ML18227A338

Issuance of Amendment No. 295, Proposed Defueled Technical Specifications and Revised License Conditions for Permanently Defueled Condition

ML18227A338
Person / Time
Site:	Oyster Creek
Issue date:	10/26/2018
From:	John Lamb Special Projects and Process Branch
To:	Bryan Hanson Entergy Nuclear Operations
Lamb J, NRR/DORL/LSPB, 415-3100
References
L-2017-LLA-0395
Download: ML18227A338 (85)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 October 26, 2018 Mr. Bryan C. Hanson President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

OYSTER CREEK NUCLEAR GENERATING STATION- ISSUANCE OF AMENDMENT RE: LICENSE AMENDMENT REQUEST FOR PROPOSED DEFUELED TECHNICAL SPECIFICATIONS AND REVISED LICENSE CONDITIONS FOR PERMANENTLY DEFUELED CONDITION (EPID L-2017-LLA-0395)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 295 to Renewed Facility Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station (Oyster Creek), in response to your application dated November 16, 2017, as supplemented by letter dated March 29, 2018.

The amendment revises the Oyster Creek renewed facility operating license and the associated technical specifications to permanently defueled technical specifications consistent with the permanent cessation of operations and permanent removal of fuel from the reactor vessel.

A copy of the related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Jo G. Lamb, Senior Project Manager Sp cial Projects and Process Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-219

Enclosures:

1. Amendment No. 295 to Renewed DPR-16

2. Safety Evaluation cc: Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 295 Renewed License No. DPR-16

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

A. The application for amendment by Exelon Generation Company, LLC (the licensee), dated November 16, 2017, as supplemented by letter dated March 29, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's 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 Commission's 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 Commission's regulations and all applicable requirements have been satisfied.

Enclosure 1

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraphs 1.B, 1.D, 1.E, 2.B.(1 ), 2.B.(2), 2.B.(3), 2.B.(5), 2.C.(1 ), 2.C.(2), 2.C.(3), 2.C.(5), 2.C.(6), 2.C.(7),

2.C.(10) through 2.C.(15), 2.D, 2.E, 3.A through 3.K, 3.M, and 4 of the Renewed Facility Operating License No. DPR-16 are hereby amended to read as follows:

1.B. DELETED 1.D. 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; 1.E. There is reasonable assurance (i) that the activities authorized by this 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 rules and regulations set forth in 10 CFR Chapter I; 2.B.(1) Pursuant to Section 104b of the Act and 10 CFR Part 50, to possess and use Oyster Creek Nuclear Generation Station at the designated location on the Oyster Creek site in Ocean County, New Jersey, in accordance with the procedures and limitations set forth in this renewed license; 2.B.(2) 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 Updated Final Safety Analysis Report, as supplemented and amended; 2.B.(3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, or special nuclear materials as sealed neutron sources that were used for reactor startup, sealed sources that were used for calibration of reactor instrumentation and are used in radiation monitoring equipment, and as fission detectors in amounts as required; 2.B.(5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate such byproduct, source, or special nuclear materials that were produced by the operation of the facility.

2.C.(1) DELETED 2.C.(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 295, are hereby incorporated in the license. Exelon Generation Company shall

maintain the facility in accordance with the Permanently Defueled Technical Specifications (POTS).

2.C.(3) DELETED 2.C.(5) DELETED 2.C.(6) DELETED 2.C.(7) DELETED 2.C.(10) DELETED 2.C.(11) DELETED 2.C.(12) DELETED 2.C.(13) DELETED 2.C.(14) DELETED 2.C.(15) DELETED 2.D DELETED 2.E DELETED 3.A DELETED 3.B DELETED 3.C DELETED 3.D DELETED 3.E DELETED 3.F DELETED 3.G DELETED 3.H DELETED 3.1 DELETED 3.J DELETED 3.K DELETED 3.M DELETED

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

3. This license amendment is effective on November 16, 2018, and shall be implemented in 60 days from the effective date.

FOR THE NUCLEAR REGULATORY COMMISSION

-121-Douglas A. Broaddus, Chief Special Projects and Process Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to Renewed Facility Operating License No. DPR-16 and Technical Specifications Date of Issuance: October 26, 2018

ATTACHMENT TO LICENSE AMENDMENT NO. 295 OYSTER CREEK NUCLEAR GENERATING STATION RENEWED FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Replace the following pages of Renewed Facility Operating License No. DPR-16 and the technical specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Renewed Facility License No. DPR-16 REMOVE INSERT to to Appendix A- Technical Specifications REMOVE INSERT i to iii i 1.0-1 to 1.0-9 1.0-1 2.1-1 to 2.1-3 2.2-1 2.3-1 to 2.3-8 3.0-1 to 3.0-4 3/4.0-1 to 3/4.0-2 3.1-1 to 3.1-21 3/4.1-1 3.2-1 to 3.2-12 3/4.2-1 3.3-1 to 3.3-10 3.4-1 to 3.4-9 3.5-1 to 3.5-12 3.6-1 to 3.6-6 3.7-1 to 3.7-5 3.8-1 to 3.8-3 3.9-1 to 3.9-3 3.10-1 to 3.10-4 3.11-1 3.12-1 to 3.12-2 3.13-1 to 3.13-5 3.14-1 3.15-1 to 3.15-3 3.17-1 4.0-1 to 4.0-4 4.1-1 to 4.1-10 4.2-1 to 4.2-4 4.3-1 to 4.3-4 4.4-1 to 4.4-3 4.5-1 to 4.5-18

Appendix A - Technical Specifications REMOVE INSERT 4.6-1 to 4.6-2 4.7-1 to 4.7-6 4.8-1 to 4.8-2 4.9-1 to 4.9-2 4.10-1 to 4.10-2 4.11-1 4.12-1 to 4.12-2 4.13-1 to 4.13-2 4.14-1 4.15-1 to 4.15-2 4.16-1 4.17-1 5.1-1 to 5.1-2 5.1-1 5.2-1 5.3-1 to 5.3-2 6-3 to 6-14 6-3 to 6-7

EXELON GENERATION COMPANY, LLC DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION RENEWED FACILITY OPERATING LICENSE Renewed License No. DPR-16

1. The Nuclear Regulatory Commission (the Commission) having previously made the findings set forth in License No. DPR-16, has now found that:

A. The application for a Renewed Facility Operating License No. DPR-16 filed by the applicant complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I and all required notifications to other agencies or bodies have been duly made; B. DELETED C. Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the term of this Renewed Facility Operating License No. DPR-16 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 the renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commission's regulations; D. 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; E. There is reasonable assurance (i) that the activities authorized by this 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 rules and regulations set forth in 10 CFR Chapter I; F. Exelon Generation Company, LLC (Exelon Generation Company) is technically qualified to engage in the activities authorized by this license in accordance with the rules and regulations of the Commission; Renewed License No. DPR-16 Amendment No. 295

G. Exelon Generation Company has satisfied the applicable provisions of 10 CFR Part 140, "Financial Protection Requirements and Indemnity Agreements," of the Commission's regulations; H. The issuance of this license will not be inimical to the common defense and security or to the health and safety of the public; I. The receipt, possession and use of source, byproduct, and special nuclear materials as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Parts 30, 40, and 70; and J. The issuance of this license is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Facility Operating License No. DPR-16, dated July 2, 1991, as amended, is superseded in its entirety by Renewed Facility Operating License No. DPR-16, hereby issued to Exelon Generation Company, to read as follows:

A. This renewed license applies to the Oyster Creek Nuclear Generating Station, a boiling-water reactor and associated equipment (the facility). The facility is located in Ocean County, New Jersey, and is described in the licensee's Updated Final Safety Analysis Report, as supplemented and amended, and in the licensee's Environmental Report, as supplemented and amended.

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

(1) Pursuant to Section 104b of the Act and 10 CFR Part 50, to possess and use Oyster Creek Nuclear Generation Station at the designated location on the Oyster Creek site in Ocean County, New Jersey, in accordance with the procedures and limitations set forth in this renewed license; (2) 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 Updated Final Safety Analysis Report, as supplemented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, or special nuclear materials as sealed neutron sources that were used for reactor startup, sealed sources that were used for calibration of reactor instrumentation and are used in radiation monitoring equipment, and as fission detectors in amounts as required; (4) 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 materials without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; and Renewed License No. DPR-16 Amendment No. 295

( 5) Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate such byproduct, source, or special nuclear materials that were produced by the operation of the facility.

C. This license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations set forth in 10 CFR Chapter I 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 or incorporated below:

(1) DELETED (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 295, are hereby incorporated in the license. Exelon Generation Company shall maintain the facility in accordance with the Permanently Defueled Technical Specifications (POTS).

(3) DELETED

( 4) Exelon Generation Company 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 the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The combined set of plans 1 , submitted by letter dated May 17, 2006, is entitled: "Oyster Creek Nuclear Generating Station Security Plan, Training and Qualification Plan, and Safeguards Contingency Plan, Revision 5."

The set contains Safeguards Information protected under 10 CFR 73.21.

Exelon Generation Company 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 Exelon Generation Company CSP was approved by License Amendment No. 280 and modified by License Amendment Nos. 288 and 292.

(5) DELETED (6) DELETED (7) DELETED 1

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

Renewed License No. DPR-16 Amendment No. 295

(8) Mitigation Strategy License Condition 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 (9) The licensee shall implement and maintain all Actions required by Attachment 2 to NRC Order EA-06-137, issued June 20, 2006, except the last action that requires incorporation of the strategies into the site security plan, contingency plan, emergency plan and/or guard training and qualification plan, as appropriate.

(10) DELETED

( 11) DELETED (12) DELETED (13) DELETED

( 14) DELETED (15) DELETED Renewed License No. DPR-16 Amendment No. 295

(16) License Renewal Commitments The UFSAR supplement, as revised, describes certain future activities to be completed prior to April 9, 2009, and during the term of this renewed operating license No. DPR-16. Exelon Generation Company shall complete these activities in accordance with Appendix A of NUREG-1875, "Safety Evaluation Report Related to the License Renewal of Oyster Creek Generating Station," dated March 2007, as supplemented on September 19, 2008, and shall notify the NRC in writing when implementation of those activities required prior to April 9, 2009 are complete and can be verified by NRC inspection.

(17) Biological Opinion Within 30 days from the issuance date of the renewed license, Exelon Generation Company shall comply with the terms and conditions of the Incidental Take Statement associated with certain sea turtles in the Biological Opinion in effect or as subsequently issued by the National Marine Fisheries Service regarding operation of the facility.

D. DELETED E. DELETED F. The licensee shall have and maintain financial protection of such type and in such amounts as the Commission shall require in accordance with Section 170 of the Atomic Energy Act of 1954, as amended, to cover public liability claims.

3. Sale and License Transfer Conditions:

A. DELETED B. DELETED C. DELETED D. DELETED E. DELETED F. DELETED G. DELETED H. DELETED I. DELETED J. DELETED Renewed License No. DPR-16 Amendment No. 295

K. DELETED L. DELETED M. DELETED

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

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/

Bruce S. Mallett Deputy Executive Director for Reactor and Preparedness Programs Office of the Executive Director for Operations

Attachment:

Appendices A and B -

Technical Specifications Date of Issuance: April 8, 2009 Renewed License No. DPR-16 Amendment No. 295

TABLE OF CONTENTS Section 1 DEFINITIONS Page 1.1 Actions 1.0-1 1.2 Certified Fuel Handler 1.0-1 1.3 Non-Certified Operator 1.0-1 Section 2.0 DELETED Section 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 Limiting Conditions for Operation and Surveillance Requirement Applicability 3/4.0-1 3/4.1 Spent Fuel Storage 3/4.1-1 3/4.2 Radioactive Liquid Storage 3/4.2-1 Section 5 DESIGN FEATURES 5.1 Site 5.1-1 5.2 Spent Fuel Storage 5.1-1 Section 6 ADMINISTRATIVE CONTROLS 6.1 Responsibility 6-1 6.2 Organization 6-1 6.3 Facility Staff Qualifications 6-2 6.4 DELETED 6-2 6.5 DELETED 6-2 6.6 DELETED 6-2 6.7 DELETED 6-2 6.8 Procedures and Programs 6-3 6.9 Reporting Requirements 6-5 6.10 Record Retention 6-6 6.11 DELETED 6-6 6.12 DELETED 6-6 6.13 High Radiation Area 6-6 6.14 DELETED 6-6 6.15 DELETED 6-7 6.16 DELETED 6-7 6.17 DELETED 6-7 6.18 DELETED 6-7 6.19 Offsite Dose Calculation Manual 6-7 6.20 DELETED 6-7 6.21 Technical Specification (TS) Bases Control Program 6-7 OYSTER CREEK Amendment No.: 161,106,205,241,276

~295

SECTION I DEFINITIONS The following frequently used terms are defined to aid in the uniform interpretation of the specifications.

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

1.2 CERTIFIED FUEL HANDLER A CERTIFIED FUEL HANDLER is an individual who complies with provisions of the CERTIFIED FUEL HANDLER training program required by Specification 6.3.2.

1.3 NON-CERTIFIED OPERATOR A NON-CERTIFIED OPERATOR is a non-licensed operator who complies with the qualification requirements of Specification 6.3.1, but is not a CERTIFIED FUEL HANDLER.

OYSTER CREEK 1.0-1 Amendment No.: 20,44,64,167,178, 295

SECTION 3/4 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 3/4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENT APPLICABILITY Applicability: Applies to all Limiting Conditions for Operation and Surveillance Requirements.

Objective: To preserve the single failure criterion for safety systems.

LCO Applicability LCO 3.0.1 LCOs shall be met during the specified conditions in the TS, 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.

Surveillance Requirement Applicability SR 4.0.1 Surveillance requirements shall be met during the specified conditions in the applicability for individual LCOs, unless otherwise stated in the surveillance requirements. 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 4.0.2. Surveillances do not have to be

• performed on variables outside specified limits.

SR 4.0.2 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.

OYSTER CREEK 3/4.0-1 Amendment No. 295

SR 4.0.3 Entry into a specified condition in the Applicability ofan LCO shall only be made when the LCO's Surveillance has been met within its specified frequency, except as provided by 4.0.2.

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

SR 4.0.4 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.

OYSTER CREEK 3/4.0-2 Amendment No.: 295

3/4.1 SPENT FUEL STORAGE Applicability: During movement of irradiated fuel assemblies in the spent fuel pool.

Objective: To assure safe storage of spent fuel.

LCO: 3.1 Spent Fuel Pool Water Level Whenever irradiated fuel is stored in the spent fuel storage pool, water level shall be maintained at a level~ 117 feet 8 inches (elevation above sea level) with the exception of planned cask movements.

ACTIONS:

Condition Required Action Completion Time Spent fuel pool water Suspend movement of irradiated fuel Immediately level is not within assemblies and movement of loads over limit. the storage racks containing fuel.

SURVEILLANCE REQUIREMENTS Surveillance Freguency 4.1 Verify the spent fuel pool water level is ~ 117 feet 8 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> inches.

OYSTER CREEK 3/4.1-1 Amendment No.: 295

3/4.2 RADIOACTIVE LIQUID STORAGE Applicability: Applies at all times to outdoor tanks used to store radioactive liquids.

Objective: To assure that radioactive effluents are not released to the environment in an uncontrolled manner and to assure that the radioactive concentrations of any material released is kept as low as is reasonably achievable and, in any event, within the limits of 10 CFR Part 20.1301 and 40 CFR Part 190.1 O(a).

LCO: 3.2 The quantity of radioactive material, excluding tritium, noble gases, and radionuclides having half-lives shorter than three days, contained in outdoor storage tanks shall not exceed 10.0 curies. Included in this specification are all outdoor storage tanks that contain radioactivity that are not surrounded by liners, dikes, or walls capable of holding the tank contents, or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

ACTIONS:

Condition Required Action Completion Time In the event the Begin treatment and continue it until As soon as quantity of the total quantity of radioactive reasonably radioactive material material in the tank is 10 curies or achievable in any applicable less, and describe the reason for storage tank exceeding the limit in the next Annual exceeds 10.0 curies. Effluent Release Report.

SURVEILLANCE REQUIREMENTS Surveillance Frequency 4.2 Liquids contained in outdoor storage tanks included Once per 7 days in this specification shall be sampled and analyzed when radioactive for radioactivity. liquid is being added to the tank OYSTER CREEK 3/4.2-1 Amendment No.: 295

SECTION 5 DESIGN FEATURES 5.1 SITE A. The reactor (center line) is located 1,358 feet west of the east boundary of New Jersey State Highway Route 9 which is the minimum exclusion distance as defined in 10 CFR 100.3. The licensee will at all times retain the complete authority to determine and maintain sufficient control of all activities through ownership, easement, contract and/or other legal instruments on property which is closer to the reactor (center line) than 1,358 feet. This includes the authority to exclude or remove personnel and property within the minimum exclusion distance.

