCs regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36, Technical specifications. Pursuant to 10 CFR 50.36, each operating license issued by the Commission includes TSs and includes items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs);

(3) surveillance requirements (SRs); (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports.

10 CFR 50.36(c)(1) states, in part, that safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. . . .

Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions.

The regulation at 10 CFR 50.36(c)(2) states, in part, Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. When a [LCO] of a nuclear reactor is not met, the licensee shall shut down the reactor or follow any remedial action permitted by the [TSs] until the condition can be met. The regulation provides four criteria to define the scope of equipment and parameters to be included in the TS LCOs. These criteria were developed for licenses authorizing operation (i.e.,

operating reactors) and focused on instrumentation to detect degradation of the reactor coolant system pressure boundary, process variables and equipment, design features, or operating restrictions that affect the integrity of fission product barriers during design-basis accidents (DBAs) or transients. A fourth criterion refers to the use of operating experience and probabilistic risk assessment to identify and include in the TS those SSCs shown to be significant to public health and safety.

The regulation at 10 CFR 50.36(c)(3) states, in part, Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met.

The regulation at 10 CFR 50.36(c)(4) states, in part, Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c) (1), (2), and (3) of this section.

The regulation at 10 CFR 50.36(c)(5) states, in part, Administrative controls 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.

The regulations in 10 CFR 50.36(c)(6) further state that TSs involving safety limits, limiting safety system settings, and limiting control system settings; LCOs; SRs; design features; and administrative controls for decommissioning facilities will be developed on a case-by-case basis.

A general discussion of the criteria that were used by the NRC staff in its evaluation to ensure that TS LCOs proposed for deletion are no longer required to be included in TSs is provided below. These criteria were also used in the evaluation of the proposed changes to existing TSs and the proposed new TSs.

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 is present in the reactor or reactor coolant system at the IP1 facility in the permanently shutdown and defueled condition, this criterion is not applicable.

Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) states that TS LCOs must be established for a process variable, design feature, or operating restriction that is an initial condition of a DBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The purpose of this criterion is to capture those process variables that have initial values assumed in the DBA and transient analyses, and which are monitored and controlled during power operation. The scope of DBAs applicable to a reactor permanently shut down and defueled is reduced from those postulated for an operating reactor. Because the IP1 facility is permanently shut down and defueled, with its spent fuel stored in the IPEC ISFSI, where operation is controlled in accordance with the applicable TSs for the storage system, there are no IP1 DBAs associated with IP1 that could impact the integrity of a fission product barrier. Therefore, this criterion is not applicable.

Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) states that TS LCOs must be established for an SSC that is part of the primary success path and which functions or actuates to mitigate a design basis accident 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 TSs 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 IP1 facility is permanently shut down and defueled, with its spent fuel stored in the IPEC ISFSI, where the operation of the ISFSI is controlled in accordance with the applicable TSs for the storage system. There are no IP1 DBAs associated with IP1 that could impact the integrity of a fission product barrier. Therefore, this criterion is not applicable.

Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) states that TS LCOs must be established for SSCs which 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. The IP1 facility has been maintained in the SAFSTOR condition for a considerable period, and the spent fuel has been transferred from the IP1 spent fuel storage pool to the IPEC ISFSI. Given the limited operating systems remaining at IP1 and reduced radioactive source term at IP1, there is limited potential impact due to a possible radiological release resulting from an event at IP1. Consequently, there are no events associated with the IP1 SSCs that remain in use that are considered significant to public health and safety from a risk perspective. Therefore, this criterion is not applicable.

The NRC staff notes that information contained in Draft NUREG-1625, Proposed Standard Technical Specifications for Permanently Defueled Westinghouse Plants, dated March 1998 (ADAMS Accession No. ML082330233), was also considered in its evaluation. Specifically, the draft NUREG provides examples of TSs that the staff found acceptable during previous TS

reviews for permanently shutdown and defueled reactors. The NRC staff also considered information contained in NUREG-1431, Revision 4, Standard Technical Specifications - Westinghouse Plants, dated April 2012 (ADAMS Accession No. ML12100A222).

3.2 Radiological Consequences from Design-Basis Accidents Radiological accidents considered in licensing nuclear power plants are classified as DBAs and severe (beyond-design-basis) accidents. DBAs are those accidents that both the licensee and the NRC staff evaluate to ensure that the plant can withstand normal and abnormal transients and a broad spectrum of postulated accidents without undue hazard to the health and safety of the public. Severe accidents are those that are beyond the design basis of the plant. They are more severe than DBAs because they may result in substantial damage to the fuel, whether or not there are serious offsite consequences. For the most part, DBAs focus on reactor operation and are not applicable to plants undergoing decommissioning. The only DBAs or severe accidents applicable to a decommissioning plant are typically those involving the spent fuel pool. These postulated accidents are not expected to occur during the life of the plant, but are evaluated to establish the design basis for the preventive and mitigative safety systems of the spent fuel storage facility.

Regulations governing accidents that must be addressed by nuclear power facilities, both operating and shutdown, are found in 10 CFR Part 50 and 10 CFR Part 100. The environmental impacts of DBAs, including those associated with the spent fuel pool, are evaluated during the initial licensing process. The ability of the plant to withstand these accidents is demonstrated to be acceptable before issuance of the operating license. The results of these evaluations are found in license documentation, such as the NRC staffs safety evaluation report, the final environmental statement, and in the licensees Updated Final Safety Analysis Report or equivalent. The consequences for these events are evaluated for the hypothetical maximally exposed individual. The licensee is required to maintain the acceptable design and performance criteria throughout the life of the plant.

As indicated above in Section 3.1 of this safety evaluation (SE), the IP1 facility is permanently shut down and defueled, and the IP1 spent fuel has been transferred from the IP1 spent fuel storage pool to the IPEC ISFSI. Additionally, the IP1 spent fuel storage pool has been drained.

