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{{#Wiki_filter:April 22, 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 UNIT NO. 3  ISSUANCE OF AMENDMENT NO. 270 RE: PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS (EPID L-2020-LLA-0090)
 
==Dear Sir or Madam:==
 
The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 270 to Renewed Facility Operating License No. DPR-64 for Indian Point Nuclear Generating Unit No. 3 (Indian Point 3). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated April 28, 2020.
The amendment revises the Indian Point 3 Renewed Facility Operating License and the associated TSs to permanently defueled TSs, consistent with the permanent cessation of operations and permanent removal of fuel from the reactor vessel.
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. 50-286
 
==Enclosures:==
: 1. Amendment No. 270 to DPR-64
: 2. Safety Evaluation cc: Listserv
 
ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.
DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 License No. DPR-64
: 1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:
A. The application for amendment by Entergy Nuclear Operations, Inc. (ENO, the licensee) dated April 28, 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 Appendix A, Technical Specifications; Appendix B, Environmental Technical Specification Requirements; and Appendix C, Inter-Unit Fuel Transfer Technical Specifications, as indicated in the attachment to this license amendment, and Renewed Facility Operating License No. DPR-64 is hereby amended to read as follows:
The title RENEWED FACILITY OPERATING LICENSE is to read RENEWED FACILITY LICENSE Paragraph 1.B is to read as follows:
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 2.A and 2.B are to read as follows:
: 2.      Accordingly, Renewed Facility License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows:
A. This renewed license applies to the Indian Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and associated equipment (the facility), owned by ENIP3 and maintained by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the Defueled Safety Analysis Report, as supplemented and amended, and the Environmental Report, as amended.
B. Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:
(1)      Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) ENIP3 to possess and use, and (b) ENO to possess and use the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; (2)      ENO pursuant to the Act and 10 CFR Part 70, to possess at any time, special nuclear material that was used as reactor fuel, in accordance with the limitations for storage, as described in the Defueled Safety Analysis Report, as supplemented and amended; (3)      ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct source and special nuclear material
 
as sealed neutron sources that were used for reactor startup, sealed sources that were used for calibration of reactor instrumentation and are used in the calibration of radiation monitoring equipment, and that were used as fission detectors in amounts as required; (4)    ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components; (5)    ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials that were produced by the operation of the facility.
Paragraph 2.C is to read as follows:
(1)  Deleted per Amendment No. 270.
(2)  Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 270, are hereby incorporated in the renewed license. ENO shall maintain the facility in accordance with the Technical Specifications.
Paragraphs 2.H, 2.O, 2.AA, 2.AB, and 2.AD are to read as follows:
H. Deleted per Amendment No. 270.
O. Deleted per Amendment No. 270.
AA. Deleted per Amendment No. 270.
AB. Deleted per Amendment No. 270.
AD. Deleted per Amendment No. 270.
Paragraph 3 is to read as follows:
: 3. This renewed license is effective as of the date of issuance, and until the Commission notifies the licensee in writing that the license is terminated.
: 3.      This license amendment is effective following the docketing of the certifications required by 10 CFR 50.82(a)(1)(i) and (ii) that Indian Point Nuclear Generating Unit No. 3 has been permanently shut down and defueled, and shall be implemented within 90 days of the effective date of the amendment.
FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Jason Jason C.        C. Paige Date: 2021.04.22 Paige          11:25:16 -04'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
 
==Attachment:==
 
Changes to the Renewed Facility Operating License, Appendix A, Technical Specifications, Appendix B, Environmental Technical Specification Requirements, and Appendix C, Inter-Unit Fuel Transfer Technical Specifications Date of Issuance: April 22, 2021
 
ATTACHMENT TO LICENSE AMENDMENT NO. 270 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 RENEWED FACILITY OPERATING LICENSE NO. DPR-64 DOCKET NO. 50-286 Replace the following pages of the Renewed Facility Operating License; Appendix A, Permanently Defueled Technical Specifications; Appendix B, Environmental Technical Specification Requirements; and Appendix C, Inter-Unit Fuel Transfer Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. For the Appendix A PDTS revised pages, vertical line revision bars were not used as the changes were considered a major rewrite.
Facility Operating License No. DPR-64 REMOVE                                    INSERT through                            through                      Appendix A, Permanently Defueled Technical Specifications REMOVE                                    INSERT Title Page                              Title Page i through iv                                    i 1.1-1 through 1.1-8                                1.1-1 1.2-1 through 1.2-3                                1.2-1 1.3-1 through 1.3-14                      1.3-1 through 1.3-2 1.4-1 through 1.4-4                                1.4-1 2.0-1                                    2.0-1 3.0-1 through 3.0-5                      3.0-1 through 3.0-2 3.1.1-1 through 3.7.13-2                              ---
3.7.14-1                                  3.7.14-1 3.7.15-1 through 3.7.15-2                            3.7.15-1 3.7.16-1 through 3.7.16-3                  3.7.16-1 through 3.7.16-3 3.7.17-1 through 3.9.6-1                              ---
4.0-1 through 4.0-3                                4.0-1 5.0-1 through 5.0-41                      5.0-1 through 5.0-15
 
Appendix B, Environmental Technical Specification Requirements REMOVE                                INSERT Part I Title Page                    Part I Title Page Table of Contents                    Table of Contents 1-1                                  1-1 3-1                                  3-1 3-2                                  3-2 3-3                                  3-3 4-1                                  4-1 4-2                                  4-2 5-1                                  5-1 5-2                                  5-2 5-4                                  5-4 Part II Title Page                    Part II Title Page Appendix C, Inter-Unit Fuel Transfer Technical Specifications REMOVE                                  INSERT Part I Title Page                    Part I Title Page 1                                    1 2                                    2 Part II Title Page                    Part II Title Page
 
ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.
DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 RENEWED FACILITY LICENSE Renewed License No. DPR-64
: 1. The Nuclear Regulatory Commission (the Commission) has found that:
A. The application for a renewed license filed by Entergy Nuclear Indian Point 3, LLC (ENIP3) (the licensee) and Entergy Nuclear Operations, Inc. (ENO)
(operator) for Indian Point Nuclear Generating Unit No. 3 (IP3 at the Indian Point Energy Center (IPEC) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will be maintained in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this renewed license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. ENIP3 and ENO are financially and technically qualified to engage in the activities authorized by this amendment; E. ENIP3 and ENO have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements" of the Commission's regulations; F. The issuance of this renewed license will not be inimical to the common defense and security or to the health and safety of the public; G. The receipt, possession and use of source, byproduct and special nuclear material as authorized by this renewed license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70 including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31; Amendment No. 270
 
H. The issuance of this renewed license is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1); and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by this renewed license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations.
: 2. Accordingly, Renewed Facility License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows:
A. This renewed license applies to the Indian Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and associated equipment (the facility),
owned by ENIP3 and maintained by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the Defueled Safety Analysis Report" as supplemented and amended, and the Environmental Report, as amended.
B. Subject to the conditions and requirements incorporated herein, the Commission licenses:
(1)    Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) ENIP3 to possess and use, and (b) ENO to possess and use the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; (2)    ENO pursuant to the Act and 10 CFR Part 70, to possess, at any time, special nuclear material that was used as reactor fuel, in accordance with the limitations for storage, as described in the Defueled Safety Analysis Report, as supplemented and amended; (3)    ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct source and special nuclear material as sealed neutron sources that were used for reactor startup, sealed sources that were used for calibration of reactor instrumentation and are used in the calibration of radiation monitoring equipment, and that were used as fission detectors in amounts as required; Amendment No. 270
 
(4)  ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components; (5)  ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials that were produced by the operation of the facility.
C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
(1)  Deleted per Amendment No. 270 (2)  Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 270, are hereby incorporated in the renewed license. ENO shall maintain the facility in accordance with the Technical Specifications.
D.    (DELETED)
E.    (DELETED)
F. This renewed license is also subject to appropriate conditions by the New York State Department of Environmental Conservation in its letter granting a Section 401 certification under the Federal Water Pollution Control Act Amendments of 1972.
G. ENO shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and to the authority of 10 CFR 50.90 and CFR 50.54(p). The combined set of plans 1 for the Indian Point Energy Center, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: Physical Security, Training and Qualification, and Safeguards Contingency Plan, Revision 0, and was submitted by {{letter dated|date=October 14, 2004|text=letter dated October 14, 2004}}, as supplemented by {{letter dated|date=May 18, 2006|text=letter dated May 18, 2006}}.
1 The Training and Qualification Plan and Safeguards Contingency Plan are Appendices to the Security Plan.
Amendment No. 270
 
ENO shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The ENO CSP was approved by License Amendment No. 243, as supplemented by changes approved by License Amendment Nos. 254, 260, and 263.
ENO has been granted Commission authorization to use stand alone preemption authority under Section 161A of the Atomic Energy Act, 42 U.S.C.
2201a with respect to the weapons described in Section II supplemented with Section Ill of Attachment 1 to its application submitted by {{letter dated|date=August 20, 2013|text=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.
H. Deleted per Amendment No. 270 I. DELETED J. DELETED K. DELETED L. DELETED M. DELETED N. DELETED O. Deleted per Amendment No. 270 P. ENIP3 and ENO shall take no action to cause Entergy Global Investments, Inc.
or Entergy International Ltd. LLC, or their parent companies to void, cancel, or modify the $70 million contingency commitment to provide funding for the facility as represented in the application for approval of the transfer of the license from PASNY to ENIP3 and ENO, without the prior written consent of the Director, Office of Nuclear Reactor Regulation.
Q. DELETED R. DELETED S. DELETED T. DELETED U. DELETED V. DELETED Amendment No. 270
 
W. For purposes of ensuring public health and safety, ENIP3, upon the transfer of this license to it, and upon transfer of decommissioning funds from PASNY to ENO, shall provide decommissioning funding assurance for the facility by the prepayment or equivalent method, to be held in a decommissioning trust fund for the facility, of no less than the amount required under NRC regulations at 10 CFR 50.75. Any amount held in any decommissioning trust maintained by ENO for the facility after the transfer of the facility license to ENIP3 may be credited towards the amount required under this paragraph.
X. ENIP3 shall take all necessary steps to ensure that the decommissioning trust is maintained in accordance with the application for the transfer of this license to ENIP3 and ENO, as modified by the request to transfer decommissioning funds from PASNY, and the requirements of the order approving the transfer and order approving the transfer of decommissioning funds from PASNY to ENO, and consistent with the safety evaluations supporting such orders.
AA. Deleted per Amendment No. 270 AB. Deleted per Amendment No. 270 AC. Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:
(a)    Fire fighting response strategy with the following elements:
: 1. Pre-defined coordinated fire response strategy and guidance
: 2. Assessment of mutual aid fire fighting assets
: 3. Designated staging areas for equipment and materials
: 4. Command and control
: 5. Training of response personnel (b)    Operations to mitigate fuel damage considering the following:
: 1. Protection and use of personnel assets
: 2. Communications
: 3. Minimizing fire spread
: 4. Procedures for implementing integrated fire response strategy
: 5. Identification of readily-available pre-staged equipment
: 6. Training on integrated fire response strategy
: 7. Spent fuel pool mitigation measures (c)    Actions to minimize release to include consideration of:
: 1. Water spray scrubbing
: 2. Dose to onsite responders AD. Deleted per Amendment No. 270 AE. ENO may transfer IP3 spent fuel to the IP2 spent fuel pit subject to the conditions listed in Appendix C. ENO is further authorized to transfer IP3 spent fuel into NRC approved storage casks for onsite storage by ENO and ENIP3.
Amendment No. 270
 
AF. License Renewal License Conditions (1) The information in the UFSAR supplement, submitted pursuant to 10 CFR 54.21(d) and as revised during the license renewal application review process, and licensee commitments as listed in Appendix A of the Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Units 2 and 3, (SER) and supplements to the SER, are collectively the License Renewal UFSAR Supplement. The UFSAR Supplement is henceforth part of the UFSAR, which will be updated in accordance with 10 CFR 50.71(e). As such, the licensee may make changes to the programs, activities, and commitments described in the UFSAR Supplement, provided the licensee evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59, Changes, Tests, and Experiments, and otherwise complies with the requirements in that section.
(2) The License Renewal UFSAR Supplement, as defined in license condition AF(1) above, describes certain programs to be implemented and activities to be completed prior to the period of extended operation (PEO).
: a. The licensee shall implement those new programs and enhancements to existing programs no later than the date specified in the License Renewal UFSAR Supplement.
: b. The licensee shall complete those activities no later than the date specified in the License Renewal UFSAR Supplement.
: 3.      This renewed license is effective as of the date of issuance, and until the Commission notifies the licensee in writing that the license is terminated.
FOR THE NUCLEAR REGULATORY COMMISSION Ho K. Nieh, Director Office of Nuclear Reactor Regulation Attachments:
Appendix A - Permanently Defueled Technical Specifications Appendix B - Environmental Technical Specification Requirements Appendix C - Inter-Unit Fuel Transfer Technical Specifications Date of Issuance: September 17, 2018 Amendment No. 270
 
APPENDIX A TO FACILITY LICENSE DPR-64 PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS AND BASES FOR THE INDIAN POINT 3 NUCLEAR GENERATING STATION UNIT NO. 3 WESTCHESTER COUNTY, NEW YORK ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)
AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)
DOCKET NO. 50-286 Date of Issuance:
April 15, 1976 Amendment No. 270
 
Facility License No. DPR-64 Appendix A - Permanently Defueled Technical Specifications TABLE OF CONTENTS 1.0        USE AND APPLICATION 1.1            Definitions 1.2            Logical Connectors 1.3            Completion Times 1.4            Frequency 2.0        DELETED 3.0        LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.7        SPENT FUEL PIT REQUIREMENTS 3.7.14          Spent Fuel Pit Water Level 3.7.15          Spent Fuel Pit Boron Concentration 3.7.16          Spent Fuel Assembly Storage 4.0        DESIGN FEATURES 4.1            Site Location 4.2            Deleted 4.3            Fuel Storage 5.0        ADMINISTRATIVE CONTROLS 5.1            Responsibility 5.2            Organization 5.3            Facility Staff Qualifications 5.4            Procedures 5.5            Programs and Manuals 5.5.1              Offsite Dose Calculation Manual (ODCM) 5.5.2              DELETED 5.5.3              NOT USED 5.5.4              Radioactive Effluent Controls Program 5.5.5              DELETED 5.5.6              DELETED 5.5.7              DELETED 5.5.8              DELETED 5.5.9              DELETED 5.5.10              DELETED 5.5.11              Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5.12              DELETED 5.5.13              Technical Specification (TS) Bases Control Program 5.6            Reporting Requirements 5.6.1              NOT USED 5.6.2              Annual Radiological Environmental Operating Report 5.6.3              Radioactive Effluent Release Report 5.7            High Radiation Area Indian Point 3                                  i                      Amendment No. 270
 
Definitions 1.1 1.0      USE AND APPLICATION 1.1      Definitions
-----------------------------------------------------------NOTE------------------------------------------------------------
The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases.
Term                                            Definition ACTIONS                                          ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times.
CERTIFIED FUEL HANDLER                          A CERTIFIED FUEL HANDLER is an individual who (CFH)                                            complies with the provisions of the CERTIFIED FUEL HANDLER training and retraining program required by TS 5.3.2.
NON-CERTIFIED OPERATOR                          A NON-CERTIFIED OPERATOR is a non-licensed operator who complies with the qualification requirements of Specification 5.3.1, but is not a CERTIFIED FUEL HANDLER.
Indian Point 3                                              1.1-1                                  Amendment No. 270
 
Logical Connectors 1.2 1.0    USE AND APPLICATION 1.2    Logical Connectors PURPOSE            The purpose of this section is to explain the meaning of logical connectors.
Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Required Actions and Surveillances. The only logical connectors that appear in TS are AND and OR. The physical arrangement of these connectors constitutes logical conventions with specific meanings.
BACKGROUND          Several levels of logic may be used to state Required Actions. These levels are identified by the placement (or nesting) of the logical connectors and by the number assigned to each Required Action. The first level of logic is identified by the first digit of the number assigned to a Required Action and the placement of the logical connector in the first level of nesting (i.e., left justified with the number of the Required Action).
The successive levels of logic are identified by additional digits of the Required Action number and by successive indentations of the logical connectors.
When logical connectors are used to state a Surveillance, only the first level of logic is used, and the logical connector is left justified with the statement of the Surveillance.
EXAMPLE            The following example illustrates the use of logical connectors.
EXAMPLE 1.2-1 ACTIONS CONDITION                REQUIRED ACTION              COMPLETION TIME A. LCO not met            A.1    Verify AND A.2    Restore In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.
Indian Point 3                                  1.2-1                            Amendment No. 270
 
Completion Times 1.3 1.0    USE AND APPLICATION 1.3    Completion Times PURPOSE            The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.
BACKGROUND          Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe handling and storage of spent nuclear fuel. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met.
Specified with each stated Condition are Required Action(s) and Completion Time(s).
DESCRIPTION        The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the facility is in a specified condition stated in the Applicability of the LCO.
Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the facility is not within the LCO Applicability.
EXAMPLE            The following example illustrates the use of Completion Times with different Required Actions.
Indian Point 3                                1.3-1                          Amendment No. 270
 
Completion Times 1.3 1.3    Completion Time EXAMPLE (continued)
EXAMPLE 1.3-1 ACTIONS CONDITION            REQUIRED ACTION            COMPLETION TIME A. Spent fuel pit      A.1  Suspend              Immediately boron                    movement of fuel concentration not        assemblies in the within limit.            spent fuel pit.
AND A.2  Initiate action to    Immediately restore spent fuel pit boron concentration to within limit.
Condition A has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion time is referenced to the time that Condition A is entered.
The Required Actions of Condition A are to immediately suspend movement of fuel assemblies in the spent fuel pit and initiate action to restore spent fuel pit boron concentration within limit.
IMMEDIATE          When "Immediately" is used as a Completion Time, the Required Action COMPLETION TIME should be pursued without delay and in a controlled manner.
Indian Point 3                              1.3-2                          Amendment No. 270
 
Frequency 1.4 1.0    USE AND APPLICATION 1.4    Frequency PURPOSE          The purpose of this section is to define the proper use and application of Frequency requirements.
DESCRIPTION      Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.
The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR)
Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR.
EXAMPLE          The following example illustrates the type of Frequency statement that appears in the Technical Specifications (TS).
EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                FREQUENCY Verify level is within limits.                          12 hours Example 1.4-1 contains the type of SR encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval.
Although the Frequency is stated as 12 hours, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when a variable is outside specified limits, or the facility is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the facility is in a specified condition in the Applicability of the LCO, then SR 3.0.3 becomes applicable.
If the interval as specified by SR 3.0.2 is exceeded while the facility is not in a specified condition in the Applicability of the LCO for which performance of the SR is required, then SR 3.0.4 becomes applicable.
The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3, prior to entry into the specified condition or the LCO is considered not met (in accordance with SR 3.0.1).
Indian Point 3                              1.4-1                            Amendment No. 270
 
Deleted 2.0 2.0    DELETED Indian Point 3 2.0-1 Amendment No. 270
 
LCO Applicability 3.0 3.0    LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1          LCOs shall be met during the specified conditions in the Applicability, except as provided in LCO 3.0.2.
LCO 3.0.2          Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met.
If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated.
Indian Point 3                              3.0-1                          Amendment No. 270
 
SR Applicability 3.0 3.0    SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1        SRs shall be met during the specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on variables outside specified limits.
SR 3.0.2        The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance.
SR 3.0.3        If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.
If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.
SR 3.0.4        Entry into a specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3.
Indian Point 3                            3.0-2                          Amendment No. 270
 
Spent Fuel Pit Water Level 3.7.14 3.7    SPENT FUEL PIT REQUIREMENTS 3.7.14 Spent Fuel Pit Water Level LCO 3.7.14              The spent fuel pit water level shall be  23 ft over the top of irradiated fuel assemblies seated in the storage racks.
APPLICABILITY:          During movement of irradiated fuel assemblies in the spent fuel pit.
ACTIONS CONDITION                      REQUIRED ACTION                    COMPLETION TIME A. Spent fuel pit water        A.1 Suspend movement of                Immediately level not within limit.          irradiated fuel assemblies in the spent fuel pit.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                        FREQUENCY SR 3.7.14.1    Verify the spent fuel pit water level is  23 ft        7 days above the top of the irradiated fuel assemblies seated in the storage racks.
Indian Point 3                                  3.7.14-1                          Amendment No. 270
 
Spent Fuel Pit Boron Concentration 3.7.15 3.7    SPENT FUEL PIT REQUIREMENTS 3.7.15 Spent Fuel Pit Boron Concentration LCO 3.7.15            The spent fuel pit boron concentration shall be  1000 ppm.
                      -----------------------------------------------NOTE--------------------------------------------
During inter-unit transfer of fuel the spent fuel pit boron concentration must also meet Appendix C LCO 3.1.1, Boron Concentration.
APPLICABILITY:        When fuel assemblies are stored in the spent fuel pit and a spent fuel pit verification has not been performed since the last movement of fuel assemblies in the spent fuel pit.
ACTIONS CONDITION                        REQUIRED ACTION                            COMPLETION TIME A. Spent fuel pit boron          A.1      Suspend movement of                  Immediately concentration not within              fuel assemblies in the limit.                                spent fuel pit.
AND A.2.1 Initiate action to restore              Immediately spent fuel pit boron concentration to within limit.
OR A.2.2 Initiate action to perform a            Immediately spent fuel pit verification.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                                    FREQUENCY SR 3.7.15.1    Verify the spent fuel pit boron concentration is                  31 days within limit.
Indian Point 3                                      3.7.15-1                                  Amendment No. 270
 
Spent Fuel Assembly Storage 3.7.16 3.7    SPENT FUEL PIT REQUIREMENTS 3.7.16 Spent Fuel Assembly Storage LCO 3.7.16          Fuel assemblies stored in the spent fuel pit shall be classified in accordance with Figure 3.7.16-1 based on initial enrichment and burnup; and, Fuel assembly storage location within the spent fuel pit shall be restricted based on the Figure 3.7.16-1 classification as follows:
: a. Fuel assemblies classified as Type 2 may be stored in any location in either Region 1 or Region 2;
: b. Fuel assemblies classified as Type 1A, 1B or 1C shall be stored in Region 1;
: c. Fuel assembly storage location within Region 1 shall be restricted as follows:
: 1. Type 1A assemblies may be stored anywhere in Region 1;
: 2. Type 1B assemblies may be stored anywhere in Region 1, except a Type 1B assembly shall not be stored face-adjacent to a Type 1C assembly;
: 3. Type 1C assemblies shall not be stored in Row 64 or in Column ZZ; and
: 4. Type 1C assemblies shall be stored in Region 1 locations where all face-adjacent locations are as follows:
a) occupied by Type 2 or Type 1A assemblies, or b) occupied by non-fuel components, or c) empty.
APPLICABILITY:      Whenever any fuel assembly is stored in the spent fuel pit.
Indian Point 3                              3.7.16-1                        Amendment No. 270
 
Spent Fuel Assembly Storage 3.7.16 ACTIONS CONDITION                    REQUIRED ACTION                  COMPLETION TIME A. Requirements of the      A.1    Initiate action to move fuel Immediately LCO not met.                    to restore compliance with LCO 3.7.16.
SURVEILLANCE REQUIREMENTS SURVEILLANCE                                      FREQUENCY SR 3.7.16.1  Verify by administrative means the initial          Prior to storing the fuel enrichment and burnup of each fuel assembly        assembly in the spent fuel and that the storage location meets LCO 3.7.16      pit requirements.
Indian Point 3                              3.7.16-2                        Amendment No. 270
 
Spent Fuel Assembly Storage 3.7.16 Figure 3.7.16-1 (Page 1 of 1)
Fuel Assembly Classification for Storage in the Spent Fuel Pit Indian Point 3                        3.7.16-3                      Amendment No. 270
 
Design Features 4.0 4.0    DESIGN FEATURES 4.1    Site Location Indian Point 3 is located on the east bank of the Hudson River at Indian Point, Village of Buchanan, in upper Westchester County, New York. The site is approximately 24 miles north of the New York City boundary line. The nearest city is Peekskill which is 2.5 miles northeast of Indian Point.
The minimum distance from the reactor center line to the boundary of the site exclusion area and the outer boundary of the low population zone as defined in 10 CFR 100.3 is 350 meters and 1100 meters, respectively.
4.2    Deleted 4.3    Fuel Storage 4.3.1  Criticality 4.3.1.1    The spent fuel storage racks are designed and shall be maintained with:
: a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;
: b. keff  0.95 if assemblies are inserted in accordance with Technical Specification 3.7.16, Spent Fuel Assembly Storage;
: c. A nominal 9.075 inch center to center distance between fuel assemblies placed in the high density fuel storage racks (Region II);
: d. A nominal 10.76 inch center to center distance between fuel assemblies placed in low density fuel storage racks (Region I).
4.3.2  Drainage The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pool below a nominal elevation of 88 ft.
4.3.3  Capacity The spent fuel pit is designed and shall be maintained with a storage capacity limited to no more than 1345 fuel assemblies.
Indian Point 3                                  4.0-1                        Amendment No. 270
 
Responsibility 5.1 5.0    ADMINISTRATIVE CONTROLS 5.1    Responsibility 5.1.1          The plant manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.
The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.
5.1.2          The shift manager (SM) shall be responsible for the shift command function.
Indian Point 3                                5.0-1                            Amendment No. 270
 
Organization 5.2 5.0    ADMINISTRATIVE CONTROLS 5.2    Organization 5.2.1          Onsite and Offsite Organizations Onsite and offsite organizations shall be established for facility staff and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear fuel.
: a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all decommissioning organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the facility-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the DSAR and Quality Assurance Plan, as appropriate;
: b. The plant manager shall be responsible for overall safe maintenance of the facility and shall have control over those onsite activities necessary for safe storage and maintenance of nuclear fuel;
: c. The corporate officer with direct responsibility for IP3 shall have corporate responsibility for the safe storage and handling of nuclear fuel and shall take any measures needed to ensure acceptable performance of the staff in maintaining and providing technical support to the facility to ensure safe management of nuclear fuel; and
: d. The individuals who train the CERTIFIED FUEL HANDLERS, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.
Indian Point 3                                  5.0-2                          Amendment No. 270
 
Organization 5.2 5.2    Organization 5.2.2          Facility Staff The facility staff organization shall include the following:
: a.      Each duty shift shall be composed of at least one shift manager and one NON-CERTIFIED OPERATOR. The NON-CERTIFIED OPERATOR position may be filled by a CERTIFIED FUEL HANDLER.
At least one person qualified to stand watch in the control room (NON-CERTIFIED OPERATOR or CERTIFIED FUEL HANDLER) shall be present in the control room when nuclear fuel is stored in the spent fuel pool.
: b.      Shift crew composition may be less than the minimum requirement of 5.2.2.a for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements and all of the following conditions are met:
: 1)    No fuel movements are in progress;
: 2)    No movement of loads over fuel are in progress; and
: 3)    No unmanned shift positions during shift turnover shall be permitted while the shift crew is less than the minimum.
: c.      An individual qualified in radiation protection procedures shall be on site during fuel handling operations and during movement of heavy loads over the fuel storage racks. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
: d.      Not Used.
: e.      The shift manager shall be a CERTIFIED FUEL HANDLER.
: f.      Deleted.
Indian Point 3                                    5.0-3                      Amendment No. 270
 
Facility Staff Qualifications 5.3 5.0    ADMINISTRATIVE CONTROLS 5.3    Facility Staff Qualifications 5.3.1          Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1-1978 for comparable positions with exceptions specified in the IPEC Quality Assurance Program Manual (QAPM).
5.3.2          An NRC approved training and retraining program for CERTIFIED FUEL HANDLERS shall be maintained.
Indian Point 3                                5.0-4                          Amendment No. 270
 
Procedures 5.4 5.0    ADMINISTRATIVE CONTROLS 5.4    Procedures 5.4.1          Written procedures shall be established, implemented, and maintained covering the following activities:
: a.      The procedures applicable to the safe storage of nuclear fuel recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 except as provided in the quality assurance program described or referenced in the DSAR;
: b.      Deleted;
: c.      Quality assurance for effluent and environmental monitoring;
: d.      Deleted; and
: e.      All programs specified in Specification 5.5.
Indian Point 3                                  5.0-5                        Amendment No. 270
 
Programs and Manuals 5.5 5.0    ADMINISTRATIVE CONTROLS 5.5    Programs and Manuals The following programs shall be established, implemented, and maintained.
5.5.1          Offsite Dose Calculation Manual (ODCM)
: a.      The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and
: b.      The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification 5.6.2 and Specification 5.6.3.
: c.      Licensee initiated changes to the ODCM:
: 1.      Shall be documented and records of reviews performed shall be retained. This documentation shall contain:
(a) Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and (b) A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;
: 2.      Shall become effective after the approval of the plant manager; and
: 3.      Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.
5.5.2          Deleted Indian Point 3                                  5.0-6                          Amendment No. 270
 
Programs and Manuals 5.5 5.5    Programs and Manuals 5.5.3          Not Used 5.5.4          Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:
: a.      Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;
: b.      Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2;
: c.      Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;
: d.      Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit/facility to unrestricted areas, conforming to 10 CFR 50, Appendix I;
: e.      Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days;
: f.      Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I; Indian Point 3                                  5.0-7                          Amendment No. 270
 
Programs and Manuals 5.5 5.5    Programs and Manuals 5.5.4          Radioactive Effluent Controls Program (continued)
: g.      Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:
: a. For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and
: b. For iodine-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to dose rate of 1500 mrem/yr to any organ.
: h.      Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit/facility to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;
: i.      Limitations on the annual and quarterly doses to a member of the public from iodine-131, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit/facility to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and
: j.      Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluents Controls Program surveillance frequency.
5.5.5 through Deleted 5.5.10 Indian Point 3                                  5.0-8                          Amendment No. 270
 
Programs and Manuals 5.5 5.5    Programs and Manuals 5.5.11        Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, Postulated Radioactive Release due to Waste Gas System Leak or Failure.
The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, Postulated Radioactive Release due to Tank Failures.
The program shall include:
: a.      The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);
: b.      A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank shall be limited to less than the amount that would result in a whole body exposure of  0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks contents; and
: c.      A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks contents.
The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.
5.5.12        Deleted Indian Point 3                                    5.0-9                          Amendment No. 270
 
Programs and Manuals 5.5 5.5    Programs and Manuals 5.5.13        Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
: a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
: b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
: 1. a change in the TS incorporated in the license; or
: 2. a change to the DSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
: c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the DSAR.
: d. Proposed changes that do not meet the criteria of Specification 5.5.13.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
Indian Point 3                              5.0-10                          Amendment No. 270
 
Reporting Requirements 5.6 5.0    ADMINISTRATIVE CONTROLS 5.6    Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.
5.6.1          Not Used 5.6.2          Annual Radiological Environmental Operating Report
              --------------------------------------------------NOTE-------------------------------------------------
A single submittal may be made for a multiple unit/facility station. The submittal should combine sections common to all units/facilities at the station.
The Annual Radiological Environmental Operating Report covering the operation of the unit/facility during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.
A full listing of the information to be contained in the Annual Radiological Environmental Operating Report is provided in the ODCM.
5.6.3          Radioactive Effluent Release Report
              --------------------------------------------------NOTE-------------------------------------------------
A single submittal may be made for a multiple unit/facility station. The submittal shall combine sections common to all units/facilities at the station; however, for units/facilities with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit/facility.
The Radioactive Effluent Release Report covering the operation of the unit/facility in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit/facility. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR Part 50.36a and 10 CFR 50, Appendix I, Section IV.B.l.
Indian Point 3                                        5.0-11                                  Amendment No. 270
 
High Radiation Area 5.7 5.0    ADMINISTRATIVE CONTROLS 5.7    High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:
5.7.1          High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation
: a.      Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
: b.      Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
: c.      Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following facility radiation protection procedures for entry to, exit from, and work in such areas.
: d.      Each individual or group entering such an area shall possess:
: 1.      A radiation monitoring device that continuously displays radiation dose rates in the area; or
: 2.      A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or
: 3.      A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or Indian Point 3                                  5.0-12                        Amendment No. 270
 
High Radiation Area 5.7 5.7    High Radiation Area 5.7.1          High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation (continued)
: 4.      A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i)    Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii)    Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.
: e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.
Indian Point 3                                5.0-13                          Amendment No. 270
 
High Radiation Area 5.7 5.7    High Radiation Area 5.7.2          High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation
: a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:
: 1.      All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or his or her designee.
: 2.      Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.
: b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
: c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following facility radiation protection procedures for entry to, exit from, and work in such areas.
: d. Each individual or group entering such an area shall possess:
: 1.      A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or
: 2.      A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or Indian Point 3                                5.0-14                          Amendment No. 270
 
High Radiation Area 5.7 5.7    High Radiation Area 5.7.2          High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued)
: 3.      A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i)    Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii)    Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.
: 4.      In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.
: e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.
: f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.
Indian Point 3                                5.0-15                          Amendment No. 270
 
APPENDIX B TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)
AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)
INDIAN POINT 3 NUCLEAR POWER PLANT ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART I: NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION PLAN FACILITY LICENSE NO. DPR-64 DOCKET NUMBER 50-286 Renewed License No. DPR-64 Amendment No. 270
 
INDIAN POINT NUCLEAR GENERATING PLANT UNIT 3 ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART I:      NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION PLAN TABLE OF CONTENTS Section                                                                                                                    Page 1.0    Objectives of the Environmental Protection Plan ................................................... 1-1 2.0    Environmental Protection Issues ........................................................................... 2-1 3.0    Consistency Requirements .................................................................................... 3-1 3.1    Plant Design and Operation................................................................................... 3-1 3.2    Reporting Related to the NPDES Permits and State Certifications ........................ 3-2 3.3    Changes Required for Compliance with Other Environmental Regulations. ........... 3-3 4.0    Environmental Conditions ...................................................................................... 4-1 4.1    Unusual or Important Environmental Events .......................................................... 4-1 4.2    Environmental Monitoring ...................................................................................... 4-1 5.0    Administrative Procedures ..................................................................................... 5-1 5.1    Review and Audit .................................................................................................. 5-1 5.2    Records Retention ................................................................................................. 5-1 5.3    Changes in Environmental Protection Plan ............................................................ 5-1 5.4    Plant Reporting Requirements ............................................................................... 5-2 Renewed License No. DPR-64 Amendment No. 270
 
1.0    Objectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) is to provide for protection of environmental values during handling and storage of spent fuel and maintenance of the nuclear facility. The principal objectives of the EPP are as follows:
(1)    Verify that the facility is maintained in an environmentally acceptable manner, as established by the FES and other NRC environmental impact assessments.
(2)    Coordinate NRC requirements and maintain consistency with other Federal, State and local requirements for environmental protection.
(3)    Keep NRC informed of the environmental effects of handling and storage of spent fuel and maintenance of the facility and of actions taken to control those effects.
Environmental concerns identified in the FES which relate to water quality matters are regulated by way of the licensee's SPDES permit.
1-1              Renewed License No. DPR-64 Amendment No. 270
 
3.0      Consistency Requirements 3.1      Plant Design and Operation ENO may make changes in facility design or operations or perform tests or experiments affecting the environment provided such changes, tests or experiments do not involve an unreviewed environmental question, and do not involve a change in the Environmental Protection Plan.* Changes in the facility design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of this EPP. Activities governed by Section 3.3 are not subject to the requirements of this section.
Before engaging in additional construction or operational activities which may affect the environment, ENO shall prepare and record an environmental evaluation of such activity.
When the evaluation indicates that such activity involves an unreviewed environmental question, ENO shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation. When such activity involves a change in the Environmental Protection Plan, such activity and change to the Environmental Protection Plan may be implemented only in accordance with an appropriate license amendment as set forth in Section 5.3.
A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the final environmental statement (FES) or final supplemental environmental impact statement (FSEIS), as modified by the staffs testimony to the Atomic Safety and Licensing Boards, supplements to the FES or FSEIS, environmental impact appraisals, or in any decision of the Atomic Safety and Licensing Board; This provision does not relieve the ENO of the requirements of 10 CFR 50.59.
3-1                Renewed License No. DPR-64 Amendment No. 270
 
or (2) a significant change in effluents; or (3) a matter not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact.
ENO shall maintain records of changes in facility design or operation and of tests and experiments carried out pursuant to this Subsection. These records shall include a written evaluation which provides a basis for the determination that the change, test, or experiment does not involve an unreviewed environmental question nor constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1.0. ENO shall include as part of its Annual Environmental Protection Plan Report per Subsection 5.4.1: brief descriptions, analyses, interpretations, and evaluations of such changes, tests and experiments.
3.2      Reporting Related to the NPDES Permits and State Certifications Violations of the NPDES Permit or the State certification (pursuant to Section 4.1 of the Clean Water Act) shall be reported to the NRC by submittal of copies of the reports required by the NPDES Permit or certification.
Changes and additions to the NPDES Permit or the State certification shall be reported to the NRC within 30 days following the date the change is approved. If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted.
The NRC shall be notified of changes to the effective NPDES Permit proposed by ENIP3 and ENO by providing NRC with a copy of the proposed change at the same time it is submitted to the permitting agency. The notification of a licensee-initiated change shall include a copy of the requested revision submitted to the permitting agency. ENO shall provide the NRC a copy of 3-2              Renewed License No. DPR-64 Amendment No. 270
 
the application for renewal of the NPDES permit at the same time the application is submitted to the permitting agency.
3.3    Changes Required for Compliance with Other Environmental Regulations Changes in facility design or operation and performance of tests or experiments which are required to achieve compliance with other Federal, State, or local environmental regulations are not subject to the requirements of Section 3.1.
3-3                Renewed License No. DPR-64 Amendment No. 270
 