5.2 SPENT FUEL STORAGE 5.2.1 Spent Fuel Storage A. The spent fuel storage facilities are designed and shall be maintained with a K-effective equivalent to less than or equal to 0.95 including all calculational uncertainties.

B. The temperature of the water in the spent fuel storage pool, measured at or near the surface, shall not exceed 125°F.

C. The maximum amount of spent fuel assemblies stored in the spent fuel storage pool shall be 3035.

OYSTER CREEK 5.1-1 Amendment No. ~ 295

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the items referenced below:

a. The procedures applicable to safe storage of nuclear fuel recommended in Appendix "A" of Regulatory Guide 1.33 as referenced in the Decommissioning Quality Assurance Program (DQAP).

b. Surveillance and test activities of equipment that affects nuclear safety and radioactive waste management equipment.

c. Fuel Handling Operations.

d. Security Plan Implementation.

e. Fire Protection Program Implementation.

f. Emergency Plan Implementation.

g. Process Control Plan Implementation.

h. Offsite Dose Calculation Manual Implementation.

i. Quality Assurance Program for effluent and environmental monitoring using the guidance in Regulatory Guide 4.15, Revision 1.

6.8.2 Each procedure required by 6.8.1 above, and substantive changes thereto, shall be reviewed and approved prior to implementation and shall be reviewed periodically as set forth in administrative procedures.

6.8.3 Temporary changes to procedures of 6.8.1, above, may be made provided:

a. The intent of the original procedure is not altered;

b. The change is approved by two members of the licensee's management staff knowledgeable in the area affected by the procedure. For changes which may affect the operational status of facility systems or equipment, at least one of these individuals shall be a member of operations management or supervision who is a CERTIFIED FUEL HANDLER.

c. The change is documented, reviewed and approved within 14 days of implementation.

OYSTER CREEK 6-3 Amendment No.: 69,78,89,108,117,125,134,161,180, 181,194,203,210,213,224,232, 251,273, 290, 295

6.8.4 The following programs shall be established, implemented and maintained:

a. Radioactive Effluent Controls Program A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluent and for maintaining the doses to members of the public from radioactive effluent as low as reasonably achievable. The program (1) shall be contained in the ODCM, (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1. Limitations on the operability of radioactive liquid and gaseous monitoring instrumentation including the surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

2. Limitations on the concentrations of radioactive material released in liquid effluent to the unrestricted area conforming to less than the concentration values in Appendix B, Table 2, Column 2 to 10 CFR 20.1001-20.2402.

3. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluent in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM.

4. Limitations on the annual and quarterly doses and dose commitment to a member of the public from radioactive materials in liquid effluent released to the unrestricted area conforming to Appendix I of 10 CFR 50,

5. 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.

6. Limitations on the operability and use of the liquid and gaseous effluent treatment systems to ensure that the appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in the 31 day period would exceed 2 percent of the guidelines for the annual dose or dose commitment conforming to Appendix I to 10 CFR 50,

7. Limitations on the dose rate resulting from radioactive materials released in gaseous effluents from the site to the unrestricted area shall be limited to the following:

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

b. For iodine-131, iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mRems/yr to any organ.

8. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents to the unrestricted area conforming to Appendix I of 10 CFR 50, OYSTER CREEK 6-4 Amendment No.: 290,295

9. Limitations on the annual and quarterly doses to a member of the public from 1-131, 1-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluent released beyond the site boundary conforming to Appendix I of 10 CFR 50, 1O. Limitations on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from Uranium fuel cycle sources conforming to 40 CFR Part 190.

b. Radiological Environmental Monitoring Program A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in the ODCM, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

1. Monitoring, sampling, analysis, and reporting of radiation and radionuclides in the environment in accordance with the methodology and parameters in the ODCM, 6.9 REPORTING REQUIREMENTS In addition to the applicable reporting requirements of 10 CFR, the following identified reports shall be submitted to the Administrator of the NRC Region I office unless otherwise noted.

6.9.1 Routine Reports

a. Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the facility during 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 effluent and solid waste released from the facility. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

b. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the facility during the previous calendar year shall be submitted prior to May 1 of each year.

The Report shall include summaries, interpretations, and an analysis 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:

(1) the ODCM; and, (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

OYSTER CREEK 6-5 Amendment No.: 2-90,- 295

6.10 RECORD RETENTION 6.10.1 Quality Assurance Records shall be retained as specified by the DQAP.

6.11 DELETED 6.12 DELETED 6.13 HIGH RADIATION AREA 6.13.1 In lieu of the "control device" or "alarm signal" required by Section 20.1601 of 10 CFR 20, each high radiation area in which the intensity of radiation at 30 cm (11.8 in.) is greater than deep dose equivalent of 100 mRem/hr but less than 1,000 mRem/hr shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP).

NOTE: Health Physics personnel shall be exempt from the RWP issuance requirement during the performance of their assigned radiation protection duties, provided they are following plant radiation protection procedures for entry into high radiation areas.

An individual or group of individuals permitted to enter such areas shall be provided with one or more of the following:

a. A radiation monitoring device which continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a pre-set integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them.

c. A health physics qualified individual (i.e., qualified in radiation protection procedures) with a radiation dose rate monitoring device who is responsible for providing positive exposure control over the activities within the area and who will perform periodic radiation surveillance at the frequency in the RWP. The surveillance frequency will be established by the management position responsible for radiological controls.

6.13.2 Specification 6.13.1 shall also apply to each high radiation area in which the intensity of radiation is greater than deep dose equivalent of 1,000 mRem/hr at 30 cm (11.8 in.) but less than 500 rads in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1 meter (3.28 ft.) from sources of radioactivity. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of operations and/or radiation protection supervision on duty.

6.14 DELETED OYSTER CREEK 6-6 Amendment No.: 2-9-G, 295

6.15 DELETED 6.16 DELETED 6.17 DELETED 6.18 DELETED 6.19 OFFSITE DOSE CALCULATION MANUAL

a. Licensee initiated changes to the ODCM shall be submitted to the NRC in the Annual Radioactive Effluent Release Report for the period in which the changes were made.

This submittal shall contain:

1. sufficiently detailed information to justify the changes without benefit of additional or supplemental information;

2. a determination that the changes did not reduce the accuracy or reliability of dose calculations or setpoint determination; and,

3. documentation that the changes have been reviewed and approved pursuant to Section 6.8.2.

b. Change(s) shall become effective upon review and approval by licensee management.

6.20 DELETED 6.21 TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1 . A change in the TS incorporated in the license or

2. A change to the updated FSAR (UFSAR) or Bases that requires NRG approval pursuant to 10 CFR 50.59.

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

d. Proposed changes that meet the criteria of Specification 6.21.b.1 or 6.21.b.2 above shall be reviewed and approved by the NRG 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).

OYSTER CREEK 6-7 Amendment No.: 69,78,84,117,134,203,210, 213,251,290, 295

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 295 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-16 EXELON GENERATION COMPANY, LLC OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219

1.0 INTRODUCTION

By application dated November 16, 2017 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML17320A411 ), as supplemented by letter dated March 29, 2018 (ADAMS Accession No. ML18088A317), Exelon Generation Company, LLC (Exelon or the licensee) requested changes to Renewed Facility Operating License (RFOL)

No. DPR-16 and the associated technical specifications (TSs) for the Oyster Creek Nuclear Generating Station (Oyster Creek). Specifically, Exelon requested an amendment to revise the Oyster Creek 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.

The supplemental letter dated March 29, 2018, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register (FR)on January 16, 2018 (83 FR 2229).

2.0 BACKGROUND

By letter dated January 23, 2002 (ADAMS Accession No. ML013410156), the NRC issued Amendment No. 223 to Oyster Creek. This amendment deleted TSs 5.3.1.B and 5.3.1.C.

These TSs restricted the handling of heavy loads over irradiated fuel stored in the spent fuel pool (SFP). The basis for deleting these TSs was the upgrade of the reactor building crane and associated handling systems to a single-failure proof system.

By letter dated April 26, 2007 (ADAMS Accession No. ML071080019), the NRC issued Amendment No. 262 to Oyster Creek. This amendment revised the Oyster Creek licensing basis in the area of radiological dose analyses for design-basis accidents (DBAs) using the alternate source term (AST) depicted in Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," July 2000 Enclosure 2

(ADAMS Accession No. ML003716792). Additionally, this amendment revised the Oyster Creek TSs consistent with the amended design basis.

By letter dated January 7, 2011 (ADAMS Accession No. ML110070507), the licensee submitted Notification of Permanent Cessation of Power Operations for Oyster Creek. In this letter, Exelon notified the NRC of its intent to permanently cease operations at Oyster Creek no later than December 31, 2019. By letter dated February 14, 2018 (ADAMS Accession No. ML18045A084), the licensee submitted its revised Notification of Permanent Cessation of Power Operations for Oyster Creek. In this letter, Exelon notified the NRC of its intent to permanently cease operations at Oyster Creek no later than October 31, 2018.

On September 17, 2018 (ADAMS Accession No. ML18263A163), Exelon permanently ceased power operations at Oyster Creek. By letter dated September 25, 2018 (ADAMS Accession No. ML18268A258), Exelon certified that all the fuel was permanently removed the Oyster Creek reactor vessel.

In accordance with Title 10 of the Code of Federal Regulations (10 CFR) Section 50.82(a)(2),

upon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, the Oyster Creek 10 CFR part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel.

By letter dated September 6, 2016 (ADAMS Accession No. ML16222A787), the NRC staff approved a Certified Fuel Handler (CFH) training program for Oyster Creek.

By letter dated March 7, 2017 (ADAMS Accession No. ML16235A413), the NRC staff issued Amendment No. 290 for Oyster Creek. This amendment revised and removed certain requirements from the Section 6, "Administrative Controls," portions of the Oyster Creek TSs that are not applicable to the facility in a permanently defueled condition. In addition, the amendment added definitions to TS Section 1, "Definitions." Also, the amendment made additions to, deletions from, and conforming administrative changes to the TSs.

By letter dated June 23, 2017 (ADAMS Accession No. ML17067A042), the NRC staff issued Amendment No. 291 for Oyster Creek. This amendment deleted certain license conditions, which imposed specific requirements on the decommissioning trust fund agreement, so that, instead, the provisions of 10 CFR 50. 75(h) that specify the regulatory requirements for decommissioning trust funds apply to Oyster Creek.

By letter dated December 22, 2017 (ADAMS Accession No. ML17289A222), the NRC staff issued Amendment No. 292 for Oyster Creek. This amendment revised the Oyster Creek RFOL for the Cyber Security Plan Milestone 8 full implementation completion date, as set forth in the Cyber Security Plan implementation schedule, and the physical protection license condition.

Specifically, this amendment revised the Cyber Security Plan Milestone 8 completion date from December 31, 2017, to August 31, 2021.

The existing Oyster Creek TSs contain limiting conditions for operation (LCOs) that provide for appropriate functional capability of equipment required for safe operation of the facility, including when the plant is in a defueled condition. Since the safety function related to safe storage and management of irradiated fuel at an operating plant is similar to the corresponding function at a permanently defueled facility, the applicable existing TSs provide an appropriate level of control.

However, the majority of the existing TSs are only applicable when the reactor is in an operational MODE. Once Exelon submits its certifications of permanent cessation of operations

and permanent removal of fuel from the reactor vessel for Oyster Creek, consistent with 10 CFR 50.82(a)(2), the Oyster Creek 10 CFR Part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel; therefore, the LCOs (and associated surveillance requirements (SRs)) that do not apply in a defueled condition are being proposed for deletion. The proposed amendment would revise the RFOL and associated TSs to reflect the permanent cessation of operations and the permanent removal of fuel from the reactor vessel at Oyster Creek. In general, the changes would eliminate those TSs applicable in operating MODES and MODES where fuel is em placed in the reactor vessel, as well as certain TSs required for the movement of irradiated fuel assemblies. Changes are also proposed to TS definitions, administrative controls, and related to programs and procedures.

The proposed amendment would also revise the RFOL to clarify or remove certain conditions not relevant to the permanently shutdown and defueled condition and would add conditions consistent with other permanently shutdown and defueled reactors.

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 NRC's 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) LCOs, (3) SRs, (4) design features, (5) administrative controls, (6) decommissioning, (7) initial notification, and (8) written reports.

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, or components (SSCs) that affect the integrity of fission product barriers during 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 are no longer required to be included in TSs, i~ provided below.

Criterion 1 of 10 CFR 50.36(c)(2)(ii)(A) states that TS LCOs must be established for "installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." Since no fuel is present in the reactor or RCS at the Oyster Creek facility, 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 OBA or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The purpose of this criterion is to capture those process variables that have initial values assumed in the OBA and transient analyses, and which are monitored and controlled during power operation. The scope of DBAs applicable to a permanently shutdown and defueled reactor is reduced from those postulated for an operating reactor, and most TSs satisfying Criterion 2 are no longer applicable. The one existing TS that defines the initial condition of the OBA associated with irradiated fuel movement is discussed in Section 3.5 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. There are no transients that continue to apply to permanently shutdown and defueled reactors. The scope of applicable DBAs that continue to apply to Oyster Creek 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 an SSC "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 are no longer any DBAs at Oyster Creek in the permanently shutdown and defueled condition that can result in a significant offsite radiological risk to public health and safety.

3.2 Radiological Consequences from Design-Basis Accidents Chapter 15, "Accident Analysis," of the Oyster Creek Updated Final Safety Analysis Report (UFSAR) describes the OBA scenarios that are applicable to Oyster Creek during power and refueling operations and the accidents with the greatest potential for radiation exposure. The most severe postulated accidents for nuclear power plants involve damage to the nuclear reactor core and the release of large quantities of fission products. When the reactor is permanently defueled and irradiated fuel assemblies are stored in the SFP and the independent spent fuel storage installation, the spectrum of credible accidents is much smaller than for an operational plant, and most of the accident scenarios postulated in the UFSAR are no longer possible.

The licensee stated that the only accident with potential offsite radiological consequences that remains applicable to Oyster Creek in the permanently shutdown and defueled condition is a fuel handling accident (FHA) in the reactor building where the SFP is located. The FHA analysis for Oyster Creek shows that following 60 days of decay time after reactor shutdown, the dose consequences from an FHA are acceptable without certain systems operable during and following the event, provided that 23 feet of water is maintained above the irradiated fuel assemblies in the SFP.

The NRG staff evaluated the radiological consequences of the postulated FHA OBA against the dose criteria specified in 10 CFR 50.67, "Accident source term," and using the guidance described in RG 1.183. The 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.

By letter dated April 26, 2007 (ADAMS Accession No. ML071080019), the NRC issued Amendment No. 262 to Oyster Creek. This amendment revised the Oyster Creek licensing basis in the area of radiological dose analyses for DBAs using the AST depicted in RG 1.183.

Additionally, this amendment revised the Oyster Creek TSs consistent with the amended design basis.

The FHA-specific dose acceptance criteria are specified in NUREG-0800, "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," Revision 0, July 2000 (ADAMS Accession No. ML003734190).

The dose acceptance criteria for the FHA are a total effective dose equivalent (TEDE) of 6.3 roentgen equivalent man (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 (CR) for the duration of the accident.

The regulations under 10 CFR 50.67 state, in part, 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) 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 [GDC] for Nuclear Power Plants,"

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. To modify their emergency plans to reflect fhe risk commensurate with power reactors that have been permanently shut down, power reactor licensees transitioning to decommissioning must seek exemptions from certain emergency planning regulatory requirements before amending these plans. The regulations under 10 CFR 50.12 provide that the NRC may, upon application by a licensee or upon its own initiative, grant exemptions from the requirements of the regulations, which are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security and when special circumstances are present, including circumstances in which application of the regulation would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule. 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, "Development of Emergency Action Levels for Non-Passive Reactors," Revision 6 (ADAMS Accession No. ML12326A805), provides guidance for the development of emergency action levels (EALs) for reactors in a permanently defueled condition. The NEI 99-01 topical report 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 EAL scheme must confirm that the source terms and release motive forces are not sufficient to warrant classification of a site area emergency (SAE) or general emergency. An SAE 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 Guides (PAGs). The EPA PAG for sheltering or evacuation of the public is a projected dose of one to five rem total effective dose (TED 1 ) in 4 days. In addition, the EPA PAG for recommending the administration of potassium iodide (Kl) (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 SAE classification threshold as 100 millirem (mrem) TEDE or 500 mrem thyroid committed dose equivalent.

The RG 1.183 provides the methodology for analyzing the radiological consequences of several DBAs to show compliance with 10 CFR 50.67. The 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.

The SRP Section 15.0.1 provides review guidance to the 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 CR for the duration of the accident.