There are limited operating systems remaining at IP1 and only those areas that either store or process radioactive materials (e.g., the Fuel Handling Building and waste storage/process areas in the Chemical Systems Building and the Integrated Liquid Radwaste Systems Building) need to be considered in evaluating the radiological hazards for IP1. The limited operating systems combined with the reduced radioactive source term result in a limited potential impact due to a possible radiological release resulting from an event at IP1. Due to its current configuration and current licensing basis, there are no postulated DBAs that remain applicable to IP1.

3.3 Fire Protection During Decommissioning The regulation at 10 CFR 50.48(a)(1) requires each holder of an operating license and holders of a combined operating license issued under Part 52 to have a fire protection program (FPP)

that satisfies General Design Criterion 3 of Appendix A to 10 CFR Part 50 and states that the FPP must:

Describe the overall FPP for the facility; Identify the various positions within the licensees organization that are responsible for the program; State the authorities that are delegated to each of these positions to implement those responsibilities; and Outline the plans for fire protection, fire detection and suppression capability, and limitation of fire damage.

For permanently shutdown reactors, 10 CFR 50.48(f) governs fire protection for plants licensed under 10 CFR Part 50 and 10 CFR Part 52. The regulation at 10 CFR 50.48(f) requires licensees that have certified the permanent cessation of operations and the removal of fuel from the reactor vessel to maintain an FPP to address the potential for fires that could result in a radiological hazard.

The objectives of the FPP in 10 CFR 50.48(f)(1) are to (1) reasonably prevent such fires from occurring; (2) rapidly detect, control, and extinguish those fires that do occur and that could result in a radiological hazard; and (3) ensure that the risk of fire-induced radiological hazards to the public, environment, and plant personnel is minimized. The regulation at 10 CFR 50.48(f)(2) requires licensees to assess the FPP on a regular basis and revise it, as appropriate, throughout the various stages of facility decommissioning. The regulation at 10 CFR 50.48(f)(3) permits licensees to make changes to the FPP without NRC approval if these changes do not reduce the effectiveness of fire protection for facilities, systems, and equipment, and that could result in a radiological hazard, taking into account the decommissioning plant conditions and activities.

The requirement to submit certifications in accordance with 10 CFR 50.82(a)(1) was established after IP1 was shut down. However, IP1 is contiguous with IP2 and some of its systems support IP3. As indicated in Section 2.0 of this SE, Entergy certified in accordance with 10 CFR 50.82(a)(1)(i) and (ii) that power operations ceased at IP2 on April 30, 2020, and that the fuel was permanently removed from the IP2 reactor vessel and placed in the IP2 spent fuel pool on May 12, 2020. With the docketing of the IP2 certifications in accordance with 10 CFR 50.82(a)(1)(i) and (ii), Entergy is required to maintain an FPP for IP2 that complies with 10 CFR 50.48(f) using the approved FPP. This program will also be applied to IP1. For IP3, the requirements for an operating nuclear power plant will continue to apply until Entergy certifies the permanent cessation of operations at IP3 and the permanent removal of fuel from the IP3 reactor vessel. While the IP1 Appendix A TSs are revised to denote that IP1 systems are utilized to support IP3, the continued operation of IP3 and planned permanent cessation of operations on April 30, 2021, do not impact the implementation of the IP1 PDTS.

3.4 Continuation of License The regulation at 10 CFR 50.51(b) states that 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.

4.0 TECHNICAL EVALUATION

4.1 Proposed Changes to the Provisional Operating License 4.1.1 License Title The current license title is Provisional Operating License.

The licensee proposed to delete Operating from the title, so that it reads, Provisional License.

The proposed change to the title to delete Operating would provide a more accurate description of the facility for the permanently shutdown and defueled condition. Consistent with License Condition 2.a) of the IP1 POL, the license no longer authorizes operation of the reactor.

In addition, IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved to the reactor containment building without prior review and authorization by the NRC. Accordingly, the NRC staff finds the proposed change to the license title acceptable.

4.1.2 License Condition 1.A Currently, License Condition 1.A reads:

The application for amendment by Entergy Nuclear Operations, Inc. (the licensee), dated May 30,2002, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commissions

[rues] and regulations set forth in 10 CFR Chapter 1; The licensee proposed to make editorial corrections to License Condition 1.A by adding a space between the date and the year, and replacing the misspelled word rues with the word rules.

These are editorial changes that do not alter the meaning of the license condition. Therefore, the NRC staff finds the proposed change to License Condition 1.A acceptable.

4.1.3 License Condition 1.B Currently, License Condition 1.B reads:

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

The licensee proposed License Condition 1.B to read:

The facility will be maintained in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission.

The proposed change to the description the facility will operate to the facility will be maintained would provide a more accurate description of the requirements during the permanently shutdown and defueled condition. Consistent with License Condition 2.a) of the IP1 POL, the license no longer authorizes operation of the reactor and IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved into the reactor containment building without prior review and authorization by the NRC. In addition, the IP1 spent fuel is stored at the IPEC ISFSI. Therefore, the NRC staff finds the proposed change to License Condition 1.B acceptable.

4.1.4 License Condition Preamble Currently, the license condition preamble reads:

Accordingly, License No. DPR-5 is hereby amended as indicated in the attachment to this license amendment, and paragraphs 3.B and 9 of Provisional Operating License No. DPR-5 are hereby amended to read as follows:

The licensee proposed the license condition preamble to read:

Accordingly, License No. DPR-5 is hereby amended as indicated in the attachment to this license amendment, and Provisional License No. DPR-5 is hereby amended to read as follows:

The proposed change to delete paragraphs 3.B and 9 of will remove an unnecessary reference contained in the 10 CFR license. Also, the deletion of the term Operating would provide a more accurate description of the facility during the permanently shutdown and defueled condition. Consistent with License Condition 2.a) of the IP1 POL, the license no longer authorizes operation of the reactor. In addition, IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved to the reactor containment building without prior review and authorization by the NRC. Therefore, the NRC staff finds the proposed change to the license condition preamble acceptable.