4.0    Environmental Conditions 4.1    Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to the handling and storage of spent fuel and maintenance of the facility shall be recorded and promptly reported to the NRC within 24 hours by telephone, telegraph, or facsimile transmissions followed by a written report per Subsection 5.4.2. The following are examples: excessive bird impaction events, onsite plant or animal disease outbreaks, unusual mortality or occurrence of any species protected by the Endangered Species Act of 1973, unusual fish kills, unusual increase in nuisance organisms or conditions, and unanticipated or emergency discharge of waste water or chemical substances.
No routine monitoring programs are required to implement this condition.
4.2    Environmental Monitoring In accordance with Section 7(a) of the Endangered Species Act, the National Marine Fisheries Service (NMFS) issued a Biological Opinion related to the continued operation of IP2 and IP3 that pertains to shortnose sturgeon (Acipenser brevirostrum) and Atlantic sturgeon (Acipenser oxyrinchus oxyrinchus). The Biological Opinion includes an Incidental Take Statement with Reasonable and Prudent Measures that the NMFS has determined to be necessary or appropriate to minimize the amount or extent of incidental take and associated Terms and Conditions, which are non-discretionary and implement the Reasonable and Prudent Measures. The currently applicable Biological Opinion concludes that continued operation of IP2 and IP3 is not likely to jeopardize the continued existence of the listed species or to adversely affect the designated critical habitat of those species. This Biological Opinion 4-1                Renewed License No. DPR-64 Amendment No. 270
 
conservatively bounds the conditions that will occur in the permanently shut down and defueled condition.
Entergy shall adhere to the requirements within the Incidental Take Statement of the currently applicable Biological Opinion. Changes to the Biological Opinion, including the Incidental Take Statement, Reasonable and Prudent Measures, and Terms and Conditions contained therein, must be preceded by consultation between the NRC, as the authorizing agency, and the NMFS.
4-2              Renewed License No. DPR-64 Amendment No. 270
 
5.0      Administrative Procedures 5.1      Review and Audit ENO shall provide a review and audit of compliance with the Environmental Protection Plan.
The audits shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure is utilized to achieve the independent review and audit function and results of the audits activities shall be maintained and made available for inspection.
5.2      Records Retention Records and logs relative to the environmental aspects of previous plant operation and the handling and storage of spent fuel and maintenance of the facility shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to the NRC on request.
Records of modifications to facility structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the facility. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.
5.3      Changes in Environmental Protection Plan Requests for changes in the Environmental Protection Plan shall include an assessment of the environmental impacts of the proposed change and a supporting justification. Implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision to the Environmental Protection Plan. This EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.
5-1                Renewed License No. DPR-64 Amendment No. 270
 
5.4      Plant Reporting Requirements 5.4.1    Routine Reports An Annual Environmental Protection Plan Report describing implementation of this EPP for the previous year shall be submitted to the NRC prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following issuance of the operating license. The period of the first report shall begin with the date of issuance of the operating license.
The report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this Environmental Protection Plan for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous non-radiological environmental monitoring reports, and an assessment of the observed impacts of the previous plant operation and the handling and storage of spent fuel and maintenance of the facility on the environment. If harmful effects or evidence of trends towards irreversible damage to the environment are observed, ENO shall provide a detailed analysis of the data and a proposed course of action to alleviate the problem.
The Annual Environmental Protection Plan Report shall also include:
(a)      A list of EPP noncompliances and the corrective actions taken to remedy them.
(b)      A list of all changes in facility design or operation, tests, and experiments made in accordance with Subsection 3.1 which involved a potentially significant unreviewed environmental issue.
(c)      A list of nonroutine reports submitted in accordance with Subsection 5.4.2.
5-2                Renewed License No. DPR-64 Amendment No. 270
 
5.4.2    Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of a nonroutine event. The report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact and facility conditions, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.
Events reportable under this subsection which also require reports to other Federal, State, or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided a copy of such report at the same time it is submitted to the other agency.
5-4                Renewed License No. DPR-64 Amendment No. 270
 
APPENDIX B TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)
AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)
INDIAN POINT 3 NUCLEAR POWER PLANT ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART II: RADIOLOGICAL ENVIRONMENTAL FACILITY LICENSE NO. DPR-64 DOCKET NUMBER 50-286 Renewed License No. DPR-64 Amendment No. 270
 
APPENDIX C TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)
AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)
INDIAN POINT NUCLEAR GENERATING UNIT No. 3 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART I: SPENT FUEL TRANSFER CANISTER AND TRANSFER CASK SYSTEM FACILITY LICENSE NO. DPR-64 DOCKET NO. 50-286 Amendment No. 270
 
Facility License Appendix C - Inter-Unit Fuel Transfer Technical Specifications SPENT FUEL SHIELDED TRANSFER CANISTER AND TRANSFER CASK SYSTEM
 
==1.0 DESCRIPTION==
 
The spent fuel transfer system consists of the following components: (1) a spent fuel shielded transfer canister (STC), which contains the fuel; (2) a transfer cask (HI-TRAC 100D) (hereafter referred to as HI-TRAC), which contains the STC during transfer operations; and (3) a bottom missile shield.
The STC and HI-TRAC are designed to transfer irradiated nuclear fuel assemblies from the Indian Point 3 (IP3) spent fuel pit to the Indian Point 2 (IP2) spent fuel pit. A fuel basket within the STC holds the fuel assemblies and provides criticality control. The shielded transfer canister provides the confinement boundary, water retention boundary, gamma radiation shielding, and heat rejection capability. The HI-TRAC provides a water retention boundary, protection of the STC, gamma and neutron radiation shielding, and heat rejection capability.
The STC contains up to 12 fuel assemblies.
The STC is the confinement system for the fuel. It is a welded, multi-layer steel and lead cylinder with a welded base-plate and bolted lid. The inner shell of the canister forms an internal cylindrical cavity for housing the fuel basket. The outer surface of the canister inner shell is buttressed with lead and steel shells for radiation shielding. The minimum thickness of the steel, lead and steel shells relied upon for shielding starting with the innermost shell are 3/4 inch steel, 2 3/4 inch lead and 3/4 inch steel, respectively. The canister closure incorporates two O-ring seals to ensure its confinement function. The confinement system consists of the canister inner shell, bottom plate, top flange, top lid, top lid O-ring seals, vent port seal and cover plate, and drain port seal and coverplate. The fuel basket, for the transfer of 12 Pressurized Water Reactor (PWR) fuel assemblies, is a fully welded, stainless steel, honeycomb structure with neutron absorber panels attached to the individual storage cell walls under stainless steel sheathing. The maximum gross weight of the fully loaded STC is 40 tons.
The HI-TRAC is a multi-layer steel and lead cylinder with a bolted bottom (or pool) and top lid.
For the fuel transfer operation the HI-TRAC is fitted with a solid top lid, an STC centering assembly, and a bottom missile shield. The inner shell of the transfer cask forms an internal cylindrical cavity for housing the STC. The outer surface of the cask inner shell is buttressed with intermediate lead and steel shells for radiation shielding. The minimum thickness of the steel, lead and steel shells relied upon for shielding starting with the innermost shell are 3/4 inch steel, 2  inch lead and 1 inch steel, respectively. An outside shell called the water jacket contains water for neutron shielding, with a minimum thickness of 5. The HI-TRAC bottom and top lids incorporate a gasket seal design to ensure its water confinement function. The water confinement system consists of the HI-TRAC inner shell, bottom lid, top lid, top lid seal, bottom lid seal, vent port seal, vent port cap and bottom drain plug.
The HI-TRAC provides a water retention boundary, protection of the STC, gamma and neutron radiation shielding, and heat rejection capability. The bottom missile shield is attached to the bottom of the HI-TRAC and provides tornado missile protection of the pool lid bolted joint. The HI-TRAC can withstand a tornado missile in other areas without the need for additional shielding. The STC centering assembly provides STC position control within the HI-TRAC and also acts as an internal impact limiter in the event of a non-mechanistic tipover accident.
INDIAN POINT 3                                      1                              Amendment No. 270
 
Facility License Appendix C - Inter-Unit Fuel Transfer Technical Specifications 2.0 CONDITIONS 2.1 OPERATING PROCEDURES Written operating procedures shall be prepared for cask handling, loading, movement, surveillance, maintenance, and recovery from off normal conditions such as crane hang-up.
The written operating procedures shall be consistent with the technical basis described in Chapter 10 of the Licensing Report (Holtec International Report HI-2094289).
2.2 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM Written cask acceptance tests and maintenance program shall be prepared consistent with the technical basis described in Chapter 8 of the Licensing Report (Holtec International Report HI-2094289).
2.3 PRE-OPERATIONAL TESTING AND TRAINING EXERCISE A training exercise of the loading, closure, handling/transfer, and unloading, of the equipment shall be conducted prior to the first transfer. The training exercise shall not be conducted with irradiated fuel. The training exercise may be performed in an alternate step sequence from the actual procedures, but all steps must be performed. The training exercise shall include, but is not limited to the following:
a) Moving the STC into the IP3 spent fuel pool.
b) Preparation of the HI-TRAC for STC loading.
c) Selection and verification of specific fuel assemblies and non-fuel hardware to ensure type conformance.
d) Loading specific assemblies and placing assemblies into the STC (using a single dummy fuel assembly), including appropriate independent verification.
e) Remote installation of the STC lid and removal of the STC from the spent fuel pool.
f) Placement of the STC into the HI-TRAC with the STC centering assembly.
g) STC closure, establishment of STC water level with steam, verification of STC water level, STC leakage testing, and operational steps required prior to transfer, as applicable.
h) Establishment and verification of HI-TRAC water level.
i) Installation of the HI-TRAC top lid.
j) HI-TRAC closure, leakage testing, and operational steps required prior to transfer, as applicable.
k) Movement of the HI-TRAC with STC from the IP3 fuel handling building to the IP2 fuel handling building along the haul route with designated devices.
l) Moving the STC into the IP2 spent fuel pool.
m) Manual crane operations for bare STC movements including demonstration of recovery from a crane hang-up with the STC suspended from the crane.
INDIAN POINT 3                                    2                              Amendment No. 270
 
APPENDIX C TO FACILITY LICENSE FOR ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)
AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)
INDIAN POINT NUCLEAR GENERATING UNIT No. 3 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART II: TECHNICAL SPECIFICATIONS FACILITY LICENSE NO. DPR-64 DOCKET NO. 50-286 Amendment No. 270
 
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-64 ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.
INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286
 
==1.0      INTRODUCTION==
 
By application dated April 28, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20132A200), Entergy Nuclear Operations, Inc. (Entergy or the licensee), requested changes to Renewed Facility Operating License (RFOL) No. DPR-64 and the associated Technical Specifications (TSs) for Indian Point Nuclear Generating Unit No. 3 (IP3). Specifically, Entergy requested an amendment to revise the IP3 RFOL and the associated TSs to Permanently Defueled Technical Specifications (PDTS), consistent with the permanent cessation of operations and permanent removal of fuel from the reactor vessel.
 
==2.0      BACKGROUND==
 
By {{letter dated|date=February 8, 2017|text=letter dated February 8, 2017}} (ADAMS Accession No. ML17044A004), the licensee submitted a Notification of Permanent Cessation of Power Operations for Indian Point Nuclear Generating Unit No. 2 (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, 2024 and 2025, respectively.
Once the licensee submits to the NRC the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.82(a)(1)(i) and (ii), and the certifications are docketed, the 10 CFR Part 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel.
By {{letter dated|date=May 12, 2020|text=letter dated May 12, 2020}} (ADAMS Accession No. ML20133J902), Entergy certified 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 (SFP) on May 12, 2020.
Enclosure 2
 
By {{letter dated|date=April 28, 2020|text=letter dated April 28, 2020}} (ADAMS Accession No. ML20081J402), the NRC issued Amendment No. 294 to the IP2 operating license, which revised the IP2 RFOL and the associated TSs to PDTS, consistent with the permanent cessation of operations and permanent removal of fuel from the reactor vessel.1 By {{letter dated|date=April 15, 2020|text=letter dated April 15, 2020}} (ADAMS Accession No. ML20078L140), the NRC issued Amendment Nos. 293 and 268 for IP2 and IP3, respectively. This amendment revised the on-shift staffing and emergency response organization in the site emergency plan for the post-shutdown and permanently defueled condition.
By {{letter dated|date=April 10, 2020|text=letter dated April 10, 2020}} (ADAMS Accession No. ML20071Q717), the NRC issued Amendment Nos. 292 and 267 for IP2 and IP3, respectively. This amendment revised and removed certain requirements from the Section 5.0, Administrative Controls, portion of the IP2 and IP3 TSs that are not applicable to the facility in a permanently defueled condition, as well as revised and made editorial changes to the TS Table of Contents; Section 1.1, Definitions; and Section 4.0, Design Features.
By {{letter dated|date=December 18, 2019|text=letter dated December 18, 2019}} (ADAMS Accession No. ML19333B868), the NRC approved the certified fuel handler training and retraining program for IP2 and IP3.
 
==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 (SLs), 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.
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.
Section 50.36 of 10 CFR 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 focus on instrumentation to detect degradation of the reactor coolant system (RCS) pressure boundary and process variables; design features; operating restrictions; or structures, systems, and components (SSCs) that affect the integrity of fission product barriers during design-basis accidents (DBAs) or transients. They also focus on SSCs which operating experience or probabilistic risk assessment have shown to be significant to public health and safety. A general discussion of how these criteria were evaluated to ensure that the TS LCOs proposed for deletion will no longer be required to be included in the TSs is provided below.
1 Amendment No. 294 was corrected by {{letter dated|date=May 4, 2020|text=letter dated May 4, 2020}} (ADAMS Accession No. ML20122A262).
 
Criterion 1 of 10 CFR 50.36(c)(2)(ii)(A) states that TS LCOs must be established for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Since no fuel will be present in the reactor or RCS at the IP3 facility following permanent defueling, 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. The applicable DBAs for IP3 in the permanently defueled condition are the fuel handling accident (FHA) in the fuel handling building, accidental release-recycle of waste liquid, and the accidental release of waste gas, which are discussed in Section 4.0 of this safety evaluation (SE).
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 primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion) so that the plant response to DBAs and transients limits the consequences of these events to within the appropriate acceptance criteria.
Chapter 14, Safety Analysis, of the IP3 Updated Final Safety Analysis Report (UFSAR) describes the DBA scenarios that are applicable during plant operations (ADAMS Accession No. ML19282B159). After the docketed certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel are submitted to the NRC in accordance with 10 CFR 50.82(a)(1)(i) and (ii), the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). With the reactor in a permanently shutdown and defueled condition, the SFP and its cooling systems will be dedicated only to spent fuel storage. In this condition, the spectrum of credible accidents is much smaller than for an operational plant. Therefore, most of the accident scenarios postulated in Chapter 14 of the IP3 UFSAR will no longer be applicable after IP3 is in the permanently defueled condition. Note that, at Indian Point, the SFP may also be referred to as the spent fuel pit. The scope of applicable DBAs that continues to apply to IP3 is discussed in more detail in Section 4.0 of this SE.
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. There will no longer be any applicability of DBAs at IP3, once the reactor is in the permanently shutdown and defueled condition, that can result in a significant offsite radiological risk to public health and safety.
 
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.
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 SFP. 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 SFP, 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 UFSAR 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.
The regulation in 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance, states, in part:
(a)    As an aid in evaluating a proposed site, an applicant should assume a fission produce release[2] from the core, the expected demonstrable leak rate from the containment and the meteorological conditions pertinent to his site to derive an exclusion area, a low population zone and population center distance. For the purpose of this analysis, which shall set forth the basis for the numerical values used, the applicant should determine the following:
(1)    An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation 2 The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.
 
dose to the whole body in excess of 25 rem [roentgen equivalent man] [3] or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
(2)    A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.
The NRC staff evaluated the radiological consequences of the postulated FHA DBA against the dose criteria specified in 10 CFR 50.67, Accident source term, and using the guidance described in Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 (ADAMS Accession No. ML003716792). RG 1.183 provides guidance to licensees on acceptable application of alternative source term (AST) submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST.
By {{letter dated|date=March 24, 2005|text=letter dated March 24, 2005}} (ADAMS Accession No. ML050870383), the NRC issued Amendment No. 225 for IP3. This amendment revised the IP3 Operating License and TSs to increase the licensed rated thermal power by 4.85 percent from 3,067.4 megawatts thermal (MWt) to 3,216 MWt. As part of the IP3 power uprate project, a reanalysis of the several DBAs was performed, including the waste gas decay tank accident, which represents the current analysis of record.
By {{letter dated|date=March 22, 2005|text=letter dated March 22, 2005}} (ADAMS Accession No. ML050750431), the NRC issued Amendment No. 224 for IP3. This amendment approved the alternative source methodology for DBAs, including for the FHA for IP3, in accordance with 10 CFR 50.67 to perform the radiological consequences analyses of DBAs as described in RG 1.183. This approval included consideration of the aforementioned power uprate.
The FHA-specific dose acceptance criteria are specified in NUREG-0800, Revision 0, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR
[Light-Water Reactor] Edition (SRP), Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, dated July 2000 (ADAMS Accession No. ML003734190).
The dose acceptance criteria for the FHA are a total effective dose equivalent (TEDE) of 6.3 rem at the exclusion area boundary (EAB) for the worst 2 hours, 6.3 rem at the outer boundary of the low population zone (LPZ), and 5 rem in the control room for the duration of the accident.
3 The whole body dose of 25 rem referred to above corresponds numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP recommendations may be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5, 1959). However, neither its use nor that of the 300 rem value for thyroid exposure as set forth in these site criteria guides are intended to imply that these numbers constitute acceptable limits for emergency doses to the public under accident conditions. Rather, this 25 rem whole body value and the 300 rem thyroid value have been set forth in these guides as reference values, which can be used in the evaluation of reactor sites with respect to potential reactor accidents of exceedingly low probability of occurrence, and low risk of public exposure to radiation.
 
The regulations in 10 CFR 50.67(b)(2) state, in part, that the NRC may issue the amendment only if the applicants analysis demonstrates with reasonable assurance that:
(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv [Sievert]
(25 rem)4 total effective dose equivalent (TEDE).
(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).
(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.
Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC),
Criterion 19, Control room, states, in part:
A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.
Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.
The emergency planning requirements of 10 CFR 50.47, Emergency plans, and Appendix E to 10 CFR Part 50, Emergency Planning and Preparedness for Production and Utilization Facilities, continue to apply to a nuclear power reactor after permanent cessation of operations and removal of fuel from the reactor vessel. There are no explicit regulatory provisions distinguishing emergency planning requirements for a power reactor that has been permanently shut down from those for an operating power reactor. The NRC staff notes that the risk of an offsite radiological release is significantly lower, and the types of possible accidents are significantly fewer at a nuclear power reactor that has permanently ceased operations and removed fuel from the reactor vessel than at an operating power reactor.
Nuclear Energy Institute (NEI) topical report NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors, dated November 2012 (ADAMS 4
The use of 0.25 Sv (25 rem) TEDE is not intended to imply that this value constitutes an acceptable limit for emergency doses to the public under accident conditions. Rather, this 0.25 Sv (25 rem) TEDE value has been stated in this section as a reference value that can be used in the evaluation of proposed design-basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation.
 
Accession No. ML12326A805), provides guidance for the development of emergency action levels for reactors in a permanently defueled condition. NEI 99-01, Revision 6, was endorsed by the NRC in a {{letter dated|date=March 28, 2013|text=letter dated March 28, 2013}} (ADAMS Accession No. ML12346A463).
Revision 6 of NEI 99-01 states that the accident analysis necessary to adopt the permanently defueled emergency action level scheme must confirm that the source terms and release motive forces are not sufficient to warrant classification of a site area emergency or general emergency. A site area emergency would be declared for any events where exposure levels beyond the site area boundary are expected to exceed 10 percent of the Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs). The EPA PAG for sheltering or evacuation of the public is a projected dose of 1 to 5 rem total effective dose (TED5) in 4 days.
In addition, the EPA PAG for recommending the administration of potassium iodide (KI) (as a thyroid blocking agent) is a projected dose of 5 rem to the child thyroid from radioactive iodine.
Correspondingly, NEI 99-01 established the site area emergency classification threshold as 100 millirem (mrem) TEDE or 500 mrem thyroid committed dose equivalent.
RG 1.183 provides the methodology for analyzing the radiological consequences of several DBAs to show compliance with 10 CFR 50.67. RG 1.183 provides guidance to licensees on acceptable application of AST submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST.
SRP Section 15.0.1 provides review guidance to the NRC staff for the review of AST amendment requests. Section 15.0.1 states that the NRC reviewer should evaluate the proposed change against the guidance in RG 1.183. The dose acceptance criteria for the FHA are a TEDE of 6.3 rem at the EAB for the worst 2 hours, 6.3 rem at the outer boundary of the LPZ, and 5 rem in the control room for the duration of the accident.
Regulatory Issue Summary 2006-04, Experience with Implementation of Alternative Source Terms, dated March 7, 2006 (ADAMS Accession No. ML053460347), discusses experiences with analyzing an accident involving a release from off-gas or waste systems. As part of full AST implementation, some licensees have included an accident involving a release from their off-gas or waste gas system. For this type of accident, licensees have proposed acceptance criteria of 500 mrem TEDE. The acceptance criterion for this event is that associated with the dose to an individual member of the public as described in 10 CFR Part 20, Standards for Protection Against Radiation. When the NRC revised 10 CFR Part 20 to incorporate a TEDE dose, the offsite dose to an individual member of the public was changed from 500 mrem whole body to 100 mrem TEDE. Therefore, any licensee who chooses to implement AST for an off-gas or waste gas system release should base its acceptance criteria on 100 mrem TEDE.
Licensees may also choose not to implement AST for this accident and continue with their existing analysis and acceptance criteria of 500 mrem whole body.
Branch Technical Position 11-5, Postulated Radioactive Release Due to a Waste Gas System Leak or Failure, of SRP Chapter 11, Radioactive Waste Management, provides guidance to the reviewer for assessing the analysis of an accidental release from the waste gas system.
5 For the purposes of this SE, the terms TED and TEDE are used interchangeably as both describing the combined effects of internal and external radiation exposure.
 
3.3    Spent Fuel Pool Criticality GDC 61, Fuel storage and handling and radioactivity control, requires, in part, that these systems shall be designed with a capability to permit appropriate periodic inspection and testing of components important to safety.
GDC 62, Prevention of criticality in fuel storage and handling, requires that, Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.
Per 10 CFR 50.68(a), each holder of an operating license shall comply with either 10 CFR 70.24 or the requirements in 10 CFR 50.68(b). The licensee has elected to meet 10 CFR 50.68(b) and, accordingly, must comply with the following requirements:
(1) Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.
(2) The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.
(3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.
(4) If no credit for soluble boron is taken, the k-effective [estimated ratio of neutron production to neutron absorption and leakage] of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.
The regulations in 10 CFR 50.36(b) require TSs to be derived from the analyses and evaluation included in the safety analysis report and amendments thereto. As required by 10 CFR 50.36(c)(4), the TSs will include design 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 10 CFR 50.36.
 
==4.0    TECHNICAL EVALUATION==
 
4.1    Accident Analysis Chapter 14 of the IP3 UFSAR describes the postulated DBA and transient scenarios applicable to IP3 during power operations. They demonstrate that the plant could be operated safely and that radiological consequences from postulated accidents do not exceed the regulatory guidelines of 10 CFR 50.67 or 10 CFR Part 100, as applicable. Two basic groups of events are pertinent to safety, which are abnormal operational transients and postulated DBAs; these two groups were investigated separately. The analyses of the abnormal operational transients evaluate the ability of the plant protection features to ensure that during these transients, no fuel damage occurs, and the RCS pressure limit is not exceeded. The safety design limits require that damage to the fuel be limited and that no nuclear system process barrier damage results from any abnormal operational occurrence. Thus, analysis of this group of events evaluates the features that protect the first two radioactive material barriers. Analysis of the events in the second group, postulated DBAs, evaluates situations that require functioning of the engineered safeguards in order to protect the fission product barriers, including containment, in order to minimize the offsite radiological consequences.
The most severe postulated DBA involves damage to the nuclear reactor core and the release of large quantities of fission products. Many of these accident scenarios involve failures or malfunctions of systems, which could affect the fuel in the reactor vessel. With the termination of reactor operations and the permanent removal of fuel from the reactor vessel, such accidents are no longer possible. Therefore, the postulated accidents involving failure or malfunction of the reactor, reactor cooling system, steam system, or turbine generator are no longer applicable. While spent fuel remains in the SFP, the accidents that remain applicable to IP3 in the permanently shutdown and defueled condition are the FHA in the fuel handling building and the accidental release of waste liquid or waste gas. For completeness, the NRC staff also evaluated the applicability of other DBAs documented in the IP3 UFSAR to ensure that these accidents would not have consequences that could potentially exceed the 10 CFR 100.11 dose limits and RG 1.183 dose acceptance criteria.
4.1.1  Fuel Handling Accident As discussed in Section 3.2 of this SE, an analysis of the FHA utilizing the AST methodology described in RG 1.183 was previously approved by the NRC in License Amendment No. 224 on March 22, 2005. Later, as part of the IP3 power uprate project, a reanalysis of the FHA was performed utilizing the AST methodology, which is currently the analysis of record as presented in Section 14.2.1, Fuel Handling Accidents, of the IP3 UFSAR.
Section 14.2.1 of the IP3 UFSAR evaluates the various fuel handling accident scenarios. After the reactor has been completely defueled following permanent shut down, an FHA in the reactor cavity is no longer a credible accident. The DBA FHA in the SFP is applicable when IP3 is in a permanently shutdown and defueled condition. The licensees analysis applied the AST methodology outlined in RG 1.183, with the exception of fission product gap activity fractions because the licensee could not show that all fuel would meet the burnup conditions in footnote 11 of RG 1.183. Therefore, the licensees analysis assumed fission product gap release fractions to be 30 percent for Kr-85, 12 percent for I-131, and 10 percent for other iodines and noble gases. These gap release fractions are consistent with RG 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling
 
Accident in the Fuel Handling and Storage Facility for Boiling Water and Pressurized Water Reactors, dated March 1972 (ADAMS Accession No. ML083300022), as modified by NUREG/CR-5009, Assessment of the Use of Extended Burnup Fuel in Light Water Reactors, dated February 1988, which was approved in License Amendment No. 215.6 The analysis was performed to determine the radiological consequences to operators in the control room and the public at the EAB and LPZ as a function of time after shutdown. The analysis demonstrates that the dose consequences of the FHA will remain within the licensing basis dose limits, provided the SFP water level requirements of TS LCO 3.7.14 are met, without crediting fuel storage building ventilation, the station vent radiation monitors, control room isolation, and control room filtration assuming 84 hours of decay time following shutdown. In addition, after a decay time of at least 720 hours (30 days) prior to fuel movement, the analysis of the FHA results in an EAB TEDE dose of 0.47 rem.
The FHA is defined as the dropping of a single spent fuel assembly in the SFP during fuel handling activities, such that all 204 fuel rods in the assembly suffer mechanical damage to the cladding. The assembly activity inventory is based on the at-power core average power level with an additional radial peaking factor of 1.7 times. The gap activity in the damaged rods is instantaneously released into the SFP. The release occurs under 23 feet of water, which acts as a filter. Proposed TS LCO 3.7.14 will ensure the minimum water level in the SFP is established prior to fuel handling and maintained during fuel handling operations. The activity released is assumed to be reach the environment outside the building within 2 hours. As mentioned above, no credit is taken for removal of iodine by filters nor is credit taken for isolation of the release path.
The analysis concludes that without crediting mitigation by any active SSC, the calculated TEDE values to the control room, EAB, and LPZ are less than the limits set forth in 10 CFR 50.67 and RG 1.183. In addition, after a decay time of at least 720 hours (30 days) prior to fuel movement, the analysis of the FHA results in an EAB TEDE dose of 0.47 rem, which is less than the EPA PAGs recommended threshold for evacuation of 1 rem.
In performing this review, the NRC staff relied upon information provided by the licensee and NRC staff experience in performing similar reviews. The NRC staff concludes that the dose consequence from an FHA for the permanently defueled IP3 meets the applicable radiological dose criteria at the EAB, LPZ, and in the control room.
4.1.2    Accidental Release of Waste Gas Section 14.2.3, Accidental Release - Waste Gas, of the IP3 UFSAR evaluates the accidental release of waste gas. Tanks accumulating significant quantities of radioactive gases during operation are the volume control tanks, the liquid holdup tanks, and the gas decay tanks. The volume control tank accumulates gases over a core cycle by stripping action of the entering spray. During a refueling shutdown, this activity is vented to the waste gas system and stored for decay. The liquid holdup tanks receive reactor coolant after passing through demineralizers during the process of coolant purification.
The volume control tank and liquid holdup tanks were not considered in this analysis since gaseous products from these liquid tanks are collected and compressed in the waste gas decay tanks for decay prior to release. Potential liquid waste releases are considered from these tanks; however, any liquid releases are retained in the building or sumps, and only volatilized 6 Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Selective Adoption of Alternate Source Term (TAC No. MB5382), dated March 17, 2003 (ADAMS Accession No. ML030760135).
 
components would be released to the environment. These volatilized components are evaluated as part of the waste gas decay tank accident.
The waste gas tank accident assumes an instantaneous rupture with a conservative ground-level release all of noble gases. The 50,000 curie (Ci) dose equivalent Xenon (Xe)-133 waste gas tank activity assumed in this calculation bounds the current Xe-133 dose-equivalent limit of 29,761 Ci, as well as the administrative Xe-133 dose-equivalent limit of 6,000 Ci.
Once the reactor is permanently shut down and defueled, there is no mechanism to raise the primary coolant activity. Therefore, upon permanent shutdown and cooldown, the source term contained within the waste gas tanks represents the worst case source term, which is expected to be less than the assumed waste gas tank rupture analysis of record, and thus is bounded.
Subsequent additions to the waste gas tanks resulting from water management activities would be less than the final shutdown and cooldown waste gas tank source term.
The analysis concludes that without crediting mitigation by any active SSC, the calculated TEDE to the control room is less than the limit set forth in 10 CFR 50.67 and the whole-body dose value of 500 mrem at the EAB and LPZ.
In performing this review, the NRC staff relied upon information provided by the licensee and NRC staff experience in performing similar reviews, including a similar amendment for IP3. The NRC staff concludes that the dose consequence from a waste gas tank release for the permanently defueled IP3 meets the applicable radiological dose criteria at the control room, EAB, and LPZ.
4.1.3    Accidental Release of Waste Liquid The accidental release of liquid waste is discussed in Section 14.2.2, Accidental Release of Waste Liquid, of the IP3 UFSAR, which concludes that a potential liquid waste release collects in building sumps or is retained in building vaults and not released to the environment. As such, the licensee states the hazard from these releases is derived only from any volatilized components. The volatilized components are what comprise the waste gas accident discussed in the previous section of this SE. The licensee states that a separate liquid-specific release accident evaluation is not required to be performed regarding removal of supporting systems such as primary auxiliary building ventilation, station vent radiation monitors, control room isolation, and control room filtration. Based on the above, the NRC staff concludes that the dose consequence from a liquid waste incident for IP3 in the permanently shutdown and defueled condition will not approach the EPA PAGs for sheltering or evacuation and would not trigger the declaration of a site area emergency.
4.1.4    Accident Analysis Conclusions The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological impacts of the proposed changes. The staff finds that the licensees proposed changes use analysis methods and assumptions consistent with the guidance contained in RG 1.183. The staff compared the doses estimated by the licensee to the applicable criteria and to the results of confirmatory analyses performed by the staff. The staff finds that there is reasonable assurance that IP3, following adoption of the proposed amendment, will continue to provide sufficient safety margins with adequate defense in depth to address unanticipated events and to compensate for uncertainties in accident progression and in analysis assumptions and parameters. The staff concludes that the licensee has demonstrated that the dose
 
consequences for postulated accidents at the permanently shutdown and defueled plant would not have consequences that could potentially exceed the 10 CFR 50.67 dose limits and RG 1.183 dose acceptance criteria or approach the EPA PAG criteria of 1 rem TED after a 30-day fuel decay period prior to fuel movement. Therefore, the staff finds the proposed changes to be acceptable from a dose consequence perspective.
4.2    Boral Neutron Absorber Surveillance Program The spent fuel storage racks (as described in the IP2 and IP3 License Renewal Application (LRA), Appendix B, Section 2.3.3.1, Spent Fuel Pit Cooling, dated April 23, 2007 (ADAMS Accession No. ML071210517)) provide a storage location at the bottom of the spent fuel pit for spent fuel assemblies. The racks (auxiliary system according to NUREG-1801, Volume 2 Generic Aging Lessons Learned (GALL) Report: Tabulation of Results, dated September 2005 (ADAMS Accession No. ML052110006) are full length, top entry type. Spent fuel pit cooling racks are constructed of stainless steel with Boral as a neutron-absorbing material (NAM).
Reduction of neutron-absorbing capacity and loss of material due to general corrosion are aging effects requiring management for Boral spent fuel storage racks exposed to a treated borated water environment. These aging effects are managed by the Boral Surveillance Program (as described in the licensees LRA).
The Boral surveillance program at IP3, as described in Section A.2.1.3, Boral Surveillance Program, of the IP3 UFSAR Appendix A (ADAMS Accession No. ML19282B087) and in its responses to Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools" (GL 2016-01) dated November 3, 2016 (ADAMS Accession No. ML16314E266), and May 30, 2018 (ADAMS Accession No. ML18151A858), consists of a coupon surveillance program. The coupons were taken from the same lots of material used in construction of the racks and encased in a similar manner as the in-service material. The coupons are thus able to detect aging/degradation mechanisms that the in-service materials experience. Boral inspection and testing activities are conducted at a frequency of at least once every 10 years.
The continued performance of Boral as a neutron absorbing material in the SFP is managed by the aging management programs; Boral Surveillance Program as supplemented by the Water Chemistry Control - Primary and Secondary Program.
Specifically, as stated in its November 3, 2016, response letter to GL 2016-01, the Boral panels that are used are a composite material made of boron carbide and aluminum in three distinct layers. The outer layers are 0.0125 inches thick aluminum cladding (type 1100 alloy aluminum).
The center layer is a vented 0.05-inch-thick uniform aggregate of boron carbide particles held in an aluminum alloy matrix. Vented means that it is clad on the front and back only, while the sides are unclad to allow for gas migration and release. These strips of Boral are scalloped such that they may be held in place by welded coins and are not themselves welded. Strips of Boral poison are retained on the outer sides of the rack modules by a thin sheet of stainless steel. The individual SFP cells are fabricated from American Society for Testing and Materials ASTM A-240, Type 304 stainless steel.
The NRC staff verified that the existing Boral Surveillance Program is not changed by the proposed PDTS license amendment and that the inspection frequency and acceptance criteria will remain in place until such time as there is no longer fuel in the IP3 SFP. The NRC staff confirmed with the licensee that the IP3 Boral Surveillance Program as described in Appendix A of the IP3 UFSAR, will be retained in the Defueled Safety Analysis Report (DSAR) which will be
 
the document that replaces the UFSAR and will remain applicable to IP3 once it is shut down and defueled (ADAMS Accession No. ML21073A002).
During the period in which IP3 has ceased operations and is permanently defueled, there may be spent fuel in the SFP which requires criticality control. In order to ensure 10 CFR 50.68 is met, in part, the condition of the NAM is monitored to ensure that it does not degrade below the minimum Boron-10 areal density that the licensee assumes in its criticality safety analysis.
As stated above, LRA Appendix A, Section A.3.1.3, and the licensees GL 2016-01 response (See ADAMS accession numbers at Section 3.1) describes the licensees NAM monitoring program. The program contains provisions for routine monitoring and corrective actions to ensure the Boral can meet its neutron attenuation capability as assumed in the SFP criticality safety analysis.
The Boral Surveillance Program is an existing program that provides assurance the Boral neutron absorbers in the spent fuel racks maintain the validity of the criticality analysis in support of the rack design. The program relies on representative coupon samples mounted in surveillance assemblies located in the SFP to monitor performance of the absorber material without disrupting the integrity of the storage system. Coupon samples mounted in surveillance assemblies are removed from the SFP on a prescribed schedule and physical and chemical properties are measured. From these data, the stability and integrity of the Boral in the storage cells are assessed.
The licensees NAM monitoring program was incorporated into the IP3 UFSAR as approved in Supplement 2 to NUREG-1930, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Plant, Units 2 and 3, Appendix A, dated July 2015 (ADAMS Accession No. ML15188A383). The proposed IP3 PDTS amendment did not propose any revisions to the NAM monitoring program.
4.2.1    NAM Monitoring Program The NRC staff reviewed the IP3 NAM monitoring program, as there will be spent fuel in the SFP after IP3 permanently ceases operations and its reactor is defueled. The staff noted that the NAM monitoring program contains provisions to ensure the Boral does not degrade and can continue to perform its neutron attenuation function as described in the Section A.2.1.3 of the IP3 UFSAR Appendix A.
The NRC staff finds the NAM monitoring program is acceptable because the continued performance of Boral (as described in LRA Section 3.3.2.1.1) as a neutron-absorbing material in the SFP is supplemented by the Water Chemistry Control - Primary and Secondary Program, which manages the degradation of Boral including the reduction of neutron absorbing capacity consistent with NUREG-1801 material, environment, and aging effects.
The licensee has not proposed any changes to the NAM monitoring program found in LRA. The staff finds this acceptable as the NAM monitoring program was previously reviewed and approved in the LRA. The program provides reasonable assurance that the licensee will continue to monitor the condition of the NAM and take appropriate corrective actions if it degrades.
 