1 For the purposes of this safety evaluation, the terms "TED" and "TEDE" are used interchangeably as both describing the combined effects of internal and external radiation exposure.

Regulatory Issue Summary (RIS) 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.

4.0 TECHNICAL EVALUATION

4.1 Accident Analysis During normal power reactor operations, the forced inlet flow of water through the RCS removes the heat from the reactor by generating steam. The steam system, operating at high temperatures and pressures, transfers this heat to the turbine generator. The most severe postulated accidents for nuclear power plants involve damage to the nuclear reactor core and the release of large quantities of fission products to the RCS. Section 15 of the Oyster Creek UFSAR describes the OBA scenarios that are applicable to Oyster Creek. Many of the accident scenarios postulated in the UFSAR involve failures or malfunctions of systems, which could affect the reactor core. With the permanent cessation of reactor operations at Oyster Creek and the permanent removal of the fuel from the reactor vessel, most of the DBAs postulated in the UFSAR will no longer be possible. The irradiated fuel will be stored in the SFP and the independent spent fuel storage installation. The reactor, RCS, steam system, and turbine generator are no longer in operation and have no function related to the storage of the irradiated fuel. Therefore, the postulated accidents involving failure or malfunction of the reactor, RCS, steam system, or turbine generator are no longer applicable.

The licensee has stated, and the NRC staff agrees that while spent fuel remains in the SFP, the only remaining OBA at Oyster Creek accident is an FHA that takes place in the SFP located in the Reactor Building. For completeness, the NRC staff also evaluated the applicability of the other DBAs documented in the Oyster Creek UFSAR to ensure that those accidents would not have consequences that could potentially exceed the 10 CFR 50.67 dose limits and RG 1.183 dose acceptance criterion or approach the EPA PAG criteria of 1 rem total effective dose. 2 These accidents include a Postulated Radioactive Tank Failure and Release of Radioactive Liquid Waste while radioactive liquids are still present. As discussed in UFSAR Section 15. 7 .2, "Radioactive Liquid Waste System Leak or Failure," and Section 15.7.3, "Postulated Radioactive 2 Use of EPA PAGs as a threshold is consistent with the planning basis for the 10-mile EPZ provided in NUREG-0396 (EPA 520/1-78-016), "Planning Basis for the Development of State and Local Government Radiological Emergency Response Plans in Support of Light Water Nuclear Power Plants," and endorsed by the Commission in a policy statement published on October 23, 1979 (44 FR 61123).

Releases Due to Liquid Tank Failure," the analysis for the liquid release, assuming failure of all liquid radwaste equipment, would result in a computed dose due to noble gases not to exceed 500 mrem at the site boundary. These analyses remain valid for Oyster Creek in a permanently shut down and defueled condition and therefore will not be further addressed.

In addition, the licensee considered a BOBE scenario to evaluate the effects of a loss of SFP water inventory resulting in radiation exposure at the EAB, LPZ, and the CR. The purpose of this evaluation was to determine the offsite radiological impact of a complete loss of SFP water.

4.2 Fuel Handling Accident After the permanent cessation of operations and permanent removal of fuel from the reactor vessel, an FHA onto the top of the core (or elsewhere within containment) is no longer possible and, therefore, no longer part of the licensing basis. However, an FHA in the SFP (which is located in the Reactor Building) is still possible at Oyster Creek, as long as spent fuel is stored in the SFP.

The OBA FHA in the Reactor Building is applicable when Oyster Creek is in a permanently shut down and defueled condition. The licensee's analysis was performed to determine the dose to operators in the CR and the public at the EAB or "site boundary" as a function of time after shutdown. The analysis demonstrates that radiological doses at the EAB, LPZ, and the CR are within allowable limits of 10 CFR 50.67, without crediting secondary containment operability, standby gas treatment system, or CR high efficiency air filtration after a 60-day fuel decay period following permanent reactor shutdown. The analysis shows that the dose at the EAB 33 days after shutdown (with no credit for containment) is less than 1 rem TEDE.

In performing its review, the NRC staff relied upon information provided by the licensee and staff experience in performing similar reviews. The NRC staff concludes that the dose consequence from the OBA FHA for Oyster Creek in the permanently shutdown and defueled condition meets the applicable radiological dose criteria at the EAB, LPZ, and the CR, would not approach the EPA PAGs, and would not cause a declaration of an SAE after a 60-day fuel decay period following permanent reactor shutdown.

4.3 Spent Fuel Cask Drop Accident Section 15. 7.5, "Spent Fuel Cask Drop Accident," and Section 9.1.2.2.3, "Cask Drop Protection System," of the Oyster Creek UFSAR discuss the potential for a spent fuel cask drop accident in the SFP located in the reactor building. Section 9.1.2.2.3 states, in part, that:

Amendment Number 223 to Oyster Creek's Facility Operating License Number DPR-16, as accepted by the Nuclear Regulatory Commission, eliminates Technical Specifications 5.3.1.B and 5.3.1.C. These Technical Specifications restricted the movement of heavy loads over irradiated fuel stored in the Spent Fuel Pool. Elimination of these Technical Specifications also eliminates the requirements for the design functions of the Cask Drop Protection System. The justification for eliminating these Technical Specifications and the Cask Drop Protection System is that the Reactor Building Crane has been upgraded to be single-failure-proof as defined by NUREG-0612.

The NRC staff concludes that due to the fact that the Oyster Creek reaclor building crane is licensed as being single-failure-proof, a spent fuel cask drop accident is not considered a credible accident for Oyster Creek.

4.4 Liquid Tank Rupture Section 15. 7.3, "Postulated Radioactive Releases Due to Liquid Tank Rupture," of the Oyster Creek UFSAR describes the liquid tank rupture as the uncontrolled or unanticipated release of the radioactive liquids. Assuming average meteorology and assuming failure of all liquid radwaste equipment, the computed dose at the site boundary due to noble gases would not exceed 500 mrem. The dose consequence documented in the UFSAR analysis is a fraction of the EPA PAGs.

Therefore, the NRC staff concludes that the dose consequence from a liquid tank rupture for Oyster Creek in the permanently shutdown and defueled condition will not approach the EPA PAGs for sheltering or evacuation and would not trigger the declaration of an SAE.

4.5 Waste Liquid Incident Section 15. 7.2, "Radioactive Liquid Waste System Leak or Failure," of the Oyster Creek UFSAR explains that a waste liquid incident is considered to be any incident that results in the release of waste liquid, and its accompanying activity, to the environment. The Oyster Creek radioactive waste disposal system is designed such that any spillage or leakage of radioactive liquid waste would be retained within the facility. Assuming average meteorology and assuming failure of all liquid radwaste equipment, the computed dose at the site boundary due to noble gases would not exceed 500 mrem. The dose consequence documented in the UFSAR analysis is a fraction of the EPA PAGs.

Therefore, the NRC staff concludes that the dose consequence from a waste liquid incident for Oyster Creek in the permanently shutdown and defueled condition will not approach the EPA PAGs for sheltering or evacuation and would not trigger the declaration of an SAE.

4.6 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 NRC staff finds that the licensee's proposed changes use analysis methods and assumptions consistent with the guidance contained in RG 1.183. The NRC staff compared the doses estimated by the licensee to the applicable criteria and to the results of confirmatory analyses performed by the staff. The NRC staff finds that there is reasonable assurance that Oyster Creek, as modified by 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 NRC staff concludes that the licensee has demonstrated that the dose consequences for postulated accidents at Oyster Creek in the permanently shutdown and defueled condition 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 criterion of 1 rem TED after a 60-day fuel decay period following permanent reactor shutdown. Therefore, the NRC staff finds the proposed changes to be acceptable from a dose consequence perspective.

4. 7 Proposed Changes to Renewed Facility Operating License

4. 7 .1 License Condition 1. B Currently, License Condition 1.B reads:

Construction of the Oyster Creek Nuclear Generating Station (Oyster Creek or the facility) has been completed in conformity with Provisional Construction Permit No. CPPR-15; the application, as amended; the provisions of the Act; and the rules and regulations of the Commission.

The licensee proposes to delete License Condition 1.B, because the decommissioning of Oyster Creek does not depend on the conformity with the Provisional Construction Permit No. CPPR-15. By letter dated June 21, 1968 (ADAMS Accession No. ML011140370), the Atomic Energy Commission issued an Order extending the latest completion date of the Provisional Construction Permit No. CPPR-15 to June 30, 1969. By letter dated December 24, 1968 (ADAMS Accession No. ML011140441 ), the Atomic Energy Commission transmitted to the licensee a Notice of Proposed Issuance of a Provisional Operating License. Therefore, the Provisional Construction Permit No. CPPR-15 was superseded by the Provisional Operating License DPR-16, which eventually became RFOL No. DPR-16, dated April 8, 2009 (ADAMS Accession No. ML080280440). The NRC staff finds it acceptable to delete License Condition 1.B.

4.7.2 License Condition 1.C Currently, License Condition 1.C reads:

Actions have been identified and have been or will be taken with respect to (1) managing the effects of aging during the term of this Renewed Facility Operating License No. DPR-16 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 the renewed operating license will continue to be conducted in accordance with the current licensing basis, as defined in 10 CFR 54.3, for the facility, and that any changes made to the facility's current licensing basis in order to comply with 10 CFR 54.29(a) are in accordance with the Act and the Commission's regulations; The licensee originally proposed to delete this license condition. However, by letter dated March 29, 2018, Exelon withdrew the request to delete License Condition 1.C; therefore, License Condition 1.C will remain as written.

4.7.3 License Condition 1.D Currently, License Condition 1.D reads:

The facility will operate in conformity with the application, as amended; the provisions of the Act; and the rules and regulations of the Commission (except as exempted from compliance in Section 2.D. below);

The licensee proposes License Condition 1.D 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 Oyster Creek license no longer authorizes 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.

4.7.4 License Condition 1.E Currently, License Condition 1.E reads:

There is reasonable assurance (i) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public and (ii) that such activities will be conducted in compliance with the Commission's rules and regulations set forth in 10 CFR Chapter I (except as exempted from compliance in Section 2.D. below);

The licensee proposes License Condition 1.E to read:

There is reasonable assurance (i) that the activities authorized by this 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 rules and regulations set forth in 10 CFR Chapter I; Consistent with 10 CFR Part 50.82(a)(2), the Oyster Creek license no longer authorizes 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 1.E.

4.7.5 License Condition 2.B.(1)

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

Pursuant to Section 104b of the Act and 10 CFR Part 50, to possess, use, and operate Oyster Creek Nuclear Generation Station at the designated location on the Oyster Creek site in Ocean County, New Jersey, in accordance with the procedures and limitations set forth in this renewed license; The licensee proposes License Condition 2.B.(1) to read:

Pursuant to Section 104b of the Act and 10 CFR Part 50, to possess and use Oyster Creek Nuclear Generation Station at the designated location on the Oyster Creek site in Ocean County, New Jersey, in accordance with the procedures and limitations set forth in this renewed license;

Consistent with 10 CFR Part 50.82(a)(2), the Oyster Creek license no longer authorizes 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 ).

4.7.6 License Condition 2.B.(2)

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

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 Updated Final Safety Analysis Report, as supplemented and amended; The licensee proposes License Condition 2.B.(2) to read:

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 Updated Final 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 Oyster Creek no longer authorizes operation of the reactor. As such, Oyster Creek 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 Oyster Creek possesses reactor fuel that was used for past operations. Therefore, the NRC staff approves the proposed change to License Condition 2.B.(2).

4.7.7 License Condition 2.8.(3)

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

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, or special nuclear materials 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 proposes License Condition 2.B.(3) to read:

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use at any time any byproduct, source, or special nuclear materials as sealed neutron sources that were used for reactor startup, sealed sources that were

used for calibration of reactor instrumentation and are used in radiation monitoring equipment, and as fission detectors in amounts as required; The proposed change to this license condition removes the authorization for receipt and use of byproduct, source, and SNM as sealed neutron sources for reactor startup; but retains authorization to possess such sources previously used for reactor startup. The deletion of the authorization to receive and use sources for reactor startup is consistent with the fact that Oyster Creek will no longer be authorized to operate and the continued authorization to possess neutron sources that were used for reactor startup is consistent with the safe storage of byproduct, source, and SNM. The use of sources for radiation monitoring will continue to be required. Since the Oyster Creek 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 the decommissioning plant. Therefore, the NRC staff approves the proposed change to License Condition 2.8.(3).

4.7.8 License Condition 2.8.(5)

Currently, License Condition 2.8.(5) reads as follows:

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate such byproduct, source, or special nuclear materials as may be produced by the operation of the facility.

The licensee proposes License Condition 2.8.(5) to read:

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate such byproduct, source, or 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, source, 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 Oyster Creek 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 the decommissioning plant.

Therefore, the NRC staff finds the proposed change to License Condition 2.8.(5) acceptable.

4.7.9 License Condition 2.C.(1)

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

Maximum Power Level Exelon Generation Company is authorized to operate the facility at steady-state power levels not in excess of 1930 megawatts (thermal) ( 100 percent rated power) in accordance with the conditions specified herein.

The licensee proposes to delete License Condition 2.C.(1 ). Since Exelon docketed the Oyster Creek 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 Oyster Creek 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.

4.7.10 License Condition 2.C.(2)

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

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 292, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

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

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 295, are hereby incorporated in the license. Exelon Generation Company shall maintain the facility in accordance with the Permanently Oefueled Technical Specifications (POTS).

Consistent with 10 CFR Part 50.82(a)(2), the Oyster Creek license no longer authorizes 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 POTS. The license condition is changed from "operate the facility" to "maintain the facility," which describes the permanently defueled condition in which the Oyster Creek 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.

4.7.11 License Condition 2.C.(3)

Currently, License Condition 2.C.{3) reads:

Fire Protection Exelon Generation Company shall implement and maintain in effect all provisions of the approved fire protection program as described in the Updated Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Report dated March 3, 1978, and supplements thereto, subject to the following provision:

The licensee 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 proposes to delete License Condition 2.C.(3). Since Exelon docketed the certifications for permanent cessation of operations and permanent removal of fuel from the

reactor vessel, the fire protection program will be revised to take into account the facility conditions and activities during decommissioning. Oyster Creek 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 a fire protection program 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 Oyster Creek. However, many of the elements that are applicable for the operating plant fire protection program continue to be applicable during facility decommissioning. During the decommissioning process, a fire protection program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard.

The regulation is applicable regardless of whether a requirement for a fire protection program is included in the facility license. Therefore, a license condition requiring such a program for a permanently shut down and defueled facility is not necessary.

The NRC staff finds that License Condition 2.C.(3) for Oyster Creek is based on maintaining fire protection programs 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 fire protection programs at Oyster Creek once the facility is permanently shutdown and the fuel has been permanently removed from the reactor. However, elements of the fire protection program (e.g., License Condition 2.C.(8), Mitigating Strategy) continue during decommissioning to address fire events that could result in radiological hazards. The regulation in 10 CFR 50.48(f) requires Oyster Creek to address the potential for fires, which could result in a radiological hazard. The NRC staff concludes that the rule, which requires a fire protection program 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 fire protection programs 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 licensee's request to delete License Condition 2.C.(3) is acceptable.

4.7.12 License Condition 2.C.(5)

Currently, License Condition 2.C.(5) states:

Inspections of core spray spargers, piping and associated components will be performed in accordance with BWRVIP-18, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines," as approved by NRC staff's Final Safety Evaluation Report dated December 2, 1999.

The licensee proposes to delete License Condition 2.C.(5). Since Exelon docketed the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, inspections of core spray spargers, piping, and associated components will no longer be required. The core spray system will no longer be required to provide cooling to the reactor in the event of the design-basis loss-of-coolant accident (LOCA). Therefore, the NRC staff finds the deletion of License Condition 2.C.(5) acceptable.

4.7.13 License Condition 2.C.(6)

Currently, License Condition 2.C.(6) states:

Long Range Planning Program - Deleted The licensee proposes to reformat this license condition by removing the title so that the condition just states "Deleted." This license condition was previously deleted in License Amendment No. 244, dated July 13, 2004 (ADAMS Accession No. ML041560041 ). The NRC staff finds the reformatting of this license condition administrative in nature and, therefore, acceptable.

4.7.14 License Condition 2.C.(7)

Currently, License Condition 2.C.(7) states:

Reactor Vessel Integrated Surveillance Program Exelon Generation Company is authorized to revise the Updated Final Safety Analysis Report (UFSAR) to allow implementation of the Boiling Water Reactor Vessel and Internals Project reactor pressure vessel Integrated Surveillance Program as the basis for demonstrating compliance with the requirements of Appendix H to Title 10 of the Code of Federal Regulations Part 50, "Reactor Vessel Material Surveillance Program Requirements," as set forth in the licensee's application dated December 20, 2002, and as supplemented on May 30, September 10, and November 3, 2003.