4.1.5 License Condition 1 Currently, License Condition 1 reads:

This license applies to the utilization facility consisting of a pressurized water reactor (hereinafter referred to as 'the reactor), and associated components and equipment hereinafter specified, which is owned by ENIP2, located in Westchester County, New York, and described in the amended and Substituted Application for Licenses dated November 30, 1960, as amended; in the Application for License amendment dated April 6, 1965, as supplemented May 6, 1965; and in the Application for License amendment dated December 3, 1965 (hereinafter referred to as 'the application'), and which is a part of the electric

generating plant which has been designated by ENIP2 as the Indian Point Station Unit No. 1.

The licensee proposed License Condition 1 to read:

This license applies to the utilization facility consisting of a pressurized water reactor (the facility) and associated components and equipment, which is owned by ENIP2. The facility is located in Westchester County, New York, and described in the Defueled Safety Analysis Report, as supplemented and amended, and which is designated by ENIP2 as the Indian Point Station Unit No. 1.

The licensee proposed to revise License Condition 1 to provide a generic description of the facility to reflect the permanently shutdown and defueled condition, with all spent fuel stored at the IPEC ISFSI. The licensee also proposed to refer to the facility as described in the Defueled Safety Analysis Report (DSAR). The NRC staff reviewed the proposed changes to License Condition 1 and concludes that the changes are editorial in nature and are applicable to a permanently shutdown and defueled facility. The DSAR is the document that will remain applicable to IP1. Therefore, the NRC staff finds the proposed change to License Condition 1 acceptable.

4.1.6 License Condition 2.a)

Currently, License Condition 2.a) reads:

ENIP2 and ENO, pursuant to Section 104b. of the Act and Title 10CFR Part 50, Licensing of Production and Utilization Facilities, to possess but not operate the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations described in the application and this license; The licensee proposed License Condition 2.a) to read:

ENIP2 and ENO, pursuant to Section 104b. of the Act and Title 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess but not operate the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations described in the application and this license; The licensee proposed to revise the license condition to correct an error regarding a reference to Title 10 CFR Part 50. This is an editorial correction that does not alter the meaning of the license condition. Therefore, the NRC staff finds the proposed change to License Condition 2.a) acceptable.

4.1.7 License Condition 2.b)

Currently, License Condition 2.b) reads as follows:

ENO, pursuant to the Act and 10 CFR Part 70, to receive and possess up to 1918 kilograms of contained uranium-235 previously received for reactor operation;

The licensee proposed License Condition 2.b) to read:

ENO, pursuant to the Act and 10 CFR Part 70, to possess up to 1918 kilograms of contained uranium-235 previously received for reactor operation; The licensee proposed to revise the license condition to remove the authorization to receive uranium-235 since IP1 has no need to receive uranium-235 as it is permanently shut down and defueled, with all of its spent fuel stored at the IPEC ISFSI. Consistent with License Condition 2.a) of the IP1 POL, the license no longer authorizes operation of the reactor. In addition, IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved into the reactor containment building without prior review and authorization by the NRC. The continued authorization to possess uranium-235 previously received for reactor operation is necessary as IP1 currently possesses the uranium-235 that was used for the past operations of the reactor. Therefore, the NRC staff finds the proposed change to License Condition 1.b) acceptable.

4.1.8 License Condition 2.e)

Currently, License Condition 2.e) reads as follows:

ENO, pursuant to the Act and 10 CFR Parts 30 and 70, to receive and possess, but not to separate, such byproduct and special materials as were produced by the prior operation of the facility; The licensee proposed License Condition 2.e) to read:

ENO, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not to separate, such byproduct and special materials as were produced by the prior operation of the facility; The licensee proposed to revise the license condition to remove the authorization to receive byproduct and special materials that were produced by the prior operation of the IP1 facility which is permanently shut down and defueled, with all of its spent fuel stored at the IPEC ISFSI.

Consistent with License Condition 2.a) of the IP1 POL, the license no longer authorizes operation of the reactor. In addition, IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved into the reactor containment building without prior review and authorization by the NRC. IP1 has no need to receive byproduct and special materials that were produced by the prior operation of the facility. The continued authorization to possess byproduct and special materials that were produced by the prior operation of the facility is necessary as IP1 and the ISFSI currently possess the material that was used for the past operations of the reactor. Therefore, the NRC staff finds the proposed change to License Condition 2.e) acceptable.

4.1.9 License Condition 3.b)

Currently, License Condition 3.b) reads:

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 59, are hereby incorporated in the license. ENO shall maintain the facility in accordance with the Technical Specifications.

The licensee proposed License Condition 3.b) to read:

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 63, are hereby incorporated in the license. ENO shall maintain the facility in accordance with the Technical Specifications.

The licensee proposed to revise the license condition to replace the license amendment number to reflect the issuance of the revised TSs. This is an administrative change reflecting the current license amendment for IP1 as Amendment No. 63. Therefore, the NRC staff finds the proposed change to License Condition 3.b) acceptable.

4.1.10 License Condition 3.c)

Currently, License Condition 3.c) reads:

Records In addition to those otherwise required under this license and applicable regulations, ENO shall keep the following records:

1. Reactor operating records, including power levels and period of operation at each power level.

2. Records showing the radioactivity released or discharged into the air or water beyond the effective control of ENO as measured at or prior to the point of such release or discharge.

3. Records of scrams, including reasons therefor.

4. Records of principal maintenance operations involving substitution or replacement of facility equipment or components and the reasons therefor.

5. Records of radioactivity measurements at on-site and off-site monitoring stations.

6. Records of facility tests and measurements performed pursuant to the requirements of the Technical Specifications.

The licensee proposed to delete this license condition in its entirety on the basis that the subject records will be maintained by the Quality Assurance Program Manual (i.e., as required by 10 CFR Part 50, Appendix B, Criterion XVII, Quality Assurance Records) and because provisions relating to recordkeeping do not assure safe operation of a facility in a permanently defueled condition. The NRC staff reviewed the proposed deletion of License Condition 3.c) and concludes that the proposed changes are applicable for a defueled facility. Therefore, the NRC staff finds the deletion of License Condition 3.c) acceptable.

