ML16235A413

Issuance of Amendment No. 290, Changes to Administrative Controls Section of the Technical Specifications No Longer Applicable to Facility in Permanently Defueled Condition

ML16235A413
Person / Time
Site:	Oyster Creek
Issue date:	03/07/2017
From:	John Lamb Special Projects and Process Branch
To:	Bryan Hanson Exelon Nuclear
Lamb J
References
CAC MF8108
Download: ML16235A413 (52)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 7, 2017 Mr. Bryan C. Hanson President and Chief Nuclear Officer Exelon Generation Company, LLC Oyster Creek Nuclear Generating Station 4300 Winfield Road Warrenville, IL 60555

SUBJECT:

OYSTER CREEK NUCLEAR GENERATING STATION - ISSUANCE OF AMENDMENT REGARDING CHANGES TO THE ADMINISTRATIVE CONTROLS SECTION OF THE TECHNICAL SPECIFICATIONS (CAC NO. MF8108)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 290 to Renewed Facility Operating License No. DPR-16 for the Oyster Creek Nuclear Generating Station (OCNGS), in response to your application dated May 17, 2016, as supplemented by letters dated November 2, 2016, and March 1, 2017.

The amendment revises and removes certain requirements from the Section 6, "Administrative Controls," portions of the OCNGS Technical Specifications (TSs) that are not applicable to the facility in a permanently defueled condition. In addition, the amendment adds definitions to TS Section 1, "Definitions." Also, the amendment makes additions to, deletions from, and conforming administrative changes to the TSs.

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

G. Lamb, Senior Project Manager cial Projects and Process Branch ision of Operating Reactor Licensing ice of Nuclear Reactor Regulation Docket No. 50-219

Enclosures:

1. Amendment No. 290 to DPR-16

2. Safety Evaluation cc w/encls: Distribution via 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. 290 Renewed License No. DPR-16

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

A. The application for amendment by Exelon Generation Company, LLC (the licensee) dated May 17, 2016, as supplemented by letters dated November 2, 2016, and March 1, 2017, 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 paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-16 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 290, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

3. This license amendment will be effective upon the licensee's submittal of the certifications required by 10 CFR 50.82(a)(1 )(i) and (ii), and shall be implemented within 60 days of the effective date of the amendment, but may not exceed March 29, 2020.

FOR THE NUCLEAR REGULATORY COMMISSION

/RA/ Eva A. Brown for Douglas A. Broaddus, Chief Special Projects and Process Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance: March 7, 2017

ATTACHMENT TO LICENSE AMENDMENT NO. 290 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-16 OYSTER CREEK NUCLEAR GENERATING STATION DOCKET NO. 50-219 Replace the following pages of the Renewed Facility Operating License and Appendix A, 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 Operating License No. DPR-16 Remove Insert 3 3 Appendix A. Technical Specifications Remove ii ii iii iii 1.0-9 1.0-9 6-1 6-1 6-2 6-2 6-2a 6-3 6-9 6-10 6-3 6-11 6-4 6-12 6-5 6-13 6-6 6-14 6-14a 6-15 6-16 6-7 6-17 6-8 6-18 6-9 6-19 6-10 6-20 6-11 6-21 6-12 6-22 6-13 6-23 6-14

(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 material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; (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 (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 as may be 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) 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.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 290, are hereby incorporated in the license. Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

(3) 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.