All capsules in the reactor vessel that are removed and tested must meet the test procedures and reporting requirements of the most recent NRC-approved version of the Boiling Water Reactor Vessel and Internals Project Integrated Surveillance Program appropriate for the configuration of the specimens in the capsule. Any changes to the capsule withdrawal schedule, including spare capsules, must be approved by the NRC prior to implementation. All capsules placed in storage must be maintained for future insertion. Any changes to storage requirements must be approved by the NRC, as required by 10 CFR Part 50, Appendix H.

The licensee proposes to delete License Condition 2.C.(7). This license condition was issued concurrent with the RFOL on April 8, 2009. This license condition is described in Section 1.7, "Summary of Proposed License Conditions," of NUREG-1875, "Safety Evaluation Report Related to the License Renewal of Oyster Creek Generating Station," issued April 2007 (Volumes 1 and 2 available at ADAMS Accession Nos. ML071290023 and ML071310246, respectively).

The regulation 10 CFR Part 50 Appendix H requires that the design of the reactor vessel surveillance capsule program and withdrawal schedules meet the requirements in the version of American Society for Testing and Materials (ASTM} Standard Practice E 185 that is current on the issue date of the American Society of Mechanical Engineers (ASME) Code to which the RPV was purchased. The rule also requires the licensee to perform capsule testing and to report the test results in accordance with the requirements of ASTM E 185-82 to the extent practicable for the configuration of the test specimen in the RPV surveillance capsules.

The requirements in 10 CFR Part 50, Appendix Hare only relevant to nuclear power plants that are authorized to operate in the reactor-critical operating mode because this is the plant operating mode that produces high energy neutrons as a result of the reactor's nuclear fission process, and the requirements are set in place to provide assurance that the RPV will maintain adequate levels of fracture toughness throughout the operating life of the reactor. This license condition was imposed with the assumption that Oyster Creek would be operating for an additional 20 years (i.e., to and inclusive of April 9, 2029) and would not be proposing to end power operations of the facility prior to that date.

This license condition is proposed for deletion in its entirety to reflect the permanently defueled condition of the facility once the certification of permanent removal of fuel from the reactor vessel is submitted. Continued implementation of the applicable surveillance capsule testing and reporting requirements is no longer necessary for Oyster Creek because Exelon has decided to permanently cease power operations at Oyster Creek, and from a fracture toughness perspective, the Oyster Creek RPV will cease to be exposed to further irradiation by high energy neutrons or subjected to any high thermal stress environments. Further, 10 CFR 50.60(a) stipulates that reactor facilities for which the certifications required under 10 CFR 50.82(a)(1) have been submitted, no longer need to meet the fracture toughness and material surveillance program requirements for the reactor coolant pressure boundary set forth in 10 CFR Part 50, Appendices G and H.

The physical and radiological control of the remaining surveillance capsules that are located in the RPV will be managed in accordance with the applicable radiological control requirements of 10 CFR Part 20 and with any applicable security or physical protection requirements for components in either 10 CFR Part 37 or 10 CFR Part 73. Therefore, the removal, testing, reporting, and storage requirements for reactor vessel surveillance capsules and their test specimens do not need to be implemented further once Oyster Creek permanently ceases power operations. There will no longer be any need to remove the remaining surveillance capsules from the RPV or perform material testing of the test specimens in those capsules.

Based on its review of the proposed deletion, the NRC staff concludes that continued implementation of the applicable surveillance capsule testing and reporting requirements is no longer necessary for Oyster Creek because power operation is no longer authorized at Oyster Creek since the 10 CFR 50.82(a)(1) certifications have been docketed, and from a fracture toughness perspective, the Oyster Creek RPV will cease to be exposed to further irradiation by high energy neutrons or subjected to any high thermal stress environments. The removal, testing, reporting, and storage requirements for reactor vessel surveillance capsules and their test specimens do not need to be implemented further since Oyster Creek permanently ceased power operations on September 17, 2018. Therefore, the NRC staff finds the deletion of License Condition 2.C.(7) acceptable.

4.7.15 License Condition 2.C.(10)

Currently, License Condition 2.C.(10) states:

Upon implementation of Amendment No. 265 adopting TSTF-448, Revision 3, the assessment of CRE [control room envelope] habitability as required by

Specification 6.22.c.(ii), and the measurement of CRE pressure as required by Specification 6.22.d, shall be considered met. Following implementation:

(a) The first performance of the periodic assessment of CRE habitability, Specification 6.22.c.(ii), shall be within 3 years, plus the 9-month allowance of Specification 1.24.

{b) The first performance of the periodic measurement of CRE pressure, Specification 6.22.d, shall be within 24 months, plus the 180 days allowed by Specification 1.24, as measured from the date -of the most recent successful pressure measurement test, or within 180 days if not performed previously.

The licensee proposes to delete License Condition 2.C.(10). The proposed change would remove the requirements of TSTF-448 that involve 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 New Power Reactors," May 2003 (ADAMS Accession No. ML031490664). These assessments were completed in accordance with the schedule specified in the license condition.

Exelon analyzed the FHA for dose results for the CR after permanent shutdown. The analysis accounts for radioactive material inventory in the most recently irradiated elements in the SFP after 60 days of decay. For the analysis, Exelon took no credit for CR isolation or filtered recirculation of the CR air. The results of the calculation showed that the dose consequences to occupants in the CR were below acceptable limits. The dose at the CR would be 2.235 rem, which is less than the 10 CFR 50.67 dose limit of 5 rem. Based on the fact that the dose at the CR is less than the 10 CFR 50.67 dose limit and that no credit was taken for CR isolation or filtered recirculation, the CRE habitability program is not required to provide airborne radiological protection for the CR operators.

The NRC staff reviewed the calculation results. After 60 days of decay and an FHA, the dose at the CR is 2.235 rem, with no credit for CR isolation or filtered recirculation of the CR air. This is less than the 10 CFR 50.67 dose limit of 5 rem. Therefore, the NRC staff finds the deletion of License Condition 2.C.(10) acceptable.

4.7.16 License Condition 2.C.(11)

Currently, License Condition 2.C.(11) states:

Inspection of Drywell Sand Bed Region The licensee shall perform full scope inspections (as defined in Appendix A of the license renewal safety evaluation report dated March 20, 2007, and summarized in the Updated Final Safety Analysis Report (UFSAR)) of the drywell sand bed region every other refueling outage beginning in the refueling outage prior to April 9, 2009.

The licensee proposes to delete License Condition 2.C.(11 ). This license condition was issued concurrent with the RFOL on April 8, 2009. This license condition is described in Section 1. 7 of NUREG-1875.

Since Exelon docketed the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, the inspection of drywall sand bed region will no longer be required. Refueling outages will no longer occur nor will Oyster Creek operate during the remaining period of extended operation (ending April 9, 2029), and these activities that are unique to the renewed license are not necessary. The decommissioning of Oyster Creek is not dependent on the requirements of 10 CFR Part 54 for a renewed license.

The NRC staff reviewed the proposed deletion of License Condition 2.C.(11 ). As noted above, power operation is no longer authorized at Oyster Creek since the licensee's 10 CFR 50.82(a)(1) certifications have been docketed. Therefore, the NRC staff finds the deletion of License Condition 2.C.(11) acceptable.

4.7.17 License Condition 2.C.(12)

Currently, License Condition 2.C.(12) states:

Drywall Trenches The licensee shall monitor the drywall trenches (as defined in Appendix A of the license renewal safety evaluation report dated March 20, 2007, and summarized in the UFSAR) every refueling outage to identify and eliminate the sources of water and shall receive NRC approval prior to restoring the trenches to their original design configuration.

The licensee proposes to delete License Condition 2.C.(12). This license condition was issued concurrent with the RFOL on April 8, 2009. This license condition is described in Section 1. 7 of NUREG-1875. Since the licensee docketed the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, monitoring the two trenches for the presence of water during refueling outages will no longer be necessary. There will no longer be refueling outages nor a need for the drywall shell (i.e., primary containment). The decommissioning of Oyster Creek is not dependent on the requirements of 10 CFR Part 54 for a renewed license.

The NRC staff reviewed the proposed deletion of License Condition 2.C.(12). As noted above, power operation is no longer authorized at Oyster Creek since the licensee's 10 CFR 50.82(a)(1) certifications have been docketed. Therefore, the NRC.stafffinds the deletion of License Condition 2.C.(12) acceptable.

4.7.18 License Condition 2.C.(13)

Currently, License Condition 2.C.(13) states:

Engineering Study of Refueling Cavity Liner The licensee shall perform an engineering study prior to April 9, 2009 to identify options to eliminate or reduce the leakage in the facility cavity liner.

The licensee proposes to delete License Condition 2.C.(13). This license condition was issued concurrent with the RFOL on April 8, 2009. This license condition is described in Section 1. 7 of NUREG-1875. This license condition is a one-time requirement that has been completed and is proposed for deletion in its entirety. On March 27, 2009, the NRC staff completed a license

renewal follow-up inspection at Oyster Creek (ADAMS Accession No. ML091380379). The NRC staff did not identify any significant problems or concerns. Based on the conclusion of the NRC's review, this license condition has been completed in its entirety and may be eliminated.

Therefore, the NRC staff finds the deletion of License Condition 2. C.( 13) acceptable.

4.7.19 License Condition 2.C.(14)

Currently, License Condition 2.C.(14) states:

Three-Dimensional Finite-Element Analysis of Drywell Shell The licensee shall perform a three-dimensional finite-element analysis of the drywell shell and shall provide to the NRC staff a report of the results prior to April 9, 2009.

The licensee proposes to delete License Condition 2.C.(14). This license condition was issued concurrent with the RFOL on April 8, 2009. This license condition is described in Section 1. 7 of NUREG-1875. This license condition is a one-time requirement that has been completed and is proposed for deletion in its entirety. On March 27, 2009, the NRC staff completed a license renewal follow-up inspection at Oyster Creek. The NRC staff did not identify any significant problems or concerns. Based on the conclusion of the NRC's review, this license condition has been completed in its entirety and may be eliminated. Therefore, the NRC staff finds the deletion of License Condition 2.C.(14) acceptable.

4.7.20 License Condition 2.C.(15)

Currently, License Condition 2.C.(15) states:

UFSAR Supplement Changes The UFSAR supplement, as revised, submitted pursuant to 10 CFR 54.21(d),

shall be included in the next scheduled update to the UFSAR required by 10 CFR 50.71(e)(4), as modified by an exemption granted by letter dated July 7, 2004 (ADAMS Accession No. ML041340673), following the issuance of this renewed operating license. Until that update is complete, Exelon Generation Company may make changes to the programs and activities described in the supplement without prior Commission approval, provided that Exelon Generation Company evaluates such changes pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

The licensee proposes to delete License Condition 2.C.(15). This license condition was issued concurrent with the RFOL on April 8, 2009. This license condition is described in Section 1. 7 of NUREG-1875. This license condition is a one-time requirement to update the UFSAR to include the UFSAR supplement required by 10 CFR 54.21(d) in the next UFSAR update as required by 10 CFR 50.71(e) and allows changes to be made to that supplement under the provisions of 10 CFR 50.59 until the UFSAR update is completed. Oyster Creek UFSAR, Revision 16, which included the supplement (Appendix A) for the License Renewal Application, was submitted to the NRC on December 23, 2009 (ADAMS Accession No. ML110691283). This action satisfied the requirements of Oyster Creek License Condition 2.C.(15); therefore, the NRC staff finds the deletion of License Condition 2.C.(15) acceptable.

4.7.21 License Condition 2.D Currently, License Condition 2.D states:

The facility has been granted certain exemptions from the requirements of Section 111.G of Appendix R to 10 CFR Part 50, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979."

This section relates to fire protection features for ensuring the systems and associated circuits used to achieve and maintain safe shutdown are free of fire damage. These exemptions were granted and sent to the licensee in letters dated March 24, 1986 and June 25, 1990.

The facility has also been granted certain exemptions from the requirements of Section 111.J of Appendix R to 10 CFR Part 50, "Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979." This section relates to emergency lighting that shall be provided in all areas needed for operation of safe shutdown equipment and in access and egress routes thereto.

This exemption was granted and sent to the licensee in a letter dated February 12, 1990.

In addition, the facility has been granted certain exemptions from Section 55.45(b)(2)(iii) and (iv) of 10 CFR Part 55, "Operators' Licenses." These sections contain requirements related to site-specific simulator certification and require that operating tests will not be administered on other than a certified or an approved simulation facility after May 26, 1991. These exemptions were granted and sent to the licensee in a letter dated March 25, 1991.

These exemptions granted pursuant to 10 CFR 50.12 are authorized by law, will not present an undue risk to the public health and safety, and are consistent with the common defense and security. With these exemptions, the facility will operate, to the extent authorized herein, in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission.

The licensee proposes to delete License Condition 2.D. This license condition documents specific exemptions from 10 CFR Part 50 and 10 CFR Part 55, "Operators' Licenses," as approved by the NRC, specifically, exemptions from the requirements of 10 CFR Part 50, Appendix R, Sections 111.G and 111.J and of 10 CFR 55.45(b)(2)(iii) and (iv).

The requirements of 10 CFR Part 50, Appendix Rare required to mitigate the consequences of a OBA under post-fire conditions and to limit fire damage to systems required to achieve and maintain safe shutdown conditions. Since Exelon docketed the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, the requirements of 10 CFR Part 50, Appendix R will no longer apply. Therefore, Oyster Creek will no longer need these exemptions to 10 CFR Part 50, Appendix R. During the decommissioning process, a fire protection program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. The regulation is applicable regardless of whether a requirement for a fire protection program is included in the facility license. Therefore, a license condition requiring such a program for a permanently shutdown and defueled facility is not necessary. The fire protection program will be revised to take into account the facility

conditions and activities during decommissioning. Oyster Creek 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. Similarly, since Exelon docketed the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, the requirements of 10 CFR Part 55 relating to the Oyster Creek simulator will no longer apply. Therefore, the NRC staff finds the deletion of License Condition 2.D acceptable.

4.7.22 License Conditions 2.E and 3.A through 3.K Currently, License Conditions 2.E and 3.A through 3.K state:

Deleted The licensee proposes to change these license conditions from "Deleted" to "DELETED." This is an editorial change that does not make any technical changes. Therefore, the NRC staff finds the proposed changes to License Condition 2.E and 3.A through 3.K acceptable.

4.7.23 License Condition 3.M Currently, License Condition 3.M states:

At the time of the closing of the transfer of Oyster Creek, and the respective license from AmerGen Energy Company, LLC (AmerGen) to Exelon Generation Company, AmerGen shall transfer to Exelon Generation Company ownership and control of AmerGen Oyster Creek NQF, LLC, and AmerGen Consolidation, LLC shall be merged into Exelon Generation Consolidation, LLC. Also at the time of the closing, decommissioning funding assurance provided by Exelon Generation Company, using an additional method allowed under 10 CFR 50. 75 if necessary, must be equal to or greater than the minimum amount calculated on that date pursuant to, and required by 10 CFR 50. 75 for Oyster Creek.

Furthermore, funds dedicated for Oyster Creek prior to closing shall remain dedicated to Oyster Creek following the closing. The name of AmerGen Oyster Creek NQF, LLC shall be changed to Exelon Generation Oyster Creek NQF, LLC at the time of the closing.

The licensee proposes to delete License Condition 3.M. This license condition eliminates references to AmerGen Energy Company, LLC (AmerGen), and replaces them with references to Exelon Generation Company, LLC, to reflect the results of the license transfer. AmerGen transferred to Exelon ownership and control of AmerGen Oyster Creek NQF, LLC and AmerGen Consolidation, LLC merged into Exelon Generation Consolidation, LLC. On December 23, 2008, the NRC approved the transfer of the license and the ownership of Oyster Creek to Exelon (ADAMS Accession No. ML082750072). The name of AmerGen Oyster Creek NQF, LLC was changed to Exelon Generation Oyster Creek NQF, LLC at the time of the closing. In a letter dated March 31, 2009, Exelon reported to the NRC that the decommissioning trust agreements for Oyster Creek had been modified to reflect the change in ownership from AmerGen to Exelon (ADAMS Accession No. ML090900463). The requirements of this license condition have been completed; therefore, the NRC staff finds the deletion of License Condition 3.M acceptable.

4. 7.24 License Condition 4 Currently, License Condition 4 states:

This license is effective as of the date of issuance and shall expire at midnight on April 9, 2029.

The licensee proposes License Condition 4 to read:

This license is effective as of the date of issuance and is effective 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 Oyster Creek RFOL no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel. The proposed change would revise License Condition 4 to conform with 10 CFR 50.51, "Continuation of license," in that the license authorizes ownership and possession by Exelon until the Commission notifies the licensee in writing that the license is terminated.