4.1.11 License Condition 4 Currently, License Condition 4 reads:

Deleted by Amendment No.7, dated 11-14-74.

The licensee proposed to delete License Condition 4 and remove the historical reference to the license amendment that deleted the content of the original license condition. This change is an administrative change to provide consistency with regard to how deleted license conditions are delineated in the license. Therefore, the NRC staff finds the deletion of License Condition 4 acceptable.

4.1.12 License Condition 5 Currently, License Condition 5 reads:

Definitions - As used in this license the term facility means the following systems and components as described in the application The licensee proposed to delete License Condition 5 to remove the statement that associated components and equipment are specified within the license and therefore, reflect the permanently shutdown and defueled condition. With IP1 permanently shut down and defueled, License Condition 2.a) of the IP1 POL no longer authorizes operation of the reactor. In addition, IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved into the reactor containment building without prior review and authorization by the NRC. The NRC staff reviewed the proposed changes to License Condition 5 and concludes that the changes are applicable for a permanently shutdown and defueled facility. Therefore, the NRC staff finds the deletion of License Condition 5 acceptable.

4.1.13 License Condition 9 Currently, License Condition 9 states:

The approved Decommissioning Plan supplements the Final Safety Analysis Report (FSAR) and the licensee may (i) make changes in the facility or procedures as described in the FSAR or the Decommissioning Plan and (ii) conduct tests, or experiments not described in the FSAR or Decommissioning Plan, without prior Commission approval, provided the requirements of 10 CFR 50.59 and 10 CFR 50.82(a)(6) and (7) are satisfied.

The licensee proposed License Condition 9 to read:

The approved Decommissioning Plan supplements the Defueled Safety Analysis Report (DSAR) and the licensee may (i) make changes in the facility or procedures as described in the DSAR or the Decommissioning Plan and (ii) conduct tests, or experiments not described in the DSAR or Decommissioning Plan, without prior Commission approval, provided the requirements of 10 CFR 50.59 and 10 CFR 50.82(a)(6) and (7) are satisfied.

The licensee proposed to revise License Condition 9 to retitle the FSAR as the DSAR. The NRC staff reviewed the proposed change to License Condition 9 and concludes that the changes are editorial in nature and are applicable to a permanently shutdown and defueled facility. The DSAR is the document that will remain applicable to IP1. Therefore, the NRC staff finds the proposed change to License Condition 9 acceptable.

4.1.14 License Condition 10 Currently, License Condition 10 states:

The amended license is effective as of the date of its issuance, shall be implemented within 30 days, and shall expire at midnight, September 28, 2013.

The licensee proposed License Condition 10 to read:

The amended license is effective as of the date of its issuance, and until the Commission notifies the licensee in writing that the license is terminated.

The proposed change would modify this license condition to reflect the permanently shutdown and defueled condition of the facility. Consistent with License Condition 2.a) of the IP1 POL, the license no longer authorizes operation of the reactor. In addition, IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved into the reactor containment building without prior review and authorization by the NRC. 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. The NRC staff reviewed the proposed change to License Condition 10. The current License Condition 10, which documents the date of the expiration of the POL, is no longer necessary for the permanently shutdown and defueled condition of the facility in the process of decommissioning. The revised License Condition 10 documents the current condition of the facility and summarizes the actions and requirements applicable to the facility by 10 CFR 50.51.

Therefore, the NRC staff finds the proposed change to License Condition 10 acceptable.

4.2 Changes to Appendix A to the Provisional Operating License DPR-05 4.2.1 Title Page The current title page states, in part:

PROVISIONAL OPERATING LICENSE DPR-5 AMENDMENT NO. 50

The licensee proposed to change the title page to state, in part:

PROVISIONAL LICENSE DPR-5 AMENDMENT NO. 63 The licensee proposed to delete Operating from the title, so that it reads Provisional License DPR-5. The proposed change to the title to delete Operating would provide a more accurate description of the facility that is not licensed to operate and in the permanently shutdown and defueled condition. Consistent with License Condition 2.a) of the IP1 POL, the license no longer authorizes operation of the reactor. In addition, IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved to the reactor containment building without prior review and authorization by the NRC. Accordingly, the NRC staff finds the proposed change to the License Title acceptable.

4.2.2 Table of Contents The licensee proposed to revise the Table of Contents of Appendix A, to reflect proposed additions, deletions, and changes to the TSs, as described in Sections 4.2.3 through 4.2.9 of this SE, and as detailed in Attachment 1 to the LAR dated June 30, 2020. The changes to the Table of Contents are editorial and do not change any technical content. The NRC staff finds the changes to the Table of Contents acceptable.

4.2.3 TS Section 1.0, General Information The licensee proposed a revision to TS 1.0. The current TS 1.0 states the following:

the reactor remains in a defueled status and the unit continues to operate as a support facility for overall Indian Point Units 1 and 2 operations. Unit No. 1 and Unit No. 2 are physically contiguous and share a number of systems and facilities as well as a common operating organization. The technical specifications contained herein recognize this commonality as well as the intended use of the Unit No. 1 facilities to support Unit No. 2 until retirement of that unit, and contain specific references to Appendix A to the Indian Point Unit No. 2 Facility Operating License No. DPR-26. Unit No. 1 contains radioactive waste processing facilities, which provide waste processing services for both Unit No. 1 and Unit No. 2.

Radiological effluent The licensee proposed to revise TS 1.0 to state the following:

the reactor remains in a defueled status and some of the facilities structures, systems, and components support activities at Indian Point Units 1, 2, and 3.

Unit No. 1 and Unit No. 2 are physically contiguous and share a number of systems and facilities as well as a common organization. The technical specifications contained herein recognize this commonality as well as the intended use of the Unit No. 1 facilities to support Unit Nos. 2 and 3 until the fuel for these units is transferred from the spent fuel pits to an offsite storage facility, and contain specific references to Appendix A to the Indian Point Unit No. 2 Facility License No. DPR-26. Unit No. 1 contains radioactive waste processing facilities, which provide waste processing services for Unit Nos. 1, 2, and 3.