4.2.2    NAM Monitoring Program Conclusion The NRC staff reviewed the portion of the LAR related to the IP3 NAM monitoring program and has determined that the monitoring program as described in the IP3 LRA provides reasonable assurance that the licensee will be able to detect degradation of the neutron-absorbing material before its ability to perform its intended safety function is impacted. On this basis, the staff has concluded that the continuation of the NAM monitoring program and the contents of the program after permanent cessation of reactor operations, meet the applicable requirements of 10 CFR 50.68 and GDCs 61 and 62, and are therefore, acceptable.
4.3      Proposed Changes to the Renewed Facility Operating License 4.3.1    License Title The current license title is Renewed Facility Operating License.
The licensee proposed to delete Operating from the title, so that it reads Renewed Facility License.
The proposed change to the title to delete Operating would provide a more accurate description of the facility during the permanently shutdown and defueled condition. Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable to the NRC staff.
4.3.2    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, as amended, 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. Since, consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize use of the facility for power operation or emplacement or retention of fuel into the reactor vessel, this change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.
 
4.3.3  License Condition 2 Currently, License Condition 2 reads:
Accordingly, Renewed Facility Operating License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows:
The licensee proposed License Condition 2 to read:
Accordingly, Renewed Facility License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows:
Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. The removal of the discussion of Operating license would provide accuracy in the 10 CFR Part 50 license description. Therefore, the NRC staff approves the proposed change to License Condition 2.
4.3.4  License Condition 2.A Currently, License Condition 2.A reads:
This renewed license applies to the Indian Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and associated equipment (the facility), owned by ENIP3 and operated by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the Final Facility Description and Safety Analysis Report, as supplemented and amended, and the Environmental Report, as amended.
The licensee proposed License Condition 2.A to read:
This renewed license applies to the Indian Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and associated equipment (the facility), owned by ENIP3 and maintained by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the Defueled Safety Analysis Report, as supplemented and amended, and the Environmental Report, as amended.
Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. The removal of the discussion of Operating license would provide accuracy in the 10 CFR Part 50 license description. In addition, it is modified to reflect that a Defueled Safety Analysis Report will be prepared to address the permanently shutdown and defueled condition. Therefore, the NRC staff approves the proposed change to License Condition 2.A.
4.3.5  License Condition 2.B.(1)
Currently, License Condition 2.B.(1) reads:
Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) ENIP3 to possess and use, and (b) ENO to possess, use and operate, the facility at the designated location in
 
Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; The licensee proposed License Condition 2.B.(1) to read:
Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) ENIP3 to possess and use, and (b) ENO to possess and use the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. The facility would remain authorized to possess the existing spent fuel and use the systems required to support safe fuel storage (e.g., the SFP) during the decommissioning period in accordance with the specified limitations for storage. The removal of the discussion of operating would provide accuracy in the 10 CFR Part 50 license description. Therefore, the NRC staff approves the proposed change to License Condition 2.B.(1).
4.3.6    License Condition 2.B.(2)
Currently, License Condition 2.B.(2) reads as follows:
ENO pursuant to the Act and 10 CFR Part 70, to receive, possess, and use, at any time, special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Facility Description and Safety Analysis Report, as supplemented and amended; The licensee proposed License Condition 2.B.(2) to read:
ENO pursuant to the Act and 10 CFR Part 70, to possess, at any time, special nuclear material that was used as reactor fuel, in accordance with the limitations for storage, as described in the Defueled Safety Analysis Report, as supplemented and amended; The proposed change to this license condition would remove the authorization for receipt and use of special nuclear material (SNM) as reactor fuel. It would eliminate the reference to use of the SNM for reactor operations and limit the possession of SNM to SNM that was used as reactor fuel. Pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for IP3 will no longer authorize operation of the reactor. As such, IP3 has no need to receive SNM in the form of reactor fuel and cannot use SNM as reactor fuel for reactor operations. The continued authorization to possess SNM that was used as reactor fuel is necessary, as IP3 possesses reactor fuel that was used for past operations. Therefore, the NRC staff approves the proposed change to License Condition 2.B.(2).
 
4.3.7    License Condition 2.B.(3)
Currently, License Condition 2.B.(3) reads as follows:
ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; The licensee proposed License Condition 2.B.(3) to read:
ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct source and special nuclear material as sealed neutron sources that were used for reactor startup, sealed sources that were used for calibration of reactor instrumentation and are used in the calibration of radiation monitoring equipment, and that were used as fission detectors in amounts as required; The proposed change to this license condition revises the IP3 authorization to receive, possess and use any byproduct, source and special nuclear material as sealed neutron sources to clarify that these sealed neutron sources previously were used for reactor startup, calibration of reactor instrumentation, and as fission detectors. The proposed license condition is also modified to reflect that the use of neutron sources for calibration of radiation monitoring equipment will continue to be authorized. Since the IP3 license will no longer authorize operation of the facility pursuant to 10 CFR 50.82(a)(2), this revised license condition is consistent with the requirements associated with the decommissioning plant. Therefore, the NRC staff approves the proposed change to License Condition 2.B.(3).
4.3.8    License Condition 2.B.(4)
Currently, License Condition 2.B.(4) reads as follows:
ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components.
The licensee proposed License Condition 2.B.(4) to read:
ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components; The proposed change to this license condition is a grammatical correction by replacing the ending period with a semi-colon. The change is editorial and does not change any technical requirement. The NRC staff approves the proposed change to License Condition 2.B.(3).
 
4.3.9    License Condition 2.B.(5)
Currently, License Condition 2.B.(5) reads as follows:
ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.
The licensee proposed License Condition 2.B.(5) to read:
ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials that were produced by the operation of the facility.
This license condition is proposed for revision to allow possession, but not separation, of byproduct, and SNM that were produced by the operation of the facility, as opposed to those materials as may be produced by the operation of the facility. Since the IP3 license will no longer authorize operation of the facility pursuant to 10 CFR 50.82(a)(2), this license condition is consistent with the requirements associated with a permanently shutdown and defueled facility in the process of decommissioning. Therefore, the NRC staff finds the proposed change to License Condition 2.B.(5) appropriate and acceptable.
4.3.10    License Condition 2.C.(1)
Currently, License Condition 2.C.(1) reads:
Maximum Power Level ENO is authorized to operate the facility at steady state reactor core power levels not in excess of 3216 megawatts thermal (100% of rated power).
The licensee proposed to delete License Condition 2.C.(1). Since Entergy docketed the IP3 certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, reference to operation of the facility would be inconsistent with 10 CFR 50.82(a)(2).
The NRC staff reviewed the proposed deletion of License Condition 2.C.(1) and determined that operation would not be authorized at IP3 at any power level since its 10 CFR 50.82(a)(1) certifications were docketed. Therefore, the NRC staff finds the proposed deletion of License Condition 2.C.(1) acceptable.
4.3.11      License Condition 2.C.(2)
Currently, License Condition 2.C.(2) reads:
Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 269, are hereby incorporated in the Renewed License.
ENO shall operate the facility in accordance with the Technical Specifications.
 
The licensee proposed License Condition 2.C.(2) to read:
Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 270, are hereby incorporated in the Renewed License.
ENO shall maintain the facility in accordance with the Technical Specifications.
Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. This license condition is proposed for revision to account for the permanently defueled condition of the facility and to incorporate the permanently defueled TSs. The license condition is changed from operate the facility to maintain the facility, which describes the permanently defueled condition in which the IP3 license will no longer authorize the use of the facility for power operation. Therefore, the NRC staff finds the proposed change to License Condition 2.C.(2) acceptable.
4.3.12      License Condition 2.C.(3) and 2.C.(4)
Currently, License Conditions 2.C.(3) and 2.C.(4) both read: (DELETED).
License Condition 2.C.(3) and 2.C.(4) are historical references to deleted license conditions.
The licensee proposed to delete these license conditions in their entirety. The change to the license conditions is administrative and does not change any technical requirement. The NRC staff approves the proposed change to delete License Conditions 2.C.(3) and 2.C.(4) in their entirety.
4.3.13      License Condition 2.H Currently, License Condition 2.H reads:
ENO shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for Indian Point Nuclear Generating Unit No. 3 and as approved in NRC fire protection safety evaluations (SEs) dated September 21, 1973, March 6, 1979, May 2, 1980, November 18, 1982, December 30, 1982, February 2, 1984, April 16, 1984, January 7, 1987, September 9, 1988, October 21, 1991, April 20, 1994, January 5, 1995, and supplements thereto, subject to the following provision:
ENO may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
The licensee proposed to delete License Condition 2.H. The licensee stated that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the fire protection program (FPP) will be revised to take into account the facility conditions and activities during decommissioning. IP3 will continue to utilize the defense-in-depth concept, placing special emphasis on detection and suppression in order to minimize radiological releases to the environment. This license condition, which is based on maintaining an FPP at an operating reactor in accordance with 10 CFR 50.48 with the ability to achieve and maintain safe shutdown of the reactor in the event
 
of a fire, will no longer be applicable at IP3. However, many of the elements that are applicable for the operating plant FPP continue to be applicable during facility decommissioning. During the decommissioning process, an FPP is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. The regulation is applicable regardless of whether a requirement for an FPP is included in the facility license. Therefore, a license condition requiring such a program for a permanently shutdown and defueled facility is not necessary.
The NRC staff finds that License Condition 2.H for IP3 is based on maintaining FPPs that provide reasonable assurance of the ability to achieve and maintain safe shutdown in the event of a fire in accordance with 10 CFR 50.48. Achieving and maintaining safe shutdown in the event of a fire is no longer applicable to the decommissioned FPPs at IP3 once the facility is permanently shut down and the fuel has been permanently removed from the reactor.
However, elements of the FPP (e.g., License Condition 2.AC, Mitigation Strategy License Condition) continue during decommissioning to address fire events that could result in radiological hazards. The regulation in 10 CFR 50.48(f) requires IP3 to address the potential for fires, which could result in a radiological hazard. The NRC staff concludes that the rule, which requires an FPP for licenses that have submitted the certifications under 10 CFR 50.82(a)(1), is sufficient to ensure that a program is maintained. Therefore, a license condition that also requires FPPs for the permanently shutdown and defueled unit is redundant. Based on the above, the NRC staff concludes that reliance on 10 CFR 50.48(f) is appropriate and that the licensees request to delete License Condition 2.H is acceptable.
4.3.14      License Condition 2.O Currently, License Condition 2.O reads:
Evaluation, status and schedule for completion of balance of plant modifications as outlined in {{letter dated|date=February 12, 1983|text=letter dated February 12, 1983}}, shall be forwarded to the NRC by January 1, 1984.
License Condition 2.O is a reference to a historical obligation that was previously met. The licensee proposed to delete this license condition in its entirety. The proposed revision to License Condition 2.O is an administrative change and does not change any technical requirement. The NRC staff approves the proposed change to delete License Condition 2.O in its entirety.
4.3.15      License Condition 2.AA Currently, License Condition 2.AA reads:
The following conditions relate to the amendment approving the conversion to Improved Standard Technical Specifications:
: 1.      This amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee-controlled documentsThe relocation of requirements and detailed information shall be completed on or before the implementation of this amendment.
: 2.      The following is a schedule for implementing surveillance requirements (SRs):
For SRs that are new in this amendment, the first performance is due at the end of the first surveillance interval that begins on the date of the implementation of this amendment.
For SRs that existed prior to the amendment whose intervals of performance are being extended, the first extended surveillance interval begins upon completion of the last surveillance performed prior to the date of implementation of this amendment.
The licensee proposed to delete this license condition in its entirety. This is a historical license condition, the activities of which were completed in accordance with a license condition that is no longer applicable at IP3 as a permanently shutdown and defueled facility. Therefore, the NRC staff finds the proposed deletion of License Condition 2.AA acceptable.
4.3.16    License Condition 2.AB Currently, License Condition 2.AB reads:
The following conditions relate to the amendment approving the conversion to Improved Standard Technical Specifications:
With the reactor critical, Entergy shall maintain the reactor coolant system cold leg at a temperature (Tcold) greater than or equal to 525 °F. Entergy shall maintain a record of the cumulative time that the plant is operated with the reactor critical while Tcold is below 525 °F. Upon determination by Entergy that the cumulative time of plant operation with the reactor critical while Tcold is below 525 °F has exceeded one (1) year, Entergy must:
(a) within one (1) month, inform the NRC, in writing, and (b) within six (6) months submit the results of an analysis of the impact of the operation with Tcold below 525 °F on the pressurized thermal shock reference temperature (RTpts).
Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. This license condition, which is based on operations with the reactor critical, will no longer be applicable at IP3 as a permanently shutdown and defueled facility. Based on its review of the proposed deletion, the NRC staff concludes that continued implementation of License Condition 2.AB will no longer be applicable for IP3 because power operation will no longer be authorized once the 10 CFR 50.82(a)(1) certifications have been docketed. Therefore, the NRC staff finds the deletion of License Condition 2.AB acceptable.
 
4.3.17    License Condition 2.AD Currently, License Condition 2.AD reads:
Upon implementation of Amendment No. 239 adopting TSTF-448, Revision 3 (as supplemented), the determination of control room envelope (CRE) unfiltered air inleakage as required by Technical Specification (TS) Surveillance Requirement (SR) 3. 7.11.4, in accordance with TS 5.5.16.c.(i), the assessment of CRE habitability as required by TS 5.5.16.c.(ii), and the measurement of CRE pressure as required by TS 5.5.16.d, shall be considered met. Following implementation:
(a) The first performance of SR 3.7.11.4, in accordance with TS 5.5.16.c.(i),
shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measure from February 1, 2005, the date of the most recent successful tracer gas test, as stated in the {{letter dated|date=June 28, 2005|text=June 28, 2005, letter}} response to Generic Letter 2003-01.
(b) The first performance of the periodic assessment of CRE habitability, TS 5.5.16.c.(ii), shall be within the next 9 months since the time period since the most recent successful tracer gas test is greater than 3 years.
(c) The first performance of the periodic measurement of CRE pressure, TS 5.5.16.d, shall be within 24 months, plus the 182 days allowed by SR 3.0.2, as measured from January 4, 2007, the date of the most recent successful pressure measurement test.
The licensee proposed to delete License Condition 2.AD. The proposed change would remove the requirements of Technical Specifications Task Force (TSTF) Traveler TSTF-448, Control Room Habitability, which involves assessing the CRE habitability at the frequencies specified in Sections C.1 and C.2 of RG 1.197, Revision 0, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, dated May 2003 (ADAMS Accession No. ML031490664). The licensee proposed to not retain this license condition in the proposed PDTS, as it is a historical license condition, and the associated test, assessment, and measurement of the defined TSTF-448 requirements were completed in accordance with the schedule specified in the license condition. The NRC staff finds the deletion of License Condition 2.AD acceptable.
4.3.18    License Condition 2.AF.(2).c Currently, License Conditions 2.AF.(2).c reads:
The licensee shall notify the NRC in writing within 30 days after having accomplished item (2)a above and include the status of those activities that have been or remain to be completed in item (2)b above.
The licensee proposed to delete this license condition in its entirety. This is a historical license condition regarding the licensees implementation of license renewal programs and enhancements (Item (2)a), the activities of which were completed in accordance with the schedule specified (item (2)b) in the license condition. The removal of this license condition is administrative. The NRC staff finds the proposed deletion of License Condition 2.AF.(2).c acceptable.
 
4.3.19      Deleted License Conditions with Historical Amendment References The licensee proposed to delete numerous historical license amendment number and date references for previously deleted license conditions in the Renewed Facility Operating License.
Specifically, the reference to previously approved license amendments listed on the right-hand margin of the pages for License Conditions 1.D, 1.E, 2.D, 2.E, 2.I through 2.N, and 2.P are proposed for deletion. This revision is administrative and does not change any technical requirement. The NRC staff approves the proposed change to delete the historical references for the aforementioned license conditions.
4.3.20      License Condition 3 Currently, License Condition 3 states:
This renewed license is effective as of the date of issuance, and shall expire at midnight April 30, 2025.
The licensee proposed License Condition 3 to read:
This renewed license is effective as of the date of issuance, and until the Commission notifies the licensee in writing that the license is terminated.
The proposed change would modify this license condition to reflect the permanently shutdown and defueled condition of the facility. Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel. The proposed change would revise License Condition 3 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 3. The current License Condition 3, which documents the date of the expiration of the RFOL, is no longer necessary for the permanently shutdown and defueled facility in the process of decommissioning. The revised License Condition 3 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 3 acceptable.
4.4      Changes to Appendix A, Technical Specification 4.4.1    Title Page The current title page states, in part:
FACILITY OPERATING LICENSE DPR-64 TECHNICAL SPECIFICATIONS AND BASES
 
The licensee proposed to change the title page to state, in part:
FACILITY LICENSE DPR-64 PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS AND BASES Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed, pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable to the NRC staff.
4.4.2    Table of Contents The licensee proposed to revise the Table of Contents to reflect proposed additions, deletions, and changes to the TSs, as described in Sections 4.4.3 through 4.4.7 of this SE, and as detailed in Attachment 1 to the LAR dated April 28, 2020. The changes to the Table of Contents are editorial and do not change any technical requirement. The NRC staff finds the changes to the Table of Contents acceptable.
4.4.3    TS Section 1.0, Definitions The licensee proposed maintaining the definitions for Certified Fuel Handler and Non-certified Operator, proposed for addition in its LAR dated April 15, 2019 (ADAMS Accession No. ML19105B236), which proposed changes to the Administrative Controls section of the TSs.
Additionally, the term and definition for Actions would be retained.
The licensee proposed deletion of the following definitions since the terms are not used in any PDTS LCO and do not apply to a facility in the permanently defueled condition. In addition to the definitions listed below, the licensee proposed deletion of Table 1.1-1, MODES. TS Table 1 1-1 provides as follows:
 
Table 1.1-1 MODES 4.4.3.1  Definitions Proposed for Deletion The licensee proposed to delete the following definitions in TS Section 1.0, Use and Application, which currently state:
ACTUATION LOGIC TEST An ACTUATION LOGIC TEST shall be the application of various simulated or actual input combinations in conjunction with each possible interlock logic state required for OPERABILITY of a logic circuit and the verification of the required logic output. The ACTUATION LOGIC TEST, as a minimum, shall include a continuity check of output devices.
AXIAL FLUX DIFFERENCE (AFD)
AFD shall be the difference in normalized flux signals between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, display, and trip functions. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the
 
next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.
CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.
CHANNEL OPERATIONAL TEST (COT)
A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.
CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS REPORT (COLR)
The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 5.6.5. Plant operation within these limits is addressed in individual Specifications.
DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present.
If a specific isotope is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT I-131 shall be performed using Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988.
DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as
 
the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE Equivalent XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.
La [MAXIMUM ALLOWABLE PRIMARY CONTAINMENT LEAKAGE RATE]
The maximum allowable primary containment leakage rate, La, shall be 0.1% of primary containment air weight per day at the calculated peak containment pressure (Pa).
LEAKAGE LEAKAGE shall be:
: a.      Identified LEAKAGE
: 1. LEAKAGE, such as that from pump seals or valve packing (except for leakage into closed systems and reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (Leakage into closed systems is leakage that can be accounted for and contained by a system not directly connected to the atmosphere.
Leakage past the pressurizer safety valve seats and leakage past the safety injection pressure isolation valves are examples of reactor coolant system leakage into closed systems.)
: 2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
: 3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
: b.      Unidentified LEAKAGE All LEAKAGE (except for leakage into closed systems and RCP seal water injection or leakoff) that is not identified LEAKAGE;
: c.      Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.
 
MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.
MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant loop temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.
OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).
PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:
: a.      Described in FSAR Chapter 13, Tests and Operations;
: b.      Authorized under the provisions of 10 CFR 50.59; or
: c.      Otherwise approved by the Nuclear Regulatory Commission.
QUADRANT POWER TILT RATIO (QPTR)
QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.
RATED THERMAL POWER (RTP)
RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3216 MWt.
 
SHUTDOWN MARGIN (SDM)
SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:
: a.      All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
: b.      In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.
SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each required slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated required testable actuation devices.
STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.
THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.
TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)
A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy.
The NRC staff examined the licensees proposal to maintain the definitions for Actions, Certified Fuel Handler and Non-certified Operator and reformat them on one page as the remaining TS definitions. The NRC staff finds that these terms appropriately apply to the safe storage and handling of spent fuel in the SFP and are to be retained in the PDTS.
 
The NRC staff reviewed the TS definitions proposed for deletion and concludes that all of the terms listed above are only meaningful to a reactor authorized to operate. Once IP3 submits the docketed certifications required by 10 CFR 50.82(a)(1), the 10 CFR Part 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds the deletion of these definitions and Table 1.1-1 associated with the MODES definition from the TSs acceptable.
4.4.3.2    TS 1.2, Logical Connectors TS Section 1.2, Logical Connectors, in the IP3 TSs explains the meaning of logical connectors. The current description for TS Section 1.2 states, in part:
PURPOSE Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies BACKGROUND When logical connectors are used to state a Condition, Completion Time, Surveillance, or Frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the Condition, Completion Time, Surveillance, or Frequency.
EXAMPLES The following examples illustrate the use of logical connectors.
Example 1.2-2 The licensee proposed to modify the TSs to reflect the logical connectors that continue to exist in the TSs. Specifically, the purpose and background sections are revised to remove the terms Conditions, Completion Times, and Frequencies. Also, the licensee proposed to revise the current examples by making them a singular example and deleting Example 1.2-2. The revised TS Section 1.2, Logical Connectors, will state the following:
PURPOSE Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Required Actions and Surveillances
 
BACKGROUND When logical connectors are used to state a Surveillance, only the first level of logic is used, and the logical connector is left justified with the statement of the Surveillance.
EXAMPLE The following example illustrates the use of logical connectors
[Example 1.2-2 is proposed for deletion]
The NRC staff examined the licensees proposed revisions to TS Section 1.2 and concludes that the revisions are administrative in nature and reflect the logical connectors used in TS 3.7.15, which is the only LCO that contain logical connectors in the PDTS. Therefore, the staff finds these revisions acceptable.
4.4.3.3    TS 1.3, Completion Times TS Section 1.3, Completion Times, in the IP3 TSs establishes the completion time convention and provides guidance for its use. The current background for TS Section 1.3 states the following:
Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit The licensee proposed to revise the background to replace the phrase operation of the unit with handling and storage of spent nuclear fuel. The revised background will state the following:
Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe handling and storage of spent nuclear fuel The current description for TS Section 1.3 states the following:
The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO.
Unless otherwise specified, ...
Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability.
If situations are discovered
 
The licensee proposed to revise the description to remove the terms inoperable equipment or and MODE, replace the term unit with facility, remove the entire discussion of this section starting from the phrase Unless otherwise specified and remove the entire subsequent discussion, starting from the phrase If situations are discovered. The revised description will state the following:
The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the facility is in a specified condition stated in the Applicability of the LCO.
Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the facility is not within the LCO Applicability.
The current examples for TS Section 1.3 state the following:
The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.
Example 1.3-1 Example 1.3-2 Example 1.3-3 Example 1.3-4 Example 1.3-5 Example 1.3-6 Example 1.3-7 The licensee proposed to revise the examples by making them a singular example, replacing the phrase types of Conditions and changing Conditions with Required Actions, revising Example 1.3-1, and deleting Examples 1.3-2 through 1.3-7. The revised example will state the following:
The following example illustrates the use of Completion Times with different Required Actions.
Example 1.3-1 is modified to address completion times as utilized by TS 3.7.15.
Accordingly, the description of Immediate Completion Time, which previously followed the Example 1.3-7, is retained.
The NRC staff examined the licensees proposal to modify the background, description, and examples of TS Section 1.3. Once IP3 permanently shuts down and defuels, (1) the primary mission will change from safe operation of the unit to the safe handling and storage of spent nuclear fuel, which is reflected in the revised background; (2) the license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, and therefore, will no longer have operability requirements for any equipment, which is reflected in the revised description; and (3) the only TSs with completion times in the PDTS will be TSs 3.7.14, 3.7.15, and 3.7.16, which are reflected in the revised example. The term Inoperable equipment refers
 
to equipment relied upon in the TS for an operating reactor but is not relied upon in the PDTS.
There are no systems or components required to be operable in the PDTS as they are not required to function to mitigate any of the remaining DBAs. The NRC staff reviewed the licensees basis for the revisions to TS Section 1.3 and concludes that the proposed changes are acceptable, as they reflect IP3s permanently shutdown and defueled condition, as well as the completion times that will continue to exist in the PDTS.
4.4.3.4    TS 1.4, Frequency TS Section 1.4, Frequency, in the IP3 TSs defines the proper use and application of frequency requirements. The current description for TS Section 1.4 states the following:
The specified Frequency is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The specified Frequency consists of the requirements of the Frequency column of each SR as well as certain Notes in the surveillance column that modify performance requirements.
Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only required when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.
The licensee proposed to revise the description to remove the entire subsequent discussion, starting from the phrase as well as certain Notes in the Surveillance column. The revised description will state the following:
The specified Frequency is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The specified Frequency consists of the requirements of the Frequency column of each SR.
The current examples for TS Section 1.4 state the following:
The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3.
Example 1.4-1 Example 1.4-2 Example 1.4-3 The licensee proposed to revise Example 1.4-1 and the descriptive paragraph below it and proposed to delete Examples 1.4-2 and 1.4-3.
 
The revised example will state the following:
The following example illustrates the type of Frequency statement that appears in the Technical Specifications (TS).
The NRC staff examined the licensees proposal to modify the description and examples of TS Section 1.4. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). The only SRs that will remain in the TSs are limited to those in TSs 3.7.14, 3.7.15, and 3.7.16. The revisions to TS Section 1.4 provide the rules of usage and an example that continue to be applicable to those TSs. Example 1.4-1 is modified to address an example of a Frequency that is utilized by TS 3.7.14. The NRC staff reviewed the licensees provided basis for these revisions and concludes that the revisions are acceptable, as they reflect the condition of the facility after shutdown and defueling occur.
4.4.4  TS Section 2.0, Safety Limits (SLs)
TS Section 2.0, Safety Limits (SLs), in the IP3 TSs contains SLs that were necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity from the reactor core and the RCS pursuant to 10 CFR 50.36(c)(1).
The licensee proposed to delete TS 2.0, Safety Limits (SLs); TS 2.1, Safety Limits; and TS 2.2, Safety Limit Violations, in their entirety, since the SLs do not apply to a reactor that is in a permanently defueled condition. The SLs established in TS Section 2.1 prevent overheating of the fuel and possible cladding perforation, which would result in the release of fission products to the reactor coolant, and protect the integrity of the RCS from over-pressurization, thereby preventing the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. Safety limit violations in TS Section 2.2 are values of various parameters for which automatic protective action is needed during normal operations or anticipated transients to prevent exceeding an SL.
Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). The specifications in TS Section 2.0 do not apply to the safe storage and handling of spent fuel in the SFP. The NRC staff reviewed the proposed deletion and finds that the SLs are not applicable to a reactor in a permanently defueled condition. Therefore, the NRC staff finds that the proposed deletions of TS Sections 2.0, 2.1, and 2.2, in their entirety, are acceptable.
4.4.5  TS Section 3.0, Limiting Condition for Operation (LCO) Applicability TS Section 3.0, Limiting Condition for Operation (LCO) Applicability, in the IP3 TSs contains LCOs, which specify the lowest functional capability or performance levels of equipment required for safe operation of the facility, and contain the general requirements applicable to all LCOs and apply at all times unless otherwise stated in the TSs.
The licensee proposed revisions to LCOs 3.0.1 and 3.0.2 and deletion of LCOs 3.0.3 through 3.0.8.
 
The current LCO 3.0.1 states the following:
LCOs shall be met during the MODES or other specified conditions in the Applicability, except as provided in LCO 3.0.2, LCO 3.0.7 and LCO 3.0.8.
The licensee proposed to revise LCO 3.0.1 to state the following:
LCOs shall be met during the specified conditions in the Applicability, except as provided in LCO 3.0.2.
The current LCO 3.0.2 states, in part, the following:
Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met, except as provided in LCO 3.0.5 and LCO 3.0.6.
The licensee proposed to revise LCO 3.0.2 to state, in part, the following:
Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met.
The NRC staff reviewed the proposed changes to LCO 3.0.1 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the plant or placing fuel in the reactor vessel, the references to modes are no longer applicable. In addition, the references to LCOs 3.0.7 and 3.0.8 pertain to special tests and operations required for an operating reactor and actions required for equipment, respectively, that will no longer be operating once IP3 permanently shuts down and defuels. Therefore, the NRC staff finds the changes acceptable.
The NRC staff reviewed the proposed changes to LCO 3.0.2 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. The references to LCOs 3.0.5 and 3.0.6 pertain to restoring equipment to service under administrative controls and allowing performance of SRs on equipment declared inoperable, respectively, on equipment that will no longer be operating once IP3 permanently shuts down and defuels. Therefore, the NRC staff finds the changes acceptable.
The NRC staff reviewed the proposed deletions of LCOs 3.0.3, 3.0.4, 3.0.5, 3.0.6, 3.0.7, and 3.0.8. These LCOs pertain to an operating reactor. Once IP3 permanently shuts down and defuels, these LCOs will no longer be applicable. Therefore, the NRC staff finds the deletions acceptable.
4.4.5.1  TS Section 3.0, Surveillance Requirement (SR) Applicability TS Section 3.0, Surveillance Requirement (SR) Applicability, in the IP3 TSs contains the general requirements applicable to all SRs and applies at all times unless otherwise stated in a TS.
 
The licensee proposed revisions to SRs 3.0.1, 3.0.2, 3.0.3, and 3.0.4. The current SR 3.0.1 states the following:
SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SRSurveillances do not have to be performed on inoperable equipment or variables outside specified limits.
The licensee proposed to revise SR 3.0.1 to state the following:
SRs shall be met during the specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SRSurveillances do not have to be performed on variables outside specified limits.
The current SR 3.0.2 states the following:
The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.
For Frequencies specified as once, the above interval extension does not apply.
If a Completion Time requires periodic performance on a once per . . . basis, the above Frequency extension applies to each performance after the initial performance.
Exceptions to this Specification are stated in the individual Specifications.
The licensee proposed to revise SR 3.0.2 to state the following:
The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance.
The current SR 3.0.3 states, in part, the following:
If it is discovered that a Surveillance was not performed within its specified performed within its specified Frequency, then compliance with the requirement to declare the requirement to declare the LCO not met may be delayed, from the time of discovery, discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. The delay period is only applicable when there is a reasonable expectation the surveillance will be met when performed. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.
 
The licensee proposed to revise SR 3.0.3 to state, in part, the following:
If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours and the risk impact shall be managed.
The current SR 3.0.4 states the following:
Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCOs Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.
This provision shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.
The licensee proposed to revise SR 3.0.4 to state the following:
Entry into a specified condition in the Applicability of an LCO shall only be made when the LCOs Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3.
The NRC staff reviewed the proposed changes to SRs 3.0.1, 3.0.2, 3.0.3, and 3.0.4 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the plant or placing fuel in the reactor vessel, the references to modes and the discussions about shutting down the unit are no longer applicable. Further, the NRC staff agrees that the statements to be deleted are no longer necessary because the defueled TSs do not contain frequencies of the type described in the statements being deleted. Therefore, the staff finds that the proposed changes to delete these references reflect the plant status and are appropriate and acceptable.
4.4.5.2    TS Section 3.1, Reactivity Control Systems TS Section 3.1, Reactivity Control Systems, in the IP3 TSs contains requirements to assure and verify operability of reactivity control systems. A description of each TS and the NRC staffs evaluation of each of the specifications proposed for deletion are as follows:
TS 3.1.1, Shutdown Margin (SDM), establishes shutdown margin as the minimum shutdown margin in the reactor core. The shutdown margin limits are specified in the Core Operating Limits Report (COLR). TS 3.1.1 is proposed for deletion and is applicable in MODE 2 with keff  1.0 and MODES 3 through 5. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, the TS will not be required because the operation in the
 
applicable modes and specified conditions will no longer occur. TS 3.1.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.1 acceptable.
TS 3.1.2, Core Reactivity, establishes that core reactivity be within +/- 1 percent k/k of predicted values. TS 3.1.2 is proposed for deletion and is applicable in MODES 1 and 2.
The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in MODES 1 and 2 will no longer occur and TS 3.1.2 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.2 acceptable.
TS 3.1.3, Moderator Temperature Coefficient (MTC), establishes that MTC be maintained within the limits specified in the COLR. The maximum upper limit shall be 0.0 k/k degrees Fahrenheit (°F) at hot zero power. TS 3.1.3 is proposed for deletion and is applicable in MODES 1 and 2 with keff  1.0 for the upper MTC limit and MODES 1, 2, and 3 for the lower MTC limit. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in the applicable modes and specified conditions will no longer occur. TS 3.1.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.3 acceptable.
TS 3.1.4, Rod Group Alignment Limits, establishes that all shutdown and control rods shall be operable, and the difference between each individual indicated rod position and its group step counter demand position shall be within the limits. TS 3.1.4, including Table 3.1.4-1, is proposed for deletion and is applicable in MODES 1 and 2. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in MODES 1 and 2 will no longer occur, and TS 3.1.4 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.4 acceptable.
TS 3.1.5, Shutdown Bank Insertion Limits, establishes that each shutdown bank shall be within insertion limits specified in the COLR. TS 3.1.5 is proposed for deletion and is applicable in MODES 1 and 2. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in MODES 1 and 2 will no longer occur, and TS 3.1.5 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.5 acceptable.
TS 3.1.6, Control Bank Insertion Limits, establishes that control banks shall be within the insertion, sequence, and overlap limits specified in the COLR. TS 3.1.6 is proposed for deletion and is applicable in MODES 1 and 2 with keff  1.0. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with
 
10 CFR 50.82(a)(2). Therefore, operation in the applicable modes and specified conditions will no longer occur. TS 3.1.6 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.6 acceptable.
TS 3.1.7, Rod Position Indication, establishes that the individual rod position indication system and the demand position indication shall be operable. TS 3.1.7 is proposed for deletion and is applicable in MODES 1 and 2. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in MODES 1 and 2 will no longer occur, and TS 3.1.7 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.7 acceptable.
TS 3.1.8, Physics Tests Exceptions - Mode 2, establishes that during the performance of physics tests, the requirements of Moderator Temperature Coefficient; Rod Group Alignment Limits; Shutdown Bank Insertion Limits; Control Bank Insertion Limits; and RCS Minimum Temperature for Criticality, may be suspended, provided that certain conditions as specified in LCO 3.1.8 are met. TS 3.1.8 is proposed for deletion and is applicable during physics tests initiated in MODE 2. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in the applicable mode and specified condition will no longer occur. TS 3.1.8 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.6 acceptable.
4.4.5.3    TS Section 3.2, Power Distribution Limits TS Section 3.2, Power Distribution Limits, in the IP3 TSs contains power distribution limits that provide assurance that fuel design criteria are not exceeded, and the accident analysis assumptions remain valid. A description of each of the specifications proposed for deletion is provided as follows:
TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)), establishes that FQ(Z) shall be within the limits specified in the COLR. As a result, this TS will not apply in the permanently defueled condition. TS 3.2.1 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (FNH), establishes that FNH shall be within the limits specified in the COLR. TS 3.2.2 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.2.3, Axial Flux Difference (AFD) (Constant Axial Offset Control (CAOC)
Methodology, establishes that the AFD shall be maintained within the target band about the target flux difference. The target band is specified in the COLR. AFD may deviate outside the target band with thermal power less than (<) 90 percent RTP but greater than or equal () 50 percent RTP, provided AFD is within the acceptable operation limits, and cumulative penalty deviation time is less than or equal to () 1 hour during the previous 24 hours. The acceptable operation limits are specified in the COLR. TS 3.2.3
 
does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.2.4, Quadrant Power Tilt Ratio (QPTR), establishes that the QPTR shall be 1.02. TS 3.2.4 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
The licensee proposed to delete TS Section 3.2 in its entirety. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2).
Therefore, the specifications addressed in TS Section 3.2 will not be required, and these requirements will not apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS Section 3.2 acceptable.
4.4.5.4  TS Section 3.3, Instrumentation TS Section 3.3, Instrumentation, in the IP3 TSs contains operability requirements for sensing and control instrumentation required for safe operation of the facility. Below are the specifications in TS Section 3.3.
TS 3.3.1, Reactor Protection System (RPS) Instrumentation, establishes that the RPS instrumentation for each function in Table 3.3.1-1 shall be operable.
TS 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, establishes that the ESFAS instrumentation for each function in Table 3.3.2-1 shall be operable.
TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation, establishes that the PAM instrumentation for each function in Table 3.3.3-1 shall be operable.
TS 3.3.4, Remote Shutdown, establishes that the remote shutdown functions shall be operable.
TS 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation, establishes that certain LOP DG start instrumentation conditions as specified in LCO 3.3.5 shall be operable.
TS 3.3.6, Containment Purge System and Pressure Relief Line Isolation Instrumentation, establishes that the containment purge system and pressure relief line isolation instrumentation for each function in Table 3.3.6-1 shall be operable.
TS 3.3.7, Control Room Ventilation System (CRVS) Actuation Instrumentation, establishes that the CRVS actuation instrumentation for each function in Table 3.3.7-1 shall be operable.
TS 3.3.8, Fuel Storage Building Emergency Ventilation System (FSBEVS) Actuation Instrumentation, establishes that the FSBEVS manual and automatic instrumentation shall be operable.
 