The NRC staff reviewed the proposed change to License Condition 4. The current License Condition 4, which documents the date of the expiration of the RFOL, is no longer necessary for the permanently shutdown and defueled condition of the plant in the process of decommissioning. The revised License Condition 4 documents the current condition of the plant 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 4 acceptable.

4.8 Proposed TS Changes - TS Section 1 - Definitions 4.8.1 Definitions Proposed for Deletion The licensee proposes deleting the following definitions from TS 1.0, because they pertain to an operating reactor. Once Oyster Creek is permanently shut down and defueled, the definitions no longer apply:

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

A verification of operability is an administrative check, by examination of appropriate plant records (logs, surveillance test records) to determine that a system, subsystem, train, component or device is not inoperable.

Such verification does not preclude the demonstration (testing) of a given system, subsystem, train, component or device to determine operability.

1.2 OPERATING Operating means that a system or component is performing its required function.

1.3 POWER OPERATION Power operation is any operation when the reactor is in the startup mode or run mode except when primary containment integrity is not required.

1.4 STARTUP MODE The reactor is in the startup mode when the reactor mode switch is in the startup mode position. In this mode, the reactor protection system scram trips initiated by condenser low vacuum and main steam line isolation valve closure are bypassed when reactor pressure is less than 600 psig; the low pressure main steam line isolation valve closure is bypassed; the IRM trips for rod block and scram are operable; and the SRM trips for rod block are operable.

1.5 RUN MODE The reactor is in the run mode when the reactor mode switch is in the run mode position. In this mode, the reactor protection system is energized with APRM protection and the control rod withdrawal interlocks are in service.

1.6 SHUTDOWN CONDITION The reactor is in the SHUTDOWN CONDITION when there is fuel in the reactor vessel, the reactor is subcritical, all operable control rods are fully inserted, and the mode switch is in the shutdown mode position. In this position, a control rod block is initiated.

1. 7 COLD SHUTDOWN CONDITION The reactor is in the COLD SHUTDOWN CONDITION when the reactor is in the SHUTDOWN CONDITION, and (except during REACTOR VESSEL PRESSURE TESTING), the reactor coolant system is maintained at less than 212°F and vented.

1.8 PLACE IN SHUTDOWN CONDITION Proceed with and maintain an uninterrupted normal plant shutdown operation until the SHUTDOWN CONDITION is met.

1.9 PLACE IN COLD SHUTDOWN CONDITION Proceed with and maintain an uninterrupted normal plant shutdown operation until the COLD SHUTDOWN CONDITION is met.

1.10 PLACE IN ISOLATED CONDITION Proceed with and maintain an uninterrupted normal isolation of the reactor from the turbine condenser system including closure of the main steam isolation valves.

1.11 REFUEL MODE The reactor is in the REFUEL MODE when the reactor mode switch is in the REFUEL MODE position and there is fuel in the reactor vessel. In this mode the refueling platform interlocks are in operation.

1.12 REFUELING OUTAGE For the purpose of designating frequency of testing and surveillance, a REFUELING OUTAGE shall mean a regularly scheduled REFUELING OUTAGE. Following the first REFUELING OUTAGE, successive tests or surveillances shall be performed at least once per 24 months.

1.13 PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY means that the drywell and adsorption chamber are closed and all of the following conditions are satisfied:

A. All non-automatic primary containment isolation valves which are not required to be open for plant operation are closed.

B. At least one door in the airlock is closed and sealed.

C. All automatic primary containment isolation valves are OPERABLE or the affected penetration is isolated.

D. All blind flanges and manways are closed.

1.14 SECONDARY CONTAINMENT INTEGRITY Secondary containment integrity means that the reactor building is closed and the following conditions are met:

A. At least one door at each access opening is closed. (Note:

Momentary opening and closing of the trunnion room door does not constitute a loss of secondary containment integrity. In COLD SHUTDOWN CONDITION or REFUEL MODE, the trunnion room door may remain open provided the trunnion room is isolated from the secondary containment through the reactor building walls, penetrations and either the inboard or outboard valves to the main steam and feedwater piping being secured in the closed position.)

B. The standby gas treatment system is operable.

C. All automatic secondary containment isolation valves are operable or are secured in the closed position.

1.15 (DELETED) 1.16 RATED FLUX Rated flux is the neutron flux that corresponds to a steady state power level of 1930 MW(t). Use of the term 100 percent also refers to the 1930 thermal megawatt power level.

1.17 REACTOR THERMAL POWER-TO-WATER Reactor thermal power-to-water is the sum of ( 1) the instantaneous integral over the entire fuel clad outer surface of the product of heat transfer area increment and position dependent heat flux and (2) the instantaneous rate of energy deposition by neutron and gamma reactions in all the water and core components except fuel rods in the cylindrical volume defined by the active core height and the inner surface of the core shroud.

1.18 PROTECTIVE INSTRUMENTATION LOGIC DEFINITIONS A. Instrument Channel An instrument channel means an arrangement of a sensor and auxiliary equipment required to generate and transmit to a trip system a single trip signal related to the plant parameter monitored by that instrument channel.

B. Trip System A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective trip function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action. Initiation of protective action may require the tripping of a single trip system (e.g., initiation of a core spray loop, automatic depressurization, isolation of an isolation condenser, offgas system isolation, reactor building isolation, standby gas treatment and rod block) or the coincident tripping of two trip systems (e.g., initiation of scram, isolation condenser, reactor isolation, and primary containment isolation).

1.19 INSTRUMENT SURVEILLANCE DEFINITIONS A. 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.

B. CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY of all devices in the channel required for channel OPERABILITY. The CHANNEL FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total channel steps.

C. CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass all devices in the channel required for channel OPERABILITY and the CHANNEL FUNCTIONAL TEST.

Calibration of instrument channels with resistance temperature detector (RTD} or thermocouple sensors may consist of an in place qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel.

The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps.

D. Source Check A SOURCE CHECK is the qualitative assessment of channel response when the channel sensor is exposed to a source of radioactivity.

1.20 FDSAR Oyster Creek Unit No. 1 Facility Description and Safety Analysis Report as amended by revised pages and figure changes contained in Amendments 14, 31 and 45* and continuing through Amendment 79.

• Per Erata dtd. 4-9-69 1.21 COREALTERATION A core alteration is the addition, removal, relocation or other manual movement of fuel or controls in the reactor core. Control rod movement with the control rod drive hydraulic system is not defined as a core alteration.

1.22 CRITICAL POWER RATIO The critical power ratio is the ratio of that power in a fuel assembly which is calculated, by application of an NRC approved CPR correlation, to cause some point in that assembly to experience boiling transition divided by the actual assembly operating power.

1.23 (DELETED) 1.25 APPENDIX J TEST PRESSURE For the purpose of conducting leak rate tests to meet 10 CFR 50 Appendix J, Pa = 35 psig.

1.26 FRACTION OF LIMITING POWER DENSITY (FLPD)

The fraction of limiting power density is the ratio of the linear heat generation rate (LHGR) existing at a given location to the design LHGR for that bundle type.

1.27 MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD)

The maximum fraction of limiting power density is the highest value existing in the core of the fraction of limiting power density (FLPD).

1.28 FRACTION OF RATED POWER (FRP)

The FRACTION OF RATED POWER is the ratio of core THERMAL POWER to RATED THERMAL POWER.

1.29 TOP OF ACTIVE FUEL (TAF) 353.3 inches above vessel zero.

1.30 REPORTABLE EVENT A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.

1.31 IDENTIFIED LEAKAGE IDENTIFIED LEAKAGE is that leakage which is collected in the primary containment equipment drain tank and eventually transferred to radwaste for processing.

1.32 UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE is all measured leakage that is other than identified leakage.

1.33 PROCESS CONTROL PLAN The PROCESS CONTROL PLAN shall contain the current formulas, sampling, analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR Parts 20, 61 and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

1.34 AUGMENTED OFFGAS SYSTEM (AOG)

The AUGMENTED OFFGAS SYSTEM is a system designed and installed to holdup and/or process radioactive gases from the main condenser offgas system for the purpose of reducing the radioactive material content of the gases before release to the environs.

1.35 MEMBER OF THE PUBLIC A MEMBER OF THE PUBLIC is a person who is not occupationally associated with Exelon Generation Company, LLC and who does not normally frequent the Oyster Creek Nuclear Generating Station site. The category does not include contractors, contractor employees, vendors, or persons who enter the site to make deliveries, to service equipment, work on the site, or for other purposes associated with plant functions.

1.37 PURGE PURGE OR PURGING is the controlled process of discharging air or gas from a confinement and replacing it with air or gas.

1.38 SITE BOUNDARY The SITE BOUNDARY is the perimeter line around the Oyster Creek beyond which the land is neither owned, leased nor otherwise subject to control by Exelon Generation Company, LLC (ref. ODCM). The area outside the SITE BOUNDARY is termed OFFSITE or UNRESTRICTED AREA.

1.39 REACTOR VESSEL PRESSURE TESTING System pressure testing required by ASME Code Section XI, Article IWA-5000, including system leakage and hydrostatic test, with reactor vessel completely water solid, core not critical and section 3.2.A satisfied.

1.40 SUBSTANTIVE CHANGES SUBSTANTIVE CHANGES are those which affect the activities associated with a document or the document's meaning or intent.

Example of non-substantive changes are: (1) correcting spelling, (2) adding (but not deleting) sign-off spaces, (3) blocking in notes, cautions, etc., (4) changes in corporate and personnel titles which do not reassign responsibilities and which are not referenced in the Appendix A Technical Specifications, and (5) changes in nomenclature or editorial changes which clearly do not change function, meaning or intent.

1.41 DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 microcuries per gram which alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 or Regulatory Guide 1.109, "Calculation of Annual Doses to Man from Routine Releases of Reactor Effluences for the Purpose of Evaluating Compliance with 10 CFR Par 40 Appendix I."

1.42 AVERAGE PLANAR LINEAR HEAT GENERATION RATE The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be applicable to a specific planar height and is equal to the sum of the heat generation rate per unit length of fuel rod for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle at that height.

1.43 CORE OPERATING LIMITS REPORT The Oyster Creek CORE OPERATING LIMITS REPORT (COLR) is the document that provides core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specification 6.9.1.f.

Plant operation within these operating limits is addressed in individual specifications.

1.44 LOCAL LINEAR HEAT GENERATION RATE The LOCAL LINEAR HEAT GENERATION RATE (LLHGR) shall be applicable to a specific planar height and is equal to the AVERAGE PLANAR LINEAR GENERATION RATE (APLHGR) at the specified height multiplied by the local peaking factor at that height.

1.45 SHUTDOWN MARGIN (SDM)

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical throughout the operating cycle assuming that:

a. The reactor is xenon free;

b. The moderator temperature is~ 68°F, corresponding to the most reactive state; and

c. All control rods are fully inserted except for the single control rod of highest reactivity worth, which is assumed to be fully withdrawn. With control rods not capable of being fully inserted, the reactivity worth of these control rods must be accounted for in the determination of SDM.

1.46 IDLE RECIRCULATION LOOP A recirculation loop is idle when its discharge valve is in the closed position and its discharge bypass valve and suction valve are in the open position.

1.47 ISOLATED RECIRCULATION LOOP A recirculation loop is fully isolated when the suction valve, discharge valve and discharge bypass valve are in the closed position.

1.48 OPERATIONAL CONDITION The reactor plant operational status as to criticality, reactor mode switch position, reactor coolant temperature, and/or specific system status.

These conditions consist of POWER OPERATION, STARTUP MODE, SHUTDOWN CONDITION, COLD SHUTDOWN CONDITION, and REFUEL MODE. A change or entry into an operating condition is Signified by movement of the reactor mode switch or a change in reactor coolant Temperature from <212°F to ~212°F.

1.49 RATED THERMAL POWER (RTP)

RTP shall be a total reactor core eat transfer rate to the reactor coolant of 1930 MWth.

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

1.51 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The PTLR is the unit specific document that provides the reactor vessel pressure and temperature limits, including heat up and cooldown rates, for the current reactor vessel fluence period. These pressure and temperature limits shall be determined for each fluence period in accordance with Specification 6.23.

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. Therefore, once Oyster Creek is permanently shut down and defueled, the NRC staff finds the deletion of these definitions from the TS acceptable.

4.8.2 Definitions Proposed for Relocation The current definition in TS Section 1.24 for SURVEILLANCE REQUIREMENTS is:

Surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within the safety limits, and that the limiting conditions of operation will be met. Each surveillance requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the surveillance interval.

Surveillance requirements for systems and components are applicable only during the modes of operation for which the system or components are required to be operable, unless otherwise stated in the specification.

This definition establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance, e.g., transient conditions or other ongoing surveillance or maintenance activities. It also provides flexibility to accommodate the length of a fuel cycle for surveillances that are performed at each refueling outage and are specified with a fuel cycle length surveillance interval. It is not intended that this provision be used repeatedly as a convenience to extend surveillance intervals beyond that specified for the surveillance that are not performed during refueling outages. The limitation of this definition is based on engineering judgement and the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

The current definition in TS Section 1.36 for OFFSITE DOSE CALCULATION MANUAL (ODCM) is:

The OFFSITE DOSE CALCULATION MANUAL shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluent, in the calculation of gaseous and liquid effluent monitoring Alarm/trip Setpoints, and in the conduct of the Environmental Radiological Monitoring Program. The ODCM shall also contain (1) the Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Section 6.8.4; and (2) descriptions of the information that should be included in the Annual Radioactive Effluent Release Report AND Annual Radiological Environmental Operating Report required by Specifications 6.9.1.d and 6.9.1.e, respectively.

The SURVEILLANCE REQUIREMENTS definition is proposed to be reformatted, revised, and relocated from TS Section 1.24 to TS 3/4.0, "Limiting Conditions for Operation and Surveillance Requirement Applicability," as SR 4.0.4. The proposed change would ensure that the appropriate requirements for the 25-percent grace period are maintained (see discussion of SR 4.0.4). The portion of the current SURVEILLANCE REQUIREMENTS definition with respect to modes is proposed to be deleted due to the elimination of the reactor modes.

The OFFSITE DOSE CALCULATION MANUAL definition is proposed to be relocated from TS Section 1.36 to the ODCM.

The NRC staff reviewed the proposed deletion of the TS definitions SURVEILLANCE REQUIREMENTS and OFFSITE DOSE CALCULATION MANUAL. The NRC staff finds this acceptable, because the SURVEILLANCE REQUIREMENTS definition is being relocated to SR 4.0.4, and the OFFSITE DOSE CALCULATION MANUAL definition is being relocated to the ODCM.

4.8.3 Definition Proposed for Addition The licensee proposes to add an ACTIONS definition. The licensee proposes it to read:

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

The definition for ACTIONS would be added in order to clarify a term used in the POTS sections. The definition is based on the definition in NUREG-1433, Revision 4.0, "Standard Technical Specifications, General Electric BWR/4 Plants," April 2012 (ADAMS Accession No. ML12104A192). The definition is proposed to be numbered as TS Section 1.1 to place it in alphabetical order with the remaining TS definitions. The NRC staff finds the addition of the ACTIONS definition acceptable because this is an editorial change that does not make any technical changes.

4.8.4 Definitions Proposed for Renumbering The licensee proposes to maintain the current definitions for CERTIFIED FUEL HANDLER and NON-CERTIFIED OPERATOR. These definitions were added as a result of Amendment No. 290, dated March 7, 2017 (ADAMS Accession No. ML16235A413). The licensee proposes to renumber the definitions for CERTIFIED FUEL HANDLER and NON-CERTIFIED OPERATOR from TS Sections 1.52 and 1.53 to TS Sections 1.2 and 1.3, respectively, in order to account for the other definitions that are proposed to be deleted. The NRC staff finds the proposed renumbering of the definitions for CERTIFIED FUEL HANDLER and NON-CERTIFIED OPERATOR acceptable because this is an editorial change that does not make any technical changes.

4.9 Proposed Deletion of TS Section 2 Section 2, "Safety Limits and Limiting Safety System Settings," of the Oyster Creek TSs establishes Safety Limits (SLs), which preclude violation of the fuel cladding integrity and RCS design pressure. TS Section 2 contains three specifications:

• TS 2.1, "Safety Limit - Fuel Cladding Integrity"

• TS 2.2, "Safety Limit - Reactor Coolant System Pressure"

• TS 2.3, "Safety Limiting Safety System Settings" Safety Limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. Limiting safety system settings for nuclear reactors are settings for automatic protective devices related to those variables having significant safety functions. The requirements in TSs 2.1 and 2.2 prevent overheating of the fuel and cladding, as well as possible cladding perforation that would result in the release of fission products to the reactor coolant. Technical Specification 2.1 is applicable in the Modes of Run and Startup/Hot Standby. Technical Specifications 2.1 and 2.2 promulgate requirements on parameters to protect the integrity of the RCS against overpressure. Technical Specification 2.2 is applicable in all Modes. The licensee proposes to delete the SLs and limiting safety system settings specified in TS Sections 2.1, 2.2, and 2.3, because they are not applicable to a reactor that is permanently shut down and defueled. The licensee states that the SLs and limiting safety system settings TSs limit important process variables that are necessary to reasonably protect the integrity of certain physical barriers required for safe operation of the reactor in all Modes. However, 10 CFR 50.82(a)(2) prohibits operation of the reactor or emplacement or retention of fuel in the reactor vessel. Therefore, the SL and limiting safety system settings TSs only address specific process variables that are no longer applicable to Oyster Creek since the certifications of permanent cessation of operations and permanent removal of fuel from the reactor vessel have been docketed.