Radiological effluent

The licensee proposed modifying the general information to eliminate the reference to a shared operating organization between IP1 and IP2 since IP1 is permanently shut down and defueled, with all its spent fuel stored at the IPEC ISFSI. Since IP2 is no longer authorized to operate the reactor or place or retain fuel in the reactor vessel, personnel required for IP1 and IP2 are substantially reduced, and an operating organization will not be required. These modifications delineate that some of the IP1 structures, systems, and components continue to fulfill a support function for IP1, IP2, and IP3 until the fuel for these units is transferred from the spent fuel pits to an offsite storage facility.

The NRC staff examined the licensees proposed modifications to TS Section1.0. The NRC staff reviewed the licensees basis for these revisions and concludes that the revisions are acceptable, as they reflect the decommissioning facility.

4.2.4 TS 1.1, Definitions The licensee proposed deletion of the terms and definitions in TS 1.1.1, 1.1.2, 1.1.4, 1.1.5, and 1.3, stating in part, that the terms are no longer used in any TSs, or the definitions reside elsewhere. The licensee proposed to retain TS 1.1.3 and 1.2 because it remains applicable to a defueled facility.

4.2.4.1 TS 1.1.1, Final Safety Analysis Report The current TS 1.1.1 definition states the following:

1.1.1 Final Safety Analysis Report The Final Safety Analysis Report (FSAR) for Indian Point Unit No. 1, shall be deemed to refer to, as appropriate, the Final Hazards Summary Report for the Consolidated Edison Indian Point Reactor Core B and the following exhibits, which are a part of the original license application for IP1:

Docket 50-3 Exhibit K-5 (Rev. 1), Hazards Summary Report Consolidated Edison Thorium Reactor. (January, 1960) Figures 1-2, 1-3, 3-14 only.

Docket 50-3 Exhibit K-5A1 1, Supplementary Information on Plant Design of Consolidated Edison Nuclear Steam Generating Station, (August 1960) Section 3.7.1, pages 171 through 176 only and Section 3.7.2.

The licensee proposed to delete TS 1.1.1 in its entirety. The licensee stated that the FSAR will be replaced with the DSAR, that changes to the DSAR will be reviewed in accordance with 10 CFR 50.59, and that updates will be available for review by NRC inspectors. The NRC staff reviewed the licensees basis for the deletion of TS 1.1.1. as a term under TS Section 1.1, Definitions, and concludes that it is appropriate and acceptable.

4.2.4.2 TS 1.1.2, Operable-Operability The current TS 1.1.2 definition states the following:

1.1.2 Operable-Operability A system, subsystem, train, component or device shall be operable or have operability when it is capable of performing its intended safety function(s).

Implicit in this definition shall be the assumption that necessary instrumentation, controls, electrical power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its safety function(s) are also capable of performing their related support functions.

The licensee proposed to delete TS 1.1.2 in its entirety. The licensee stated the term operable-operability is not used in any TS and that there are no systems or components required to be operable in the TS, because there are no active SSCs required to function to mitigate any DBAs. The NRC staff reviewed the licensees basis for the deletion of TS 1.1.2 as a term under TS Section 1.1, Definitions, and concludes that it is appropriate and acceptable.

4.2.4.3 TS 1.1.3, Offsite Dose Calculation Manual (ODCM)

The licensee proposed to retain the TS 1.1.3 definition with revision. The current TS 1.1.3 definition states the following:

1.1.3 Offsite Dose Calculation Manual (ODCM)

The Offsite Dose Calculation Manual contains the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the environmental radiological monitoring program. Requirements for the ODCM are specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility Operating License No. DPR-26.

The licensee proposed to revise TS 1.1.3 to state the following:

The Offsite Dose Calculation Manual contains the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm/trip setpoints, and in the conduct of the environmental radiological monitoring program. Requirements for the ODCM are specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

The licensee proposed to retain the TS 1.1.3 definition and to revise the definition by removing the term Operating and deleting the specific section number (1.1.3). The licensee stated that since the proposed limited information will remain in IP1 PDTS section 1.1 through 1.3, the remaining information is consolidated in Section 1.1 as definitions. Also, consistent with License Condition 2.a) of the IP1 POL, the license no longer authorizes operation of the reactor.

Accordingly, the NRC staff finds the proposed change to TS 1.1.3 is acceptable.

4.2.4.4 TS 1.1.4, Site Boundary The current TS 1.1.4 definition states the following:

1.1.4 Site Boundary The Site Boundary is that line beyond which the land is neither owned, leased, nor otherwise controlled by either ENIP2, ENO, or other site licensee.

The licensee proposed to delete TS 1.1.4 in its entirety. The licensee stated that the term, site boundary is defined in 10 CFR 20.1003, and therefore, the definition does not need to be repeated in the PDTS. The NRC staff reviewed the licensees basis for the deletion of TS 1.1.4 as a term under TS Section 1.1, Definitions, and concludes that it is acceptable.

4.2.4.5 TS 1.1.5, Unrestricted Area The current TS 1.1.5 definition states the following:

1.1.5 Unrestricted Area An Unrestricted Area is any area at or beyond the Site Boundary, access to which is not controlled by either ENIP2, ENO, or other site licensee for purposes of protection of individuals from radiation and radioactive materials.

The licensee proposed to delete TS 1.1.5 in its entirety. The licensee stated that an unrestricted area is defined in 10 CFR 20.1003, and therefore, the definition does not need to be repeated in the PDTS. The NRC staff reviewed the licensees basis for the deletion of TS 1.1.5 as a term under TS Section 1.1, Definitions, and concludes that it is acceptable.

4.2.4.6 TS 1.2, Exclusion Distance The licensee proposed to retain the TS 1.2 definition with an editorial revision. The current TS 1.2 definition states the following:

1.2 Exclusion Distance 1.2.1 The minimum distance from the reactor facility to the nearest land boundary of the exclusion area, as defined in 10CFR100 shall be 1400 feet.