The licensee proposed to delete TS Section 3.3 in its entirety. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. TS 3.3.1 through 3.3.7 apply during operation of the reactor; TS 3.3.8 concerns the FSBEVS, and applies during movement of recently irradiated fuel in the Fuel Storage Building. As discussed below with respect to TS 3.3.17, the licensee stated that the Bases for TS 3.7.13, define recently irradiated fuel as fuel that has occupied part of a critical reactor within the previous 84 hours. The IP3 PDTS will not be implemented until after 84 hours following permanent shutdown, so TS 3.3.8 will no longer be needed once the PDTS have been implemented. Further, once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS Section 3.3 will not be required, and these requirements will not apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS Section 3.3 acceptable.
4.4.5.5    TS Section 3.4, Reactor Coolant System (RCS)
TS Section 3.4, Reactor Coolant System (RCS), in the IP3 TSs contains requirements that provide for appropriate control of process variables, design features, or operating restrictions needed for appropriate functional capability of RCS equipment required for safe operation of the facility. The licensee proposed to delete TS Section 3.4 in its entirety. A description of each of the specifications proposed for deletion is provided as follows:
TS 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits, establishes RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:
: a.      Pressurizer pressure is greater than or equal to the limit specified in the COLR;
: b.      RCS average temperature is less than or equal to the limit specified in the COLR; and
: c.      RCS total flow rate  354,400 gpm and greater than or equal to the limit specified in the COLR.
TS 3.4.1 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.2, RCS Minimum Temperature for Criticality, establishes that each RCS loop average temperature shall be  540 &deg;F. TS 3.4.2 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.3, RCS Pressure and Temperature (P/T) Limits, establishes that RCS pressure, temperature, heatup and cooldown rates shall be maintained within the limits specified.
TS 3.4.3 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.4, RCS Loops - Modes 1 and 2, establishes that four RCS loops shall be operable and in operation. TS 3.4.4 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
 
TS 3.4.5, RCS Loops - Mode 3, establishes that two RCS loops shall be operable and either two RCS loops shall be in operation when the rod control system is capable of rod withdrawal or one RCS loop shall be in operation when the rod control system is not capable of rod withdrawal. TS 3.4.5 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.6, RCS Loops - Mode 4, establishes that two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops shall be operable, and one loop shall be in operation. TS 3.4.6 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.7, RCS Loops - Mode 5, Loops Filled, establishes that one RHR loop shall be operable and in operation, and either one additional RHR loop shall be operable or the secondary side water level of at least two steam generators (SGs) shall be  71 percent wide range. TS 3.4.7 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.8, RCS Loops - Mode 5, Loops Not Filled, establishes that two RHR loops shall be operable, and one RHR loop shall be in operation. TS 3.4.8 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.9, Pressurizer, establishes that the pressurizer shall be operable with pressurizer water level  54.3 percent in MODES 1 and 2, or  90 percent in MODE 3, and two groups of pressurizer heaters operable with the capacity of each group 150 kilowatts (kW) and capable of being powered from an emergency power supply.
TS 3.4.9 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.10, Pressurizer Safety Valves, establishes that three pressurizer safety valves shall be operable with lift settings set  2,460 pounds per square inch gauge (psig) and 2,510 psig. TS 3.4.10 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.11, Pressurizer Power Operated Relief Valves (PORVs), establishes that each PORV and associated block valve shall be operable. TS 3.4.11 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.12, Low Temperature Overpressure Protection (LTOP), establishes that LTOP shall be operable with no high head safety injection pumps capable of injecting into the RCS and the accumulator discharge isolation valves closed and de-energized, and either of the specified options of LCOs 3.4.12(a) or 3.4.12(b) are met. TS 3.4.12 does not apply once the reactor is permanently defueled. The NRC staff also notes that these sections are specifically related to the requirements for the reactor coolant pressure boundary set forth in Appendices G and H to 10 CFR Part 50. Section 50.60(a) of 10 CFR stipulates that reactor facilities that have submitted the certifications required under 10 CFR 50.82(a)(1) no longer need to meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in Appendices G and H to 10 CFR Part 50. TS 3.4.12 will no longer be necessary for IP3 in accordance with 10 CFR 50.60(a) because these LCOs will not apply in a permanently defueled condition. Therefore, the TS is proposed to be deleted.
 
TS 3.4.13, RCS Operational Leakage, establishes that RCS operational leakage shall be limited to:
: a.      No pressure boundary leakage;
: b.      1 gpm unidentified leakage;
: c.      10 gpm identified leakage; and
: d.      150 gallons per day primary to secondary leakage through any one steam generator (SG).
TS 3.4.13 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.14, RCS Pressure Isolation Valve (PIV) Leakage, establishes that leakage from each RCS PIV shall be within limit. TS 3.4.14 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.15, RCS Leakage Detection Instrumentation, establishes that the following RCS leakage detection instrumentation shall be operable:
: a.      One containment sump discharge flow monitor;
: b.      One containment atmosphere radioactivity monitor (gaseous or particulate); and
: c.      One containment fan cooler unit condensate measuring system.
TS 3.4.15 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.16, RCS Specific Activity, establishes that RCS dose equivalent I-131 and dose equivalent Xe-133 specific activity shall be within limits. TS 3.4.16 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.4.17, Steam Generator (SG) Tube Integrity, establishes that SG tube integrity shall be maintained, and all SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the SG program. TS 3.4.17 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
As noted above, the licensee proposed to delete TS Section 3.4 in its entirety. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS Section 3.4 will not be required, and these requirements will not apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS Section 3.4 acceptable.
 
4.4.5.6    TS Section 3.5, Emergency Core Cooling Systems (ECCS)
TS Section 3.5, Emergency Core Cooling Systems (ECCS), in the IP3 TSs contains requirements that provide for appropriate functional capability of ECCS equipment required for mitigation of DBAs or transients to protect the integrity of a fission product barrier. The licensee proposed to delete TS Section 3.5 in its entirety. A description of each of the specifications proposed for deletion is provided as follows:
TS 3.5.1, Accumulators, establishes that four ECCS accumulators shall be operable.
TS 3.5.1 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.5.2, ECCS - Operating, establishes that three ECCS trains shall be operable in MODES 1 through 3. TS 3.5.2 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.5.3, ECCS - Shutdown, establishes that two ECCS high head safety injection subsystems and one ECCS RHR subsystem shall be operable. TS 3.5.3 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
TS 3.5.4, Refueling Water Storage Tank (RWST), establishes that the RWST and two channels of RWST low level alarm shall be operable. TS 3.5.4 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.
As noted above, the licensee proposed to delete TS Section 3.5 in its entirety. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS Section 3.5 will not be required and these requirements will not apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS Section 3.5 acceptable.
4.4.5.7    TS Section 3.6, Containment Systems TS Section 3.6, Containment Systems, in the IP3 TSs contain requirements that assure the integrity of the containment, depressurization and cooling systems, and containment isolation valves. The licensee proposed to delete TS Section 3.6 in its entirety. Below are the specifications proposed for deletion.
TS 3.6.1, Containment, establishes that containment shall be operable in MODES 1 through 4.
TS 3.6.2, Containment Air Locks, establishes that two containment air locks shall be operable in MODES 1 through 4.
TS 3.6.3, Containment Isolation Valves, establishes that each containment isolation valve shall be operable in MODES 1 through 4.
TS 3.6.4, Containment Pressure, establishes that containment pressure shall be maintained within the specified limits in MODES 1 through 4.
 
TS 3.6.5, Containment Air Temperature, establishes that containment average air temperature shall be > 50 &deg;F and  130 &deg;F in MODES 1 through 4.
TS 3.6.6, Containment Spray System and Containment Fan Cooler System, establishes that two trains of containment spray and three trains of containment fan cooler shall be operable in MODES 1 through 4.
TS 3.6.7, Recirculation pH Control System, establishes that the recirculation pH control system shall be operable in MODES 1 through 4.
TS 3.6.8, Not Used.
TS 3.6.9, Isolation Valve Seal Water (IVSW) System, establishes that the IVSW system shall be operable in MODES 1 through 4.
TS 3.6.10, Weld Channel and Penetration Pressurization System (WC&PPS),
establishes that the WC&PPS shall be operable in MODES 1 through 4.
As noted above, the licensee proposed to delete TS Section 3.6 in its entirety. These specifications do not apply in a defueled condition or for SSCs that are not needed for accident mitigation in the defueled condition. Once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS Section 3.6 will not be required because the operation in the applicable MODES and specified conditions will no longer occur. The NRC staff therefore finds the deletion of TS Section 3.6 acceptable.
4.4.5.8  TS Section 3.7, Plant Systems TS Section 3.7, Plant Systems, in the IP3 TSs provides requirements for the appropriate functional capability of plant equipment required for safe operation of the facility, including the plant being in a defueled condition. A description of each TS and the NRC staffs evaluation of each of the specifications proposed for deletion or modification is provided as follows:
TS 3.7, Plant Systems, is proposed to be renamed to Spent Fuel Pit Requirements.
The licensee states that this change is to reflect the remaining TSs in the section that deal with spent fuel pit requirements in a permanently shutdown and defueled facility.
The NRC staff has determined that the title change aligns with the specific remaining TS requirements for this section; therefore, the NRC staff finds this administrative change acceptable.
TS 3.7.1, Main Steam Safety Valves (MSSVs), including Tables 3.7.1-1 and 3.7.1-2, is proposed for deletion; this TS is applicable in MODES 1 through 3. The NRC staff has determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 3 will no longer occur, TS 3.7.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.1 acceptable.
 
TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Main Steam Check Valves (MSCVs), is proposed for deletion; this TS is applicable in MODE 1 and MODES 2 and 3, except when all MSIVs are closed. The NRC staff has determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 3 will no longer occur, TS 3.7.2 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.2 acceptable.
TS 3.7.3, Main Boiler Feedpump Discharge Valves (MBFPDVs), Main Feedwater Regulation Valves (MFRVs), Main Feedwater Inlet Isolation Valves (MFIIVs) and Main Feedwater (MF) Low Flow Bypass Valves, is proposed for deletion; this TS is applicable in MODES 1 through 3, except when each main feedwater and bypass line is isolated by a closed and deactivated motor/air operated valve or isolated by a closed manual valve.
After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 3 will no longer occur, TS 3.7.3 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.3 acceptable.
TS 3.7.4, Atmospheric Dump Valves (ADVs); TS 3.7.5; Auxiliary Feedwater (AFW)
System; TS 3.7.6, Condensate Storage Tank (CST); and TS 3.7.7, City Water (CW),
are applicable in MODES 1 through 3 and MODE 4 when the SG is relied upon for heat removal. All four specifications are proposed for deletion. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, the three specifications will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TSs 3.7.4, 3.7.5, 3.7.6, and 3.7.7 acceptable.
TS 3.7.8, Component Cooling Water (CCW) System; TS 3.7.9, Service Water System (SWS); and TS 3.7.10, Ultimate Heat Sink (UHS), are applicable in MODES 1 through 4 and are proposed for deletion. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, the three specifications will not be applicable in a permanently defueled condition.
The NRC staff therefore finds the deletion of TSs 3.7.8, 3.7.9, and 3.7.10 acceptable.
TS 3.7.11, Control Room Ventilation System (CRVS), is proposed for deletion; this TS is applicable in MODES 1 through 4 and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).
Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.7.11 will not be applicable in a permanently defueled condition. In addition, the licensee states that the Bases for TS 3.7.11, define recently irradiated fuel as fuel that has occupied part of a critical reactor within the previous 84 hours. The IP3 PDTS will not be implemented
 
until after 84 hours following permanent shutdown, so the specific condition of applicability will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition. Therefore, the NRC staff finds the deletion of TS 3.7.11 acceptable.
TS 3.7.12, Control Room Air Conditioning System (CRACS), is applicable in MODES 1 through 4 and is proposed for deletion. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.7.12 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.12 acceptable.
TS 3.7.13, Fuel Storage Building Emergency Ventilation System (FSBEVS), is proposed for deletion; this TS is applicable during movement of recently irradiated fuel assemblies in the fuel storage building. The licensee states that the Bases for TS 3.7.13, define recently irradiated fuel as fuel that has occupied part of a critical reactor within the previous 84 hours. The IP3 PDTS will not be implemented until after 84 hours following permanent shutdown, so the specific condition of applicability will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition. Therefore, the NRC staff finds the deletion of TS 3.7.13 acceptable.
TS 3.7.14, Spent Fuel Pit Water Level; TS 3.7.15, Spent Fuel Pit Boron Concentrations; and TS 3.7.16, Spent Fuel Assembly Storage, including Figure 3.7.16-1, are being retained in the proposed PDTS. The licensee proposed an administrative change replacing the TS Section 3.7 title Plant Systems with Spent Fuel Pit Requirements in the TS heading. Additionally, for each section, the note contained within Required Action A.1, which states LCO 3.0.3 is not applicable, is proposed for deletion to align with the proposed deletion of TS LCO 3.0.3. Therefore, the NRC staff finds the modifications of TS 3.7.14, TS 3.7.15, and 3.7.16 acceptable.
TS 3.7.17, Secondary Specific Activity, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.7.17 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.17 acceptable.
4.4.5.9    TS Section 3.8, Electrical Power Systems TS Section 3.8, Electrical Power Systems, in the IP3 TSs contains operability requirements that provide for appropriate functional capability of plant electrical equipment required for safe operation of the facility. The licensee proposes to delete TS Section 3.8 in its entirety, as these TSs will not apply to IP3 in the permanently defueled condition. A description of each TS and the NRC staffs evaluation of each of the specifications proposed for deletion is provided as follows:
TS 3.8.1, AC [Alternating Current] Sources - Operating, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer
 
authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.8.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.1 acceptable.
TS 3.8.2, AC Sources - Shutdown, is proposed for deletion; this TS is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the movement of recently irradiated fuel assemblies will not occur, and the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 5 and 6 and the movement of recently irradiated fuel assemblies will no longer occur, TS 3.8.2 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.2 acceptable.
TS 3.8.3, Diesel Fuel Oil and Starting Air, is proposed for deletion; this TS is applicable when the associated diesel generator is required to be operable. The operability requirements of the diesel generators are contained in TS 3.8.1 and TS 3.8.2. Since these two TSs are being proposed for deletion, TS 3.8.3 is not included in the proposed PDTS because the two TSs that it supports will no longer be required after IP3 is in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3.8.3 acceptable.
TS 3.8.4, DC Sources - Operating, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.8.4 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.4 acceptable.
TS 3.8.5, DC [Direct Current] Sources - Shutdown, is proposed for deletion; this TS is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).
Therefore, since operation in MODES 5 and 6 and the movement of recently irradiated fuel assemblies will no longer occur, TS 3.8.5 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.5 acceptable.
TS 3.8.6, Battery Cell Parameters, including Table 3.8.6-1, is proposed for deletion; this TS is applicable when the associated direct current (DC) electrical power subsystems are required to be operable. The operability requirements regarding the DC sources are contained in TS 3.8.4 and TS 3.8.5. Since these two TSs are being proposed for deletion, TS 3.8.6 is not included in the proposed PDTS because the two TSs that it supports will no longer be required after IP3 is in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3.8.6 acceptable.
TS 3.8.7, Inverters - Operating, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are
 
docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.8.7 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.7 acceptable.
TS 3.8.8, Inverters - Shutdown, is proposed for deletion; this TS is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 5 and 6 and movement of recently irradiated fuel assemblies will no longer occur, TS 3.8.8 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.8 acceptable.
TS 3.8.9, Distribution Systems - Operating, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.8.7 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.9 acceptable.
TS 3.8.10, Distribution Systems - Shutdown, is proposed for deletion; this TS is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).
Therefore, since operation in MODES 5 and 6 and movement of recently irradiated fuel assemblies will no longer occur, TS 3.8.10 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.10 acceptable.
4.4.5.10 TS Section 3.9, Refueling Operations TS Section 3.9, Refueling Operations, in the IP3 TSs contains requirements that provide for appropriate functional capability of parameters and equipment that are required for mitigation of DBAs during refueling operations (moving irradiated fuel to or from the reactor core). A description of each TS and the NRC staffs evaluation of each of the specifications proposed for deletion is provided as follows:
TS 3.9.1, Boron Concentration, establishes that boron concentrations of the RCS, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the COLR. TS 3.9.1 is proposed for deletion; this TS is applicable in MODE 6. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODE 6 will no longer occur, TS 3.9.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.9.1 acceptable.
TS 3.9.2, Nuclear Instrumentation, establishes that two source range neutron flux monitors shall be operable, and one source range audible count rate circuit shall be
 
operable. TS 3.9.2 is proposed for deletion; this TS is applicable in MODE 6. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODE 6 will no longer occur, TS 3.9.2 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.9.2 acceptable.
TS 3.9.3, Containment Penetrations, establishes requirements pertaining to movement of recently irradiated fuel assemblies within containment. TS 3.9.3 is proposed for deletion; this TS is applicable during movement of recently irradiated fuel assemblies within containment. The licensee states that the Bases for TS 3.9.3 define recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours. The PDTS will not be implemented until after 84 hours following permanent shutdown, so the specific condition of applicability will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition. Therefore, the NRC staff finds the deletion of TS 3.9.3 acceptable.
TS 3.9.4, Residual Heat Removal (RHR) and Coolant Circulation - High Water Level, establishes that one RHR loop shall be operable and in operation. TS 3.9.4 is proposed for deletion; this TS is applicable in MODE 6 with water level  23 feet above the top of the reactor vessel flange. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODE 6 will no longer occur, TS 3.9.4 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.9.4 acceptable.
TS 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level, establishes that two RHR loops shall be operable, and one RHR loop shall be in operation. TS 3.9.5 is proposed for deletion; this TS is applicable in MODE 6 with water level  23 feet above the top of the reactor vessel flange. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODE 6 will no longer occur, TS 3.9.5 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.9.5 acceptable.
TS 3.9.6, Refueling Cavity Water Level, is proposed for deletion; this TS is applicable in during movement of irradiated fuel assemblies within containment. The IP3 license will no longer authorize use of the facility for power operation, emplacement, or retention of fuel in the reactor vessel as provided in 10 CFR 50.82(a)(2). Therefore, this TS will no longer apply once the unit is in a permanent shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 3.9.6 acceptable.
4.4.6  TS Section 4.0, Design Features TS Section 4.0, Design Features, provides information and design requirements associated with plant systems. The licensee proposed grammatical corrections to TS 4.3.1.1 by exchanging the period and semicolon to the end of both TS 4.3.1.1.b and TS 4.3.1.1.d. The licensee also proposed the deletion of TS 4.3.1.2, which is related to the storage of new fuel.
 
Once the unit is permanently shut down and defueled, IP3 will not acquire new fuel for storage.
The TS is not applicable in the permanently shutdown and defueled condition. The NRC staff therefore finds the grammatical correction to TS 4.3.1.1 and the deletion of TS 4.3.1.2 acceptable.
4.4.7    TS Section 5.0, Administrative Controls TS Section 5.0, Administrative Controls, establishes the requirements associated with staffing, training, procedures, programs, and reporting requirements. This section is proposed to be revised to include only those administrative requirements needed for safe storage and movement of fuel in the SFP.
4.4.7.1      TS 5.2, Organization The current TS 5.2.1, Onsite and Offsite Organizations, states the following:
: a.      Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all decommissioning organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the facility specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the FSAR and Quality Assurance Plan, as appropriate; The licensee proposed to revise TS 5.2.1 by replacing the term FSAR with DSAR. The revised description will state the following:
: a.      Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all decommissioning organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the facility-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the DSAR and Quality Assurance Plan, as appropriate; The NRC staff reviewed the proposed changes to TS 5.2.1 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. The DSAR is the document that will remain applicable to IP3 once it is shut down and defueled. Therefore, the staff finds the proposed change is acceptable.
 
4.4.7.2      TS 5.4, Procedures The current TS 5.4.1 states the following:
: a.      The procedures applicable to the safe storage of nuclear fuel recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 except as provided in the quality assurance program described or referenced in the Updated FSAR;
: d.      Fire Protection Program implementation; and...
The licensee proposed to revise TS 5.4.1.a by replacing the term Updated FSAR with DSAR and delete TS 5.4.1.d in its entirety to state the following:
: a.      The procedures applicable to the safe storage of nuclear fuel recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 except as provided in the quality assurance program described or referenced in the DSAR;
: d.      Deleted; and...
The NRC staff reviewed the proposed changes to TS 5.4.1.a and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. The DSAR is the document that will remain applicable to IP3 once it is shut down and defueled. Therefore, the staff finds the proposed change is acceptable.
The NRC staff reviewed the proposed change to TS 5.4.1.d and concludes that the change is consistent with the transition to a permanently shutdown and defueled facility and the proposed deletion of License Condition 2.H. Because an FPP is required by 10 CFR 50.48(f) during the decommissioning process regardless of whether the TSs contain a requirement to establish, implement, and maintain procedures for an FPP, this TS is not needed. Therefore, the staff finds the proposed change is acceptable.
4.4.7.3      TS 5.5.2, Primary Coolant Sources Outside Containment The licensee proposed to delete the title for TS 5.5.2.
This is an administrative change as the TS 5.5.2 requirement was deleted in a previous license amendment (Amendment No. 267). Therefore, the NRC staff finds this editorial change acceptable.
4.4.7.4      TS 5.5.4, Radioactive Effluent Controls Program The licensee proposed to retain TS 5.5.4; however, TS paragraphs 5.5.4.d, 5.5.4.h, and 5.5.4.i are proposed to be modified by replacing unit with unit/facility.
Since IP3 will be permanently shutdown and defueled, the proposed change will appropriately reflect the status of IP3. Therefore, the NRC staff finds this administrative change acceptable.
 
4.4.7.5      TS 5.5.5, Component Cyclic or Transient Limit The licensee proposed to delete TS 5.5.5 in its entirety.
TS 5.5.5 provides controls to track the UFSAR, Section 4.1.5, cyclic and transient occurrences to ensure that components are maintained within the design limits. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Since operation in MODES 1 through 6 will no longer occur, TS 5.5.5 will not be applicable in a permanently defueled condition. Therefore, the NRC staff finds the deletion of TS 5.5.5 acceptable.
4.4.7.6      TS 5.5.6, Reactor Coolant Pump Flywheel Inspection Program The licensee proposed to delete TS 5.5.6 in its entirety.
TS 5.5.6 provides the inspection frequencies and acceptance criteria for the reactor coolant pump flywheel inspection program. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).
Since the reactor coolant pumps will no longer perform a function in the permanently shutdown and defueled condition, TS 5.5.6 is no longer applicable. Therefore, the NRC staff finds the deletion of TS 5.5.6 acceptable.
4.4.7.7      TS 5.5.7, Inservice Testing Program The licensee proposed to delete TS 5.5.7 in its entirety.
TS 5.5.7 provides controls for inservice testing of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 pumps and valves in the IP3 Inservice Testing Program that continue to operate and perform a specific function in mitigating the consequences of an accident due to the permanently shutdown and defueled status of the plant. Because the licensee is prohibited from operating the plant or placing fuel in the reactor vessel, in accordance with 10 CFR 50.82(a)(2), there are no longer any ASME Code class pumps and valves that remain in operation and are relied upon to mitigate a DBA. As such, TS 5.5.7 will no longer be relevant in the permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 5.5.7 acceptable.
4.4.7.8      TS 5.5.8, Steam Generator (SG) Program The licensee proposed to delete TS 5.5.8 in its entirety.
TS 5.5.8 requires that an SG program shall be established and implemented to ensure that SG tube integrity is maintained. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).
Therefore, TS 5.5.8 will no longer be applicable, as the SGs will no longer perform a function in the permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.8 acceptable.
 
4.4.7.9    TS 5.5.9, Secondary Water Chemistry Program The licensee proposed to delete TS 5.5.9 in its entirety.
TS 5.5.9 provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, TS 5.5.9 will no longer be applicable, as there will be no need to monitor secondary water chemistry to inhibit SG tube degradation in the permanently shutdown and defueled condition.
The NRC staff therefore finds the deletion of TS 5.5.9 acceptable.
4.4.7.10    TS 5.5.10, Ventilation Filter Testing Program (VFTP)
The licensee proposed to delete TS 5.5.10 in its entirety.
TS 5.5.10 requires that a program shall be established to implement certain testing procedures for the control room ventilation system. As discussed above, TS 3.7.11, Control Room Ventilation System (CRVS), is proposed for deletion, which the NRC staff has found acceptable. Therefore, the VFTP as a support program is not required at IP3 in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 5.5.10 acceptable.
4.4.7.11    TS 5.5.12, Diesel Fuel Oil Testing Program The licensee proposed to delete TS 5.5.12 in its entirety.
TS 5.5.12 requires that a diesel fuel oil testing program to implement testing of both new fuel oil and stored fuel oil shall be established for the onsite diesel generator fuel oil storage tanks and the reserve fuel oil storage tanks. As discussed above, TSs 3.8.1, 3.8.2, and 3.8.3, which define the operability requirements regarding the diesel generators, are proposed for deletion.
Therefore, this support program is not required in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.12 acceptable.
4.4.7.12    TS 5.5.13, Technical Specifications (TS) Bases Control Program The licensee proposed to retain TS 5.5.13; however, the references to the updated FSAR and FSAR in TS 5.5.13.b.2 and TS 5.5.13.c, respectively, would be replaced with references to the DSAR.
The NRC staff reviewed the proposed changes to TS 5.5.13 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. The DSAR is the document that will remain applicable to IP3 once it is shut down and defueled. Therefore, the NRC staff finds the proposed change to TS 5.5.13 is acceptable.
4.4.7.13    TS 5.5.14, Safety Function Determination Program (SFDP)
The licensee proposed to delete TS 5.5.14 in its entirety.
The SFDP was established to ensure that loss of a safety function is detected and appropriate actions taken in accordance with the LCO pertaining to the loss of safety function. The LCOs remaining in the PDTS do not rely on the operability of any active equipment or redundant
 
safety systems. Also, because 10 CFR 50.82(a)(2) prohibits operation of the plant or placing fuel in the reactor vessel, there is no longer a need for redundant systems. Therefore, the requirements of the SFDP, which direct cross-train checks of multiple and redundant safety systems, no longer apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.14 acceptable.
4.4.7.14    TS 5.5.15, Containment Leakage Rate Testing Program The licensee proposed to delete TS 5.5.15 in its entirety.
TS 5.5.15 requires that a program shall establish the leakage rate testing of the containment.
Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Containment integrity is not credited in the analysis of the accidents that remains credible in the permanently defueled condition. Therefore, TS 5.5.15 will no longer be applicable in the permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.15 acceptable.
4.4.7.15    TS 5.5.16, Control Room Envelope Habitability Program The licensee proposed to delete TS 5.5.16 in its entirety.
TS 5.5.16 provides the required elements of the control room envelope habitability program. As discussed above, TS 3.7.11, Control Room Ventilation System (CRVS), is proposed for deletion. Therefore, this support program is not required in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.16 acceptable.
4.4.7.16    TS 5.6.4, Not Used TS 5.6.4 is a placeholder in TS 5.6, Reporting Requirements, and is proposed for deletion.
This is an administrative change to reflect reorganization of the TSs. Therefore, the NRC staff finds this administrative change acceptable.
4.4.7.17    TS 5.6.5, Core Operating Limits Report (COLR)
The licensee proposed to delete TS 5.6.5 in its entirety.
TS 5.6.5 provides the required documentation and analytical methods used to determine the reactor core operating limits. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).
TS 5.6.5 will no longer be applicable in the permanently shutdown and defueled condition, as there will no longer be a need to establish core operating limits. Therefore, the NRC staff finds the deletion of TS 5.6.5 acceptable.
4.4.7.18    TS 5.6.6, Not Used TS 5.6.6 is a placeholder and is proposed for deletion.
 
This is an administrative change to reflect reorganization of the TSs. Therefore, the NRC staff finds this administrative change acceptable.
4.4.7.19      TS 5.6.7, Post Accident Monitoring Instrumentation (PAM) Report The licensee proposed to delete TS 5.6.7 in its entirety.
TS 5.6.7 provides the reporting requirements associated with post-accident monitoring, which applies to an operating reactor, not to the safe storage and handling of spent fuel in the SFP.
Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). TS 5.6.7 will no longer be applicable in the permanently shutdown and defueled condition, as there will no longer be a need to establish post-accident monitoring. Therefore, the NRC staff finds the deletion of TS 5.6.7 acceptable.
4.4.7.20      TS 5.6.8, Steam Generator Tube Inspection Report The licensee proposed to delete TS 5.6.8 in its entirety.
TS 5.6.8 provides the reporting requirements associated with the SG tube inspection program.
Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Since the SGs will no longer perform a function in the permanently shutdown and defueled condition, TS 5.6.8 is no longer applicable. Therefore, the NRC staff finds the deletion of TS 5.6.8 acceptable.
4.5      Changes to Appendix B, Environmental Technical Specification Requirements 4.5.1    Title Page, Part I The current title page for Part I of Appendix B states, in part:
APPENDIX B TO FACILITY OPERATING LICENSE The licensee proposed to change the title page for Part I of Appendix B to state, in part:
APPENDIX B TO FACILITY LICENSE Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed, pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable to the NRC staff.
 
4.5.2    Appendix B, Section 1.0 Currently, Appendix B, Section 1.0 reads:
The Environmental Protection Plan (EPP) is to provide for protection of environmental values during construction and operation of the nuclear facility.
The principal objectives of the EPP are as follows:
(1)    Verify that the plant is operated in an environmentally acceptable manner, as established by the FES and other NRC environmental impact assessments.
(2)    Coordinate NRC requirements and maintain consistency with other Federal, State and local requirements for environmental protection.
(3)    Keep NRC informed of the environmental effects of facility construction and operation and of actions taken to control those effects.
Environmental concerns identified in the FES which relate to water quality matters are regulated by way of the licensee's SPDES permit.
The licensee proposed Appendix B, Section 1.0 to read:
The Environmental Protection Plan (EPP) is to provide for protection of environmental values during handling and storage of spent fuel and maintenance of the nuclear facility. The principal objectives of the EPP are as follows:
(1)    Verify that the facility is maintained in an environmentally acceptable manner, as established by the FES and other NRC environmental impact assessments.
(2)    Coordinate NRC requirements and maintain consistency with other Federal, State and local requirements for environmental protection.
(3)    Keep NRC informed of the environmental effects of handling and storage of spent fuel and maintenance of the facility and of actions taken to control those effects.
Environmental concerns identified in the FES which relate to water quality matters are regulated by way of the licensees SPDES permit.
The proposed changes to Appendix B, Section 1.0 replace a reference to construction and operation with a reference to handling and storage of spent fuel and maintenance, replace a reference to plant is operated with facility is maintained, and replace a reference to facility construction and operation with handling and storage of spent fuel and maintenance of the facility. These proposed changes reflect the revised mission of the facility in the permanently shutdown and defueled condition. Since, consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize use of the facility for power operation or emplacement or retention of fuel into the reactor vessel, this change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable.
 
4.5.3  Appendix B, Section 3.1 Currently, Appendix B, Section 3.1 reads:
The licensee may make changes in station design or operation or perform tests or experimentsChanges in plant design or operation or performance of tests or experiments A proposed change, test, or experiment shall(2) a significant change in effluents or power level; The licensee proposed Appendix B, Section 3.1 to read:
The licensee may make changes in facility design or operation or perform tests or experimentsChanges in facility design or operation or performance of tests or experiments A proposed change, test or experiment shall(2) a significant change in effluents; The proposed changes to Appendix B, Section 3.1 replace references to station and plant with references to facility. These proposed changes reflect the revised mission of the facility in the permanently shutdown and defueled condition. The proposed change to Section 3.1 to eliminate the reference to power level reflects the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable.
4.5.4  Appendix B, Section 3.3 Currently, Appendix B, Section 3.3 reads:
Changes in plant design or operation and The licensee proposed Appendix B, Section 3.3 to read:
Changes in facility design or operation and The proposed change to Appendix B, Section 3.3 replaces the reference to plant with a reference to facility. This proposed change reflects the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.
 
4.5.5  Appendix B, Section 4.1 Currently, Appendix B, Section 4.1 reads:
Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and The licensee proposed Appendix B, Section 4.1 to read:
Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to the handling and storage of spent fuel and maintenance of the facility shall be recorded and The proposed change to Appendix B, Section 4.1 replaces the reference to plant operation with a reference to the handling and storage of spent fuel and maintenance of the facility. This proposed change reflects the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.
4.5.6  Appendix B, Section 4.2 Currently, Appendix B, Section 4.2 reads:
The currently applicable Biological Opinion concludes that continued operation of IP2 and IP3 is not likely to jeopardize the continued existence of the listed species or to adversely affect the designated critical habitat of those species.
The licensee proposed Appendix B, Section 4.2 to read:
The currently applicable Biological Opinion concludes that continued operation of IP2 and IP3 is not likely to jeopardize the continued existence of the listed species or to adversely affect the designated critical habitat of those species.
This Biological Opinion conservatively bounds the conditions that will occur in the permanently shut down and defueled condition.
The proposed change to Appendix B, Section 4.2 concludes that the biological opinion rendered during the evaluation of the continued operation of IP2 and IP3 conservatively bounds the conditions that will occur in the permanently shutdown and defueled condition. The NRC staff notes that the biological opinion addresses the permanent shutdown and defueling of IP3 regarding the shortnose sturgeon and Atlantic sturgeon. Therefore, the NRC staff finds the proposed change acceptable.
 
4.5.7    Appendix B, Section 5.2 Currently, Appendix B, Section 5.2 reads:
Records and logs relative to the environmental aspects of plant operation shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to the NRC on request.
Records of modifications to plant structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the plant. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.
The licensee proposed Appendix B, Section 5.2 to read:
Records and logs relative to the environmental aspects of previous plant operation and the handling and storage of spent fuel and maintenance of the facility shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to the NRC on request.
Records of modifications to facility structures, systems and components determined to potentially affect the continued protection of the environment shall be retained for the life of the facility. All other records, data and logs relating to this EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.
The proposed changes to Appendix B, Section 5.2 clarify that the reference to plant operation refers to plant operations prior to the permanent shutdown and includes a reference to the handling and storage of spent fuel and maintenance of the facility. In addition, references to plant are replaced with facility. These proposed changes reflect the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable.
4.5.8    Appendix B, Section 5.4.1 Currently, Appendix B, Section 5.4.1 reads:
and an assessment of the observed impacts of the plant operation on the environment (b)      A list of all changes in station design or operation, tests, and experiments made in accordance with subsection 3.1 which involved a potentially significant unreviewed environmental issue.
 
The licensee proposed Section 5.4.1 to read:
        ... and an assessment of the observed impacts of the previous plant operation and the handling and storage of spent fuel and maintenance of the facility on the environment...
(b)      A list of all changes in facility design or operation, tests, and experiments made in accordance with subsection 3.1 which involved a potentially significant unreviewed environmental issue.
The proposed changes to Appendix B, Section 5.4.1 clarify that the reference to plant operation refers to plant operations prior to the permanent shutdown and includes a reference to the handling and storage of spent fuel and maintenance of the facility. In addition, a reference to station is replaced with a reference to facility. These proposed changes reflect the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable.
4.5.9    Appendix B, Section 5.4.2 Currently, Section 5.4.2 reads:
The report shall (1) describe, analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (2)
The licensee proposed Section 5.4.2 to read:
The report shall (1) describe, analyze, and evaluate the event, including extent and magnitude of the impact and facility conditions, (2)
The proposed change to Appendix B, Section 5.4.2 replaces a reference to plant operating characteristics with facility conditions. This proposed change reflects the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.
4.5.10 Title Page, Part II The current title page for Part II of Appendix B states, in part:
APPENDIX B TO FACILITY OPERATING LICENSE The licensee proposed to change the title page for Part II of Appendix B to state, in part:
APPENDIX B TO FACILITY LICENSE
 
Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed, pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the NRC staff finds that the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.
4.6      Changes to Appendix C, Inter-Unit Fuel Transfer Technical Specifications The current title page for Part I and Part II of Appendix C states, in part:
APPENDIX C TO FACILITY OPERATING LICENSE The licensee proposed to change the title page for Part I and Part II of Appendix C to state, in part:
APPENDIX C TO FACILITY LICENSE Also, the current header for Part I of Appendix C states, in part:
Facility Operating License The licensee proposed to change the header for Part 1 of Appendix C to state, in part:
Facility License Appendix C is modified by replacing the references to Facility Operating License with Facility License. The proposed changes reflect the upcoming change in status regarding IP3. Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed, pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description.
Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are 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 36435; June 16, 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 C. Jackson D. Scully E. Stutzcage L. Alvarado J. Robinson S. Mehta J. Tsao B. Wolfgang K. West Date: April 22, 2021
 
ML21074A000 OFFICE      NRR/DORL/LPL1/PM      NRR/DORL/LPL1/LA          NRR/DRA/APLB/BC(A)
NAME        RGuzman                JBurkhardt                JBorromeo DATE        3/22/2021              3/19/2021                08/25/2020 OFFICE      NRR/DNRL/NCSG/BC      NRR/DSS/SNSB/BC          NRR/DNRL/NVIB/BC NAME        SBloom                SKrepel                  HGonzalez DATE        09/17/2020            11/30/2020                12/9/2020 OFFICE      NRR/DRA/ARCB/BC        NRR/DSS/SCPB/BC          NRR/DSS/SFNB/BC NAME        KHsueh                BWittick                  RLukes DATE        12/17/2020            12/20/2020                3/22/2021 OFFICE      NRR/DSS/STSB/BC        OGC-NLO w/revisions      NRR/DORL/LPL1/PM NAME        VCusumano              STurk                    JDanna (JPaige for)
DATE        3/19/2021              4/15/2021                4/22/2021 OFFICE      NRR/DORL/LPL1/PM NAME        RGuzman DATE        4/22/2021}}

Latest revision as of 19:36, 19 January 2022

Issuance of Amendment No. 270 Permanently Defueled Technical Specifications
ML21074A000
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 04/22/2021
From: Richard Guzman
NRC/NRR/DORL/LPL1
To:
Entergy Nuclear Operations
Guzman R
References
EPID L-2020-LLA-0090
Download: ML21074A000 (124)
v • d • e


Text

April 22, 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 UNIT NO. 3 ISSUANCE OF AMENDMENT NO. 270 RE: PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS (EPID L-2020-LLA-0090)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 270 to Renewed Facility Operating License No. DPR-64 for Indian Point Nuclear Generating Unit No. 3 (Indian Point 3). The amendment consists of changes to the Technical Specifications (TSs) in response to your application dated April 28, 2020.