The licensee proposed the deletion of TS Section 2 in its entirety.

The NRC staff examined the SLs, limiting safety system settings, and their Bases. There are two SLs in TS Section 2: TS 2.1, "Safety Limit - Fuel Cladding Integrity," and TS 2.2, "Safety Limit - Reactor Coolant System Pressure." The fuel cladding integrity SL is set such that no significant fuel damage is calculated to occur if the limit is not violated. The RCS pressure limit is set such that the integrity of the RCS is not threatened due to an overpressure condition. As stated in the Bases for the fuel cladding integrity SL, "the fuel cladding integrity [SL] is defined as the critical power ratio in the limiting fuel assembly, for which more than 99.9 [percent] of the fuel rods in the core are expected to avoid boiling transition, considering the power distribution within the core and all uncertainties." A limit is placed on the minimum critical power ratio. The Bases for the RCS pressure SL state that "reactor coolant system [integrity is] an important barrier in the prevention of the uncontrolled release of fission products."

Consistent with 10 CFR 50.82(a)(2), the Oyster Creek RFOL no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. In this permanently shutdown and defueled condition, there will be no minimum critical power ratio to be monitored and there will be no challenge to the RCS integrity. Based on these findings, the NRC staff concludes that the SLs and, consequently, limiting safety system settings no longer apply.

Therefore, the NRC staff finds that the proposed deletions of TS Sections 2.1, 2.2, and 2.3 are acceptable.

4.1 O Proposed Changes to TS Section 3 Section 3, "Limiting Conditions for Operation," of the Oyster Creek TS contains the general requirements applicable to all LCOs and applies at all times unless otherwise stated in TSs.

Due to the limited number of LCOs in the licensee-proposed POTS, a number of the provisions in this section are no longer necessary or applicable to the Oyster Creek facility and the licensee proposes their deletion, as indicated in the submittal, dated November 16, 2017. The licensee also proposes the addition of LCOs 3.0.1 and 3.0.2 to the POTS. These LCOs are based on NUREG-1433 and have been modified to reflect the permanently shutdown and defueled condition.

The licensee proposes to combine TS Section 3, "Limiting Conditions for Operation," and Section 4, "Surveillance Requirements," into a common specification. The proposed title of the combined TS Section would be TS Section 3/4, "LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS."

The licensee proposes to delete LCOs 3.0.A, 3.0.B, and 3.0.C.

The current LCO 3.0.A states:

In the event Limiting Conditions for Operation (LCOs) and/or associated action requirements cannot be satisfied because of circumstances in excess of those addressed in the specification, the unit shall be placed in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> unless corrective measures are completed that permit operation under the permissible action statements for the specified time interval as measured from initial discovery or until the reactor is placed in a condition in which the specification is not applicable. Exceptions to the requirements shall be stated in the individual specifications.

The current LCO 3.0.B states:

When a system, subsystem, train, component or device is determined to be inoperable solely because its emergency power source is inoperable, or solely because its normal power source is inoperable, it may be considered OPERABLE for the purpose of satisfying the requirements of applicable LCOs, provided (1) its corresponding normal or emergency power source is OPERABLE; and (2) all of its redundant system(s), subsystem(s), train(s),

component(s) and device(s) are OPERABLE, or likewise satisfy the requirements of this specification. Unless both conditions (1) and (2) are satisfied, the unit shall be placed in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> or within the time specified in the applicable specification. This specification is not applicable in COLD SHUTDOWN or the REFUEL MODE.

The current LCO 3.0.C states:

When an LCO is not met, entry into an OPERATIONAL CONDITION or other specified condition in the Applicability shall only be made:

1. When the associated LCO requirements permit continued operation in the OPERATIONAL CONDITION or other specified condition in the Applicability for an unlimited period of time; or

2. After performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering the OPERATIONAL CONDITION or other specified condition in the applicability, and the establishment of risk management actions, if appropriate; exceptions to this specification are stated in the individual Specifications, or

3. When an allowance is stated in the individual value, parameter, or other Specification.

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

Consistent with 10 CFR 50.82(a)(2), the Oyster Creek RFOL no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. In this permanently shutdown and defueled condition, there will be no power operation, no need to shut down the reactor, and no Modes of operation. Based on these findings, the NRC staff concludes that LCOs 3.0.A, 3.0.B, and 3.0.C no longer apply. Therefore, the NRC staff finds that the proposed deletions of LCOs 3.0.A, 3.0.B, and 3.0.C are acceptable.

4.11 Proposed Changes to LCO 3.0.1 The licensee proposes to add a new LCO 3.0.1 that states:

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

Proposed LCO 3.0.1 establishes the applicability statement within each individual TS as the requirement for when the LCO is required to be met (i.e., when the facility is in the specified conditions of the applicability statement of each specification). This statement is consistent with the permanently shutdown and defueled condition of Oyster Creek and LCO 3.0.1 in NUREG-1433. Therefore, the NRC staff finds the addition of the proposed LCO 3.0.1 acceptable.

4.12 Proposed Changes to LCO 3.0.2 The licensee proposes to add a new LCO 3.0.2 that states:

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. ,

Proposed LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated action shall be met. The completion time of each required action for an action condition is applicable from the point in time that an action condition is entered. The required actions establish those remedial measures that must be taken within specified completion times when the requirement of an LCO is not met. This statement is consistent with the permanently shutdown and defueled condition of Oyster Creek and LCO 3.0.2 in NUREG-1433. Therefore, the NRC staff finds the addition of the proposed LCO 3.0.2 acceptable.

4.13 Proposed Changes to SR 4.0.1 The licensee proposes to revise and relocate SR 4.0.1, such that it states:

Surveillance requirements shall be met during the specified conditions in the applicability for individual LCOs, unless otherwise stated in the surveillance requirements. 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 4.0.2. Surveillances do not have to be performed on variables outside specified limits.

The licensee proposes to relocate SR 4.0.1 from current TS Section 4, "Surveillance Requirements," to immediately follow the LCO statement in proposed POTS Section 3/4.

The SR 4.0.1 establishes the requirement that SRs must be met during the modes or other specified conditions in the applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. The SR 4.0.1 is to ensure that surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits.

The SR 4.0.1 is proposed for revision to remove references to operating modes and inoperable equipment. Because 10 CFR 50.82(a)(2) prohibits operation of the plant or placing fuel in the reactor vessel, the reference to operating modes is no longer relevant and is therefore proposed to be deleted. Since there are no LCOs for equipment to be operable or in operation in the POTS, the exception to not perform surveillances on inoperable equipment is also proposed to be deleted.

The proposed changes to SR 4.0.1 are either consistent with the permanently shutdown and defueled condition of Oyster Creek or editorial; therefore, the NRC staff finds them acceptable.

4.14 Proposed Changes to SR 4.0.2 The licensee proposes to relocate SR 4.0.2 that currently states:

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 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.

The SR 4.0.2 would be relocated from current TS Section 4 to immediately follow the LCO statement in proposed POTS Section 3/4. This proposed relocation is editorial and does not make any technical changes; therefore, the NRC staff finds it acceptable.

4.15 Proposed Changes to SR 4.0.3 The licensee proposes to revise and relocate SR 4.0.3, such that it states:

Entry into a specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillance has been met within its specified frequency, except as provided by 4.0.2.

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

The SR 4.0.3 would be relocated from current TS Section 4 to immediately follow the LCO statement in proposed POTS Section 3/4. The SR 4.0.3 establishes the requirements that all applicable SRs must be met before entry into an operational mode or other specified condition in the applicability. The SR 4.0.3 would be modified, such that, the SRs in proposed POTS 3/4.1 for the SFP must be met prior to entry into the specified condition. The remaining language is not necessary to preclude this and is proposed to be deleted. The proposed revision includes grammatical corrections. Because 10 CFR 50.82(a)(2) will prohibit operation of the plant or placing fuel in the reactor vessel, the reference to OPERATIONAL CONDITION and shutdown of the unit are no longer relevant and are proposed to be deleted. Additionally, the reference to exceptions and allowances stated in the TS LCO would be deleted since these items are not applicable in the POTS. The proposed changes to SR 4.0.3 are either consistent with the permanently shutdown and defueled condition of Oyster Creek or editorial; therefore, the NRC staff finds them acceptable.

4.16 Proposed New SR 4.0.4 The licensee proposes a new SR 4.0.4 that states:

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 SR 4.0.4 would immediately follow the LCO statement in proposed PDTS Section 3/4. The SR 4.0.4 is based upon the Oyster Creek TS Definition for "Surveillance Requirement." The wording of the proposed specification is from NUREG-1433, except that it is modified for a facility in a permanently shutdown and defueled condition. There is no change in intent for this statement and the Oyster Creek TS definition; both statements provide an allowance for extending the frequency for performance of an SR to 1.25 times the interval specified in the frequency to facilitate scheduling or unforeseen problems that may prevent performance during normal intervals.

The proposed SR is consistent with the permanently shutdown and defueled condition of Oyster Creek and NUREG-1433; therefore, the NRC staff finds its addition acceptable.

4.17 Proposed TS 3/4.1 - SPENT FUEL STORAGE Section 3.1 of the Oyster Creek TSs, "Protective Instrumentation," contains the LCOs and actions that provide for appropriate functional capability of sensing and control instrumentation required for safe operation of the facility. This section contains the following LCOs:

• TS 3.1.A - Operating Requirements for Plant Protective Instrumentation

• TS 3.1.B - Average Power Range Monitor Operability

• TS 3.1.C- Local Power Range monitors (LPRMs) and Traversing In-Core Probes (TIPs)

• Table 3.1.1 - Section A- Scram

• Table 3.1.1 - Section B - Reactor Isolation

• Table 3.1.1 - Section C- Isolation Condenser Initiation

• Table 3.1.1 - Section D - Core Spray

• Table 3.1.1 - Section E - Containment Spray

• Table 3.1.1 - Section F - Primary Containment Isolation

• Table 3.1.1 - Section G - Automatic Depressurization

• Table 3.1.1 - Section H - Isolation Condenser Isolation

• Table 3.1.1 - Section I - Off-gas System Isolation

• Table 3.1.1 - Section J - Reactor Building Isolation and Standby Gas Treatment System Initiation

• Table 3.1.1 - Section K- Rod Block

• Table 3.1.1 - Section L - Condenser Vacuum Pump Isolation

• Table 3.1.1 - Section M - Diesel Generator Load Sequence Timers

• Table 3.1.1 - Section N - Loss of Power

• Table 3.1.1 - Section O - Containment Vent and Purge Isolation

• Table 3.1.1 - Section P - RWCU [Reactor Water Cleanup] HELB [High Energy Line Break] Isolation

• Table 3.1.1 - Notes The TS Section 3.1 contains LCOs to ensure the operability of protective instrumentation. The LCOs are related to plant instrumentation that performs protective and monitoring functions to ensure safe operation of the reactor and to mitigate the effects of reactor-related DBAs. The licensee proposes to delete the entire above bulleted TSs. These TSs do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2).

Therefore, the protective functions addressed in TS Section 3.1 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply to the reactor in the permanently

shutdown and defueled condition. Based on the above, the NRC staff finds the proposed deletion of current TS 3.1 acceptable.

The licensee proposes to renumber and retitle TS 3.1 to TS 3/4.1, "Spent Fuel Storage," and add a new specification to address operability requirements for the SFP level. Specifically, the licensee proposes the following:

Applicability: During movement of irradiated fuel assemblies in the spent fuel pool.

Objective: To assure safe storage of spent fuel.

LCO: 3.1 Spent Fuel Pool Water Level Whenever irradiated fuel is stored in the spent fuel storage pool, water level shall be maintained at a level ~ 117 feet 8 inches (elevation above sea level) with the exception of planned cask movements.

ACTIONS:

Condition Required Action Completion Time Spent fuel pool water Suspend movement of irradiated Immediately level is not within fuel assemblies and movement limit. of loads over the storage racks containino fuel.

SURVEILLANCE REQUIREMENTS Surveillance Freauencv 4.1 I Verify the spent fuel pool water level is ~117 feet 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 8 inches.

The TS 3/4.1 for SFP water level is proposed to ensure safe storage and management of the spent fuel. This specification is proposed to be numbered as TS 3/4.1. The table of contents is also proposed to be revised to reflect these changes. Proposed LCO 3.1, "Spent Fuel Pool Water Level," specifies requirements to ensure that the minimum water level in the SFP meets the assumptions of iodine decontamination factors following an FHA in the SFP. The required minimum water level in the SFP also provides shielding during the movement of spent fuel and meets the assumptions of the FHA described in calculation C-1302-226-E310-460 and UFSAR Section 15.7.4, "Design Basis Fuel Handling Accidents in the Containments." The resultant dose limits at the exclusion area boundary are within the criteria of RG 1.183.

A general description of the SFP design is found in UFSAR Section 9.1.2, "Spent Fuel Storage."

The assumptions of the FHA are found in the UFSAR, Section 15. 7.4. The FHA is evaluated for dropping an irradiated fuel assembly onto irradiated fuel bundles stored in the SFP. The consequences of an FHA in the SFP are documented in UFSAR Chapter 15, "Accident Analysis." The water level in the SFP provides for absorption of water soluble fission product gases and transport delays of soluble and insoluble gases that must pass through the water

before being released to the building atmosphere. This absorption and transport delay reduces the potential radioactivity of the release during an FHA.

The SFP water level is monitored in terms of elevation above mean sea level. The elevation of 117 feet 8 inches corresponds to the SFP low level alarm in the CR. Since the pool has no installed drains, level cannot be lowered by the cooling system below the level of the weirs. At the normal 400-gallon per minute (gpm) flow rate, the SFP level is about three inches above the weir level, and the overflow just equals the 400 gpm being supplied to the SFP from the diffusers. At the SFP low level alarm level, the SFP contains a depth of approximately 37 feet of water (approximately 23 feet above active fuel), providing adequate shielding for normal building occupancy by operating personnel.

Proposed LCO 3.1 requires that when the water level in the SFP is lower than the required level, the movement of irradiated fuel assemblies and movement of loads over the storage racks containing fuel is to be "immediately" suspended. "Immediately" as used in this completion time means that the required action should be pursued without delay and in a controlled manner, such that the suspension of this activity shall not preclude completion of movement of an irradiated fuel assembly to a safe position. This effectively precludes a spent fuel handling accident from occurring in the SFP when the level is below the required elevation.

This proposed specification is not meant to affect spent fuel cask movements during planned SFP level adjustments. The UFSAR Chapter 15 analysis states that a spent fuel cask drop accident is no longer credible since the reactor building crane has been upgraded to be single-failure proof.

Proposed SR 4.1 verifies that sufficient SFP water is available in the event of an FHA. The water level in the SFP must be checked periodically. The frequency of every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable based on operating experience, considering that the water volume in the pool is normally stable and that water level changes are controlled by unit procedures.

Based on the above, the NRC staff finds that the SFP water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii) and that the proposed addition of TS 3/4.1 is acceptable.

4.18 Proposed Change to TS 3.2 - Reactivity Control The TS 3.2, "Reactivity Control," contains LCOs related to reactivity control capability and applies to core reactivity and the reactivity control systems to protect the integrity of the fission product barrier. Below are the specifications in this section.

• TS 3.2.A - Core Reactivity

• TS 3.2.B- Control Rod System

• TS 3.2.C - Standby Liquid Control System

• TS 3.2.D- Reactivity Anomalies The licensee proposes to delete this section. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2).

Therefore, the reactivity control functions addressed in TS 3.2 will not be required and these

LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3.2 acceptable.

4.19 Proposed Change to TS 3.3 - Reactor Coolant The TS 3.3, "Reactor Coolant," contains LCOs that provide assurance of the reactor coolant pressure boundary (RCPB) integrity and safe operation of the RCS. The protection and monitoring functions of the RCS have been designed to ensure safe operation of the reactor required to protect the integrity of a fission product barrier. The RCS is a primary barrier against the release of fission products to the environs.