1.2.2 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 10CFR100.3 is 460 meters and 1100 meters, respectively.

The licensee proposed to revise TS 1.2 to state the following:

1.2 Exclusion Distance 1.2.1 The minimum distance from the reactor facility to the nearest land boundary of the exclusion area, as defined in 10 CFR 100 shall be 1400 feet.

1.2.2 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 460 meters and 1100 meters, respectively.

The licensee proposed to retain TS 1.2 in the IP1 PDTS and to revise the definition by adding an appropriate space in 10 CFR 100 and 10 CFR 100.3 for TS 1.2.1 and TS 1.2.2, respectively.

The NRC staff examined the licensees proposed revisions to TS Section 1.2 and concludes that the revisions are editorial in nature. Therefore, the NRC staff finds the proposed change to TS 1.2 is acceptable.

4.2.4.7 TS 1.3, Principal Activities The current TS 1.3 definition states the following:

1.3 Principal Activities The principal activities carried on within the Exclusion Area shall be the generation, transmission and distribution of steam and electrical energy (except by gas-fired power plant); associated service activities; activities relating to the controlled conversion of the atomic energy of fuel to heat energy by the process of nuclear fission; and the storage, utilization and production of special nuclear, source and byproduct materials. Transmission and distribution of natural gas shall be through the use of facilities located as described in the application as amended.

The licensee proposed to delete TS 1.3 in its entirety. The licensee stated the term principal activities is not used in the PDTS. The NRC staff reviewed the licensees basis for the deletion of TS 1.3 as a term under TS Section 1.1, Definitions, and concludes that it is appropriate and acceptable.

4.2.5 TS Section 2.0, Reactor Facility Design Performance Requirements 4.2.5.1 TS 2.1, Electrical Power Supply Current TS 2.1 states the following:

2.1 Electrical Power Supply Power for electrical equipment shall normally be supplied by at least two independent transmission feeders from the Consolidated Edison system. If power is lost to the spent fuel storage area radiation monitor, a portable monitor will be promptly set up in the spent fuel storage area.

The licensee proposed to delete TS 2.1 in its entirety. The licensee stated that no IP1 SSCs are required to perform a function to mitigate any DBA. In addition, the spent fuel storage pool for IP1 is no longer in use, because the spent fuel has been transferred to the IPEC ISFSI. The NRC staff reviewed the licensees basis for the deletion of TS 2.1 and concludes that it is acceptable.

4.2.5.2 TS 2.2, Fuel Storage The current TS 2.2 states the following:

2.2 Fuel Storage 2.2.1 No fuel other than Irradiated fuel from Indian Point Unit No. 1 shall be stored in the Unit No. 1 spent fuel storage area. No fresh fuel shall be stored at Unit No. 1.

2.2.2 Spent fuel storage shall be provided in the storage pools in the Fuel Handling Building.

2.2.3 Spent fuel storage shall be provided with racks that shall limit the effective multiplication factor to less than 0.75.

2.2.4 Radiation levels in the spent fuel storage area shall be monitored continuously with a high level alarm indication in a location manned by a licensed operator*

2.2.5 If a spent fuel pool contains spent fuel, the spent fuel cask shall not be moved over that pool or within a distance of that pool such that the cask could strike the pool if it fell or tipped.

2.2.6 A dead-load test shall be successfully performed on the fuel handling building crane before fuel movement begins

• licensed Operator for IP-2 The licensee proposed to revise TS 2.2 to state the following:

Fuel Storage No fuel shall be stored in the Unit No. 1 spent fuel storage area.

The licensee proposed to modify TS 2.2.1 in order to renumber it as TS 2.2 and to reflect that no fuel will be permitted to be stored in the IP1 spent fuel storage area. The licensee also proposed to delete TS 2.2.2 through 2.2.6 in their entirety. These TS relate to the safe storage and handling of spent fuel. As discussed above, the IP1 spent fuel storage area no longer contains irradiated fuel and the reactor has been permanently shut down. Since IP1 no longer stores spent fuel in the storage area, portions of TS 2.2.1 and TS 2.2.2 through 2.2.6 no longer meet the criteria of 10 CFR 50.36(c)(2) for inclusion in the TS. Therefore, the NRC staff finds the proposed changes and deletions to TS 2.2 acceptable.

4.2.5.3 TS 2.3, Fire Protection The current TS 2.3 states the following:

2.3 Fire Protection Overall site fire protection is provided by a fire protection system, which is common to both Unit No. 1 and Unit No. 2. Operation, maintenance and testing are controlled by common procedures.

The licensee proposed to delete TS 2.3 in its entirety. The NRC staff reviewed the proposed change to TS 2.3 and concludes that the change is consistent with the transition to a permanently shutdown and defueled facility. During the decommissioning process, an FPP is required by 10 CFR 50.48(f) and the ability to make changes to the FPP is governed by 10 CFR 50.48(f)(3). These requirements exist regardless of whether the TSs contain a requirement to provide a fire protection system, or to establish procedures for operation, maintenance, and testing of the fire protection system. Therefore, TS 2.3 is not needed when the facility is permanently shut down and defueled. Accordingly, the NRC staff finds the proposed change to TS 2.3 to be acceptable.

4.2.6 TS Section 3.0, Administrative and Procedural Safeguards 4.2.6.1 TS 3.1, Responsibility The current TS 3.1 states the following:

3.1 Responsibility Responsibilities are as specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility Operating License No. DPR-26.

The licensee proposed to revise TS 3.1 to state the following:

3.1 Responsibility Responsibilities are as specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

The licensee proposed to remove the term Operating from TS 3.1 regarding responsibilities.

The licensee stated that License Condition 2.a) of the IP1 POL no longer authorizes operation of the reactor and IP1 TS 3.3.1 does not permit fuel to be loaded into the reactor core or moved into the reactor containment building without prior NRC authorization. For IP2, its 10 CFR Part 50 license no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2), and therefore, the responsibilities for facility operation are revised to account for the decommissioning facility conditions and activities. The revised IP2 requirements will continue to apply to IP1. The NRC staff reviewed the licensees basis for the proposed revision to TS 3.1 and concludes that it is acceptable.