The amendment revises the Indian Point 3 Renewed Facility Operating License and the associated TSs to permanently defueled TSs, consistent with the permanent cessation of operations and permanent removal of fuel from the reactor vessel.

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. 50-286

Enclosures:

1. Amendment No. 270 to DPR-64
2. Safety Evaluation cc: Listserv

ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-286 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 270 License No. DPR-64

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

A. The application for amendment by Entergy Nuclear Operations, Inc. (ENO, the licensee) dated April 28, 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 Appendix A, Technical Specifications; Appendix B, Environmental Technical Specification Requirements; and Appendix C, Inter-Unit Fuel Transfer Technical Specifications, as indicated in the attachment to this license amendment, and Renewed Facility Operating License No. DPR-64 is hereby amended to read as follows:

The title RENEWED FACILITY OPERATING LICENSE is to read RENEWED FACILITY LICENSE Paragraph 1.B is to read as follows:

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 2.A and 2.B are to read as follows:

2. Accordingly, Renewed Facility License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows:

A. This renewed license applies to the Indian Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and associated equipment (the facility), owned by ENIP3 and maintained by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the Defueled Safety Analysis Report, as supplemented and amended, and the Environmental Report, as amended.

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

(1) Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) ENIP3 to possess and use, and (b) ENO to possess and use the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; (2) ENO pursuant to the Act and 10 CFR Part 70, to possess at any time, special nuclear material that was used as reactor fuel, in accordance with the limitations for storage, as described in the Defueled Safety Analysis Report, as supplemented and amended; (3) ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct source and special nuclear material

as sealed neutron sources that were used for reactor startup, sealed sources that were used for calibration of reactor instrumentation and are used in the calibration of radiation monitoring equipment, and that were used as fission detectors in amounts as required; (4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials that were produced by the operation of the facility.

Paragraph 2.C is to read as follows:

(1) Deleted per Amendment No. 270.

(2) Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 270, are hereby incorporated in the renewed license. ENO shall maintain the facility in accordance with the Technical Specifications.

Paragraphs 2.H, 2.O, 2.AA, 2.AB, and 2.AD are to read as follows:

H. Deleted per Amendment No. 270.

O. Deleted per Amendment No. 270.

AA. Deleted per Amendment No. 270.

AB. Deleted per Amendment No. 270.

AD. Deleted per Amendment No. 270.

Paragraph 3 is to read as follows:

3. This renewed license is effective as of the date of issuance, and until the Commission notifies the licensee in writing that the license is terminated.
3. This license amendment is effective following the docketing of the certifications required by 10 CFR 50.82(a)(1)(i) and (ii) that Indian Point Nuclear Generating Unit No. 3 has been permanently shut down and defueled, and shall be implemented within 90 days of the effective date of the amendment.

FOR THE NUCLEAR REGULATORY COMMISSION Digitally signed by Jason Jason C. C. Paige Date: 2021.04.22 Paige 11:25:16 -04'00' James G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Renewed Facility Operating License, Appendix A, Technical Specifications, Appendix B, Environmental Technical Specification Requirements, and Appendix C, Inter-Unit Fuel Transfer Technical Specifications Date of Issuance: April 22, 2021

ATTACHMENT TO LICENSE AMENDMENT NO. 270 INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 RENEWED FACILITY OPERATING LICENSE NO. DPR-64 DOCKET NO. 50-286 Replace the following pages of the Renewed Facility Operating License; Appendix A, Permanently Defueled Technical Specifications; Appendix B, Environmental Technical Specification Requirements; and Appendix C, Inter-Unit Fuel Transfer Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change. For the Appendix A PDTS revised pages, vertical line revision bars were not used as the changes were considered a major rewrite.

Facility Operating License No. DPR-64 REMOVE INSERT through through Appendix A, Permanently Defueled Technical Specifications REMOVE INSERT Title Page Title Page i through iv i 1.1-1 through 1.1-8 1.1-1 1.2-1 through 1.2-3 1.2-1 1.3-1 through 1.3-14 1.3-1 through 1.3-2 1.4-1 through 1.4-4 1.4-1 2.0-1 2.0-1 3.0-1 through 3.0-5 3.0-1 through 3.0-2 3.1.1-1 through 3.7.13-2 ---

3.7.14-1 3.7.14-1 3.7.15-1 through 3.7.15-2 3.7.15-1 3.7.16-1 through 3.7.16-3 3.7.16-1 through 3.7.16-3 3.7.17-1 through 3.9.6-1 ---

4.0-1 through 4.0-3 4.0-1 5.0-1 through 5.0-41 5.0-1 through 5.0-15

Appendix B, Environmental Technical Specification Requirements REMOVE INSERT Part I Title Page Part I Title Page Table of Contents Table of Contents 1-1 1-1 3-1 3-1 3-2 3-2 3-3 3-3 4-1 4-1 4-2 4-2 5-1 5-1 5-2 5-2 5-4 5-4 Part II Title Page Part II Title Page Appendix C, Inter-Unit Fuel Transfer Technical Specifications REMOVE INSERT Part I Title Page Part I Title Page 1 1 2 2 Part II Title Page Part II Title Page

ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

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

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

A. The application for a renewed license filed by Entergy Nuclear Indian Point 3, LLC (ENIP3) (the licensee) and Entergy Nuclear Operations, Inc. (ENO)

(operator) for Indian Point Nuclear Generating Unit No. 3 (IP3 at the Indian Point Energy Center (IPEC) complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B. The facility will be maintained in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this renewed license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. ENIP3 and ENO are financially and technically qualified to engage in the activities authorized by this amendment; E. ENIP3 and ENO have satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements" of the Commission's regulations; F. The issuance of this renewed license will not be inimical to the common defense and security or to the health and safety of the public; G. The receipt, possession and use of source, byproduct and special nuclear material as authorized by this renewed license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40 and 70 including 10 CFR Sections 30.33, 40.32, 70.23, and 70.31; Amendment No. 270

H. The issuance of this renewed license is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied; and I. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the period of extended operation on the functionality of structures and components that have been identified to require review under 10 CFR 54.21(a)(1); and (2) time-limited aging analyses that have been identified to require review under 10 CFR 54.21(c), such that there is reasonable assurance that the activities authorized by this renewed license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facilitys current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commissions regulations.

2. Accordingly, Renewed Facility License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows:

A. This renewed license applies to the Indian Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and associated equipment (the facility),

owned by ENIP3 and maintained by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the Defueled Safety Analysis Report" as supplemented and amended, and the Environmental Report, as amended.

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

(1) Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) ENIP3 to possess and use, and (b) ENO to possess and use the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; (2) ENO pursuant to the Act and 10 CFR Part 70, to possess, at any time, special nuclear material that was used as reactor fuel, in accordance with the limitations for storage, as described in the Defueled Safety Analysis Report, as supplemented and amended; (3) ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct source and special nuclear material as sealed neutron sources that were used for reactor startup, sealed sources that were used for calibration of reactor instrumentation and are used in the calibration of radiation monitoring equipment, and that were used as fission detectors in amounts as required; Amendment No. 270

(4) ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components; (5) ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials that were produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:

(1) Deleted per Amendment No. 270 (2) Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 270, are hereby incorporated in the renewed license. ENO shall maintain the facility in accordance with the Technical Specifications.

D. (DELETED)

E. (DELETED)

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

G. ENO shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822), and to the authority of 10 CFR 50.90 and CFR 50.54(p). The combined set of plans 1 for the Indian Point Energy Center, which contain Safeguards Information protected under 10 CFR 73.21, is entitled: Physical Security, Training and Qualification, and Safeguards Contingency Plan, Revision 0, and was submitted by letter dated October 14, 2004, as supplemented by letter dated May 18, 2006.

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

Amendment No. 270

ENO shall fully implement and maintain in effect all provisions of the Commission-approved cyber security plan (CSP), including changes made pursuant to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The ENO CSP was approved by License Amendment No. 243, as supplemented by changes approved by License Amendment Nos. 254, 260, and 263.

ENO has been granted Commission authorization to use stand alone preemption authority under Section 161A of the Atomic Energy Act, 42 U.S.C.

2201a with respect to the weapons described in Section II supplemented with Section Ill of Attachment 1 to its application submitted by letter dated August 20, 2013, as supplemented by letters dated November 21, 2013, and July 24, 2014, and citing letters dated April 27, 2011, and January 4, 2012. ENO shall fully implement and maintain in effect the provisions of the Commission-approved authorization.

H. Deleted per Amendment No. 270 I. DELETED J. DELETED K. DELETED L. DELETED M. DELETED N. DELETED O. Deleted per Amendment No. 270 P. ENIP3 and ENO shall take no action to cause Entergy Global Investments, Inc.

or Entergy International Ltd. LLC, or their parent companies to void, cancel, or modify the $70 million contingency commitment to provide funding for the facility as represented in the application for approval of the transfer of the license from PASNY to ENIP3 and ENO, without the prior written consent of the Director, Office of Nuclear Reactor Regulation.

Q. DELETED R. DELETED S. DELETED T. DELETED U. DELETED V. DELETED Amendment No. 270

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

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

AA. Deleted per Amendment No. 270 AB. Deleted per Amendment No. 270 AC. Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

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

1. Pre-defined coordinated fire response strategy and guidance
2. Assessment of mutual aid fire fighting assets
3. Designated staging areas for equipment and materials
4. Command and control
5. Training of response personnel (b) Operations to mitigate fuel damage considering the following:
1. Protection and use of personnel assets
2. Communications
3. Minimizing fire spread
4. Procedures for implementing integrated fire response strategy
5. Identification of readily-available pre-staged equipment
6. Training on integrated fire response strategy
7. Spent fuel pool mitigation measures (c) Actions to minimize release to include consideration of:
1. Water spray scrubbing
2. Dose to onsite responders AD. Deleted per Amendment No. 270 AE. ENO may transfer IP3 spent fuel to the IP2 spent fuel pit subject to the conditions listed in Appendix C. ENO is further authorized to transfer IP3 spent fuel into NRC approved storage casks for onsite storage by ENO and ENIP3.

Amendment No. 270

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

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

a. The licensee shall implement those new programs and enhancements to existing programs no later than the date specified in the License Renewal UFSAR Supplement.
b. The licensee shall complete those activities no later than the date specified in the License Renewal UFSAR Supplement.
3. This renewed license is effective as of the date of issuance, and until the Commission notifies the licensee in writing that the license is terminated.

FOR THE NUCLEAR REGULATORY COMMISSION Ho K. Nieh, Director Office of Nuclear Reactor Regulation Attachments:

Appendix A - Permanently Defueled Technical Specifications Appendix B - Environmental Technical Specification Requirements Appendix C - Inter-Unit Fuel Transfer Technical Specifications Date of Issuance: September 17, 2018 Amendment No. 270

APPENDIX A TO FACILITY LICENSE DPR-64 PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS AND BASES FOR THE INDIAN POINT 3 NUCLEAR GENERATING STATION UNIT NO. 3 WESTCHESTER COUNTY, NEW YORK ENTERGY NUCLEAR INDIAN POINT 3, LLC (ENIP3)

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

DOCKET NO. 50-286 Date of Issuance:

April 15, 1976 Amendment No. 270

Facility License No. DPR-64 Appendix A - Permanently Defueled Technical Specifications TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1 Definitions 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency 2.0 DELETED 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.7 SPENT FUEL PIT REQUIREMENTS 3.7.14 Spent Fuel Pit Water Level 3.7.15 Spent Fuel Pit Boron Concentration 3.7.16 Spent Fuel Assembly Storage 4.0 DESIGN FEATURES 4.1 Site Location 4.2 Deleted 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Facility Staff Qualifications 5.4 Procedures 5.5 Programs and Manuals 5.5.1 Offsite Dose Calculation Manual (ODCM) 5.5.2 DELETED 5.5.3 NOT USED 5.5.4 Radioactive Effluent Controls Program 5.5.5 DELETED 5.5.6 DELETED 5.5.7 DELETED 5.5.8 DELETED 5.5.9 DELETED 5.5.10 DELETED 5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program 5.5.12 DELETED 5.5.13 Technical Specification (TS) Bases Control Program 5.6 Reporting Requirements 5.6.1 NOT USED 5.6.2 Annual Radiological Environmental Operating Report 5.6.3 Radioactive Effluent Release Report 5.7 High Radiation Area Indian Point 3 i Amendment No. 270

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions

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

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

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

CERTIFIED FUEL HANDLER A CERTIFIED FUEL HANDLER is an individual who (CFH) complies with the provisions of the CERTIFIED FUEL HANDLER training and retraining program required by TS 5.3.2.

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

Indian Point 3 1.1-1 Amendment No. 270

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

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

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

The successive levels of logic are identified by additional digits of the Required Action number and by successive indentations of the logical connectors.

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

EXAMPLE The following example illustrates the use of logical connectors.

EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met A.1 Verify AND A.2 Restore In this example the logical connector AND is used to indicate that when in Condition A, both Required Actions A.1 and A.2 must be completed.

Indian Point 3 1.2-1 Amendment No. 270

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

BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe handling and storage of spent nuclear fuel. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met.

Specified with each stated Condition are Required Action(s) and Completion Time(s).

DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the facility is in a specified condition stated in the Applicability of the LCO.

Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the facility is not within the LCO Applicability.

EXAMPLE The following example illustrates the use of Completion Times with different Required Actions.

Indian Point 3 1.3-1 Amendment No. 270

Completion Times 1.3 1.3 Completion Time EXAMPLE (continued)

EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pit A.1 Suspend Immediately boron movement of fuel concentration not assemblies in the within limit. spent fuel pit.

AND A.2 Initiate action to Immediately restore spent fuel pit boron concentration to within limit.

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

The Required Actions of Condition A are to immediately suspend movement of fuel assemblies in the spent fuel pit and initiate action to restore spent fuel pit boron concentration within limit.

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

Indian Point 3 1.3-2 Amendment No. 270

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR)

Applicability. The "specified Frequency" consists of the requirements of the Frequency column of each SR.

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

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify level is within limits. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Example 1.4-1 contains the type of SR encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subsequent interval.

Although the Frequency is stated as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for flexibility. The measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when a variable is outside specified limits, or the facility is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the facility is in a specified condition in the Applicability of the LCO, then SR 3.0.3 becomes applicable.

If the interval as specified by SR 3.0.2 is exceeded while the facility is not in a specified condition in the Applicability of the LCO for which performance of the SR is required, then SR 3.0.4 becomes applicable.

The Surveillance must be performed within the Frequency requirements of SR 3.0.2, as modified by SR 3.0.3, prior to entry into the specified condition or the LCO is considered not met (in accordance with SR 3.0.1).

Indian Point 3 1.4-1 Amendment No. 270

Deleted 2.0 2.0 DELETED Indian Point 3 2.0-1 Amendment No. 270

LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during the specified conditions in the Applicability, except as provided in LCO 3.0.2.

LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met.

If the LCO is met or is no longer applicable prior to expiration of the specified Completion Time(s), completion of the Required Action(s) is not required unless otherwise stated.

Indian Point 3 3.0-1 Amendment No. 270

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3. Surveillances do not have to be performed on variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

SR 3.0.4 Entry into a specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3.

Indian Point 3 3.0-2 Amendment No. 270

Spent Fuel Pit Water Level 3.7.14 3.7 SPENT FUEL PIT REQUIREMENTS 3.7.14 Spent Fuel Pit Water Level LCO 3.7.14 The spent fuel pit water level shall be 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel pit.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pit water A.1 Suspend movement of Immediately level not within limit. irradiated fuel assemblies in the spent fuel pit.

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

Indian Point 3 3.7.14-1 Amendment No. 270

Spent Fuel Pit Boron Concentration 3.7.15 3.7 SPENT FUEL PIT REQUIREMENTS 3.7.15 Spent Fuel Pit Boron Concentration LCO 3.7.15 The spent fuel pit boron concentration shall be 1000 ppm.

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

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

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pit boron A.1 Suspend movement of Immediately concentration not within fuel assemblies in the limit. spent fuel pit.

AND A.2.1 Initiate action to restore Immediately spent fuel pit boron concentration to within limit.

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

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the spent fuel pit boron concentration is 31 days within limit.

Indian Point 3 3.7.15-1 Amendment No. 270

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

a. Fuel assemblies classified as Type 2 may be stored in any location in either Region 1 or Region 2;
b. Fuel assemblies classified as Type 1A, 1B or 1C shall be stored in Region 1;
c. Fuel assembly storage location within Region 1 shall be restricted as follows:
1. Type 1A assemblies may be stored anywhere in Region 1;
2. Type 1B assemblies may be stored anywhere in Region 1, except a Type 1B assembly shall not be stored face-adjacent to a Type 1C assembly;
3. Type 1C assemblies shall not be stored in Row 64 or in Column ZZ; and
4. Type 1C assemblies shall be stored in Region 1 locations where all face-adjacent locations are as follows:

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

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

Indian Point 3 3.7.16-1 Amendment No. 270

Spent Fuel Assembly Storage 3.7.16 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Initiate action to move fuel Immediately LCO not met. to restore compliance with LCO 3.7.16.

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

Indian Point 3 3.7.16-2 Amendment No. 270

Spent Fuel Assembly Storage 3.7.16 Figure 3.7.16-1 (Page 1 of 1)

Fuel Assembly Classification for Storage in the Spent Fuel Pit Indian Point 3 3.7.16-3 Amendment No. 270

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location Indian Point 3 is located on the east bank of the Hudson River at Indian Point, Village of Buchanan, in upper Westchester County, New York. The site is approximately 24 miles north of the New York City boundary line. The nearest city is Peekskill which is 2.5 miles northeast of Indian Point.

The minimum distance from the reactor center line to the boundary of the site exclusion area and the outer boundary of the low population zone as defined in 10 CFR 100.3 is 350 meters and 1100 meters, respectively.

4.2 Deleted 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;
b. keff 0.95 if assemblies are inserted in accordance with Technical Specification 3.7.16, Spent Fuel Assembly Storage;
c. A nominal 9.075 inch center to center distance between fuel assemblies placed in the high density fuel storage racks (Region II);
d. A nominal 10.76 inch center to center distance between fuel assemblies placed in low density fuel storage racks (Region I).

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

4.3.3 Capacity The spent fuel pit is designed and shall be maintained with a storage capacity limited to no more than 1345 fuel assemblies.

Indian Point 3 4.0-1 Amendment No. 270

Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The plant manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during his absence.

The plant manager or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety.

5.1.2 The shift manager (SM) shall be responsible for the shift command function.

Indian Point 3 5.0-1 Amendment No. 270

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for facility staff and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear fuel.

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all decommissioning organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the facility-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the DSAR and Quality Assurance Plan, as appropriate;
b. The plant manager shall be responsible for overall safe maintenance of the facility and shall have control over those onsite activities necessary for safe storage and maintenance of nuclear fuel;
c. The corporate officer with direct responsibility for IP3 shall have corporate responsibility for the safe storage and handling of nuclear fuel and shall take any measures needed to ensure acceptable performance of the staff in maintaining and providing technical support to the facility to ensure safe management of nuclear fuel; and
d. The individuals who train the CERTIFIED FUEL HANDLERS, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

Indian Point 3 5.0-2 Amendment No. 270

Organization 5.2 5.2 Organization 5.2.2 Facility Staff The facility staff organization shall include the following:

a. Each duty shift shall be composed of at least one shift manager and one NON-CERTIFIED OPERATOR. The NON-CERTIFIED OPERATOR position may be filled by a CERTIFIED FUEL HANDLER.

At least one person qualified to stand watch in the control room (NON-CERTIFIED OPERATOR or CERTIFIED FUEL HANDLER) shall be present in the control room when nuclear fuel is stored in the spent fuel pool.

b. Shift crew composition may be less than the minimum requirement of 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements and all of the following conditions are met:
1) No fuel movements are in progress;
2) No movement of loads over fuel are in progress; and
3) No unmanned shift positions during shift turnover shall be permitted while the shift crew is less than the minimum.
c. An individual qualified in radiation protection procedures shall be on site during fuel handling operations and during movement of heavy loads over the fuel storage racks. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.
d. Not Used.
e. The shift manager shall be a CERTIFIED FUEL HANDLER.
f. Deleted.

Indian Point 3 5.0-3 Amendment No. 270

Facility Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Facility Staff Qualifications 5.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1-1978 for comparable positions with exceptions specified in the IPEC Quality Assurance Program Manual (QAPM).

5.3.2 An NRC approved training and retraining program for CERTIFIED FUEL HANDLERS shall be maintained.

Indian Point 3 5.0-4 Amendment No. 270

Procedures 5.4 5.0 ADMINISTRATIVE CONTROLS 5.4 Procedures 5.4.1 Written procedures shall be established, implemented, and maintained covering the following activities:

a. The procedures applicable to the safe storage of nuclear fuel recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978 except as provided in the quality assurance program described or referenced in the DSAR;
b. Deleted;
c. Quality assurance for effluent and environmental monitoring;
d. Deleted; and
e. All programs specified in Specification 5.5.

Indian Point 3 5.0-5 Amendment No. 270

Programs and Manuals 5.5 5.0 ADMINISTRATIVE CONTROLS 5.5 Programs and Manuals The following programs shall be established, implemented, and maintained.

5.5.1 Offsite Dose Calculation Manual (ODCM)

a. The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and
b. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities, and descriptions of the information that should be included in the Annual Radiological Environmental Operating, and Radioactive Effluent Release Reports required by Specification 5.6.2 and Specification 5.6.3.
c. Licensee initiated changes to the ODCM:
1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

(a) Sufficient information to support the change(s) together with the appropriate analyses or evaluations justifying the change(s), and (b) A determination that the change(s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations;

2. Shall become effective after the approval of the plant manager; and
3. Shall be submitted to the NRC in the form of a complete, legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (i.e., month and year) the change was implemented.

5.5.2 Deleted Indian Point 3 5.0-6 Amendment No. 270

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.3 Not Used 5.5.4 Radioactive Effluent Controls Program This program conforms to 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM;
b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to 10 times the concentration values in 10 CFR 20, Appendix B, Table 2, Column 2;
c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM;
d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit/facility to unrestricted areas, conforming to 10 CFR 50, Appendix I;
e. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days. Determination of projected dose contributions from radioactive effluents in accordance with the methodology in the ODCM at least every 31 days;
f. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50, Appendix I; Indian Point 3 5.0-7 Amendment No. 270

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

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:
a. For noble gases: Less than or equal to a dose rate of 500 mrem/yr to the whole body and less than or equal to a dose rate of 3000 mrem/yr to the skin, and
b. For iodine-131, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to dose rate of 1500 mrem/yr to any organ.
h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit/facility to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;
i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit/facility to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and
j. Limitations on the annual dose or dose commitment to any member of the public, beyond the site boundary due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Radioactive Effluents Controls Program surveillance frequency.

5.5.5 through Deleted 5.5.10 Indian Point 3 5.0-8 Amendment No. 270

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTP) ETSB 11-5, Postulated Radioactive Release due to Waste Gas System Leak or Failure.

The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, Postulated Radioactive Release due to Tank Failures.

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);
b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank shall be limited to less than the amount that would result in a whole body exposure of 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks contents; and
c. A surveillance program to ensure that the quantity of radioactivity contained in all outdoor liquid radwaste tanks that are not surrounded by liners, dikes, or walls, capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

5.5.12 Deleted Indian Point 3 5.0-9 Amendment No. 270

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Technical Specifications (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
b. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:
1. a change in the TS incorporated in the license; or
2. a change to the DSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the DSAR.
d. Proposed changes that do not meet the criteria of Specification 5.5.13.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

Indian Point 3 5.0-10 Amendment No. 270

Reporting Requirements 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Reporting Requirements The following reports shall be submitted in accordance with 10 CFR 50.4.

5.6.1 Not Used 5.6.2 Annual Radiological Environmental Operating Report

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

A single submittal may be made for a multiple unit/facility station. The submittal should combine sections common to all units/facilities at the station.

The Annual Radiological Environmental Operating Report covering the operation of the unit/facility during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

A full listing of the information to be contained in the Annual Radiological Environmental Operating Report is provided in the ODCM.

5.6.3 Radioactive Effluent Release Report

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

A single submittal may be made for a multiple unit/facility station. The submittal shall combine sections common to all units/facilities at the station; however, for units/facilities with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit/facility.

The Radioactive Effluent Release Report covering the operation of the unit/facility in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit/facility. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR Part 50.36a and 10 CFR 50, Appendix I, Section IV.B.l.

Indian Point 3 5.0-11 Amendment No. 270

High Radiation Area 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 High Radiation Area As provided in paragraph 20.1601(c) of 10 CFR Part 20, the following controls shall be applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR Part 20:

5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment.
b. Access to, and activities in, each such area shall be controlled by means of Radiation Work Permit (RWP) or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
c. Individuals qualified in radiation protection procedures and personnel continuously escorted by such individuals may be exempted from the requirement for an RWP or equivalent while performing their assigned duties provided that they are otherwise following facility radiation protection procedures for entry to, exit from, and work in such areas.
d. Each individual or group entering such an area shall possess:
1. A radiation monitoring device that continuously displays radiation dose rates in the area; or
2. A radiation monitoring device that continuously integrates the radiation dose rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or
3. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area, or Indian Point 3 5.0-12 Amendment No. 270

High Radiation Area 5.7 5.7 High Radiation Area 5.7.1 High Radiation Areas with Dose Rates Not Exceeding 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation (continued)

4. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with individuals in the area who are covered by such surveillance.
e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.

Indian Point 3 5.0-13 Amendment No. 270

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation

a. Each entryway to such an area shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate that prevents unauthorized entry, and, in addition:
1. All such door and gate keys shall be maintained under the administrative control of the shift supervisor, radiation protection manager, or his or her designee.
2. Doors and gates shall remain locked except during periods of personnel or equipment entry or exit.
b. Access to, and activities in, each such area shall be controlled by means of an RWP or equivalent that includes specification of radiation dose rates in the immediate work area(s) and other appropriate radiation protection equipment and measures.
c. Individuals qualified in radiation protection procedures may be exempted from the requirement for an RWP or equivalent while performing radiation surveys in such areas provided that they are otherwise following facility radiation protection procedures for entry to, exit from, and work in such areas.
d. Each individual or group entering such an area shall possess:
1. A radiation monitoring device that continuously integrates the radiation rates in the area and alarms when the device's dose alarm setpoint is reached, with an appropriate alarm setpoint, or
2. A radiation monitoring device that continuously transmits dose rate and cumulative dose information to a remote receiver monitored by radiation protection personnel responsible for controlling personnel radiation exposure within the area with the means to communicate with and control every individual in the area, or Indian Point 3 5.0-14 Amendment No. 270

High Radiation Area 5.7 5.7 High Radiation Area 5.7.2 High Radiation Areas with Dose Rates Greater than 1.0 rem/hour at 30 Centimeters from the Radiation Source or from any Surface Penetrated by the Radiation, but less than 500 rads/hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation (continued)

3. A self-reading dosimeter (e.g., pocket ionization chamber or electronic dosimeter) and, (i) Be under the surveillance, as specified in the RWP or equivalent, while in the area, of an individual qualified in radiation protection procedures, equipped with a radiation monitoring device that continuously displays radiation dose rates in the area; who is responsible for controlling personnel exposure within the area, or (ii) Be under the surveillance as specified in the RWP or equivalent, while in the area, by means of closed circuit television, of personnel qualified in radiation protection procedures, responsible for controlling personnel radiation exposure in the area, and with the means to communicate with and control every individual in the area.
4. In those cases where options (2) and (3), above, are impractical or determined to be inconsistent with the "As Low As is Reasonably Achievable" principle, a radiation monitoring device that continuously displays radiation dose rates in the area.
e. Except for individuals qualified in radiation protection procedures, or personnel continuously escorted by such individuals, entry into such areas shall be made only after dose rates in the area have been determined and entry personnel are knowledgeable of them. These continuously escorted personnel will receive a pre-job briefing prior to entry into such areas. This dose rate determination, knowledge, and pre-job briefing does not require documentation prior to initial entry.
f. Such individual areas that are within a larger area where no enclosure exists for the purpose of locking and where no enclosure can reasonably be constructed around the individual area need not be controlled by a locked door or gate, nor continuously guarded, but shall be barricaded, conspicuously posted, and a clearly visible flashing light shall be activated at the area as a warning device.

Indian Point 3 5.0-15 Amendment No. 270

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

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT 3 NUCLEAR POWER PLANT ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART I: NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION PLAN FACILITY LICENSE NO. DPR-64 DOCKET NUMBER 50-286 Renewed License No. DPR-64 Amendment No. 270

INDIAN POINT NUCLEAR GENERATING PLANT UNIT 3 ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART I: NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION PLAN TABLE OF CONTENTS Section Page 1.0 Objectives of the Environmental Protection Plan ................................................... 1-1 2.0 Environmental Protection Issues ........................................................................... 2-1 3.0 Consistency Requirements .................................................................................... 3-1 3.1 Plant Design and Operation................................................................................... 3-1 3.2 Reporting Related to the NPDES Permits and State Certifications ........................ 3-2 3.3 Changes Required for Compliance with Other Environmental Regulations. ........... 3-3 4.0 Environmental Conditions ...................................................................................... 4-1 4.1 Unusual or Important Environmental Events .......................................................... 4-1 4.2 Environmental Monitoring ...................................................................................... 4-1 5.0 Administrative Procedures ..................................................................................... 5-1 5.1 Review and Audit .................................................................................................. 5-1 5.2 Records Retention ................................................................................................. 5-1 5.3 Changes in Environmental Protection Plan ............................................................ 5-1 5.4 Plant Reporting Requirements ............................................................................... 5-2 Renewed License No. DPR-64 Amendment No. 270

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

(1) Verify that the facility is maintained in an environmentally acceptable manner, as established by the FES and other NRC environmental impact assessments.

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

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

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

1-1 Renewed License No. DPR-64 Amendment No. 270

3.0 Consistency Requirements 3.1 Plant Design and Operation ENO may make changes in facility design or operations or perform tests or experiments affecting the environment provided such changes, tests or experiments do not involve an unreviewed environmental question, and do not involve a change in the Environmental Protection Plan.* Changes in the facility design or operation or performance of tests or experiments which do not affect the environment are not subject to the requirements of this EPP. Activities governed by Section 3.3 are not subject to the requirements of this section.

Before engaging in additional construction or operational activities which may affect the environment, ENO shall prepare and record an environmental evaluation of such activity.

When the evaluation indicates that such activity involves an unreviewed environmental question, ENO shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor Regulation. When such activity involves a change in the Environmental Protection Plan, such activity and change to the Environmental Protection Plan may be implemented only in accordance with an appropriate license amendment as set forth in Section 5.3.

A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns (1) a matter which may result in a significant increase in any adverse environmental impact previously evaluated in the final environmental statement (FES) or final supplemental environmental impact statement (FSEIS), as modified by the staffs testimony to the Atomic Safety and Licensing Boards, supplements to the FES or FSEIS, environmental impact appraisals, or in any decision of the Atomic Safety and Licensing Board; This provision does not relieve the ENO of the requirements of 10 CFR 50.59.

3-1 Renewed License No. DPR-64 Amendment No. 270

or (2) a significant change in effluents; or (3) a matter not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact.

ENO shall maintain records of changes in facility design or operation and of tests and experiments carried out pursuant to this Subsection. These records shall include a written evaluation which provides a basis for the determination that the change, test, or experiment does not involve an unreviewed environmental question nor constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1.0. ENO shall include as part of its Annual Environmental Protection Plan Report per Subsection 5.4.1: brief descriptions, analyses, interpretations, and evaluations of such changes, tests and experiments.

3.2 Reporting Related to the NPDES Permits and State Certifications Violations of the NPDES Permit or the State certification (pursuant to Section 4.1 of the Clean Water Act) shall be reported to the NRC by submittal of copies of the reports required by the NPDES Permit or certification.

Changes and additions to the NPDES Permit or the State certification shall be reported to the NRC within 30 days following the date the change is approved. If a permit or certification, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted.

The NRC shall be notified of changes to the effective NPDES Permit proposed by ENIP3 and ENO by providing NRC with a copy of the proposed change at the same time it is submitted to the permitting agency. The notification of a licensee-initiated change shall include a copy of the requested revision submitted to the permitting agency. ENO shall provide the NRC a copy of 3-2 Renewed License No. DPR-64 Amendment No. 270

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

3.3 Changes Required for Compliance with Other Environmental Regulations Changes in facility design or operation and performance of tests or experiments which are required to achieve compliance with other Federal, State, or local environmental regulations are not subject to the requirements of Section 3.1.

3-3 Renewed License No. DPR-64 Amendment No. 270

4.0 Environmental Conditions 4.1 Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to the handling and storage of spent fuel and maintenance of the facility shall be recorded and promptly reported to the NRC within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by telephone, telegraph, or facsimile transmissions followed by a written report per Subsection 5.4.2. The following are examples: excessive bird impaction events, onsite plant or animal disease outbreaks, unusual mortality or occurrence of any species protected by the Endangered Species Act of 1973, unusual fish kills, unusual increase in nuisance organisms or conditions, and unanticipated or emergency discharge of waste water or chemical substances.

No routine monitoring programs are required to implement this condition.

4.2 Environmental Monitoring In accordance with Section 7(a) of the Endangered Species Act, the National Marine Fisheries Service (NMFS) issued a Biological Opinion related to the continued operation of IP2 and IP3 that pertains to shortnose sturgeon (Acipenser brevirostrum) and Atlantic sturgeon (Acipenser oxyrinchus oxyrinchus). The Biological Opinion includes an Incidental Take Statement with Reasonable and Prudent Measures that the NMFS has determined to be necessary or appropriate to minimize the amount or extent of incidental take and associated Terms and Conditions, which are non-discretionary and implement the Reasonable and Prudent Measures. The currently applicable Biological Opinion concludes that continued operation of IP2 and IP3 is not likely to jeopardize the continued existence of the listed species or to adversely affect the designated critical habitat of those species. This Biological Opinion 4-1 Renewed License No. DPR-64 Amendment No. 270

conservatively bounds the conditions that will occur in the permanently shut down and defueled condition.

Entergy shall adhere to the requirements within the Incidental Take Statement of the currently applicable Biological Opinion. Changes to the Biological Opinion, including the Incidental Take Statement, Reasonable and Prudent Measures, and Terms and Conditions contained therein, must be preceded by consultation between the NRC, as the authorizing agency, and the NMFS.

4-2 Renewed License No. DPR-64 Amendment No. 270

5.0 Administrative Procedures 5.1 Review and Audit ENO shall provide a review and audit of compliance with the Environmental Protection Plan.

The audits shall be conducted independently of the individual or groups responsible for performing the specific activity. A description of the organization structure is utilized to achieve the independent review and audit function and results of the audits activities shall be maintained and made available for inspection.

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

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

5.3 Changes in Environmental Protection Plan Requests for changes in the Environmental Protection Plan shall include an assessment of the environmental impacts of the proposed change and a supporting justification. Implementation of such changes in the EPP shall not commence prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision to the Environmental Protection Plan. This EPP shall be retained for five years or, where applicable, in accordance with the requirements of other agencies.

5-1 Renewed License No. DPR-64 Amendment No. 270

5.4 Plant Reporting Requirements 5.4.1 Routine Reports An Annual Environmental Protection Plan Report describing implementation of this EPP for the previous year shall be submitted to the NRC prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following issuance of the operating license. The period of the first report shall begin with the date of issuance of the operating license.

The report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this Environmental Protection Plan for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous non-radiological environmental monitoring reports, and an assessment of the observed impacts of the previous plant operation and the handling and storage of spent fuel and maintenance of the facility on the environment. If harmful effects or evidence of trends towards irreversible damage to the environment are observed, ENO shall provide a detailed analysis of the data and a proposed course of action to alleviate the problem.

The Annual Environmental Protection Plan Report shall also include:

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

(b) A list of all changes in facility design or operation, tests, and experiments made in accordance with Subsection 3.1 which involved a potentially significant unreviewed environmental issue.

(c) A list of nonroutine reports submitted in accordance with Subsection 5.4.2.