The regulation under 10 CFR 50.60, "Acceptance criteria for fracture prevention measures for lightwater nuclear power reactors for normal operation," stipulates that reactor facilities which 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 RCPB set forth in 10 CFR Part 50, Appendices G and H. The maintenance rule (10 CFR 50.65) will be used to monitor the performance or condition of the SSCs associated with the storage, control, and maintenance of spent fuel in a safe condition. Below are the specifications in this section.

• TS 3.3.A - Pressure Temperature Relationships

• TS 3.3.B- Reactor Vessel Closure Head Boltdown

• TS 3.3.C - Thermal Transients

• TS 3.3.D- Reactor Coolant System Leakage

• TS 3.3.E - Reactor Coolant Quality

• TS 3.3.F - Recirculation Loop Operability

• TS 3.3.G - Primary Coolant System Pressure Isolation Valves

• TS 3.3.H - Required Minimum Recirculation Flow Rate for Operation in IRM Range 10 The licensee proposes to delete this section. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2).

Therefore, the reactor coolant specifications addressed in TS 3.3 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3.3 acceptable.

4.20 Proposed Change to TS 3.4 - Emergency Cooling The TS 3.4, "Emergency Cooling," contains LCOs to ensure the operability of the emergency cooling systems and to provide assurance of adequate cooling capability for heat removal in the event of a LOCA or isolation from the normal reactor heat sink. Below are the specifications in this section.

• TS 3.4.A - Core Spray System

• TS 3.4.B- Automatic Depressurization System

• TS 3.4.C - Containment Spray System and Emergency Service Water System

• TS 3.4.D- Control Rod Drive Hydraulic System

• TS 3.4.E- Core Spray and Containment Spray Pump Compartments Doors

• TS 3.4. F - Fire Protection System

The regulation under 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," specifies that light-water nuclear power reactors must be provided with an emergency core cooling system (ECCS) designed to meet the cooling performance requirements following postulated LOCAs. Section 10 CFR 50.46(a)(1 )(i) states, in part, that "[t]his section does not apply to a nuclear power reactor facility for which the certifications required under [10 CFR] 50.82(a)(1) have been submitted."

The licensee proposes to delete this section. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)( 1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2).

Therefore, the core and containment cooling specifications addressed in TS 3.4 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3.4 acceptable.

4.21 Proposed Changes to TS 3.5 - Containment The TS 3.5, "Containment," contains LCOs that ensure the integrity of the primary and secondary containment systems. The primary containment system provides a barrier against uncontrolled release of fission products to the environs in the event of a LOCA. Secondary containment is designed to minimize any ground level release of radioactive materials that might result from an accident. Below are the specifications in this section.

• TS 3.5.A - Primary Containment

• TS 3.5.B- Secondary Containment The OBA that remains applicable to Oyster Creek in the permanently shutdown and defueled condition is an FHA in the SFP. A calculation (C-1302-226-E310-460) was performed to assess the dose consequences of a postulated FHA after the permanent cessation of power operations.

The calculation demonstrates that radiological doses at the EAB, LPZ, and in the CR are within allowable limits of 10 CFR 50.67 without crediting secondary containment operability, standby gas treatment system, or CR high efficiency air filtration after a 60-day fuel decay period following permanent reactor shutdown.

The licensee proposes to delete this section. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2).

Therefore, the specifications for the primary and secondary containment systems addressed in TS 3.5 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3.5 acceptable.

4.22 Proposed Changes to TS 3.6 - Radioactive Effluents The TS 3.6, "Radioactive Effluents," contains LCOs to ensure that radioactive material is not released to the environment in an uncontrolled manner and to ensure that the radioactive concentrations of any material released is kept as low as is reasonably achievable and, in any

event, within the limits of 10 CFR 20.1301 and 40 CFR Part 190.10(a). The LCOs in this section apply to radioactive effluents of the facility. Below are the specifications in this section.

• TS 3.6.A- Reactor Coolant Radioactivity

• TS 3.6.B - Liquid Radwaste Treatment

• TS 3.6.C- Radioactive Liquid Storage

• TS 3.6.D - Condenser Off-gas Treatment

• TS 3.6.E - Main Condenser Off-gas Radioactivity

• TS 3.6.F - Condenser Off-gas Hydrogen Concentration

• TS 3.6.G - Not Used

• TS 3.6.H - Not Used

• TS 3.6.1- Radioactivity Concentration in Liquid Effluent

• TS 3.6.J - Limit on Dose due to Liquid Effluent

• TS 3.6.K - Dose Rate due to Gaseous Effluent

• TS 3.6.L - Air Dose due to Noble Gas in Gaseous Effluent

• TS 3.6.M - Dose due to Radioiodine and Particulates in Gaseous Effluent

• TS 3.6.N - Annual Total Dose due to Radioactive Effluents The licensee proposes to delete this section, except for TS 3.6.C, which remains applicable and is proposed to be renumbered as POTS TS 3.2. Many of the specifications in TS 3.6 were previously relocated to the ODCM in Amendment No. 166, dated December 13, 1993 (ADAMS Accession No. ML011200256). The following specifications have been relocated to the ODCM:

TS 3.6.B, TS 3.6.D, TS 3.6.1, TS 3.6.J, TS 3.6.K, TS 3.6.L, TS 3.6.M, and TS 3.6.N. The placeholders associated with these specifications are proposed to be removed in the proposed POTS; this proposed change is editorial. Technical Specification 3.6.G and TS 3.6.H currently read "not used" and are also proposed to be removed as an editorial change. The remainder of the specifications in TS Section 3.6 do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the remainder of the radioactive effluents specifications addressed in TS Section 3.6 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TSs 3.6.A, B, D, E, F, G, H, I, J, K, L, M, and N acceptable.

The licensee proposes to revise and renumber TS 3.6.C to the following.

3/4.2 RADIOACTIVE LIQUID STORAGE Applicability: Applies at all times to outdoor tanks used to store radioactive liquids.

Objective: To assure that radioactive effluents are not released to the environment in an uncontrolled manner and to assure that the radioactive concentrations of any material released is kept as low as is reasonably achievable and, in any event, within the limits of 10 CFR Part 20.1301 and 40 CFR Part 190.1 O(a).

LCO: 3.2 The quantity of radioactive material, excluding tritium, noble gases, and radionuclides having half-lives shorter than three days,

contained in outdoor storage tanks shall not exceed 10.0 curies.

Included in this specification are all outdoor storage tanks that contain radioactivity that are not surrounded by liners, dikes, or walls capable of holding the tank contents, or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system.

ACTIONS:

Condition Required Action Completion Time In the event the Begin treatment and continue it As soon as quantity of until the total quantity of reasonably radioactive material radioactive material in the tank achievable.

in any applicable is 10 curies or less, and storage tanks describe the reason for exceeds 10.0 curies. exceeding the limit in the next Annual Effluent Release Report.

SURVEILLANCE REQUIREMENTS Surveillance Frequency 4.2 Liquids contained in outdoor Once per 7 days when storage tanks included in radioactive liquid is this specification shall be being added to the sampled and analyzed for tank.

radioactivity.

The current TS 3.6.C for Radioactive Liquid Storage is proposed to be modified and renumbered as proposed TS 3/4.2 to ensure the safe storage and management of radioactive liquids contained in outdoor storage tanks. The table of contents is also proposed to be revised to reflect these changes.

The current specification's identification of two specific tanks is proposed to be deleted in order to broaden the definition of tanks. The proposed TS would include all outdoor storage tanks that contain radioactivity that are not surrounded by liners, dikes, or walls capable of holding the tank contents, or that do not have tank overflows and surrounding area drains connected to the liquid radwaste treatment system. The associated SR is proposed to be relocated as part of the reformatting of the TS to combine TS Sections 3 and 4. The Surveillance Frequency is proposed to be changed from "the frequency specified in the Surveillance Frequency Control Program," to "Once per 7 days when radioactive liquid is being added to the tank."

Based on the above, the NRC staff finds that radioactive liquid storage satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii) and that the proposed changes to TS 3/4.2 are acceptable.

4.23 Proposed Changes to TS 3. 7 - Auxiliary Electrical Power The TS 3.7, "Auxiliary Electrical Power," contains LCOs related to the operability of alternating current (AC) and direct current (DC) electrical systems. This section establishes the requirements for appropriate functional capability of plant electrical equipment required for safe

operation of the facility. This section specifies requirements to ensure that the station safety-related electrical bussing and distribution system, offsite power sources, and the onsite standby power sources (emergency diesel generators (EDGs)), provide sufficient capacity, capability, redundancy, and reliability to ensure the availability of necessary power to engineered safety features systems so that the fuel, RCS, and containment design limits are not exceeded. The requirements for EDG fuel oil storage are included for each EDG. Also included in this section are the requirements for DC power. It specifies requirements to ensure that the DC electrical power subsystems are operable. Below are the specifications in this section.

• TS 3.7.A- Required Electrical Sources

• TS 3.7.B - Required Actions for 3.7.A

• TS 3.7.C - Standby Diesel Generators

• TS 3.7.D - Station Batteries and Associated Battery Chargers The DBAs and transients analyzed in UFSAR Chapter 15 will no longer be applicable in the permanently shutdown and defueled condition, with the exception of the FHA in the SFP.

Exelon performed a calculation for an FHA in the SFP that shows that the dose consequences are acceptable without relying on any SSCs to remain functional during and following the event after 60 days of irradiated fuel decay time after permanent reactor shutdown and compliance with the SFP water level requirements in proposed TS 3/4.1.

Because the FHA analysis does not rely on normal or emergency power for accident mitigation (including any need for providing airborne radiological protection), the AC sources are not required during movement of irradiated fuel assemblies in the SFP for mitigation of a potential FHA. Therefore, during movement of irradiated fuel assemblies in the SFP, there are no active systems credited as part of the initial conditions of an analysis or as part of the primary success path for mitigation of the FHA with the unit permanently shutdown and defueled. As such, there are no DBAs that rely on AC and DC sources for mitigation.

The licensee proposes to delete this section. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2).

Therefore, the specifications addressed in TS 3.7 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3. 7 acceptable.

4.24 Proposed Changes to TS 3.8 - Isolation Condenser The TS 3.8, "Isolation Condenser," contains LCOs related to the operability requirements for the isolation condenser and its isolation valves. The isolation condenser assures decay heat removal from the reactor core under conditions when the reactor vessel is isolated from its normal heat sink. Below are the specifications in this section.

• TS 3.8.A - Two Isolation Condenser Loops

• TS 3.8.B - Minimum Water Volume - Condenser Shell Side

• TS 3.8.C - With One Isolation Condenser Inoperable in Run Mode

• TS 3.8.D - Required Action if Specification 3.8.A and 3.8.B not met

• TS 3.8.E - Inoperable Isolation Condenser Inlet (Steam Side) Isolation Valve

• TS 3.8.F - Inoperable AC Motor-Operated Isolation Condenser Outlet (Condensate Return) Isolation Valve The licensee proposes to delete TS 3.8. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2}. Therefore, the specifications addressed in TS 3.8 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition.

Therefore, the NRC staff finds the deletion of TS 3.8 acceptable.

4.25 Proposed Changes to TS 3.9- Refueling TS 3.9, "Refueling," contains LCOs related to fuel handling operations during refueling. During refueling, the reactivity potential of the core is being altered. It is necessary to require certain interlocks and restrict certain refueling procedures such that there is assurance that inadvertent criticality does not occur. Below are the specifications in this section.

• TS 3.9.A- Control Rod Position

• TS 3.9.B- Reactor Mode Switch

• TS 3.9.C - Refueling Equipment Interlocks

• TS 3.9.D - Source Range Monitors

• TS 3.9.E - Removal of Single Control Rod

• TS 3.9.F - Removal of Multiple Control Rods

• TS 3.9.G -Any Refueling Requirement Not Met The licensee proposes to delete TS 3. 9. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS 3.9 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition.

Therefore, the NRC staff finds the deletion of TS 3.9 is acceptable.

4.26 Proposed Changes to TS 3.10 - Core Limits The TS 3.10, "Core Limits," contains the LCOs to ensure that power distribution limits are met.

The LCOs will not apply to a reactor that is in a permanently shutdown and defueled condition.

Below are the specifications in this section.

• TS 3.1 O.A - Average Planar Linear Heat Generation Rate (APLHGR)

• TS 3.10. B - Local Linear Heat Generation Rate (LHGR)

• TA 3.10.C- Minimum Critical Power Ration (MCPR)

The licensee proposes to delete TS 3.10. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS 3.10 will not be required and these LCOs (and associated

SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition.

Therefore, the NRC staff finds the deletion of TS 3.10 acceptable.

4.27 Proposed Changes to TS 3.11 The TS 3.11, "Intentionally Left Blank," was added in License Amendment No. 29 dated March 3, 1978 (ADAMS Accession No. ML011150417). The licensee proposes to delete this unused section from the TS. The NRC staff finds the deletion of TS 3.11 acceptable.

4.28 Proposed Changes to TS 3.12 -Alternate Shutdown Monitoring Instrumentation The TS 3.12, "Alternate Shutdown Monitoring Instrumentation," contains the LCOs related to alternate shutdown monitoring instrumentation from outside the main CR The TS Table 3.12-1 lists the required instrumentation in this section. The instrumentation identified in this specification ensure that sufficient capability is available to permit shutdown and maintenance of hot shutdown of the plant from locations outside the CR. The specifications apply only when the plant is at power operation and when reactor coolant temperature is above 212 degrees Fahrenheit (°F). Below are the specifications in this section.

• TS 3.12.A

• TS 3.12.B

• TS Table 3.12-1 The licensee proposes to delete TS 3.12. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS 3.12 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition.

Therefore, the NRC staff finds the deletion of TS 3.12 acceptable.

4.29 Proposed Changes to TS 3.13 - Accident Monitoring Instrumentation The TS 3.13, "Accident Monitoring Instrumentation," contains LCOs related to the operability during power operation or when primary containment integrity is required to monitor the course of reactor accidents. Below are the specifications in this section.

• TS 3.13.A - Relief Valve Position Indicators

• TS 3.13.B - Safety Valve Position Indicators

• TS 3.13.C - Required Action for 3.13.A and 3.13.B

• TS 3.13.D-Wide Range Torus Water Level Monitor

• TS 3.13.E - Wide Range Drywell Pressure Monitor

• TS 3.13. F - Deleted

• TS 3.13.G - Containment High-Range Radiation Monitor

• TS 3.13.H - High-Range Radioactive Noble Gas Effluent Monitor The licensee proposes to delete TS 3.13. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore,

the specifications addressed in TS 3.13, including TS Table 3.13.1, will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of TS 3.13 acceptable.

4.30 .Proposed Changes to TS 3.15 - Explosive Gas Monitoring Instrumentation The TS 3.15, "Explosive Gas Monitoring Instrumentation," contains LCOs related to the operability of the instrumentation that monitors the hydrogen concentration in the augmented off-gas treatment system. Hydrogen is a byproduct of the reactor fission process. Below are the specifications in this section.

• TS 3.15.A - Explosive Gas Instrumentation

• TS Table 3.15.2 The licensee proposes to delete TS 3.15. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS 3.15 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition.

Therefore, the NRC staff finds the deletion of TS 3.15 acceptable.

4.31 Proposed Changes to TS 3.17 - Control Room Heating, Ventilation, and Air-Conditioning System The TS 3.17, "Control Room Heating, Ventilation, and Air-Conditioning System," contains the LCOs related to the CR Heating, Ventilation, and Air-Conditioning (HVAC) system. The operability of the CR HVAC system ensures that the CR will remain habitable for operations personnel during a postulated OBA. The operability of the CRE boundary must be maintained to protect the CR occupants during normal and accident conditions. The CRE and its boundary are defined in the Control Room Envelope Habitability Program. In order for the CR HVAC System to be considered operable, the CRE boundary must be maintained such that the CR occupant dose from a large radioactive release does not exceed the calculated dose in the licensing basis consequence analyses for DBAs, and the CR occupants are protected from hazardous chemicals and smoke. Since CR HVAC systems A and B do not have high-efficiency particulate air (HEPA) filters or charcoal absorbers, the supply fan and dampers for each system minimize the beta and gamma doses to the operators by providing positive pressurization and limiting the makeup and infiltration air into the CRE. For the supply of 100-percent outside unfiltered air to the CRE under DBA conditions, personnel occupying the CR shall not receive radiation exposure in excess of a 30-day integrated dose of 5 rem TEDE.