4.2.6.2 TS 3.2, Organization The current TS 3.2 states the following:

3.2 Organization The organization requirements are as specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility Operating License No. DPR-26.

a. All fuel handling shall be under the direct supervision of a licensed operator.*

b. The Shift Manager is responsible for operations at the Unit No. 1 facility.

• Licensed operator for IP-2 The licensee proposed to revise TS 3.2 to state the following:

3.2 Organization The organization requirements are as specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

The licensee proposed to remove the term Operating from TS 3.2 and eliminate the reference to fuel handling activities at IP1. Consistent with License Condition 2.a) of the IP1 POL, the license no longer authorizes operation of the reactor. As discussed above, the IP1 spent fuel storage area no longer contains irradiated fuel and the reactor has been permanently shut down. For IP2, its 10 CFR Part 50 license no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2), and therefore, a licensed operator is no longer needed for IP2 as the responsibilities for facility operation are revised to account for the decommissioning facility conditions and activities. The NRC staff reviewed the licensees basis for the proposed revision to TS 3.2 and concludes that it is acceptable.

4.2.6.3 TS 3.3, Operating Instructions and Procedures The current TS 3.3 states the following:

3.3 Operating Instructions and Procedures 3.3.2 Detailed written instruction setting forth procedures used in connection with the operation and maintenance of the nuclear power plant shall conform to the requirements specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility Operating License No. DPR-26.

3.3.3 Operation and maintenance of equipment related to safety when there is no fuel in the reactor shall be in accordance with written instructions.

The licensee proposed to revise TS 3.3 to state the following:

3.3 Operating Instructions and Procedures 3.3.2 Detailed written instruction setting forth procedures used in connection with the facility shall conform to the requirements specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

The licensee proposed to remove the reference to operation and maintenance of the nuclear power plant with a reference to facility and to eliminate the word Operating from TS 3.3.2.

The licensee stated that since the IP1 facility is permanently shut down and defueled, and the spent fuel has been transferred to the IPEC ISFSI, it is no longer a plant that generates electricity; thus, the use of the term facility is more appropriate. For IP2, its 10 CFR Part 50 license no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). The licensee also proposed to delete TS 3.3.3 in its entirety. The licensee stated that this requirement is a generic requirement that is duplicated by requirements in the Quality Assurance Program Manual.

The NRC staff examined the licensees proposal to replace operation and maintenance of the nuclear power plant with facility, remove Operating, and delete TS 3.3.3. The NRC staff reviewed the licensees basis for the proposed revision to TS 3.3.2 and deletion of TS 3.3.3 and concludes that these revisions are appropriate and acceptable.

4.2.7 TS Section 4.0, Operating Limitations 4.2.7.1 TS 4.1, General The current TS 4.1 states the following:

4.1 General Whenever any operation is being performed that could result in the release of radioactivity or create a change in radiation levels, supporting facilities shall be maintained and operated as required in these Technical Specifications.

The licensee proposed to delete TS 4.1 in its entirety. This TS provides a general requirement regarding activities at IP1 involving any operation that could result in the release of radioactivity or create a change in radiation levels. It applies to TS 4.2 through TS 4.6. The licensee stated that although TS 4.2 and 4.5 will be maintained in the IP1 TSs, TS 4.1 is not required to ensure that IP1 complies with these TSs. Thus, TS 4.1 will no longer serve a purpose in the IP1 TSs.

The NRC staff examined the licensees proposal to delete TS 4.1 in its entirety. The NRC staff reviewed the licensees basis for the deletion of TS 4.1 and concludes that it is acceptable.

4.2.7.2 TS 4.3, Radioactive Waste The current TS 4.3 states the following:

4.3 Radioactive Waste All radioactive waste material shall be handled in accordance with 10 CFR Part 20. In addition, solid radioactive waste shall be controlled as specified in the Process Control Program.

The licensee proposed to delete TS 4.3 in its entirety. The licensee stated that this TS requires radioactive waste to be handled and controlled in accordance with 10 CFR Part 20 and Process Control Program procedures. This TS implements the requirements of 10 CFR Parts 20, 61, and 71, and it is not necessary to restate these requirements in the IP1 TSs. The Process Control Program is conducted under standard procedures with revisions approved by facility processes, and program changes are reported to the NRC by the Annual Radioactive Effluent Report. The NRC staff reviewed the licensees basis for the deletion of TS 4.3 and concludes that it is acceptable.

4.2.7.3 TS 4.4, Radiation Monitoring The current TS 4.4 states the following:

4.4 Radiation Monitoring Radiation monitoring systems shall be maintained operable for: (1) sphere foundation sump, (2) secondary purification blowdown cooling water, and (3) area radiation monitors. If monitoring systems are not operable, effluent sampling and/or local monitoring shall be accomplished to replace the non-operating system. In addition, Unit 1 radioactive effluent monitoring instrumentation shall be operable as specified in the ODCM.

The licensee proposed to delete TS 4.4 in its entirety. The licensee stated that the requirements for radiation monitoring will be relocated to the DSAR for the sphere foundation sump and any changes to the DSAR will be reviewed in accordance with 10 CFR 50.59 and the updates will be available for review by NRC inspectors. The IP1 facility is permanently shut down and defueled, and the IP1 spent fuel has been transferred to the IPEC ISFSI. The monitoring requirements for the area radiation monitors and secondary purification blowdown cooling water will be eliminated and are no longer required. Additionally, the radiation monitors for the sphere foundation sump do not meet any of the criteria in 10 CFR 50.36(c)(2) for inclusion in the TS. The NRC staff reviewed the licensees basis for the deletion of TS 4.4 and concludes that it is acceptable.

4.2.7.4 TS 4.6, Spent Fuel Storage and Handling The current TS 4.6 states the following:

4.6 Spent Fuel Storage Handling 4.6.1 All irradiated fuel shall be stored in the racks provided in the Fuel Handling Building Storage pools, with sufficient shielding that ensures that the radiation level on the operating deck is 15 mr/hr...