5-2 Renewed License No. DPR-64 Amendment No. 270

5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of a nonroutine event. The report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact and facility conditions, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses.

Events reportable under this subsection which also require reports to other Federal, State, or local agencies shall be reported in accordance with those reporting requirements in lieu of the requirements of this subsection. The NRC shall be provided a copy of such report at the same time it is submitted to the other agency.

5-4 Renewed License No. DPR-64 Amendment No. 270

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

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT 3 NUCLEAR POWER PLANT ENVIRONMENTAL TECHNICAL SPECIFICATION REQUIREMENTS PART II: RADIOLOGICAL ENVIRONMENTAL FACILITY LICENSE NO. DPR-64 DOCKET NUMBER 50-286 Renewed License No. DPR-64 Amendment No. 270

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

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 3 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART I: SPENT FUEL TRANSFER CANISTER AND TRANSFER CASK SYSTEM FACILITY LICENSE NO. DPR-64 DOCKET NO. 50-286 Amendment No. 270

Facility License Appendix C - Inter-Unit Fuel Transfer Technical Specifications SPENT FUEL SHIELDED TRANSFER CANISTER AND TRANSFER CASK SYSTEM

1.0 DESCRIPTION

The spent fuel transfer system consists of the following components: (1) a spent fuel shielded transfer canister (STC), which contains the fuel; (2) a transfer cask (HI-TRAC 100D) (hereafter referred to as HI-TRAC), which contains the STC during transfer operations; and (3) a bottom missile shield.

The STC and HI-TRAC are designed to transfer irradiated nuclear fuel assemblies from the Indian Point 3 (IP3) spent fuel pit to the Indian Point 2 (IP2) spent fuel pit. A fuel basket within the STC holds the fuel assemblies and provides criticality control. The shielded transfer canister provides the confinement boundary, water retention boundary, gamma radiation shielding, and heat rejection capability. The HI-TRAC provides a water retention boundary, protection of the STC, gamma and neutron radiation shielding, and heat rejection capability.

The STC contains up to 12 fuel assemblies.

The STC is the confinement system for the fuel. It is a welded, multi-layer steel and lead cylinder with a welded base-plate and bolted lid. The inner shell of the canister forms an internal cylindrical cavity for housing the fuel basket. The outer surface of the canister inner shell is buttressed with lead and steel shells for radiation shielding. The minimum thickness of the steel, lead and steel shells relied upon for shielding starting with the innermost shell are 3/4 inch steel, 2 3/4 inch lead and 3/4 inch steel, respectively. The canister closure incorporates two O-ring seals to ensure its confinement function. The confinement system consists of the canister inner shell, bottom plate, top flange, top lid, top lid O-ring seals, vent port seal and cover plate, and drain port seal and coverplate. The fuel basket, for the transfer of 12 Pressurized Water Reactor (PWR) fuel assemblies, is a fully welded, stainless steel, honeycomb structure with neutron absorber panels attached to the individual storage cell walls under stainless steel sheathing. The maximum gross weight of the fully loaded STC is 40 tons.

The HI-TRAC is a multi-layer steel and lead cylinder with a bolted bottom (or pool) and top lid.

For the fuel transfer operation the HI-TRAC is fitted with a solid top lid, an STC centering assembly, and a bottom missile shield. The inner shell of the transfer cask forms an internal cylindrical cavity for housing the STC. The outer surface of the cask inner shell is buttressed with intermediate lead and steel shells for radiation shielding. The minimum thickness of the steel, lead and steel shells relied upon for shielding starting with the innermost shell are 3/4 inch steel, 2 inch lead and 1 inch steel, respectively. An outside shell called the water jacket contains water for neutron shielding, with a minimum thickness of 5. The HI-TRAC bottom and top lids incorporate a gasket seal design to ensure its water confinement function. The water confinement system consists of the HI-TRAC inner shell, bottom lid, top lid, top lid seal, bottom lid seal, vent port seal, vent port cap and bottom drain plug.

The HI-TRAC provides a water retention boundary, protection of the STC, gamma and neutron radiation shielding, and heat rejection capability. The bottom missile shield is attached to the bottom of the HI-TRAC and provides tornado missile protection of the pool lid bolted joint. The HI-TRAC can withstand a tornado missile in other areas without the need for additional shielding. The STC centering assembly provides STC position control within the HI-TRAC and also acts as an internal impact limiter in the event of a non-mechanistic tipover accident.

INDIAN POINT 3 1 Amendment No. 270

Facility License Appendix C - Inter-Unit Fuel Transfer Technical Specifications 2.0 CONDITIONS 2.1 OPERATING PROCEDURES Written operating procedures shall be prepared for cask handling, loading, movement, surveillance, maintenance, and recovery from off normal conditions such as crane hang-up.

The written operating procedures shall be consistent with the technical basis described in Chapter 10 of the Licensing Report (Holtec International Report HI-2094289).

2.2 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM Written cask acceptance tests and maintenance program shall be prepared consistent with the technical basis described in Chapter 8 of the Licensing Report (Holtec International Report HI-2094289).

2.3 PRE-OPERATIONAL TESTING AND TRAINING EXERCISE A training exercise of the loading, closure, handling/transfer, and unloading, of the equipment shall be conducted prior to the first transfer. The training exercise shall not be conducted with irradiated fuel. The training exercise may be performed in an alternate step sequence from the actual procedures, but all steps must be performed. The training exercise shall include, but is not limited to the following:

a) Moving the STC into the IP3 spent fuel pool.

b) Preparation of the HI-TRAC for STC loading.

c) Selection and verification of specific fuel assemblies and non-fuel hardware to ensure type conformance.

d) Loading specific assemblies and placing assemblies into the STC (using a single dummy fuel assembly), including appropriate independent verification.

e) Remote installation of the STC lid and removal of the STC from the spent fuel pool.

f) Placement of the STC into the HI-TRAC with the STC centering assembly.

g) STC closure, establishment of STC water level with steam, verification of STC water level, STC leakage testing, and operational steps required prior to transfer, as applicable.

h) Establishment and verification of HI-TRAC water level.

i) Installation of the HI-TRAC top lid.

j) HI-TRAC closure, leakage testing, and operational steps required prior to transfer, as applicable.

k) Movement of the HI-TRAC with STC from the IP3 fuel handling building to the IP2 fuel handling building along the haul route with designated devices.

l) Moving the STC into the IP2 spent fuel pool.

m) Manual crane operations for bare STC movements including demonstration of recovery from a crane hang-up with the STC suspended from the crane.

INDIAN POINT 3 2 Amendment No. 270

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

AND ENTERGY NUCLEAR OPERATIONS, INC. (ENO)

INDIAN POINT NUCLEAR GENERATING UNIT No. 3 INTER-UNIT FUEL TRANSFER TECHNICAL SPECIFICATIONS PART II: TECHNICAL SPECIFICATIONS FACILITY LICENSE NO. DPR-64 DOCKET NO. 50-286 Amendment No. 270

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 270 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-64 ENTERGY NUCLEAR INDIAN POINT 3, LLC AND ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 3 DOCKET NO. 50-286

1.0 INTRODUCTION

By application dated April 28, 2020 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML20132A200), Entergy Nuclear Operations, Inc. (Entergy or the licensee), requested changes to Renewed Facility Operating License (RFOL) No. DPR-64 and the associated Technical Specifications (TSs) for Indian Point Nuclear Generating Unit No. 3 (IP3). Specifically, Entergy requested an amendment to revise the IP3 RFOL and the associated TSs to Permanently Defueled Technical Specifications (PDTS), consistent with the permanent cessation of operations and permanent removal of fuel from the reactor vessel.

2.0 BACKGROUND

By letter dated February 8, 2017 (ADAMS Accession No. ML17044A004), the licensee submitted a Notification of Permanent Cessation of Power Operations for Indian Point Nuclear Generating Unit No. 2 (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, 2024 and 2025, respectively.

Once the licensee submits to the NRC the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel in accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.82(a)(1)(i) and (ii), and the certifications are docketed, the 10 CFR Part 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel.

By letter dated May 12, 2020 (ADAMS Accession No. ML20133J902), Entergy certified 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 (SFP) on May 12, 2020.

Enclosure 2

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 RFOL and the associated TSs to PDTS, consistent with the permanent cessation of operations and permanent removal of fuel from the reactor vessel.1 By letter dated April 15, 2020 (ADAMS Accession No. ML20078L140), the NRC issued Amendment Nos. 293 and 268 for IP2 and IP3, respectively. This amendment revised the on-shift staffing and emergency response organization in the site emergency plan for the post-shutdown and permanently defueled condition.

By letter dated April 10, 2020 (ADAMS Accession No. ML20071Q717), the NRC issued Amendment Nos. 292 and 267 for IP2 and IP3, respectively. This amendment revised and removed certain requirements from the Section 5.0, Administrative Controls, portion of the IP2 and IP3 TSs that are not applicable to the facility in a permanently defueled condition, as well as revised and made editorial changes to the TS Table of Contents; Section 1.1, Definitions; and Section 4.0, Design Features.

By letter dated December 18, 2019 (ADAMS Accession No. ML19333B868), the NRC approved the certified fuel handler training and retraining program for IP2 and IP3.

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 (SLs), 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.

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.

Section 50.36 of 10 CFR 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 focus on instrumentation to detect degradation of the reactor coolant system (RCS) pressure boundary and process variables; design features; operating restrictions; or structures, systems, and components (SSCs) that affect the integrity of fission product barriers during design-basis accidents (DBAs) or transients. They also focus on SSCs which operating experience or probabilistic risk assessment have shown to be significant to public health and safety. A general discussion of how these criteria were evaluated to ensure that the TS LCOs proposed for deletion will no longer be required to be included in the TSs is provided below.

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

Criterion 1 of 10 CFR 50.36(c)(2)(ii)(A) states that TS LCOs must be established for installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. Since no fuel will be present in the reactor or RCS at the IP3 facility following permanent defueling, 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. The applicable DBAs for IP3 in the permanently defueled condition are the fuel handling accident (FHA) in the fuel handling building, accidental release-recycle of waste liquid, and the accidental release of waste gas, which are discussed in Section 4.0 of this safety evaluation (SE).

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 primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion) so that the plant response to DBAs and transients limits the consequences of these events to within the appropriate acceptance criteria.

Chapter 14, Safety Analysis, of the IP3 Updated Final Safety Analysis Report (UFSAR) describes the DBA scenarios that are applicable during plant operations (ADAMS Accession No. ML19282B159). After the docketed certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel are submitted to the NRC in accordance with 10 CFR 50.82(a)(1)(i) and (ii), the 10 CFR Part 50 license will no longer permit operation of the reactor or placement of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). With the reactor in a permanently shutdown and defueled condition, the SFP and its cooling systems will be dedicated only to spent fuel storage. In this condition, the spectrum of credible accidents is much smaller than for an operational plant. Therefore, most of the accident scenarios postulated in Chapter 14 of the IP3 UFSAR will no longer be applicable after IP3 is in the permanently defueled condition. Note that, at Indian Point, the SFP may also be referred to as the spent fuel pit. The scope of applicable DBAs that continues to apply to IP3 is discussed in more detail in Section 4.0 of this SE.

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. There will no longer be any applicability of DBAs at IP3, once the reactor is in the permanently shutdown and defueled condition, that can result in a significant offsite radiological risk to public health and safety.

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.

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 SFP. 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 SFP, 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 UFSAR 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.

The regulation in 10 CFR 100.11, Determination of exclusion area, low population zone, and population center distance, states, in part:

(a) As an aid in evaluating a proposed site, an applicant should assume a fission produce release[2] from the core, the expected demonstrable leak rate from the containment and the meteorological conditions pertinent to his site to derive an exclusion area, a low population zone and population center distance. For the purpose of this analysis, which shall set forth the basis for the numerical values used, the applicant should determine the following:

(1) An exclusion area of such size that an individual located at any point on its boundary for two hours immediately following onset of the postulated fission product release would not receive a total radiation 2 The fission product release assumed for these calculations should be based upon a major accident, hypothesized for purposes of site analysis or postulated from considerations of possible accidental events, that would result in potential hazards not exceeded by those from any accident considered credible. Such accidents have generally been assumed to result in substantial meltdown of the core with subsequent release of appreciable quantities of fission products.

dose to the whole body in excess of 25 rem [roentgen equivalent man] [3] or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

(2) A low population zone of such size that an individual located at any point on its outer boundary who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage) would not receive a total radiation dose to the whole body in excess of 25 rem or a total radiation dose in excess of 300 rem to the thyroid from iodine exposure.

The NRC staff evaluated the radiological consequences of the postulated FHA DBA against the dose criteria specified in 10 CFR 50.67, Accident source term, and using the guidance described in Regulatory Guide (RG) 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, dated July 2000 (ADAMS Accession No. ML003716792). RG 1.183 provides guidance to licensees on acceptable application of alternative source term (AST) submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

By letter dated March 24, 2005 (ADAMS Accession No. ML050870383), the NRC issued Amendment No. 225 for IP3. This amendment revised the IP3 Operating License and TSs to increase the licensed rated thermal power by 4.85 percent from 3,067.4 megawatts thermal (MWt) to 3,216 MWt. As part of the IP3 power uprate project, a reanalysis of the several DBAs was performed, including the waste gas decay tank accident, which represents the current analysis of record.

By letter dated March 22, 2005 (ADAMS Accession No. ML050750431), the NRC issued Amendment No. 224 for IP3. This amendment approved the alternative source methodology for DBAs, including for the FHA for IP3, in accordance with 10 CFR 50.67 to perform the radiological consequences analyses of DBAs as described in RG 1.183. This approval included consideration of the aforementioned power uprate.

The FHA-specific dose acceptance criteria are specified in NUREG-0800, Revision 0, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR

[Light-Water Reactor] Edition (SRP), Section 15.0.1, Radiological Consequence Analyses Using Alternative Source Terms, dated July 2000 (ADAMS Accession No. ML003734190).

The dose acceptance criteria for the FHA are a total effective dose equivalent (TEDE) of 6.3 rem at the exclusion area boundary (EAB) for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 6.3 rem at the outer boundary of the low population zone (LPZ), and 5 rem in the control room for the duration of the accident.

3 The whole body dose of 25 rem referred to above corresponds numerically to the once in a lifetime accidental or emergency dose for radiation workers which, according to NCRP recommendations may be disregarded in the determination of their radiation exposure status (see NBS Handbook 69 dated June 5, 1959). However, neither its use nor that of the 300 rem value for thyroid exposure as set forth in these site criteria guides are intended to imply that these numbers constitute acceptable limits for emergency doses to the public under accident conditions. Rather, this 25 rem whole body value and the 300 rem thyroid value have been set forth in these guides as reference values, which can be used in the evaluation of reactor sites with respect to potential reactor accidents of exceedingly low probability of occurrence, and low risk of public exposure to radiation.

The regulations in 10 CFR 50.67(b)(2) state, in part, that the NRC may issue the amendment only if the applicants analysis demonstrates with reasonable assurance that:

(i) An individual located at any point on the boundary of the exclusion area for any 2-hour period following the onset of the postulated fission product release, would not receive a radiation dose in excess of 0.25 Sv [Sievert]

(25 rem)4 total effective dose equivalent (TEDE).

(ii) An individual located at any point on the outer boundary of the low population zone, who is exposed to the radioactive cloud resulting from the postulated fission product release (during the entire period of its passage), would not receive a radiation dose in excess of 0.25 Sv (25 rem) total effective dose equivalent (TEDE).

(iii) Adequate radiation protection is provided to permit access to and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 0.05 Sv (5 rem) total effective dose equivalent (TEDE) for the duration of the accident.

Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC),

Criterion 19, Control room, states, in part:

A control room shall be provided from which actions can be taken to operate the nuclear power unit safely under normal conditions and to maintain it in a safe condition under accident conditions, including loss-of-coolant accidents.

Adequate radiation protection shall be provided to permit access and occupancy of the control room under accident conditions without personnel receiving radiation exposures in excess of 5 rem whole body, or its equivalent to any part of the body, for the duration of the accident. Equipment at appropriate locations outside the control room shall be provided (1) with a design capability for prompt hot shutdown of the reactor, including necessary instrumentation and controls to maintain the unit in a safe condition during hot shutdown, and (2) with a potential capability for subsequent cold shutdown of the reactor through the use of suitable procedures.

The emergency planning requirements of 10 CFR 50.47, Emergency plans, and Appendix E to 10 CFR Part 50, Emergency Planning and Preparedness for Production and Utilization Facilities, continue to apply to a nuclear power reactor after permanent cessation of operations and removal of fuel from the reactor vessel. There are no explicit regulatory provisions distinguishing emergency planning requirements for a power reactor that has been permanently shut down from those for an operating power reactor. The NRC staff notes that the risk of an offsite radiological release is significantly lower, and the types of possible accidents are significantly fewer at a nuclear power reactor that has permanently ceased operations and removed fuel from the reactor vessel than at an operating power reactor.

Nuclear Energy Institute (NEI) topical report NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors, dated November 2012 (ADAMS 4

The use of 0.25 Sv (25 rem) TEDE is not intended to imply that this value constitutes an acceptable limit for emergency doses to the public under accident conditions. Rather, this 0.25 Sv (25 rem) TEDE value has been stated in this section as a reference value that can be used in the evaluation of proposed design-basis changes with respect to potential reactor accidents of exceedingly low probability of occurrence and low risk of public exposure to radiation.

Accession No. ML12326A805), provides guidance for the development of emergency action levels for reactors in a permanently defueled condition. NEI 99-01, Revision 6, was endorsed by the NRC in a letter dated March 28, 2013 (ADAMS Accession No. ML12346A463).

Revision 6 of NEI 99-01 states that the accident analysis necessary to adopt the permanently defueled emergency action level scheme must confirm that the source terms and release motive forces are not sufficient to warrant classification of a site area emergency or general emergency. A site area emergency would be declared for any events where exposure levels beyond the site area boundary are expected to exceed 10 percent of the Environmental Protection Agency (EPA) Protective Action Guidelines (PAGs). The EPA PAG for sheltering or evacuation of the public is a projected dose of 1 to 5 rem total effective dose (TED5) in 4 days.

In addition, the EPA PAG for recommending the administration of potassium iodide (KI) (as a thyroid blocking agent) is a projected dose of 5 rem to the child thyroid from radioactive iodine.

Correspondingly, NEI 99-01 established the site area emergency classification threshold as 100 millirem (mrem) TEDE or 500 mrem thyroid committed dose equivalent.

RG 1.183 provides the methodology for analyzing the radiological consequences of several DBAs to show compliance with 10 CFR 50.67. RG 1.183 provides guidance to licensees on acceptable application of AST submittals, including acceptable radiological analysis assumptions for use in conjunction with the accepted AST.

SRP Section 15.0.1 provides review guidance to the NRC staff for the review of AST amendment requests. Section 15.0.1 states that the NRC reviewer should evaluate the proposed change against the guidance in RG 1.183. The dose acceptance criteria for the FHA are a TEDE of 6.3 rem at the EAB for the worst 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 6.3 rem at the outer boundary of the LPZ, and 5 rem in the control room for the duration of the accident.

Regulatory Issue Summary 2006-04, Experience with Implementation of Alternative Source Terms, dated March 7, 2006 (ADAMS Accession No. ML053460347), discusses experiences with analyzing an accident involving a release from off-gas or waste systems. As part of full AST implementation, some licensees have included an accident involving a release from their off-gas or waste gas system. For this type of accident, licensees have proposed acceptance criteria of 500 mrem TEDE. The acceptance criterion for this event is that associated with the dose to an individual member of the public as described in 10 CFR Part 20, Standards for Protection Against Radiation. When the NRC revised 10 CFR Part 20 to incorporate a TEDE dose, the offsite dose to an individual member of the public was changed from 500 mrem whole body to 100 mrem TEDE. Therefore, any licensee who chooses to implement AST for an off-gas or waste gas system release should base its acceptance criteria on 100 mrem TEDE.

Licensees may also choose not to implement AST for this accident and continue with their existing analysis and acceptance criteria of 500 mrem whole body.

Branch Technical Position 11-5, Postulated Radioactive Release Due to a Waste Gas System Leak or Failure, of SRP Chapter 11, Radioactive Waste Management, provides guidance to the reviewer for assessing the analysis of an accidental release from the waste gas system.

5 For the purposes of this SE, the terms TED and TEDE are used interchangeably as both describing the combined effects of internal and external radiation exposure.

3.3 Spent Fuel Pool Criticality GDC 61, Fuel storage and handling and radioactivity control, requires, in part, that these systems shall be designed with a capability to permit appropriate periodic inspection and testing of components important to safety.

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

Per 10 CFR 50.68(a), each holder of an operating license shall comply with either 10 CFR 70.24 or the requirements in 10 CFR 50.68(b). The licensee has elected to meet 10 CFR 50.68(b) and, accordingly, must comply with the following requirements:

(1) Plant procedures shall prohibit the handling and storage at any one time of more fuel assemblies than have been determined to be safely subcritical under the most adverse moderation conditions feasible by unborated water.

(2) The estimated ratio of neutron production to neutron absorption and leakage (k-effective) of the fresh fuel in the fresh fuel storage racks shall be calculated assuming the racks are loaded with fuel of the maximum fuel assembly reactivity and flooded with unborated water and must not exceed 0.95, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such flooding or if fresh fuel storage racks are not used.

(3) If optimum moderation of fresh fuel in the fresh fuel storage racks occurs when the racks are assumed to be loaded with fuel of the maximum fuel assembly reactivity and filled with low-density hydrogenous fluid, the k-effective corresponding to this optimum moderation must not exceed 0.98, at a 95 percent probability, 95 percent confidence level. This evaluation need not be performed if administrative controls and/or design features prevent such moderation or if fresh fuel storage racks are not used.

(4) If no credit for soluble boron is taken, the k-effective [estimated ratio of neutron production to neutron absorption and leakage] of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water.

The regulations in 10 CFR 50.36(b) require TSs to be derived from the analyses and evaluation included in the safety analysis report and amendments thereto. As required by 10 CFR 50.36(c)(4), the TSs will include design 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 10 CFR 50.36.

4.0 TECHNICAL EVALUATION

4.1 Accident Analysis Chapter 14 of the IP3 UFSAR describes the postulated DBA and transient scenarios applicable to IP3 during power operations. They demonstrate that the plant could be operated safely and that radiological consequences from postulated accidents do not exceed the regulatory guidelines of 10 CFR 50.67 or 10 CFR Part 100, as applicable. Two basic groups of events are pertinent to safety, which are abnormal operational transients and postulated DBAs; these two groups were investigated separately. The analyses of the abnormal operational transients evaluate the ability of the plant protection features to ensure that during these transients, no fuel damage occurs, and the RCS pressure limit is not exceeded. The safety design limits require that damage to the fuel be limited and that no nuclear system process barrier damage results from any abnormal operational occurrence. Thus, analysis of this group of events evaluates the features that protect the first two radioactive material barriers. Analysis of the events in the second group, postulated DBAs, evaluates situations that require functioning of the engineered safeguards in order to protect the fission product barriers, including containment, in order to minimize the offsite radiological consequences.

The most severe postulated DBA involves damage to the nuclear reactor core and the release of large quantities of fission products. Many of these accident scenarios involve failures or malfunctions of systems, which could affect the fuel in the reactor vessel. With the termination of reactor operations and the permanent removal of fuel from the reactor vessel, such accidents are no longer possible. Therefore, the postulated accidents involving failure or malfunction of the reactor, reactor cooling system, steam system, or turbine generator are no longer applicable. While spent fuel remains in the SFP, the accidents that remain applicable to IP3 in the permanently shutdown and defueled condition are the FHA in the fuel handling building and the accidental release of waste liquid or waste gas. For completeness, the NRC staff also evaluated the applicability of other DBAs documented in the IP3 UFSAR to ensure that these accidents would not have consequences that could potentially exceed the 10 CFR 100.11 dose limits and RG 1.183 dose acceptance criteria.

4.1.1 Fuel Handling Accident As discussed in Section 3.2 of this SE, an analysis of the FHA utilizing the AST methodology described in RG 1.183 was previously approved by the NRC in License Amendment No. 224 on March 22, 2005. Later, as part of the IP3 power uprate project, a reanalysis of the FHA was performed utilizing the AST methodology, which is currently the analysis of record as presented in Section 14.2.1, Fuel Handling Accidents, of the IP3 UFSAR.

Section 14.2.1 of the IP3 UFSAR evaluates the various fuel handling accident scenarios. After the reactor has been completely defueled following permanent shut down, an FHA in the reactor cavity is no longer a credible accident. The DBA FHA in the SFP is applicable when IP3 is in a permanently shutdown and defueled condition. The licensees analysis applied the AST methodology outlined in RG 1.183, with the exception of fission product gap activity fractions because the licensee could not show that all fuel would meet the burnup conditions in footnote 11 of RG 1.183. Therefore, the licensees analysis assumed fission product gap release fractions to be 30 percent for Kr-85, 12 percent for I-131, and 10 percent for other iodines and noble gases. These gap release fractions are consistent with RG 1.25, Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling

Accident in the Fuel Handling and Storage Facility for Boiling Water and Pressurized Water Reactors, dated March 1972 (ADAMS Accession No. ML083300022), as modified by NUREG/CR-5009, Assessment of the Use of Extended Burnup Fuel in Light Water Reactors, dated February 1988, which was approved in License Amendment No. 215.6 The analysis was performed to determine the radiological consequences to operators in the control room and the public at the EAB and LPZ as a function of time after shutdown. The analysis demonstrates that the dose consequences of the FHA will remain within the licensing basis dose limits, provided the SFP water level requirements of TS LCO 3.7.14 are met, without crediting fuel storage building ventilation, the station vent radiation monitors, control room isolation, and control room filtration assuming 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> of decay time following shutdown. In addition, after a decay time of at least 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> (30 days) prior to fuel movement, the analysis of the FHA results in an EAB TEDE dose of 0.47 rem.

The FHA is defined as the dropping of a single spent fuel assembly in the SFP during fuel handling activities, such that all 204 fuel rods in the assembly suffer mechanical damage to the cladding. The assembly activity inventory is based on the at-power core average power level with an additional radial peaking factor of 1.7 times. The gap activity in the damaged rods is instantaneously released into the SFP. The release occurs under 23 feet of water, which acts as a filter. Proposed TS LCO 3.7.14 will ensure the minimum water level in the SFP is established prior to fuel handling and maintained during fuel handling operations. The activity released is assumed to be reach the environment outside the building within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. As mentioned above, no credit is taken for removal of iodine by filters nor is credit taken for isolation of the release path.

The analysis concludes that without crediting mitigation by any active SSC, the calculated TEDE values to the control room, EAB, and LPZ are less than the limits set forth in 10 CFR 50.67 and RG 1.183. In addition, after a decay time of at least 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> (30 days) prior to fuel movement, the analysis of the FHA results in an EAB TEDE dose of 0.47 rem, which is less than the EPA PAGs recommended threshold for evacuation of 1 rem.

In performing this review, the NRC staff relied upon information provided by the licensee and NRC staff experience in performing similar reviews. The NRC staff concludes that the dose consequence from an FHA for the permanently defueled IP3 meets the applicable radiological dose criteria at the EAB, LPZ, and in the control room.

4.1.2 Accidental Release of Waste Gas Section 14.2.3, Accidental Release - Waste Gas, of the IP3 UFSAR evaluates the accidental release of waste gas. Tanks accumulating significant quantities of radioactive gases during operation are the volume control tanks, the liquid holdup tanks, and the gas decay tanks. The volume control tank accumulates gases over a core cycle by stripping action of the entering spray. During a refueling shutdown, this activity is vented to the waste gas system and stored for decay. The liquid holdup tanks receive reactor coolant after passing through demineralizers during the process of coolant purification.

The volume control tank and liquid holdup tanks were not considered in this analysis since gaseous products from these liquid tanks are collected and compressed in the waste gas decay tanks for decay prior to release. Potential liquid waste releases are considered from these tanks; however, any liquid releases are retained in the building or sumps, and only volatilized 6 Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Selective Adoption of Alternate Source Term (TAC No. MB5382), dated March 17, 2003 (ADAMS Accession No. ML030760135).

components would be released to the environment. These volatilized components are evaluated as part of the waste gas decay tank accident.

The waste gas tank accident assumes an instantaneous rupture with a conservative ground-level release all of noble gases. The 50,000 curie (Ci) dose equivalent Xenon (Xe)-133 waste gas tank activity assumed in this calculation bounds the current Xe-133 dose-equivalent limit of 29,761 Ci, as well as the administrative Xe-133 dose-equivalent limit of 6,000 Ci.

Once the reactor is permanently shut down and defueled, there is no mechanism to raise the primary coolant activity. Therefore, upon permanent shutdown and cooldown, the source term contained within the waste gas tanks represents the worst case source term, which is expected to be less than the assumed waste gas tank rupture analysis of record, and thus is bounded.

Subsequent additions to the waste gas tanks resulting from water management activities would be less than the final shutdown and cooldown waste gas tank source term.

The analysis concludes that without crediting mitigation by any active SSC, the calculated TEDE to the control room is less than the limit set forth in 10 CFR 50.67 and the whole-body dose value of 500 mrem at the EAB and LPZ.

In performing this review, the NRC staff relied upon information provided by the licensee and NRC staff experience in performing similar reviews, including a similar amendment for IP3. The NRC staff concludes that the dose consequence from a waste gas tank release for the permanently defueled IP3 meets the applicable radiological dose criteria at the control room, EAB, and LPZ.

4.1.3 Accidental Release of Waste Liquid The accidental release of liquid waste is discussed in Section 14.2.2, Accidental Release of Waste Liquid, of the IP3 UFSAR, which concludes that a potential liquid waste release collects in building sumps or is retained in building vaults and not released to the environment. As such, the licensee states the hazard from these releases is derived only from any volatilized components. The volatilized components are what comprise the waste gas accident discussed in the previous section of this SE. The licensee states that a separate liquid-specific release accident evaluation is not required to be performed regarding removal of supporting systems such as primary auxiliary building ventilation, station vent radiation monitors, control room isolation, and control room filtration. Based on the above, the NRC staff concludes that the dose consequence from a liquid waste incident for IP3 in the permanently shutdown and defueled condition will not approach the EPA PAGs for sheltering or evacuation and would not trigger the declaration of a site area emergency.

4.1.4 Accident Analysis Conclusions The NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological impacts of the proposed changes. The staff finds that the licensees proposed changes use analysis methods and assumptions consistent with the guidance contained in RG 1.183. The staff compared the doses estimated by the licensee to the applicable criteria and to the results of confirmatory analyses performed by the staff. The staff finds that there is reasonable assurance that IP3, following adoption of the proposed amendment, will continue to provide sufficient safety margins with adequate defense in depth to address unanticipated events and to compensate for uncertainties in accident progression and in analysis assumptions and parameters. The staff concludes that the licensee has demonstrated that the dose

consequences for postulated accidents at the permanently shutdown and defueled plant would not have consequences that could potentially exceed the 10 CFR 50.67 dose limits and RG 1.183 dose acceptance criteria or approach the EPA PAG criteria of 1 rem TED after a 30-day fuel decay period prior to fuel movement. Therefore, the staff finds the proposed changes to be acceptable from a dose consequence perspective.

4.2 Boral Neutron Absorber Surveillance Program The spent fuel storage racks (as described in the IP2 and IP3 License Renewal Application (LRA), Appendix B, Section 2.3.3.1, Spent Fuel Pit Cooling, dated April 23, 2007 (ADAMS Accession No. ML071210517)) provide a storage location at the bottom of the spent fuel pit for spent fuel assemblies. The racks (auxiliary system according to NUREG-1801, Volume 2 Generic Aging Lessons Learned (GALL) Report: Tabulation of Results, dated September 2005 (ADAMS Accession No. ML052110006) are full length, top entry type. Spent fuel pit cooling racks are constructed of stainless steel with Boral as a neutron-absorbing material (NAM).

Reduction of neutron-absorbing capacity and loss of material due to general corrosion are aging effects requiring management for Boral spent fuel storage racks exposed to a treated borated water environment. These aging effects are managed by the Boral Surveillance Program (as described in the licensees LRA).

The Boral surveillance program at IP3, as described in Section A.2.1.3, Boral Surveillance Program, of the IP3 UFSAR Appendix A (ADAMS Accession No. ML19282B087) and in its responses to Generic Letter 2016-01, Monitoring of Neutron-Absorbing Materials in Spent Fuel Pools" (GL 2016-01) dated November 3, 2016 (ADAMS Accession No. ML16314E266), and May 30, 2018 (ADAMS Accession No. ML18151A858), consists of a coupon surveillance program. The coupons were taken from the same lots of material used in construction of the racks and encased in a similar manner as the in-service material. The coupons are thus able to detect aging/degradation mechanisms that the in-service materials experience. Boral inspection and testing activities are conducted at a frequency of at least once every 10 years.

The continued performance of Boral as a neutron absorbing material in the SFP is managed by the aging management programs; Boral Surveillance Program as supplemented by the Water Chemistry Control - Primary and Secondary Program.

Specifically, as stated in its November 3, 2016, response letter to GL 2016-01, the Boral panels that are used are a composite material made of boron carbide and aluminum in three distinct layers. The outer layers are 0.0125 inches thick aluminum cladding (type 1100 alloy aluminum).

The center layer is a vented 0.05-inch-thick uniform aggregate of boron carbide particles held in an aluminum alloy matrix. Vented means that it is clad on the front and back only, while the sides are unclad to allow for gas migration and release. These strips of Boral are scalloped such that they may be held in place by welded coins and are not themselves welded. Strips of Boral poison are retained on the outer sides of the rack modules by a thin sheet of stainless steel. The individual SFP cells are fabricated from American Society for Testing and Materials ASTM A-240, Type 304 stainless steel.

The NRC staff verified that the existing Boral Surveillance Program is not changed by the proposed PDTS license amendment and that the inspection frequency and acceptance criteria will remain in place until such time as there is no longer fuel in the IP3 SFP. The NRC staff confirmed with the licensee that the IP3 Boral Surveillance Program as described in Appendix A of the IP3 UFSAR, will be retained in the Defueled Safety Analysis Report (DSAR) which will be

the document that replaces the UFSAR and will remain applicable to IP3 once it is shut down and defueled (ADAMS Accession No. ML21073A002).

During the period in which IP3 has ceased operations and is permanently defueled, there may be spent fuel in the SFP which requires criticality control. In order to ensure 10 CFR 50.68 is met, in part, the condition of the NAM is monitored to ensure that it does not degrade below the minimum Boron-10 areal density that the licensee assumes in its criticality safety analysis.

As stated above, LRA Appendix A, Section A.3.1.3, and the licensees GL 2016-01 response (See ADAMS accession numbers at Section 3.1) describes the licensees NAM monitoring program. The program contains provisions for routine monitoring and corrective actions to ensure the Boral can meet its neutron attenuation capability as assumed in the SFP criticality safety analysis.

The Boral Surveillance Program is an existing program that provides assurance the Boral neutron absorbers in the spent fuel racks maintain the validity of the criticality analysis in support of the rack design. The program relies on representative coupon samples mounted in surveillance assemblies located in the SFP to monitor performance of the absorber material without disrupting the integrity of the storage system. Coupon samples mounted in surveillance assemblies are removed from the SFP on a prescribed schedule and physical and chemical properties are measured. From these data, the stability and integrity of the Boral in the storage cells are assessed.

The licensees NAM monitoring program was incorporated into the IP3 UFSAR as approved in Supplement 2 to NUREG-1930, Safety Evaluation Report Related to the License Renewal of Indian Point Nuclear Generating Plant, Units 2 and 3, Appendix A, dated July 2015 (ADAMS Accession No. ML15188A383). The proposed IP3 PDTS amendment did not propose any revisions to the NAM monitoring program.

4.2.1 NAM Monitoring Program The NRC staff reviewed the IP3 NAM monitoring program, as there will be spent fuel in the SFP after IP3 permanently ceases operations and its reactor is defueled. The staff noted that the NAM monitoring program contains provisions to ensure the Boral does not degrade and can continue to perform its neutron attenuation function as described in the Section A.2.1.3 of the IP3 UFSAR Appendix A.

The NRC staff finds the NAM monitoring program is acceptable because the continued performance of Boral (as described in LRA Section 3.3.2.1.1) as a neutron-absorbing material in the SFP is supplemented by the Water Chemistry Control - Primary and Secondary Program, which manages the degradation of Boral including the reduction of neutron absorbing capacity consistent with NUREG-1801 material, environment, and aging effects.

The licensee has not proposed any changes to the NAM monitoring program found in LRA. The staff finds this acceptable as the NAM monitoring program was previously reviewed and approved in the LRA. The program provides reasonable assurance that the licensee will continue to monitor the condition of the NAM and take appropriate corrective actions if it degrades.

4.2.2 NAM Monitoring Program Conclusion The NRC staff reviewed the portion of the LAR related to the IP3 NAM monitoring program and has determined that the monitoring program as described in the IP3 LRA provides reasonable assurance that the licensee will be able to detect degradation of the neutron-absorbing material before its ability to perform its intended safety function is impacted. On this basis, the staff has concluded that the continuation of the NAM monitoring program and the contents of the program after permanent cessation of reactor operations, meet the applicable requirements of 10 CFR 50.68 and GDCs 61 and 62, and are therefore, acceptable.

4.3 Proposed Changes to the Renewed Facility Operating License 4.3.1 License Title The current license title is Renewed Facility Operating License.