The FHA in the SFP is the only DBA that can occur with the facility in the permanently shutdown and defueled condition. In dose calculations in Reference 6 of the letter dated November 16, 2017, 3 Exelon provided an FHA-based analysis using AST methodology. The analysis determined the projected dose due to the drop of a fuel assembly onto other fuel assemblies as a function of time after shutdown. The analysis demonstrates that radiological doses at the EAB, LPZ, and in the CR from an FHA after 60 days following permanent shutdown are within allowable limits without crediting secondary containment operability and operation of the 3 Calculation C-1302-226-E310-460, "EAB, LPZ, and CR Dose Due to Fuel Handling Accident (FHA)-Post Cessation of Power Operations," dated August 9, 2017.

standby gas treatment system. No equipment is required to mitigate the effects of this event beyond the administrative controls described in Reference 6 of the letter dated November 16, 2017. Below are the specifications in this section.

• TS 3.17.A

• TS 3.17.B

• TS 3.17.C

• TS 3.17.D The licensee proposes to delete TS 3.17. These specifications do not apply to the safe storage and handling of spent fuel in the SFP. Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL license no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS 3.17 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition.

Therefore, the NRC staff finds the deletion of TS 3.17 acceptable.

4.32 Proposed Changes to SR 4.0.1 The SR 4.0.1 currently reads:

Surveillance requirements shall be met during the modes or other specified conditions in the applicability for individual LCOs, unless otherwise stated in the surveillance requirements. 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 4.0.2. Surveillances do not have to be performed on inoperable equipment or variables outside specified limits.

The licensee proposes SR 4.0.1 to read:

Surveillance requirements shall be met during the specified conditions in the applicability for individual LCOs, unless otherwise stated in the surveillance requirements. 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 4.0.2. Surveillances do not have to be performed on variables outside specified limits.

The SR 4.0.1 establishes the requirement that SRs must be met during the modes or other specified conditions in the applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits.

The SR 4.0.1 is proposed for revision to remove references to operating modes and inoperable equipment. Because 10 CFR 50.82(a)(2) prohibits operation of the plant or placing fuel in the reactor vessel, the reference to operating modes is no longer relevant and is therefore proposed

to be deleted. Since there are no LCOs for equipment to be operable or in operation in the POTS, the exception to not perform surveillances on inoperable equipment is no longer necessary.

The SR 4.0.1 is proposed to be relocated to the proposed POTS Section 3/4 as SR 4.0.1. The proposed changes to SR 4.0.1 are either consistent with the permanently shutdown and defueled condition of Oyster Creek or editorial; therefore, the NRC staff finds them acceptable.

4.33 Proposed Changes to SR 4.0.2 The licensee proposes to relocate SR 4.0.2 to the proposed POTS Section 3/4 as SR 4.0.2.

This proposed relocation is editorial and does not make any technical changes; therefore, the NRC staff finds it acceptable.

4.34 Proposed Changes to SR 4.0.3 The SR 4.0.3 currently reads:

Entry into an OPERATIONAL CONDITION or other specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillance have been met within their specified frequency, except as provided by 4.0.2. When an LCO is not met due to surveillances not having been met, entry into an OPERATIONAL CONDITION or other specified condition in the Applicability shall only be made in accordance with LCO 3.0.C.

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

The licensee proposes SR 4.0.3 to read:

Entry into a specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillance has been met within its specified frequency, except as provided by 4.0.2.

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

The SR 4.0.3 establishes the requirements that all applicable SRs must be met before entry into an operational mode or other specified condition in the applicability. The SR 4.0.3 is proposed to be modified, such that, the SRs in proposed PDTS 3/4.1 for SFP must be met prior to entry in to the specified condition. The remaining language is not necessary to preclude this and is proposed to be deleted. The proposed revision includes grammatical corrections. Because 10 CFR 50.82(a)(2) will prohibit operation of the plant or placing fuel in the reactor vessel, the reference to OPERATIONAL CONDITION and shutdown of the unit are no longer relevant and are proposed to be deleted. Additionally, the reference to exceptions and allowances stated in the TS LCO would be deleted since these items are not applicable in PDTS. The SR 4.0.3 is proposed to be relocated to POTS Section 3/4 as SR 4.0.3. The proposed changes to SR 4.0.3 are either consistent with the permanently shutdown and defueled condition of Oyster Creek or editorial; therefore, the NRC staff finds them acceptable.

• 4.35 Proposed Changes to SRs 4.1 through TS 4.17 Technical Specification Section 4 describes SRs associated with the TS Section 3 LCOs. In accordance with 10 CFR 50.36(c)(3), SRs are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the LCOs will be met.

Since there are no safety limits that apply to Oyster Creek with the reactor in a permanently shutdown and defueled condition, and since there are relatively few remaining LCOs, the number of corresponding SRs is proposed to be greatly reduced.

Specifically, the licensee proposes to delete the following SRs:

• SR 4.1 - Protective Instrumentation

• SR 4.2 - Reactivity Control

• SR 4.3 - Reactor Coolant

• SR 4.4 - Emergency Cooling

• SR 4.5 - Containment System

• SR 4. 7 - Auxiliary Electrical Power

• SR 4.8 - Isolation Condenser

• SR 4.9- Refueling

• SR 4.1 O - ECCS Related Core Limits

• SR 4.11 - Sealed Source Contamination

• SR 4.12 - Alternate Shutdown Monitoring Instrumentation

• SR 4.13 - Accident Monitoring Instrumentation

• SR 4.14 - Solid Radioactive Waste

• SR 4.15 - Explosive Gas Monitoring Instrumentation

• SR 4.17 - Control Room Heating, Ventilation, and Air-Conditioning System These specifications do not apply to the safe storage and handling of spent fuel in the SFP.

Since the certifications required by 10 CFR 50.82(a)(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in SR 4.1 through SR 4.5, SR 4. 7 through SR 4.15, and SR 4.17 will not be required and these SRs will not apply in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of SR 4.1 through SR 4.5, SR 4. 7 through SR 4.15, and SR 4.17 acceptable.

The SR 4.16, "Radiological Environmental Surveillance," was previously relocated to the ODCM in Amendment No. 166, dated December 13, 1993 (ADAMS Accession No. ML011200256).

Currently, SR 4.16 is a placeholder. Therefore, the NRC staff finds the deletion of SR 4.16 acceptable.

The SR 4.6, "Radioactive Effluent," was analyzed in SE Section 4.23. The SR 3.6.C is being maintained and renumbered. The NRC staff finds this change to proposed PDTS 3/4.2 acceptable.

Based on the above, the NRC staff finds the changes to SR 4.1 through SR 4.17 acceptable.

4.36 Proposed Changes to TS Section 5 - Design Features The TS 5.1.B currently reads:

The reactor building, standby gas treatment system and stack shall comprise a secondary containment in such fashion to enclose the primary containment in order to provide for controlled elevated release of the reactor building atmosphere under accident conditions.

The TS 5.2 currently reads:

CONTAINMENT A. The primary containment shall be of the pressure suppression type having a drywall and an absorption chamber constructed of steel. The drywall shall have a volume of approximately 180,000 ft 3 and conforms to the ASME Boiler and Pressure Vessel Code,Section VIII, for an internal pressure of 44 psig at 292°F and an external pressure of 2 psig at 150°F to 205°F. The absorption chamber shall have a total volume of approximately 210,000 ft 3 . It is designed to conform to ASME Boiler and Pressure Vessel Code,Section VIII, for an internal pressure of 35 psig at 150°F and an external pressure of 1 psig at 150°F.

B. Penetrations added to the primary containment shall be designed in accordance with standards set forth in Section V-1.5 of the Facility Description and Safety Analysis Report. Piping passing through such penetrations shall have isolation valves in accordance with standards set forth in Section V-1.6 of the Facility Description and Safety Analysis Report.

The TS 5.1.B and TS 5.2 involve design features that do not apply in the permanently shutdown and defueled condition and are proposed to be deleted. Because the Oyster Creek RFOL will no longer authorize emplacement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2), the NRC staff finds the deletion of TS 5.1.B and TS 5.2 acceptable.

The TS 5.3 currently reads:

5.3 AUXILIARY EQUIPMENT 5.3.1 Fuel Storage A. The fuel storage facilities are designed and shall be maintained with a K-effective equivalent to less than or equal to 0.95 including all calculational uncertainties.

B. Deleted C. Deleted D. The temperature of the water in the spent fuel storage pool, measured at or near the surface, shall not exceed 125°F.

E. The maximum amount of spent fuel assemblies stored in the spent fuel storage pool shall be 3035.

The licensee proposes TS 5.3 to read:

5.2 SPENT FUEL STORAGE 5.2.1 Spent Fuel Storage A. The spent fuel storage facilities are designed and shall be maintained with a K-effective equivalent to less than or equal to 0.95 including all calculational uncertainties.

B. The temperature of the water in the spent fuel storage pool, measured at or near the surface, shall not exceed 125°F.

C. The maximum amount of spent fuel assemblies stored in the spent fuel storage pool shall be 3035.

Current TS 5.3.A, D, and E describe and provide the requirements regarding prevention of criticality of spent fuel, temperature limitation of SFP water, and SFP capacity limitations.

Current TS 5.3.1 and 5.3.1.A are proposed to be modified to clarify that the requirements are applicable to only spent fuel storage since there will be no new fuel storage maintained after the permanent cessation of operations and defueling. Current TS 5.3.1.D and E are proposed to be retained as-is in the proposed PDTS. Current TS 5.3.1.B and Care proposed to be deleted because they state only "Deleted." The remaining specifications are proposed to be renumbered as TS 5.2.1.A-C. The proposed changes to TS 5.3 are either consistent with the permanently shutdown and defueled condition of Oyster Creek or editorial; therefore, the NRC staff finds them acceptable.

4.37 Proposed Changes to TS 6.8.1.a The licensee proposes to change TS 6.8.1.a. The proposed change is to replace "QATR" (Quality Assurance Topical Report) with "Decommissioning Quality Assurance Program (DQAP)."

The QATR is the Exelon quality assurance program for its operating reactors. The QATR is being replaced by a quality assurance program for decommissioning plants.

The proposed change to TS 6.8.1.a is either consistent with the permanently shutdown and defueled condition of Oyster Creek or editorial; therefore, the NRC staff finds it acceptable.

4.38 Proposed Changes to TS 6.8.4.a The licensee proposes to change TS 6.8.4.a by reformatting the uppercase defined terms of "MEMBERS OF THE PUBLIC" and "UNRESTRICTED AREA" to lowercase. This is an editorial change and does not make any technical changes; therefore, the NRC staff finds it acceptable.

4.39 Proposed Deletion of TS 6.8.4.b.2, TS 6.8.4.b.3, TS 6.8.5, TS 6.9.2, TS 6.9.3, TS 6.10.1-TS 6.10.2 1 TS 6.11, TS 6.12, TS 6.14, TS 6.15, TS 6.16, TS 6.17, TS 6.18, TS 6.20, TS 6.22, TS 6.23, TS 6.24, and TS 6.25 The licensee proposes to delete the following TSs:

• TS 6.8.4.b.2 - Land Use Census

• TS 6.8.4.b.3 - lnterlaboratory Comparison Program

• TS 6.8.5 - Station Battery Monitoring and Maintenance Program

• TS 6.9.2 - DELETED

• TS 6.9.3 - Unique Reporting Requirements

• TS 6.10.1-TS 6.10.2 - Record Retention

• TS 6.11 - Radiation Protection Program

• TS 6.12 - Deleted

• TS 6.14 - Environmental Qualification

• TS 6.15 - Integrity of Systems Outside Containment

• TS 6.16 - Iodine Monitoring

• TS 6.17 - Deleted

• TS 6.18 - Process Control Plan

• TS 6.20 - Major changes to Radioactive Waste Treatment Systems

• TS 6.22 - Control Room Envelope Habitability Program

• TS 6.23 - Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

• TS 6.24 - Surveillance Frequency Control Program

• TS 6.25 - Snubber Inspection Program These specifications do not apply to the safe storage and handling of spent fuel in the SFP.

Since the certifications required by 10 CFR 50.82(a}(1) are docketed, the Oyster Creek RFOL no longer authorizes operation of the reactor or placement or retention of fuel in the reactor vessel, pursuant to 10 CFR 50.82(a)(2). Therefore, the specifications addressed in TS 6.8.4.b.2, TS 6.8.4.b.3, TS 6.8.5, TS 6.9.2, TS 6.9.3, TS 6.10.1-TS 6.10.2, TS 6.11,

TS 6.12, TS 6.14, TS 6.15, TS 6.16, TS 6.17, TS 6.18, TS 6.20, TS 6.22, TS 6.23, TS 6.24, and TS 6.25 will not be required and these LCOs (and associated SRs in TS Section 4) will not apply in a permanently shutdown and defueled condition. Therefore, the NRC staff finds the deletion of these specifications acceptable.

NOTE: The deletion of TS 6.10.1-6.10.2, regarding record retention does not exempt Exelon from complying with 10 CFR 50. 71, "Maintenance of records, making of reports."

4.40 Proposed Changes to TS 6. 9.1 - Routine Reports The current TS 6.9.1, "Routine Reports," reads:

a. DELETED

b. DELETED

c. DELETED

d. Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the facility during 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 effluent and solid waste released from the facility. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

e. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the facility during the previous calendar year shall be submitted prior to May 1 of each year.

The Report shall include summaries, interpretations, and an analysis 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: ( 1) the ODCM; and, (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

Basis: 6.9.1.e - RELOCATED TO THE ODCM.

f. DELETED

The licensee proposes TS 6.9.1 to read:

a. Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the facility during 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 effluent and solid waste released from the facility. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

b. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the facility during the previous calendar year shall be submitted prior to May 1 of each year.

The Report shall include summaries, interpretations, and an analysis 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: (1) the ODCM; and, (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR Part 50.

There are no technical changes to TS 6.9.1; the changes are deletion and renumbering.

Therefore, the NRC staff finds the changes to TS 6.9.1 acceptable.

4.41 Proposed Change to TS 6.10, "Record Retention" The licensee proposes to delete TS Sections 6.10.1 and 6.10.2, which was discussed in SE Section 4.40. The licensee proposes to re-number 6.10.3 to 6.10.1, and to change "QATR" to "DQAP."

The proposed changes are editorial and do not make any technical changes; therefore, the NRC staff finds them acceptable.

4.42 Proposed Changes to TS 6.19 - Offsite Dose Calculation Manual The current TS 6.19, "Offsite Dose Calculation Manual," reads:

a. The ODCM shall be approved by the Commission prior to implementation.

b. Licensee initiated changes to the ODCM shall be submitted to the NRC in the Annual Radioactive Effluent Release Report for the period in which the changes were made. This submittal shall contain:

1. sufficiently detailed information to justify the changes without benefit of additional or supplemental information;

2. a determination that the changes did not reduce the accuracy or reliability of dose calculations or setpoint determination; and,

3. documentation that the changes have been reviewed and approved pursuant to Section 6.8.2.

c. Change(s) shall become effective upon review and approval by licensee management.

The licensee proposes TS 6.19 to read:

a. Licensee initiated changes to the ODCM shall be submitted to the NRC in the Annual Radioactive Effluent Release Report for the period in which the changes were made.

This submittal shall contain:

1. sufficiently detailed information to justify the changes without benefit of additional or supplemental information;

2. a determination that the changes did not reduce the accuracy or reliability of dose calculations or setpoint determination; and,

3. documentation that the changes have been reviewed and approved pursuant to Section 6.8.2.

b. Change(s) shall become effective upon review and approval by licensee management.

The licensee proposes to delete paragraph "a" since this action has been completed. The NRC staff approved the ODCM in Amendment No. 166 dated December 13, 1993 (ADAMS Accession No. ML011200256). Consequently, paragraphs "b" and "c" are proposed to be renumbered as "a" and "b," respectively. These proposed changes are editorial and do not make any technical changes; therefore, the NRC staff finds them acceptable.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendment on August 10, 2018 (ADAMS Accession No. ML18222A541 ). 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 surveillance requirements. 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, and there has been no public comment on such finding (83 FR 2229; January 16, 2018). 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: J. Lamb, A. Patel, G. Curran, E. Dickson, A. Smith, R. Grover, and S. Sheng Date: October 26, 2018

ML18227A338 *via email **via memo OFFICE NRR/DORL/LSPB/PM NRR/DORL/LSPB/LA NRR/DRA/ARCB/BC* NRR/DSS/STSB/BC**

NAME Jlamb JBurkhardt KHsueh (JParillo for) VCusumano DATE 9/14/18 9/14/18 4/6/18 4/11/18 OFFICE NRR/DSS/SRXB/BC** NRR/DSS/SCPB/BC** NRR/DMLR/MVI B/BC* NRR/DSS/SNPB/BC*

NAME JWhitman RDennig DAIiey (SRuffin for) Rlukes (SKrepel for)

DATE 6/21/18 3/20/18 3/19/18 7/27/18 OFFICE OGC- NLO* NRR/DORL/LSPB/BC NRR/DORL/LSPB/PM DBroaddus (RPascarelli NAME JWachutka Jlamb for)

DATE 9/24/18 10/24/18 10/26/18