4.6.2 Whenever, spent fuel storage pool water inventory is provided for personnel shielding, the normal water level shall be maintained at or above elevation 48 feet (approximately 6 feet above the top of the spent fuel racks) 4.6.3 Water chemistry in any spent fuel storage pool containing spent fuel shall be maintained within the following limits 4.6.4 Ventilation capable of directing all Fuel Handling Building airborne effluents through monitoring pathways shall be available during any fuel movement or other activity that might potentially damage spent fuel assemblies.

The licensee proposed to delete TS 4.6 in its entirety. As discussed above, the IP1 spent fuel storage area no longer contains irradiated fuel, all IP1 spent fuel has been transferred to the IPEC ISFSI, and the reactor has been permanently shut down. The licensee also stated there is no longer a need to address requirements regarding storage of irradiated fuel, shielding, water level, or water chemistry in the spent fuel storage pool. Given that fuel movement is complete and there are no other activities that may potentially damage spent fuel assemblies in the spent fuel storage pool, the IP1 fuel handling building ventilation system is not required to perform a function. The NRC staff finds that since IP1 spent fuel storage area no longer stores spent fuel, TS 4.6 no longer meets the criteria identified in 10 CFR 50.36(c)(2) for inclusion in the TS . The NRC staff reviewed the licensees basis for the deletion of TS 4.6 and concludes that it is acceptable.

4.2.8 TS Section 5.0, Maintenance 4.2.8.1 TS 5.2, Testing The current TS 5.2 states the following:

5.2 Testing 5.2.1 Functional radiation monitoring systems (only for the following: sphere foundation sump and secondary purification blowdown cooling water) and area radiation monitoring systems shall be:

(a) qualitatively checked daily to verify acceptable operability of instrument channel behavior during operation, and

(b) tested quarterly by injection of a simulated signal into the instrument channel to verify that it is operable, including alarm and/or trip initiating action. The quarterly interval is defined as quarterly plus or minus 25% of the quarter.

5.2.2 Unit 1 radioactive effluent monitoring instrumentation shall satisfy the surveillance requirements as specified in the ODCM.

The licensee proposed to delete TS 5.2.1 in its entirety, including the deletion of the section number 5.2.1, and proposed to renumber TS 5.2.2 as TS 5.2. As discussed above in SE section 4.2.7.3 for TS 4.4, Radiation Monitoring, the requirements for testing the radiation monitoring system for the sphere foundation sump will be relocated to the DSAR. The testing requirements for the area radiation monitoring system and the radiation monitoring system for the secondary purification blowdown cooling water will be eliminated. The radiation monitoring system for the sphere foundation sump do not satisfy any of the requirements in 10 CFR 50.36(c)(2) for inclusion in the TSs. The license stated that relocation of the sphere foundation sump testing requirements to the DSAR is appropriate, because changes to the DSAR will be reviewed in accordance with 10 CFR 50.59 and the updates will be available for review by NRC inspectors. IP1 spent fuel storage area no longer contains irradiated fuel, all IP1 spent fuel has been transferred to the IPEC ISFSI, and the reactor has been permanently shut down. Also, the renumbering of TS section number for TS 5.2.2 to TS 5.2 is an administrative change. The NRC staff reviewed the licensees basis for the deletion of TS 5.2.1 and the renumbering of TS 5.2.2 and concludes that it is acceptable.

4.2.8.2 TS 5.3, Spent Fuel Storage Pool Sampling The current TS 5.3 states the following:

5.3 Spent Fuel Storage Pool Sampling Any spent fuel storage pool containing spent fuel stored in water shall be sampled monthly for chloride level, pH and Cesium 137 activity The licensee proposed to delete TS 5.3 in its entirety. The IP1 facility is permanently shut down and defueled, and the IP1 spent fuel has been transferred to the IPEC ISFSI. The fuel is no longer stored in the IP1 spent fuel storage pool and there is no longer a need to sample the water in the spent fuel storage pool. The NRC staff reviewed the licensees basis for the deletion of TS 5.3 and concludes that it is acceptable.

4.2.9 TS Section 6.0, Plant Reporting Requirements The current TS 6.0, Plant Reporting Requirements, states the following:

Reporting Requirements are as specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility Operating License No. DPR-26.

The licensee proposed to revise TS 6.0 by eliminating the reference to Operating. The revised description will state the following:

Reporting Requirements are as specified in Appendix A to the Indian Point Nuclear Generating Unit No. 2 Facility License No. DPR-26.

Consistent with 10 CFR 50.82(a)(2), the IP2 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. The removal of the word Operating in describing the license would provide accuracy in the 10 CFR Part 50 license description. Therefore, the NRC staff finds the proposed change is acceptable.

5.0 STATE CONSULTATION

In accordance with the Commissions regulations, the New York State official was notified of the proposed issuance of the amendment on March 23, 2021. The State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment relates, in part, to changes in recordkeeping, reporting, or administrative procedures or requirements. The amendment also relates, in part, to changing requirements with respect to the installation or use of facility components located within the restricted area, as defined in 10 CFR Part 20, and changes SRs. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration (85 FR 52374; August 25, 2020), and there has been no public comment on such finding.

Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and 51.22(c)(10). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commissions 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.

Principal Contributors: R. Guzman D. Scully R. Hernandez J. Robinson S. Mehta A. Russell Date: April 14, 2021

ML21083A000 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DRA/APLB/BC(A)

NAME RGuzman JBurkhardt SVasavada DATE 4/1/2021 4/1/2021 1/25/2021 OFFICE NRR/DSS/STSB/BC NRR/DSS/SCPB/BC NMSS/DUWP/RDB/BC NAME VCusumano BWittick BWatson DATE 11/18/2020 2/22/2021 4/12/21 OFFICE OGC NRR/DORL/LPL1/BC NRR/DORL/LPL1/PM NAME STurk JDanna RGuzman DATE 4/12/2021 4/13/2021 4/14/2021