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

The proposed change to the title to delete Operating would provide a more accurate description of the facility during the permanently shutdown and defueled condition. Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable to the NRC staff.

4.3.2 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, as amended, 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. Since, consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize use of the facility for power operation or emplacement or retention of fuel into the reactor vessel, this change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.

4.3.3 License Condition 2 Currently, License Condition 2 reads:

Accordingly, Renewed Facility Operating License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows:

The licensee proposed License Condition 2 to read:

Accordingly, Renewed Facility License No. DPR-64 is hereby issued to ENIP3 and ENO to read as follows:

Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. The removal of the discussion of Operating license would provide accuracy in the 10 CFR Part 50 license description. Therefore, the NRC staff approves the proposed change to License Condition 2.

4.3.4 License Condition 2.A Currently, License Condition 2.A reads:

This renewed license applies to the Indian Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and associated equipment (the facility), owned by ENIP3 and operated by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the Final Facility Description and Safety Analysis Report, as supplemented and amended, and the Environmental Report, as amended.

The licensee proposed License Condition 2.A to read:

This renewed license applies to the Indian Point Nuclear Generating Unit No. 3, a pressurized water nuclear reactor and associated equipment (the facility), owned by ENIP3 and maintained by ENO. The facility is located in Westchester County, New York, on the east bank of the Hudson River in the Village of Buchanan, and is described in the Defueled Safety Analysis Report, as supplemented and amended, and the Environmental Report, as amended.

Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. The removal of the discussion of Operating license would provide accuracy in the 10 CFR Part 50 license description. In addition, it is modified to reflect that a Defueled Safety Analysis Report will be prepared to address the permanently shutdown and defueled condition. Therefore, the NRC staff approves the proposed change to License Condition 2.A.

4.3.5 License Condition 2.B.(1)

Currently, License Condition 2.B.(1) reads:

Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) ENIP3 to possess and use, and (b) ENO to possess, use and operate, the facility at the designated location in

Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; The licensee proposed License Condition 2.B.(1) to read:

Pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, (a) ENIP3 to possess and use, and (b) ENO to possess and use the facility at the designated location in Westchester County, New York, in accordance with the procedures and limitations set forth in this renewed license; Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. The facility would remain authorized to possess the existing spent fuel and use the systems required to support safe fuel storage (e.g., the SFP) during the decommissioning period in accordance with the specified limitations for storage. The removal of the discussion of operating would provide accuracy in the 10 CFR Part 50 license description. Therefore, the NRC staff approves the proposed change to License Condition 2.B.(1).

4.3.6 License Condition 2.B.(2)

Currently, License Condition 2.B.(2) reads as follows:

ENO pursuant to the Act and 10 CFR Part 70, to receive, possess, and use, at any time, special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Facility Description and Safety Analysis Report, as supplemented and amended; The licensee proposed License Condition 2.B.(2) to read:

ENO pursuant to the Act and 10 CFR Part 70, to possess, at any time, special nuclear material that was used as reactor fuel, in accordance with the limitations for storage, as described in the Defueled Safety Analysis Report, as supplemented and amended; The proposed change to this license condition would remove the authorization for receipt and use of special nuclear material (SNM) as reactor fuel. It would eliminate the reference to use of the SNM for reactor operations and limit the possession of SNM to SNM that was used as reactor fuel. Pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for IP3 will no longer authorize operation of the reactor. As such, IP3 has no need to receive SNM in the form of reactor fuel and cannot use SNM as reactor fuel for reactor operations. The continued authorization to possess SNM that was used as reactor fuel is necessary, as IP3 possesses reactor fuel that was used for past operations. Therefore, the NRC staff approves the proposed change to License Condition 2.B.(2).

4.3.7 License Condition 2.B.(3)

Currently, License Condition 2.B.(3) reads as follows:

ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; The licensee proposed License Condition 2.B.(3) to read:

ENO pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use, at any time, any byproduct source and special nuclear material as sealed neutron sources that were used for reactor startup, sealed sources that were used for calibration of reactor instrumentation and are used in the calibration of radiation monitoring equipment, and that were used as fission detectors in amounts as required; The proposed change to this license condition revises the IP3 authorization to receive, possess and use any byproduct, source and special nuclear material as sealed neutron sources to clarify that these sealed neutron sources previously were used for reactor startup, calibration of reactor instrumentation, and as fission detectors. The proposed license condition is also modified to reflect that the use of neutron sources for calibration of radiation monitoring equipment will continue to be authorized. Since the IP3 license will no longer authorize operation of the facility pursuant to 10 CFR 50.82(a)(2), this revised license condition is consistent with the requirements associated with the decommissioning plant. Therefore, the NRC staff approves the proposed change to License Condition 2.B.(3).

4.3.8 License Condition 2.B.(4)

Currently, License Condition 2.B.(4) reads as follows:

ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components.

The licensee proposed License Condition 2.B.(4) to read:

ENO pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration; or associated with radioactive apparatus or components; The proposed change to this license condition is a grammatical correction by replacing the ending period with a semi-colon. The change is editorial and does not change any technical requirement. The NRC staff approves the proposed change to License Condition 2.B.(3).

4.3.9 License Condition 2.B.(5)

Currently, License Condition 2.B.(5) reads as follows:

ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

The licensee proposed License Condition 2.B.(5) to read:

ENO pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials that were produced by the operation of the facility.

This license condition is proposed for revision to allow possession, but not separation, of byproduct, and SNM that were produced by the operation of the facility, as opposed to those materials as may be produced by the operation of the facility. Since the IP3 license will no longer authorize operation of the facility pursuant to 10 CFR 50.82(a)(2), this license condition is consistent with the requirements associated with a permanently shutdown and defueled facility in the process of decommissioning. Therefore, the NRC staff finds the proposed change to License Condition 2.B.(5) appropriate and acceptable.

4.3.10 License Condition 2.C.(1)

Currently, License Condition 2.C.(1) reads:

Maximum Power Level ENO is authorized to operate the facility at steady state reactor core power levels not in excess of 3216 megawatts thermal (100% of rated power).

The licensee proposed to delete License Condition 2.C.(1). Since Entergy docketed the IP3 certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, reference to operation of the facility would be inconsistent with 10 CFR 50.82(a)(2).

The NRC staff reviewed the proposed deletion of License Condition 2.C.(1) and determined that operation would not be authorized at IP3 at any power level since its 10 CFR 50.82(a)(1) certifications were docketed. Therefore, the NRC staff finds the proposed deletion of License Condition 2.C.(1) acceptable.

4.3.11 License Condition 2.C.(2)

Currently, License Condition 2.C.(2) reads:

Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 269, are hereby incorporated in the Renewed License.

ENO shall operate the facility in accordance with the Technical Specifications.

The licensee proposed License Condition 2.C.(2) to read:

Technical Specifications The Technical Specifications contained in Appendices A, B, and C, as revised through Amendment No. 270, are hereby incorporated in the Renewed License.

ENO shall maintain the facility in accordance with the Technical Specifications.

Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. This license condition is proposed for revision to account for the permanently defueled condition of the facility and to incorporate the permanently defueled TSs. The license condition is changed from operate the facility to maintain the facility, which describes the permanently defueled condition in which the IP3 license will no longer authorize the use of the facility for power operation. Therefore, the NRC staff finds the proposed change to License Condition 2.C.(2) acceptable.

4.3.12 License Condition 2.C.(3) and 2.C.(4)

Currently, License Conditions 2.C.(3) and 2.C.(4) both read: (DELETED).

License Condition 2.C.(3) and 2.C.(4) are historical references to deleted license conditions.

The licensee proposed to delete these license conditions in their entirety. The change to the license conditions is administrative and does not change any technical requirement. The NRC staff approves the proposed change to delete License Conditions 2.C.(3) and 2.C.(4) in their entirety.

4.3.13 License Condition 2.H Currently, License Condition 2.H reads:

ENO shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report for Indian Point Nuclear Generating Unit No. 3 and as approved in NRC fire protection safety evaluations (SEs) dated September 21, 1973, March 6, 1979, May 2, 1980, November 18, 1982, December 30, 1982, February 2, 1984, April 16, 1984, January 7, 1987, September 9, 1988, October 21, 1991, April 20, 1994, January 5, 1995, and supplements thereto, subject to the following provision:

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

The licensee proposed to delete License Condition 2.H. The licensee stated that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). As a result, the fire protection program (FPP) will be revised to take into account the facility conditions and activities during decommissioning. IP3 will continue to utilize the defense-in-depth concept, placing special emphasis on detection and suppression in order to minimize radiological releases to the environment. This license condition, which is based on maintaining an FPP at an operating reactor in accordance with 10 CFR 50.48 with the ability to achieve and maintain safe shutdown of the reactor in the event

of a fire, will no longer be applicable at IP3. However, many of the elements that are applicable for the operating plant FPP continue to be applicable during facility decommissioning. During the decommissioning process, an FPP is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. The regulation is applicable regardless of whether a requirement for an FPP is included in the facility license. Therefore, a license condition requiring such a program for a permanently shutdown and defueled facility is not necessary.

The NRC staff finds that License Condition 2.H for IP3 is based on maintaining FPPs that provide reasonable assurance of the ability to achieve and maintain safe shutdown in the event of a fire in accordance with 10 CFR 50.48. Achieving and maintaining safe shutdown in the event of a fire is no longer applicable to the decommissioned FPPs at IP3 once the facility is permanently shut down and the fuel has been permanently removed from the reactor.

However, elements of the FPP (e.g., License Condition 2.AC, Mitigation Strategy License Condition) continue during decommissioning to address fire events that could result in radiological hazards. The regulation in 10 CFR 50.48(f) requires IP3 to address the potential for fires, which could result in a radiological hazard. The NRC staff concludes that the rule, which requires an FPP for licenses that have submitted the certifications under 10 CFR 50.82(a)(1), is sufficient to ensure that a program is maintained. Therefore, a license condition that also requires FPPs for the permanently shutdown and defueled unit is redundant. Based on the above, the NRC staff concludes that reliance on 10 CFR 50.48(f) is appropriate and that the licensees request to delete License Condition 2.H is acceptable.

4.3.14 License Condition 2.O Currently, License Condition 2.O reads:

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

License Condition 2.O is a reference to a historical obligation that was previously met. The licensee proposed to delete this license condition in its entirety. The proposed revision to License Condition 2.O is an administrative change and does not change any technical requirement. The NRC staff approves the proposed change to delete License Condition 2.O in its entirety.

4.3.15 License Condition 2.AA Currently, License Condition 2.AA reads:

The following conditions relate to the amendment approving the conversion to Improved Standard Technical Specifications:

1. This amendment authorizes the relocation of certain Technical Specification requirements and detailed information to licensee-controlled documentsThe relocation of requirements and detailed information shall be completed on or before the implementation of this amendment.
2. The following is a schedule for implementing surveillance requirements (SRs):

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

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

The licensee proposed to delete this license condition in its entirety. This is a historical license condition, the activities of which were completed in accordance with a license condition that is no longer applicable at IP3 as a permanently shutdown and defueled facility. Therefore, the NRC staff finds the proposed deletion of License Condition 2.AA acceptable.

4.3.16 License Condition 2.AB Currently, License Condition 2.AB reads:

The following conditions relate to the amendment approving the conversion to Improved Standard Technical Specifications:

With the reactor critical, Entergy shall maintain the reactor coolant system cold leg at a temperature (Tcold) greater than or equal to 525 °F. Entergy shall maintain a record of the cumulative time that the plant is operated with the reactor critical while Tcold is below 525 °F. Upon determination by Entergy that the cumulative time of plant operation with the reactor critical while Tcold is below 525 °F has exceeded one (1) year, Entergy must:

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

Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. This license condition, which is based on operations with the reactor critical, will no longer be applicable at IP3 as a permanently shutdown and defueled facility. Based on its review of the proposed deletion, the NRC staff concludes that continued implementation of License Condition 2.AB will no longer be applicable for IP3 because power operation will no longer be authorized once the 10 CFR 50.82(a)(1) certifications have been docketed. Therefore, the NRC staff finds the deletion of License Condition 2.AB acceptable.

4.3.17 License Condition 2.AD Currently, License Condition 2.AD reads:

Upon implementation of Amendment No. 239 adopting TSTF-448, Revision 3 (as supplemented), the determination of control room envelope (CRE) unfiltered air inleakage as required by Technical Specification (TS) Surveillance Requirement (SR) 3. 7.11.4, in accordance with TS 5.5.16.c.(i), the assessment of CRE habitability as required by TS 5.5.16.c.(ii), and the measurement of CRE pressure as required by TS 5.5.16.d, shall be considered met. Following implementation:

(a) The first performance of SR 3.7.11.4, in accordance with TS 5.5.16.c.(i),

shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measure from February 1, 2005, the date of the most recent successful tracer gas test, as stated in the June 28, 2005, letter response to Generic Letter 2003-01.

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

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

The licensee proposed to delete License Condition 2.AD. The proposed change would remove the requirements of Technical Specifications Task Force (TSTF) Traveler TSTF-448, Control Room Habitability, which involves assessing the CRE habitability at the frequencies specified in Sections C.1 and C.2 of RG 1.197, Revision 0, Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors, dated May 2003 (ADAMS Accession No. ML031490664). The licensee proposed to not retain this license condition in the proposed PDTS, as it is a historical license condition, and the associated test, assessment, and measurement of the defined TSTF-448 requirements were completed in accordance with the schedule specified in the license condition. The NRC staff finds the deletion of License Condition 2.AD acceptable.

4.3.18 License Condition 2.AF.(2).c Currently, License Conditions 2.AF.(2).c reads:

The licensee shall notify the NRC in writing within 30 days after having accomplished item (2)a above and include the status of those activities that have been or remain to be completed in item (2)b above.

The licensee proposed to delete this license condition in its entirety. This is a historical license condition regarding the licensees implementation of license renewal programs and enhancements (Item (2)a), the activities of which were completed in accordance with the schedule specified (item (2)b) in the license condition. The removal of this license condition is administrative. The NRC staff finds the proposed deletion of License Condition 2.AF.(2).c acceptable.

4.3.19 Deleted License Conditions with Historical Amendment References The licensee proposed to delete numerous historical license amendment number and date references for previously deleted license conditions in the Renewed Facility Operating License.

Specifically, the reference to previously approved license amendments listed on the right-hand margin of the pages for License Conditions 1.D, 1.E, 2.D, 2.E, 2.I through 2.N, and 2.P are proposed for deletion. This revision is administrative and does not change any technical requirement. The NRC staff approves the proposed change to delete the historical references for the aforementioned license conditions.

4.3.20 License Condition 3 Currently, License Condition 3 states:

This renewed license is effective as of the date of issuance, and shall expire at midnight April 30, 2025.

The licensee proposed License Condition 3 to read:

This renewed license is effective as of the date of issuance, and until the Commission notifies the licensee in writing that the license is terminated.

The proposed change would modify this license condition to reflect the permanently shutdown and defueled condition of the facility. Consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel. The proposed change would revise License Condition 3 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 3. The current License Condition 3, which documents the date of the expiration of the RFOL, is no longer necessary for the permanently shutdown and defueled facility in the process of decommissioning. The revised License Condition 3 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 3 acceptable.

4.4 Changes to Appendix A, Technical Specification 4.4.1 Title Page The current title page states, in part:

FACILITY OPERATING LICENSE DPR-64 TECHNICAL SPECIFICATIONS AND BASES

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

FACILITY LICENSE DPR-64 PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS AND BASES Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed, pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable to the NRC staff.

4.4.2 Table of Contents The licensee proposed to revise the Table of Contents to reflect proposed additions, deletions, and changes to the TSs, as described in Sections 4.4.3 through 4.4.7 of this SE, and as detailed in Attachment 1 to the LAR dated April 28, 2020. The changes to the Table of Contents are editorial and do not change any technical requirement. The NRC staff finds the changes to the Table of Contents acceptable.

4.4.3 TS Section 1.0, Definitions The licensee proposed maintaining the definitions for Certified Fuel Handler and Non-certified Operator, proposed for addition in its LAR dated April 15, 2019 (ADAMS Accession No. ML19105B236), which proposed changes to the Administrative Controls section of the TSs.

Additionally, the term and definition for Actions would be retained.

The licensee proposed deletion of the following definitions since the terms are not used in any PDTS LCO and do not apply to a facility in the permanently defueled condition. In addition to the definitions listed below, the licensee proposed deletion of Table 1.1-1, MODES. TS Table 1 1-1 provides as follows:

Table 1.1-1 MODES 4.4.3.1 Definitions Proposed for Deletion The licensee proposed to delete the following definitions in TS Section 1.0, Use and Application, which currently state:

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

AXIAL FLUX DIFFERENCE (AFD)

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

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the required range and accuracy to known input. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, interlock, display, and trip functions. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the

next required CHANNEL CALIBRATION shall include an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping calibrations or total channel steps so that the entire channel is calibrated.

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment, by observation, of channel behavior during operation. This determination shall include, where possible, comparison of the channel indication and status to other indications or status derived from independent instrument channels measuring the same parameter.

CHANNEL OPERATIONAL TEST (COT)

A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

CORE ALTERATION CORE ALTERATION shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.

CORE OPERATING LIMITS REPORT (COLR)

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

DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcuries per gram) that alone would produce the same dose when inhaled as the combined activities of iodine isotopes I-131, I-132, I-133, I-134, and I-135 actually present.

If a specific isotope is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE EQUIVALENT I-131 shall be performed using Committed Effective Dose Equivalent (CEDE) dose conversion factors from Table 2.1 of EPA Federal Guidance Report No. 11, 1988.

DOSE EQUIVALENT XE-133 DOSE EQUIVALENT XE-133 shall be that concentration of Xe-133 (microcuries per gram) that alone would produce the same acute dose to the whole body as

the combined activities of noble gas nuclides Kr-85m, Kr-85, Kr-87, Kr-88, Xe-131m, Xe-133m, Xe-133, Xe-135m, Xe-135, and Xe-138 actually present. If a specific noble gas nuclide is not detected, it should be assumed to be present at the minimum detectable activity. The determination of DOSE Equivalent XE-133 shall be performed using effective dose conversion factors for air submersion listed in Table III.1 of EPA Federal Guidance Report No. 12, 1993, External Exposure to Radionuclides in Air, Water, and Soil.

La [MAXIMUM ALLOWABLE PRIMARY CONTAINMENT LEAKAGE RATE]

The maximum allowable primary containment leakage rate, La, shall be 0.1% of primary containment air weight per day at the calculated peak containment pressure (Pa).

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE
1. LEAKAGE, such as that from pump seals or valve packing (except for leakage into closed systems and reactor coolant pump (RCP) seal water injection or leakoff), that is captured and conducted to collection systems or a sump or collecting tank; (Leakage into closed systems is leakage that can be accounted for and contained by a system not directly connected to the atmosphere.

Leakage past the pressurizer safety valve seats and leakage past the safety injection pressure isolation valves are examples of reactor coolant system leakage into closed systems.)

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE);
b. Unidentified LEAKAGE All LEAKAGE (except for leakage into closed systems and RCP seal water injection or leakoff) that is not identified LEAKAGE;
c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay. The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

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

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

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

a. Described in FSAR Chapter 13, Tests and Operations;
b. Authorized under the provisions of 10 CFR 50.59; or
c. Otherwise approved by the Nuclear Regulatory Commission.

QUADRANT POWER TILT RATIO (QPTR)

QPTR shall be the ratio of the maximum upper excore detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER (RTP)

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 3216 MWt.

SHUTDOWN MARGIN (SDM)

SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fully withdrawn. However, with all RCCAs verified fully inserted by two independent means, it is not necessary to account for a stuck RCCA in the SDM calculation. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and
b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each required slave relays in the channel required for channel OPERABILITY and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated required testable actuation devices.

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

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

TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)

A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of all devices in the channel required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy.

The NRC staff examined the licensees proposal to maintain the definitions for Actions, Certified Fuel Handler and Non-certified Operator and reformat them on one page as the remaining TS definitions. The NRC staff finds that these terms appropriately apply to the safe storage and handling of spent fuel in the SFP and are to be retained in the PDTS.

The NRC staff reviewed the TS definitions proposed for deletion and concludes that all of the terms listed above are only meaningful to a reactor authorized to operate. Once IP3 submits the docketed certifications required by 10 CFR 50.82(a)(1), the 10 CFR Part 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds the deletion of these definitions and Table 1.1-1 associated with the MODES definition from the TSs acceptable.

4.4.3.2 TS 1.2, Logical Connectors TS Section 1.2, Logical Connectors, in the IP3 TSs explains the meaning of logical connectors. The current description for TS Section 1.2 states, in part:

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

EXAMPLES The following examples illustrate the use of logical connectors.

Example 1.2-2 The licensee proposed to modify the TSs to reflect the logical connectors that continue to exist in the TSs. Specifically, the purpose and background sections are revised to remove the terms Conditions, Completion Times, and Frequencies. Also, the licensee proposed to revise the current examples by making them a singular example and deleting Example 1.2-2. The revised TS Section 1.2, Logical Connectors, will state the following:

PURPOSE Logical connectors are used in Technical Specifications (TS) to discriminate between, and yet connect, discrete Required Actions and Surveillances

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

EXAMPLE The following example illustrates the use of logical connectors

[Example 1.2-2 is proposed for deletion]

The NRC staff examined the licensees proposed revisions to TS Section 1.2 and concludes that the revisions are administrative in nature and reflect the logical connectors used in TS 3.7.15, which is the only LCO that contain logical connectors in the PDTS. Therefore, the staff finds these revisions acceptable.

4.4.3.3 TS 1.3, Completion Times TS Section 1.3, Completion Times, in the IP3 TSs establishes the completion time convention and provides guidance for its use. The current background for TS Section 1.3 states the following:

Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit The licensee proposed to revise the background to replace the phrase operation of the unit with handling and storage of spent nuclear fuel. The revised background will state the following:

Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe handling and storage of spent nuclear fuel The current description for TS Section 1.3 states the following:

The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the unit is in a MODE or specified condition stated in the Applicability of the LCO.

Unless otherwise specified, ...

Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the unit is not within the LCO Applicability.

If situations are discovered

The licensee proposed to revise the description to remove the terms inoperable equipment or and MODE, replace the term unit with facility, remove the entire discussion of this section starting from the phrase Unless otherwise specified and remove the entire subsequent discussion, starting from the phrase If situations are discovered. The revised description will state the following:

The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the facility is in a specified condition stated in the Applicability of the LCO.

Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the facility is not within the LCO Applicability.

The current examples for TS Section 1.3 state the following:

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

Example 1.3-1 Example 1.3-2 Example 1.3-3 Example 1.3-4 Example 1.3-5 Example 1.3-6 Example 1.3-7 The licensee proposed to revise the examples by making them a singular example, replacing the phrase types of Conditions and changing Conditions with Required Actions, revising Example 1.3-1, and deleting Examples 1.3-2 through 1.3-7. The revised example will state the following:

The following example illustrates the use of Completion Times with different Required Actions.

Example 1.3-1 is modified to address completion times as utilized by TS 3.7.15.

Accordingly, the description of Immediate Completion Time, which previously followed the Example 1.3-7, is retained.

The NRC staff examined the licensees proposal to modify the background, description, and examples of TS Section 1.3. Once IP3 permanently shuts down and defuels, (1) the primary mission will change from safe operation of the unit to the safe handling and storage of spent nuclear fuel, which is reflected in the revised background; (2) the license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, and therefore, will no longer have operability requirements for any equipment, which is reflected in the revised description; and (3) the only TSs with completion times in the PDTS will be TSs 3.7.14, 3.7.15, and 3.7.16, which are reflected in the revised example. The term Inoperable equipment refers

to equipment relied upon in the TS for an operating reactor but is not relied upon in the PDTS.

There are no systems or components required to be operable in the PDTS as they are not required to function to mitigate any of the remaining DBAs. The NRC staff reviewed the licensees basis for the revisions to TS Section 1.3 and concludes that the proposed changes are acceptable, as they reflect IP3s permanently shutdown and defueled condition, as well as the completion times that will continue to exist in the PDTS.

4.4.3.4 TS 1.4, Frequency TS Section 1.4, Frequency, in the IP3 TSs defines the proper use and application of frequency requirements. The current description for TS Section 1.4 states the following:

The specified Frequency is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The specified Frequency consists of the requirements of the Frequency column of each SR as well as certain Notes in the surveillance column that modify performance requirements.

Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be performed until sometime after the associated LCO is within its Applicability, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only required when it can be and should be performed. With an SR satisfied, SR 3.0.4 imposes no restriction.

The licensee proposed to revise the description to remove the entire subsequent discussion, starting from the phrase as well as certain Notes in the Surveillance column. The revised description will state the following:

The specified Frequency is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The specified Frequency consists of the requirements of the Frequency column of each SR.

The current examples for TS Section 1.4 state the following:

The following examples illustrate the various ways that Frequencies are specified. In these examples, the Applicability of the LCO (LCO not shown) is MODES 1, 2, and 3.

Example 1.4-1 Example 1.4-2 Example 1.4-3 The licensee proposed to revise Example 1.4-1 and the descriptive paragraph below it and proposed to delete Examples 1.4-2 and 1.4-3.

The revised example will state the following:

The following example illustrates the type of Frequency statement that appears in the Technical Specifications (TS).

The NRC staff examined the licensees proposal to modify the description and examples of TS Section 1.4. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). The only SRs that will remain in the TSs are limited to those in TSs 3.7.14, 3.7.15, and 3.7.16. The revisions to TS Section 1.4 provide the rules of usage and an example that continue to be applicable to those TSs. Example 1.4-1 is modified to address an example of a Frequency that is utilized by TS 3.7.14. The NRC staff reviewed the licensees provided basis for these revisions and concludes that the revisions are acceptable, as they reflect the condition of the facility after shutdown and defueling occur.

4.4.4 TS Section 2.0, Safety Limits (SLs)

TS Section 2.0, Safety Limits (SLs), in the IP3 TSs contains SLs that were necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity from the reactor core and the RCS pursuant to 10 CFR 50.36(c)(1).

The licensee proposed to delete TS 2.0, Safety Limits (SLs); TS 2.1, Safety Limits; and TS 2.2, Safety Limit Violations, in their entirety, since the SLs do not apply to a reactor that is in a permanently defueled condition. The SLs established in TS Section 2.1 prevent overheating of the fuel and possible cladding perforation, which would result in the release of fission products to the reactor coolant, and protect the integrity of the RCS from over-pressurization, thereby preventing the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere. Safety limit violations in TS Section 2.2 are values of various parameters for which automatic protective action is needed during normal operations or anticipated transients to prevent exceeding an SL.

Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). The specifications in TS Section 2.0 do not apply to the safe storage and handling of spent fuel in the SFP. The NRC staff reviewed the proposed deletion and finds that the SLs are not applicable to a reactor in a permanently defueled condition. Therefore, the NRC staff finds that the proposed deletions of TS Sections 2.0, 2.1, and 2.2, in their entirety, are acceptable.

4.4.5 TS Section 3.0, Limiting Condition for Operation (LCO) Applicability TS Section 3.0, Limiting Condition for Operation (LCO) Applicability, in the IP3 TSs contains LCOs, which specify the lowest functional capability or performance levels of equipment required for safe operation of the facility, and contain the general requirements applicable to all LCOs and apply at all times unless otherwise stated in the TSs.

The licensee proposed revisions to LCOs 3.0.1 and 3.0.2 and deletion of LCOs 3.0.3 through 3.0.8.

The current LCO 3.0.1 states the following:

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

The licensee proposed to revise LCO 3.0.1 to state the following:

LCOs shall be met during the specified conditions in the Applicability, except as provided in LCO 3.0.2.

The current LCO 3.0.2 states, in part, the following:

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

The licensee proposed to revise LCO 3.0.2 to state, in part, the following:

Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met.

The NRC staff reviewed the proposed changes to LCO 3.0.1 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the plant or placing fuel in the reactor vessel, the references to modes are no longer applicable. In addition, the references to LCOs 3.0.7 and 3.0.8 pertain to special tests and operations required for an operating reactor and actions required for equipment, respectively, that will no longer be operating once IP3 permanently shuts down and defuels. Therefore, the NRC staff finds the changes acceptable.

The NRC staff reviewed the proposed changes to LCO 3.0.2 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. The references to LCOs 3.0.5 and 3.0.6 pertain to restoring equipment to service under administrative controls and allowing performance of SRs on equipment declared inoperable, respectively, on equipment that will no longer be operating once IP3 permanently shuts down and defuels. Therefore, the NRC staff finds the changes acceptable.

The NRC staff reviewed the proposed deletions of LCOs 3.0.3, 3.0.4, 3.0.5, 3.0.6, 3.0.7, and 3.0.8. These LCOs pertain to an operating reactor. Once IP3 permanently shuts down and defuels, these LCOs will no longer be applicable. Therefore, the NRC staff finds the deletions acceptable.

4.4.5.1 TS Section 3.0, Surveillance Requirement (SR) Applicability TS Section 3.0, Surveillance Requirement (SR) Applicability, in the IP3 TSs contains the general requirements applicable to all SRs and applies at all times unless otherwise stated in a TS.

The licensee proposed revisions to SRs 3.0.1, 3.0.2, 3.0.3, and 3.0.4. The current SR 3.0.1 states the following:

SRs shall be met during the MODES or other specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SRSurveillances do not have to be performed on inoperable equipment or variables outside specified limits.

The licensee proposed to revise SR 3.0.1 to state the following:

SRs shall be met during the specified conditions in the Applicability for individual LCOs, unless otherwise stated in the SRSurveillances do not have to be performed on variables outside specified limits.

The current SR 3.0.2 states the following:

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

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

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

Exceptions to this Specification are stated in the individual Specifications.

The licensee proposed to revise SR 3.0.2 to state the following:

The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance.

The current SR 3.0.3 states, in part, the following:

If it is discovered that a Surveillance was not performed within its specified performed within its specified Frequency, then compliance with the requirement to declare the requirement to declare the LCO not met may be delayed, from the time of discovery, discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. The delay period is only applicable when there is a reasonable expectation the surveillance will be met when performed. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

The licensee proposed to revise SR 3.0.3 to state, in part, the following:

If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the risk impact shall be managed.

The current SR 3.0.4 states the following:

Entry into a MODE or other specified condition in the Applicability of an LCO shall only be made when the LCOs Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. When an LCO is not met due to Surveillances not having been met, entry into a MODE or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.4.

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

The licensee proposed to revise SR 3.0.4 to state the following:

Entry into a specified condition in the Applicability of an LCO shall only be made when the LCOs Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3.

The NRC staff reviewed the proposed changes to SRs 3.0.1, 3.0.2, 3.0.3, and 3.0.4 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the plant or placing fuel in the reactor vessel, the references to modes and the discussions about shutting down the unit are no longer applicable. Further, the NRC staff agrees that the statements to be deleted are no longer necessary because the defueled TSs do not contain frequencies of the type described in the statements being deleted. Therefore, the staff finds that the proposed changes to delete these references reflect the plant status and are appropriate and acceptable.

4.4.5.2 TS Section 3.1, Reactivity Control Systems TS Section 3.1, Reactivity Control Systems, in the IP3 TSs contains requirements to assure and verify operability of reactivity control systems. A description of each TS and the NRC staffs evaluation of each of the specifications proposed for deletion are as follows:

TS 3.1.1, Shutdown Margin (SDM), establishes shutdown margin as the minimum shutdown margin in the reactor core. The shutdown margin limits are specified in the Core Operating Limits Report (COLR). TS 3.1.1 is proposed for deletion and is applicable in MODE 2 with keff 1.0 and MODES 3 through 5. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, the TS will not be required because the operation in the

applicable modes and specified conditions will no longer occur. TS 3.1.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.1 acceptable.

TS 3.1.2, Core Reactivity, establishes that core reactivity be within +/- 1 percent k/k of predicted values. TS 3.1.2 is proposed for deletion and is applicable in MODES 1 and 2.

The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in MODES 1 and 2 will no longer occur and TS 3.1.2 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.2 acceptable.

TS 3.1.3, Moderator Temperature Coefficient (MTC), establishes that MTC be maintained within the limits specified in the COLR. The maximum upper limit shall be 0.0 k/k degrees Fahrenheit (°F) at hot zero power. TS 3.1.3 is proposed for deletion and is applicable in MODES 1 and 2 with keff 1.0 for the upper MTC limit and MODES 1, 2, and 3 for the lower MTC limit. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in the applicable modes and specified conditions will no longer occur. TS 3.1.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.3 acceptable.

TS 3.1.4, Rod Group Alignment Limits, establishes that all shutdown and control rods shall be operable, and the difference between each individual indicated rod position and its group step counter demand position shall be within the limits. TS 3.1.4, including Table 3.1.4-1, is proposed for deletion and is applicable in MODES 1 and 2. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in MODES 1 and 2 will no longer occur, and TS 3.1.4 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.4 acceptable.

TS 3.1.5, Shutdown Bank Insertion Limits, establishes that each shutdown bank shall be within insertion limits specified in the COLR. TS 3.1.5 is proposed for deletion and is applicable in MODES 1 and 2. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in MODES 1 and 2 will no longer occur, and TS 3.1.5 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.5 acceptable.

TS 3.1.6, Control Bank Insertion Limits, establishes that control banks shall be within the insertion, sequence, and overlap limits specified in the COLR. TS 3.1.6 is proposed for deletion and is applicable in MODES 1 and 2 with keff 1.0. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with

10 CFR 50.82(a)(2). Therefore, operation in the applicable modes and specified conditions will no longer occur. TS 3.1.6 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.6 acceptable.

TS 3.1.7, Rod Position Indication, establishes that the individual rod position indication system and the demand position indication shall be operable. TS 3.1.7 is proposed for deletion and is applicable in MODES 1 and 2. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in MODES 1 and 2 will no longer occur, and TS 3.1.7 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.7 acceptable.

TS 3.1.8, Physics Tests Exceptions - Mode 2, establishes that during the performance of physics tests, the requirements of Moderator Temperature Coefficient; Rod Group Alignment Limits; Shutdown Bank Insertion Limits; Control Bank Insertion Limits; and RCS Minimum Temperature for Criticality, may be suspended, provided that certain conditions as specified in LCO 3.1.8 are met. TS 3.1.8 is proposed for deletion and is applicable during physics tests initiated in MODE 2. The NRC staff determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, operation in the applicable mode and specified condition will no longer occur. TS 3.1.8 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.1.6 acceptable.

4.4.5.3 TS Section 3.2, Power Distribution Limits TS Section 3.2, Power Distribution Limits, in the IP3 TSs contains power distribution limits that provide assurance that fuel design criteria are not exceeded, and the accident analysis assumptions remain valid. A description of each of the specifications proposed for deletion is provided as follows:

TS 3.2.1, Heat Flux Hot Channel Factor (FQ(Z)), establishes that FQ(Z) shall be within the limits specified in the COLR. As a result, this TS will not apply in the permanently defueled condition. TS 3.2.1 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.2.2, Nuclear Enthalpy Rise Hot Channel Factor (FNH), establishes that FNH shall be within the limits specified in the COLR. TS 3.2.2 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.2.3, Axial Flux Difference (AFD) (Constant Axial Offset Control (CAOC)

Methodology, establishes that the AFD shall be maintained within the target band about the target flux difference. The target band is specified in the COLR. AFD may deviate outside the target band with thermal power less than (<) 90 percent RTP but greater than or equal () 50 percent RTP, provided AFD is within the acceptable operation limits, and cumulative penalty deviation time is less than or equal to () 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> during the previous 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The acceptable operation limits are specified in the COLR. TS 3.2.3

does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.2.4, Quadrant Power Tilt Ratio (QPTR), establishes that the QPTR shall be 1.02. TS 3.2.4 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

The licensee proposed to delete TS Section 3.2 in its entirety. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2).

Therefore, the specifications addressed in TS Section 3.2 will not be required, and these requirements will not apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS Section 3.2 acceptable.

4.4.5.4 TS Section 3.3, Instrumentation TS Section 3.3, Instrumentation, in the IP3 TSs contains operability requirements for sensing and control instrumentation required for safe operation of the facility. Below are the specifications in TS Section 3.3.

TS 3.3.1, Reactor Protection System (RPS) Instrumentation, establishes that the RPS instrumentation for each function in Table 3.3.1-1 shall be operable.

TS 3.3.2, Engineered Safety Feature Actuation System (ESFAS) Instrumentation, establishes that the ESFAS instrumentation for each function in Table 3.3.2-1 shall be operable.

TS 3.3.3, Post Accident Monitoring (PAM) Instrumentation, establishes that the PAM instrumentation for each function in Table 3.3.3-1 shall be operable.

TS 3.3.4, Remote Shutdown, establishes that the remote shutdown functions shall be operable.

TS 3.3.5, Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation, establishes that certain LOP DG start instrumentation conditions as specified in LCO 3.3.5 shall be operable.

TS 3.3.6, Containment Purge System and Pressure Relief Line Isolation Instrumentation, establishes that the containment purge system and pressure relief line isolation instrumentation for each function in Table 3.3.6-1 shall be operable.

TS 3.3.7, Control Room Ventilation System (CRVS) Actuation Instrumentation, establishes that the CRVS actuation instrumentation for each function in Table 3.3.7-1 shall be operable.

TS 3.3.8, Fuel Storage Building Emergency Ventilation System (FSBEVS) Actuation Instrumentation, establishes that the FSBEVS manual and automatic instrumentation shall be operable.

The licensee proposed to delete TS Section 3.3 in its entirety. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. TS 3.3.1 through 3.3.7 apply during operation of the reactor; TS 3.3.8 concerns the FSBEVS, and applies during movement of recently irradiated fuel in the Fuel Storage Building. As discussed below with respect to TS 3.3.17, the licensee stated that the Bases for TS 3.7.13, define recently irradiated fuel as fuel that has occupied part of a critical reactor within the previous 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The IP3 PDTS will not be implemented until after 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> following permanent shutdown, so TS 3.3.8 will no longer be needed once the PDTS have been implemented. Further, once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS Section 3.3 will not be required, and these requirements will not apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS Section 3.3 acceptable.

4.4.5.5 TS Section 3.4, Reactor Coolant System (RCS)

TS Section 3.4, Reactor Coolant System (RCS), in the IP3 TSs contains requirements that provide for appropriate control of process variables, design features, or operating restrictions needed for appropriate functional capability of RCS equipment required for safe operation of the facility. The licensee proposed to delete TS Section 3.4 in its entirety. A description of each of the specifications proposed for deletion is provided as follows:

TS 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits, establishes RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure is greater than or equal to the limit specified in the COLR;
b. RCS average temperature is less than or equal to the limit specified in the COLR; and
c. RCS total flow rate 354,400 gpm and greater than or equal to the limit specified in the COLR.

TS 3.4.1 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.2, RCS Minimum Temperature for Criticality, establishes that each RCS loop average temperature shall be 540 °F. TS 3.4.2 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.3, RCS Pressure and Temperature (P/T) Limits, establishes that RCS pressure, temperature, heatup and cooldown rates shall be maintained within the limits specified.

TS 3.4.3 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.4, RCS Loops - Modes 1 and 2, establishes that four RCS loops shall be operable and in operation. TS 3.4.4 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.5, RCS Loops - Mode 3, establishes that two RCS loops shall be operable and either two RCS loops shall be in operation when the rod control system is capable of rod withdrawal or one RCS loop shall be in operation when the rod control system is not capable of rod withdrawal. TS 3.4.5 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.6, RCS Loops - Mode 4, establishes that two loops consisting of any combination of RCS loops and residual heat removal (RHR) loops shall be operable, and one loop shall be in operation. TS 3.4.6 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.7, RCS Loops - Mode 5, Loops Filled, establishes that one RHR loop shall be operable and in operation, and either one additional RHR loop shall be operable or the secondary side water level of at least two steam generators (SGs) shall be 71 percent wide range. TS 3.4.7 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.8, RCS Loops - Mode 5, Loops Not Filled, establishes that two RHR loops shall be operable, and one RHR loop shall be in operation. TS 3.4.8 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.9, Pressurizer, establishes that the pressurizer shall be operable with pressurizer water level 54.3 percent in MODES 1 and 2, or 90 percent in MODE 3, and two groups of pressurizer heaters operable with the capacity of each group 150 kilowatts (kW) and capable of being powered from an emergency power supply.

TS 3.4.9 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.10, Pressurizer Safety Valves, establishes that three pressurizer safety valves shall be operable with lift settings set 2,460 pounds per square inch gauge (psig) and 2,510 psig. TS 3.4.10 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.11, Pressurizer Power Operated Relief Valves (PORVs), establishes that each PORV and associated block valve shall be operable. TS 3.4.11 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.12, Low Temperature Overpressure Protection (LTOP), establishes that LTOP shall be operable with no high head safety injection pumps capable of injecting into the RCS and the accumulator discharge isolation valves closed and de-energized, and either of the specified options of LCOs 3.4.12(a) or 3.4.12(b) are met. TS 3.4.12 does not apply once the reactor is permanently defueled. The NRC staff also notes that these sections are specifically related to the requirements for the reactor coolant pressure boundary set forth in Appendices G and H to 10 CFR Part 50. Section 50.60(a) of 10 CFR stipulates that reactor facilities that have submitted the certifications required under 10 CFR 50.82(a)(1) no longer need to meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in Appendices G and H to 10 CFR Part 50. TS 3.4.12 will no longer be necessary for IP3 in accordance with 10 CFR 50.60(a) because these LCOs will not apply in a permanently defueled condition. Therefore, the TS is proposed to be deleted.

TS 3.4.13, RCS Operational Leakage, establishes that RCS operational leakage shall be limited to:

a. No pressure boundary leakage;
b. 1 gpm unidentified leakage;
c. 10 gpm identified leakage; and
d. 150 gallons per day primary to secondary leakage through any one steam generator (SG).

TS 3.4.13 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.14, RCS Pressure Isolation Valve (PIV) Leakage, establishes that leakage from each RCS PIV shall be within limit. TS 3.4.14 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.15, RCS Leakage Detection Instrumentation, establishes that the following RCS leakage detection instrumentation shall be operable:

a. One containment sump discharge flow monitor;
b. One containment atmosphere radioactivity monitor (gaseous or particulate); and
c. One containment fan cooler unit condensate measuring system.

TS 3.4.15 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.16, RCS Specific Activity, establishes that RCS dose equivalent I-131 and dose equivalent Xe-133 specific activity shall be within limits. TS 3.4.16 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.4.17, Steam Generator (SG) Tube Integrity, establishes that SG tube integrity shall be maintained, and all SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the SG program. TS 3.4.17 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

As noted above, the licensee proposed to delete TS Section 3.4 in its entirety. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS Section 3.4 will not be required, and these requirements will not apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS Section 3.4 acceptable.

4.4.5.6 TS Section 3.5, Emergency Core Cooling Systems (ECCS)

TS Section 3.5, Emergency Core Cooling Systems (ECCS), in the IP3 TSs contains requirements that provide for appropriate functional capability of ECCS equipment required for mitigation of DBAs or transients to protect the integrity of a fission product barrier. The licensee proposed to delete TS Section 3.5 in its entirety. A description of each of the specifications proposed for deletion is provided as follows:

TS 3.5.1, Accumulators, establishes that four ECCS accumulators shall be operable.

TS 3.5.1 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.5.2, ECCS - Operating, establishes that three ECCS trains shall be operable in MODES 1 through 3. TS 3.5.2 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.5.3, ECCS - Shutdown, establishes that two ECCS high head safety injection subsystems and one ECCS RHR subsystem shall be operable. TS 3.5.3 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

TS 3.5.4, Refueling Water Storage Tank (RWST), establishes that the RWST and two channels of RWST low level alarm shall be operable. TS 3.5.4 does not apply once the reactor is permanently defueled; therefore, the TS is proposed to be deleted.

As noted above, the licensee proposed to delete TS Section 3.5 in its entirety. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS Section 3.5 will not be required and these requirements will not apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS Section 3.5 acceptable.

4.4.5.7 TS Section 3.6, Containment Systems TS Section 3.6, Containment Systems, in the IP3 TSs contain requirements that assure the integrity of the containment, depressurization and cooling systems, and containment isolation valves. The licensee proposed to delete TS Section 3.6 in its entirety. Below are the specifications proposed for deletion.

TS 3.6.1, Containment, establishes that containment shall be operable in MODES 1 through 4.

TS 3.6.2, Containment Air Locks, establishes that two containment air locks shall be operable in MODES 1 through 4.

TS 3.6.3, Containment Isolation Valves, establishes that each containment isolation valve shall be operable in MODES 1 through 4.

TS 3.6.4, Containment Pressure, establishes that containment pressure shall be maintained within the specified limits in MODES 1 through 4.

TS 3.6.5, Containment Air Temperature, establishes that containment average air temperature shall be > 50 °F and 130 °F in MODES 1 through 4.

TS 3.6.6, Containment Spray System and Containment Fan Cooler System, establishes that two trains of containment spray and three trains of containment fan cooler shall be operable in MODES 1 through 4.

TS 3.6.7, Recirculation pH Control System, establishes that the recirculation pH control system shall be operable in MODES 1 through 4.

TS 3.6.8, Not Used.

TS 3.6.9, Isolation Valve Seal Water (IVSW) System, establishes that the IVSW system shall be operable in MODES 1 through 4.

TS 3.6.10, Weld Channel and Penetration Pressurization System (WC&PPS),

establishes that the WC&PPS shall be operable in MODES 1 through 4.

As noted above, the licensee proposed to delete TS Section 3.6 in its entirety. These specifications do not apply in a defueled condition or for SSCs that are not needed for accident mitigation in the defueled condition. Once the certifications required by 10 CFR 50.82(a)(1) are docketed, the IP3 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS Section 3.6 will not be required because the operation in the applicable MODES and specified conditions will no longer occur. The NRC staff therefore finds the deletion of TS Section 3.6 acceptable.

4.4.5.8 TS Section 3.7, Plant Systems TS Section 3.7, Plant Systems, in the IP3 TSs provides requirements for the appropriate functional capability of plant equipment required for safe operation of the facility, including the plant being in a defueled condition. A description of each TS and the NRC staffs evaluation of each of the specifications proposed for deletion or modification is provided as follows:

TS 3.7, Plant Systems, is proposed to be renamed to Spent Fuel Pit Requirements.

The licensee states that this change is to reflect the remaining TSs in the section that deal with spent fuel pit requirements in a permanently shutdown and defueled facility.

The NRC staff has determined that the title change aligns with the specific remaining TS requirements for this section; therefore, the NRC staff finds this administrative change acceptable.

TS 3.7.1, Main Steam Safety Valves (MSSVs), including Tables 3.7.1-1 and 3.7.1-2, is proposed for deletion; this TS is applicable in MODES 1 through 3. The NRC staff has determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 3 will no longer occur, TS 3.7.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.1 acceptable.

TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Main Steam Check Valves (MSCVs), is proposed for deletion; this TS is applicable in MODE 1 and MODES 2 and 3, except when all MSIVs are closed. The NRC staff has determined that after the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 3 will no longer occur, TS 3.7.2 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.2 acceptable.

TS 3.7.3, Main Boiler Feedpump Discharge Valves (MBFPDVs), Main Feedwater Regulation Valves (MFRVs), Main Feedwater Inlet Isolation Valves (MFIIVs) and Main Feedwater (MF) Low Flow Bypass Valves, is proposed for deletion; this TS is applicable in MODES 1 through 3, except when each main feedwater and bypass line is isolated by a closed and deactivated motor/air operated valve or isolated by a closed manual valve.

After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 3 will no longer occur, TS 3.7.3 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.3 acceptable.

TS 3.7.4, Atmospheric Dump Valves (ADVs); TS 3.7.5; Auxiliary Feedwater (AFW)

System; TS 3.7.6, Condensate Storage Tank (CST); and TS 3.7.7, City Water (CW),

are applicable in MODES 1 through 3 and MODE 4 when the SG is relied upon for heat removal. All four specifications are proposed for deletion. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, the three specifications will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TSs 3.7.4, 3.7.5, 3.7.6, and 3.7.7 acceptable.

TS 3.7.8, Component Cooling Water (CCW) System; TS 3.7.9, Service Water System (SWS); and TS 3.7.10, Ultimate Heat Sink (UHS), are applicable in MODES 1 through 4 and are proposed for deletion. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, the three specifications will not be applicable in a permanently defueled condition.

The NRC staff therefore finds the deletion of TSs 3.7.8, 3.7.9, and 3.7.10 acceptable.

TS 3.7.11, Control Room Ventilation System (CRVS), is proposed for deletion; this TS is applicable in MODES 1 through 4 and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).

Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.7.11 will not be applicable in a permanently defueled condition. In addition, the licensee states that the Bases for TS 3.7.11, define recently irradiated fuel as fuel that has occupied part of a critical reactor within the previous 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The IP3 PDTS will not be implemented

until after 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> following permanent shutdown, so the specific condition of applicability will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition. Therefore, the NRC staff finds the deletion of TS 3.7.11 acceptable.

TS 3.7.12, Control Room Air Conditioning System (CRACS), is applicable in MODES 1 through 4 and is proposed for deletion. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.7.12 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.12 acceptable.

TS 3.7.13, Fuel Storage Building Emergency Ventilation System (FSBEVS), is proposed for deletion; this TS is applicable during movement of recently irradiated fuel assemblies in the fuel storage building. The licensee states that the Bases for TS 3.7.13, define recently irradiated fuel as fuel that has occupied part of a critical reactor within the previous 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The IP3 PDTS will not be implemented until after 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> following permanent shutdown, so the specific condition of applicability will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition. Therefore, the NRC staff finds the deletion of TS 3.7.13 acceptable.

TS 3.7.14, Spent Fuel Pit Water Level; TS 3.7.15, Spent Fuel Pit Boron Concentrations; and TS 3.7.16, Spent Fuel Assembly Storage, including Figure 3.7.16-1, are being retained in the proposed PDTS. The licensee proposed an administrative change replacing the TS Section 3.7 title Plant Systems with Spent Fuel Pit Requirements in the TS heading. Additionally, for each section, the note contained within Required Action A.1, which states LCO 3.0.3 is not applicable, is proposed for deletion to align with the proposed deletion of TS LCO 3.0.3. Therefore, the NRC staff finds the modifications of TS 3.7.14, TS 3.7.15, and 3.7.16 acceptable.

TS 3.7.17, Secondary Specific Activity, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.7.17 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.7.17 acceptable.

4.4.5.9 TS Section 3.8, Electrical Power Systems TS Section 3.8, Electrical Power Systems, in the IP3 TSs contains operability requirements that provide for appropriate functional capability of plant electrical equipment required for safe operation of the facility. The licensee proposes to delete TS Section 3.8 in its entirety, as these TSs will not apply to IP3 in the permanently defueled condition. A description of each TS and the NRC staffs evaluation of each of the specifications proposed for deletion is provided as follows:

TS 3.8.1, AC [Alternating Current] Sources - Operating, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer

authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.8.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.1 acceptable.

TS 3.8.2, AC Sources - Shutdown, is proposed for deletion; this TS is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the movement of recently irradiated fuel assemblies will not occur, and the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 5 and 6 and the movement of recently irradiated fuel assemblies will no longer occur, TS 3.8.2 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.2 acceptable.

TS 3.8.3, Diesel Fuel Oil and Starting Air, is proposed for deletion; this TS is applicable when the associated diesel generator is required to be operable. The operability requirements of the diesel generators are contained in TS 3.8.1 and TS 3.8.2. Since these two TSs are being proposed for deletion, TS 3.8.3 is not included in the proposed PDTS because the two TSs that it supports will no longer be required after IP3 is in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3.8.3 acceptable.

TS 3.8.4, DC Sources - Operating, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.8.4 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.4 acceptable.

TS 3.8.5, DC [Direct Current] Sources - Shutdown, is proposed for deletion; this TS is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).

Therefore, since operation in MODES 5 and 6 and the movement of recently irradiated fuel assemblies will no longer occur, TS 3.8.5 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.5 acceptable.

TS 3.8.6, Battery Cell Parameters, including Table 3.8.6-1, is proposed for deletion; this TS is applicable when the associated direct current (DC) electrical power subsystems are required to be operable. The operability requirements regarding the DC sources are contained in TS 3.8.4 and TS 3.8.5. Since these two TSs are being proposed for deletion, TS 3.8.6 is not included in the proposed PDTS because the two TSs that it supports will no longer be required after IP3 is in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3.8.6 acceptable.

TS 3.8.7, Inverters - Operating, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are

docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.8.7 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.7 acceptable.

TS 3.8.8, Inverters - Shutdown, is proposed for deletion; this TS is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 5 and 6 and movement of recently irradiated fuel assemblies will no longer occur, TS 3.8.8 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.8 acceptable.

TS 3.8.9, Distribution Systems - Operating, is proposed for deletion; this TS is applicable in MODES 1 through 4. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODES 1 through 4 will no longer occur, TS 3.8.7 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.9 acceptable.

TS 3.8.10, Distribution Systems - Shutdown, is proposed for deletion; this TS is applicable in MODES 5 and 6, and during movement of recently irradiated fuel assemblies. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).

Therefore, since operation in MODES 5 and 6 and movement of recently irradiated fuel assemblies will no longer occur, TS 3.8.10 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.8.10 acceptable.

4.4.5.10 TS Section 3.9, Refueling Operations TS Section 3.9, Refueling Operations, in the IP3 TSs contains requirements that provide for appropriate functional capability of parameters and equipment that are required for mitigation of DBAs during refueling operations (moving irradiated fuel to or from the reactor core). A description of each TS and the NRC staffs evaluation of each of the specifications proposed for deletion is provided as follows:

TS 3.9.1, Boron Concentration, establishes that boron concentrations of the RCS, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the COLR. TS 3.9.1 is proposed for deletion; this TS is applicable in MODE 6. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODE 6 will no longer occur, TS 3.9.1 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.9.1 acceptable.

TS 3.9.2, Nuclear Instrumentation, establishes that two source range neutron flux monitors shall be operable, and one source range audible count rate circuit shall be

operable. TS 3.9.2 is proposed for deletion; this TS is applicable in MODE 6. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODE 6 will no longer occur, TS 3.9.2 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.9.2 acceptable.

TS 3.9.3, Containment Penetrations, establishes requirements pertaining to movement of recently irradiated fuel assemblies within containment. TS 3.9.3 is proposed for deletion; this TS is applicable during movement of recently irradiated fuel assemblies within containment. The licensee states that the Bases for TS 3.9.3 define recently irradiated fuel assemblies as fuel assemblies that have been part of a critical reactor in the previous 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br />. The PDTS will not be implemented until after 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> following permanent shutdown, so the specific condition of applicability will no longer occur. As a result, this TS will not be applicable in a permanently defueled condition. Therefore, the NRC staff finds the deletion of TS 3.9.3 acceptable.

TS 3.9.4, Residual Heat Removal (RHR) and Coolant Circulation - High Water Level, establishes that one RHR loop shall be operable and in operation. TS 3.9.4 is proposed for deletion; this TS is applicable in MODE 6 with water level 23 feet above the top of the reactor vessel flange. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODE 6 will no longer occur, TS 3.9.4 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.9.4 acceptable.

TS 3.9.5, Residual Heat Removal (RHR) and Coolant Circulation - Low Water Level, establishes that two RHR loops shall be operable, and one RHR loop shall be in operation. TS 3.9.5 is proposed for deletion; this TS is applicable in MODE 6 with water level 23 feet above the top of the reactor vessel flange. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, since operation in MODE 6 will no longer occur, TS 3.9.5 will not be applicable in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 3.9.5 acceptable.

TS 3.9.6, Refueling Cavity Water Level, is proposed for deletion; this TS is applicable in during movement of irradiated fuel assemblies within containment. The IP3 license will no longer authorize use of the facility for power operation, emplacement, or retention of fuel in the reactor vessel as provided in 10 CFR 50.82(a)(2). Therefore, this TS will no longer apply once the unit is in a permanent shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 3.9.6 acceptable.

4.4.6 TS Section 4.0, Design Features TS Section 4.0, Design Features, provides information and design requirements associated with plant systems. The licensee proposed grammatical corrections to TS 4.3.1.1 by exchanging the period and semicolon to the end of both TS 4.3.1.1.b and TS 4.3.1.1.d. The licensee also proposed the deletion of TS 4.3.1.2, which is related to the storage of new fuel.

Once the unit is permanently shut down and defueled, IP3 will not acquire new fuel for storage.

The TS is not applicable in the permanently shutdown and defueled condition. The NRC staff therefore finds the grammatical correction to TS 4.3.1.1 and the deletion of TS 4.3.1.2 acceptable.

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

4.4.7.1 TS 5.2, Organization The current TS 5.2.1, Onsite and Offsite Organizations, states the following:

a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all decommissioning organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the facility specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the FSAR and Quality Assurance Plan, as appropriate; The licensee proposed to revise TS 5.2.1 by replacing the term FSAR with DSAR. The revised description will state the following:
a. Lines of authority, responsibility, and communication shall be defined and established throughout highest management levels, intermediate levels, and all decommissioning organization positions. These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the facility-specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the DSAR and Quality Assurance Plan, as appropriate; The NRC staff reviewed the proposed changes to TS 5.2.1 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. The DSAR is the document that will remain applicable to IP3 once it is shut down and defueled. Therefore, the staff finds the proposed change is acceptable.

4.4.7.2 TS 5.4, Procedures The current TS 5.4.1 states the following:

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

The licensee proposed to revise TS 5.4.1.a by replacing the term Updated FSAR with DSAR and delete TS 5.4.1.d in its entirety to state the following:

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

The NRC staff reviewed the proposed changes to TS 5.4.1.a and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. The DSAR is the document that will remain applicable to IP3 once it is shut down and defueled. Therefore, the staff finds the proposed change is acceptable.

The NRC staff reviewed the proposed change to TS 5.4.1.d and concludes that the change is consistent with the transition to a permanently shutdown and defueled facility and the proposed deletion of License Condition 2.H. Because an FPP is required by 10 CFR 50.48(f) during the decommissioning process regardless of whether the TSs contain a requirement to establish, implement, and maintain procedures for an FPP, this TS is not needed. Therefore, the staff finds the proposed change is acceptable.

4.4.7.3 TS 5.5.2, Primary Coolant Sources Outside Containment The licensee proposed to delete the title for TS 5.5.2.

This is an administrative change as the TS 5.5.2 requirement was deleted in a previous license amendment (Amendment No. 267). Therefore, the NRC staff finds this editorial change acceptable.

4.4.7.4 TS 5.5.4, Radioactive Effluent Controls Program The licensee proposed to retain TS 5.5.4; however, TS paragraphs 5.5.4.d, 5.5.4.h, and 5.5.4.i are proposed to be modified by replacing unit with unit/facility.

Since IP3 will be permanently shutdown and defueled, the proposed change will appropriately reflect the status of IP3. Therefore, the NRC staff finds this administrative change acceptable.

4.4.7.5 TS 5.5.5, Component Cyclic or Transient Limit The licensee proposed to delete TS 5.5.5 in its entirety.

TS 5.5.5 provides controls to track the UFSAR, Section 4.1.5, cyclic and transient occurrences to ensure that components are maintained within the design limits. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Since operation in MODES 1 through 6 will no longer occur, TS 5.5.5 will not be applicable in a permanently defueled condition. Therefore, the NRC staff finds the deletion of TS 5.5.5 acceptable.

4.4.7.6 TS 5.5.6, Reactor Coolant Pump Flywheel Inspection Program The licensee proposed to delete TS 5.5.6 in its entirety.

TS 5.5.6 provides the inspection frequencies and acceptance criteria for the reactor coolant pump flywheel inspection program. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).

Since the reactor coolant pumps will no longer perform a function in the permanently shutdown and defueled condition, TS 5.5.6 is no longer applicable. Therefore, the NRC staff finds the deletion of TS 5.5.6 acceptable.

4.4.7.7 TS 5.5.7, Inservice Testing Program The licensee proposed to delete TS 5.5.7 in its entirety.

TS 5.5.7 provides controls for inservice testing of American Society of Mechanical Engineers (ASME) Code Class 1, 2, and 3 pumps and valves in the IP3 Inservice Testing Program that continue to operate and perform a specific function in mitigating the consequences of an accident due to the permanently shutdown and defueled status of the plant. Because the licensee is prohibited from operating the plant or placing fuel in the reactor vessel, in accordance with 10 CFR 50.82(a)(2), there are no longer any ASME Code class pumps and valves that remain in operation and are relied upon to mitigate a DBA. As such, TS 5.5.7 will no longer be relevant in the permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 5.5.7 acceptable.

4.4.7.8 TS 5.5.8, Steam Generator (SG) Program The licensee proposed to delete TS 5.5.8 in its entirety.

TS 5.5.8 requires that an SG program shall be established and implemented to ensure that SG tube integrity is maintained. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).

Therefore, TS 5.5.8 will no longer be applicable, as the SGs will no longer perform a function in the permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.8 acceptable.

4.4.7.9 TS 5.5.9, Secondary Water Chemistry Program The licensee proposed to delete TS 5.5.9 in its entirety.

TS 5.5.9 provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Therefore, TS 5.5.9 will no longer be applicable, as there will be no need to monitor secondary water chemistry to inhibit SG tube degradation in the permanently shutdown and defueled condition.

The NRC staff therefore finds the deletion of TS 5.5.9 acceptable.

4.4.7.10 TS 5.5.10, Ventilation Filter Testing Program (VFTP)

The licensee proposed to delete TS 5.5.10 in its entirety.

TS 5.5.10 requires that a program shall be established to implement certain testing procedures for the control room ventilation system. As discussed above, TS 3.7.11, Control Room Ventilation System (CRVS), is proposed for deletion, which the NRC staff has found acceptable. Therefore, the VFTP as a support program is not required at IP3 in a permanently defueled condition. The NRC staff therefore finds the deletion of TS 5.5.10 acceptable.

4.4.7.11 TS 5.5.12, Diesel Fuel Oil Testing Program The licensee proposed to delete TS 5.5.12 in its entirety.

TS 5.5.12 requires that a diesel fuel oil testing program to implement testing of both new fuel oil and stored fuel oil shall be established for the onsite diesel generator fuel oil storage tanks and the reserve fuel oil storage tanks. As discussed above, TSs 3.8.1, 3.8.2, and 3.8.3, which define the operability requirements regarding the diesel generators, are proposed for deletion.

Therefore, this support program is not required in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.12 acceptable.

4.4.7.12 TS 5.5.13, Technical Specifications (TS) Bases Control Program The licensee proposed to retain TS 5.5.13; however, the references to the updated FSAR and FSAR in TS 5.5.13.b.2 and TS 5.5.13.c, respectively, would be replaced with references to the DSAR.

The NRC staff reviewed the proposed changes to TS 5.5.13 and concludes that the changes are consistent with the transition to a permanently shutdown and defueled facility. The DSAR is the document that will remain applicable to IP3 once it is shut down and defueled. Therefore, the NRC staff finds the proposed change to TS 5.5.13 is acceptable.

4.4.7.13 TS 5.5.14, Safety Function Determination Program (SFDP)

The licensee proposed to delete TS 5.5.14 in its entirety.

The SFDP was established to ensure that loss of a safety function is detected and appropriate actions taken in accordance with the LCO pertaining to the loss of safety function. The LCOs remaining in the PDTS do not rely on the operability of any active equipment or redundant

safety systems. Also, because 10 CFR 50.82(a)(2) prohibits operation of the plant or placing fuel in the reactor vessel, there is no longer a need for redundant systems. Therefore, the requirements of the SFDP, which direct cross-train checks of multiple and redundant safety systems, no longer apply in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.14 acceptable.

4.4.7.14 TS 5.5.15, Containment Leakage Rate Testing Program The licensee proposed to delete TS 5.5.15 in its entirety.

TS 5.5.15 requires that a program shall establish the leakage rate testing of the containment.

Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Containment integrity is not credited in the analysis of the accidents that remains credible in the permanently defueled condition. Therefore, TS 5.5.15 will no longer be applicable in the permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.15 acceptable.

4.4.7.15 TS 5.5.16, Control Room Envelope Habitability Program The licensee proposed to delete TS 5.5.16 in its entirety.

TS 5.5.16 provides the required elements of the control room envelope habitability program. As discussed above, TS 3.7.11, Control Room Ventilation System (CRVS), is proposed for deletion. Therefore, this support program is not required in a permanently shutdown and defueled condition. The NRC staff therefore finds the deletion of TS 5.5.16 acceptable.

4.4.7.16 TS 5.6.4, Not Used TS 5.6.4 is a placeholder in TS 5.6, Reporting Requirements, and is proposed for deletion.

This is an administrative change to reflect reorganization of the TSs. Therefore, the NRC staff finds this administrative change acceptable.

4.4.7.17 TS 5.6.5, Core Operating Limits Report (COLR)

The licensee proposed to delete TS 5.6.5 in its entirety.

TS 5.6.5 provides the required documentation and analytical methods used to determine the reactor core operating limits. Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2).

TS 5.6.5 will no longer be applicable in the permanently shutdown and defueled condition, as there will no longer be a need to establish core operating limits. Therefore, the NRC staff finds the deletion of TS 5.6.5 acceptable.

4.4.7.18 TS 5.6.6, Not Used TS 5.6.6 is a placeholder and is proposed for deletion.

This is an administrative change to reflect reorganization of the TSs. Therefore, the NRC staff finds this administrative change acceptable.

4.4.7.19 TS 5.6.7, Post Accident Monitoring Instrumentation (PAM) Report The licensee proposed to delete TS 5.6.7 in its entirety.

TS 5.6.7 provides the reporting requirements associated with post-accident monitoring, which applies to an operating reactor, not to the safe storage and handling of spent fuel in the SFP.

Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). TS 5.6.7 will no longer be applicable in the permanently shutdown and defueled condition, as there will no longer be a need to establish post-accident monitoring. Therefore, the NRC staff finds the deletion of TS 5.6.7 acceptable.

4.4.7.20 TS 5.6.8, Steam Generator Tube Inspection Report The licensee proposed to delete TS 5.6.8 in its entirety.

TS 5.6.8 provides the reporting requirements associated with the SG tube inspection program.

Once the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2). Since the SGs will no longer perform a function in the permanently shutdown and defueled condition, TS 5.6.8 is no longer applicable. Therefore, the NRC staff finds the deletion of TS 5.6.8 acceptable.

4.5 Changes to Appendix B, Environmental Technical Specification Requirements 4.5.1 Title Page, Part I The current title page for Part I of Appendix B states, in part:

APPENDIX B TO FACILITY OPERATING LICENSE The licensee proposed to change the title page for Part I of Appendix B to state, in part:

APPENDIX B TO FACILITY LICENSE Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed, pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable to the NRC staff.

4.5.2 Appendix B, Section 1.0 Currently, Appendix B, Section 1.0 reads:

The Environmental Protection Plan (EPP) is to provide for protection of environmental values during construction and operation of the nuclear facility.

The principal objectives of the EPP are as follows:

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

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

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

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

The licensee proposed Appendix B, Section 1.0 to read:

The Environmental Protection Plan (EPP) is to provide for protection of environmental values during handling and storage of spent fuel and maintenance of the nuclear facility. The principal objectives of the EPP are as follows:

(1) Verify that the facility is maintained in an environmentally acceptable manner, as established by the FES and other NRC environmental impact assessments.

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

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

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

The proposed changes to Appendix B, Section 1.0 replace a reference to construction and operation with a reference to handling and storage of spent fuel and maintenance, replace a reference to plant is operated with facility is maintained, and replace a reference to facility construction and operation with handling and storage of spent fuel and maintenance of the facility. These proposed changes reflect the revised mission of the facility in the permanently shutdown and defueled condition. Since, consistent with 10 CFR 50.82(a)(2), the IP3 license will no longer authorize use of the facility for power operation or emplacement or retention of fuel into the reactor vessel, this change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable.

4.5.3 Appendix B, Section 3.1 Currently, Appendix B, Section 3.1 reads:

The licensee may make changes in station design or operation or perform tests or experimentsChanges in plant design or operation or performance of tests or experiments A proposed change, test, or experiment shall(2) a significant change in effluents or power level; The licensee proposed Appendix B, Section 3.1 to read:

The licensee may make changes in facility design or operation or perform tests or experimentsChanges in facility design or operation or performance of tests or experiments A proposed change, test or experiment shall(2) a significant change in effluents; The proposed changes to Appendix B, Section 3.1 replace references to station and plant with references to facility. These proposed changes reflect the revised mission of the facility in the permanently shutdown and defueled condition. The proposed change to Section 3.1 to eliminate the reference to power level reflects the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable.

4.5.4 Appendix B, Section 3.3 Currently, Appendix B, Section 3.3 reads:

Changes in plant design or operation and The licensee proposed Appendix B, Section 3.3 to read:

Changes in facility design or operation and The proposed change to Appendix B, Section 3.3 replaces the reference to plant with a reference to facility. This proposed change reflects the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.

4.5.5 Appendix B, Section 4.1 Currently, Appendix B, Section 4.1 reads:

Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and The licensee proposed Appendix B, Section 4.1 to read:

Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to the handling and storage of spent fuel and maintenance of the facility shall be recorded and The proposed change to Appendix B, Section 4.1 replaces the reference to plant operation with a reference to the handling and storage of spent fuel and maintenance of the facility. This proposed change reflects the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.

4.5.6 Appendix B, Section 4.2 Currently, Appendix B, Section 4.2 reads:

The currently applicable Biological Opinion concludes that continued operation of IP2 and IP3 is not likely to jeopardize the continued existence of the listed species or to adversely affect the designated critical habitat of those species.

The licensee proposed Appendix B, Section 4.2 to read:

The currently applicable Biological Opinion concludes that continued operation of IP2 and IP3 is not likely to jeopardize the continued existence of the listed species or to adversely affect the designated critical habitat of those species.

This Biological Opinion conservatively bounds the conditions that will occur in the permanently shut down and defueled condition.

The proposed change to Appendix B, Section 4.2 concludes that the biological opinion rendered during the evaluation of the continued operation of IP2 and IP3 conservatively bounds the conditions that will occur in the permanently shutdown and defueled condition. The NRC staff notes that the biological opinion addresses the permanent shutdown and defueling of IP3 regarding the shortnose sturgeon and Atlantic sturgeon. Therefore, the NRC staff finds the proposed change acceptable.

4.5.7 Appendix B, Section 5.2 Currently, Appendix B, Section 5.2 reads:

Records and logs relative to the environmental aspects of plant operation shall be made and retained in a manner convenient for review and inspection. These records and logs shall be made available to the NRC on request.

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

The licensee proposed Appendix B, Section 5.2 to read:

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

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

The proposed changes to Appendix B, Section 5.2 clarify that the reference to plant operation refers to plant operations prior to the permanent shutdown and includes a reference to the handling and storage of spent fuel and maintenance of the facility. In addition, references to plant are replaced with facility. These proposed changes reflect the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable.

4.5.8 Appendix B, Section 5.4.1 Currently, Appendix B, Section 5.4.1 reads:

and an assessment of the observed impacts of the plant operation on the environment (b) A list of all changes in station design or operation, tests, and experiments made in accordance with subsection 3.1 which involved a potentially significant unreviewed environmental issue.

The licensee proposed Section 5.4.1 to read:

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

(b) A list of all changes in facility design or operation, tests, and experiments made in accordance with subsection 3.1 which involved a potentially significant unreviewed environmental issue.

The proposed changes to Appendix B, Section 5.4.1 clarify that the reference to plant operation refers to plant operations prior to the permanent shutdown and includes a reference to the handling and storage of spent fuel and maintenance of the facility. In addition, a reference to station is replaced with a reference to facility. These proposed changes reflect the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are acceptable.

4.5.9 Appendix B, Section 5.4.2 Currently, Section 5.4.2 reads:

The report shall (1) describe, analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (2)

The licensee proposed Section 5.4.2 to read:

The report shall (1) describe, analyze, and evaluate the event, including extent and magnitude of the impact and facility conditions, (2)

The proposed change to Appendix B, Section 5.4.2 replaces a reference to plant operating characteristics with facility conditions. This proposed change reflects the revised mission of the facility in the permanently shutdown and defueled condition. After the certifications required by 10 CFR 50.82(a)(1) are docketed for IP3, the 10 CFR Part 50 license will no longer authorize operation of the reactor or placement or retention of fuel in the reactor vessel pursuant to 10 CFR 50.82(a)(2). Therefore, the NRC staff finds that the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.

4.5.10 Title Page, Part II The current title page for Part II of Appendix B states, in part:

APPENDIX B TO FACILITY OPERATING LICENSE The licensee proposed to change the title page for Part II of Appendix B to state, in part:

APPENDIX B TO FACILITY LICENSE

Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed, pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description. Therefore, the NRC staff finds that the proposed change is consistent with 10 CFR 50.82(a)(2) and is acceptable.

4.6 Changes to Appendix C, Inter-Unit Fuel Transfer Technical Specifications The current title page for Part I and Part II of Appendix C states, in part:

APPENDIX C TO FACILITY OPERATING LICENSE The licensee proposed to change the title page for Part I and Part II of Appendix C to state, in part:

APPENDIX C TO FACILITY LICENSE Also, the current header for Part I of Appendix C states, in part:

Facility Operating License The licensee proposed to change the header for Part 1 of Appendix C to state, in part:

Facility License Appendix C is modified by replacing the references to Facility Operating License with Facility License. The proposed changes reflect the upcoming change in status regarding IP3. Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed, pursuant to 10 CFR 50.82(a)(2), the IP3 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel, and this proposed change would provide accuracy in the 10 CFR Part 50 license description.

Therefore, the NRC staff finds that the proposed changes are consistent with 10 CFR 50.82(a)(2) and are 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 36435; June 16, 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 C. Jackson D. Scully E. Stutzcage L. Alvarado J. Robinson S. Mehta J. Tsao B. Wolfgang K. West Date: April 22, 2021

ML21074A000 OFFICE NRR/DORL/LPL1/PM NRR/DORL/LPL1/LA NRR/DRA/APLB/BC(A)

NAME RGuzman JBurkhardt JBorromeo DATE 3/22/2021 3/19/2021 08/25/2020 OFFICE NRR/DNRL/NCSG/BC NRR/DSS/SNSB/BC NRR/DNRL/NVIB/BC NAME SBloom SKrepel HGonzalez DATE 09/17/2020 11/30/2020 12/9/2020 OFFICE NRR/DRA/ARCB/BC NRR/DSS/SCPB/BC NRR/DSS/SFNB/BC NAME KHsueh BWittick RLukes DATE 12/17/2020 12/20/2020 3/22/2021 OFFICE NRR/DSS/STSB/BC OGC-NLO w/revisions NRR/DORL/LPL1/PM NAME VCusumano STurk JDanna (JPaige for)

DATE 3/19/2021 4/15/2021 4/22/2021 OFFICE NRR/DORL/LPL1/PM NAME RGuzman DATE 4/22/2021

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