Renewed License No. DPR-16 Amendment No. 290

TABLE OF CONTENTS (Cont'd) 1.44 Local Linear Heat Generation Rate 1.0-8 1.45 Shutdown Margin (SOM) 1.0-8 1.46 Idle Recirculation Loop 1.0-8 1.47 Isolated Recirculation Loop 1.0-8 1.48 Operational Condition 1.0-8 1.49 Rated Thermal Power 1.0-9 1.50 Thermal Power 1.0-9 1.51 Pressure and Temperature Limits Report (PTLR) 1.0-9 1.52 Certified Fuel Handler 1.0-9 1.53 Non-Certified Operator 1.0-9 Section 2 Safety Limits and Limiting Safety System Settings 2.1 Safety Limit - Fuel Cladding Integrity 2.1-1 2.2 Safety Limit - Reactor Coolant System Pressure 2.2-1 2.3 Limiting Safety System Settings 2.2-3 Section 3 Limiting Conditions for Operation 3.0 Limiting Conditions for Operation (General) 3.0-1 3.1 Protective Instrumentation 3.1-1 3.2 Reactivity Control 3.2-1 3.3 Reactor Coolant 3.3-1 3.4 Emergency Cooling 3.4-1 3.5 Containment 3.5-1 3.6 Radioactive Effluents 3.6-1 3.7 Auxiliary Electrical Power 3.7-1 3.8 Isolation Condenser 3.8-1 3.9 Refueling 3.9-1 3.10 Core Limits 3.10-1 3.11 (Not Used) 3.11-1 3.12 Alternate Shutdown Monitoring Instrumentation 3.12-1 3.13 Accident Monitoring Instrumentation 3.13-1 3.14 DELETED 3.14-1 3.15 Explosive Gas Monitoring Instrumentation 3.15-1 3.16 (Not Used) 3.16-1 3.17 Control Room Heating, Ventilating and Air Conditioning System 3.17-1 Section 4 Surveillance Requirements 4.0 Surveillance Requirement Applicability 4.0-1 4.1 Protective Instrumentation 4.1-1 4.2 Reactivity Control 4.2-1 4.3 Reactor Coolant 4.3-1 4.4 Emergency Cooling 4.4-1 4.5 Containment 4.5-1 4.6 Radioactive Effluents 4.6-1 4.7 Auxiliary Electrical Power 4.7-1 4.8 Isolation Condenser 4.8-1 4.9 Refueling 4.9-1 OYSTER CREEK 11 Amendment No.: 160, I 85, I 86, 24 I, 266, 2-69, 290

TABLE OF CONTENTS (cont'd)

Page 4.10 ECCs Related Core Limits 4.10-1 4.11 Sealed Source Contamination 4.11-1 4.12 Alternate Shutdown Monitoring Instrumentation 4.12-1 4.13 Accident Monitoring Instrumentation 4.13-1 4.14 DELETED 4.14-1 4.15 Explosive Gas Monitoring Instrumentation 4.15-1 4.16 (Deleted) 4.16-1 4.17 Control Room Heating, Ventilating and Air Conditioning System 4.17-1 Section 5 Design Features 5.1 Site 5.1-1 5.2 Containment 5.2-1 5.3 Auxiliary Equipment 5.3-1 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 6-6 6-10 Record Retention 6-8 6-11 Radiation Protection Program 6-9 6-12 (Deleted) 6-9 6-13 High Radiation Area 6-9 6-14 Environmental Qualification 6-10*

6-15 Integrity of Systems Outside Containment 6-10 6-16 Iodine Monitoring 6-10 6-17 DELETED 6-11 6-18 Process Control Plan 6-11 6-19 Offsite Dose Calculation Manual 6-11 6-20 DELETED 6-12 6-21 Technical Specification (TS) Bases Control Program 6-12 6-22 Control Room Envelope Habitability Program 6-12 6-23 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) 6-13 6.24 Surveillance Frequency Control Program 6-14 6.25 Snubber Inspection Program 6-14

• Issued by NRC Order dated 10-24-80 OYSTER CREEK lll Amendment No.: 94, 9+, 9&,-1-0&,~,

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~290

1.49 RATED THERMAL POWER (RTP)

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 1930 MWt.

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 heatup 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.

1.52 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.

1.53 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 1.0-9 Amendment No. ~. ~ 290

ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during the Plant Manager's absence.

The Plant Manager or the designee shall approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affect safe storage and maintenance of spent nuclear fuel.

6.1.2 The Shift Manager shall be responsible for the shift command function.

6.2 ORGANIZATION 6.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for facility staff and corporate management.

The onsite and offsite organization shall include the positions for activities affecting the safe storage and handling of spent nuclear fuel.

a. Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to and including facility organization positions. These relationships shall be documented and updated as appropriate, in the form of organizational descriptions. These organizational descriptions will be documented in the Updated FSAR and updated in accordance with 10 CFR 50.71e.

b. The Plant Manager shall be responsible for overall facility safe operation and shall have control over those onsite activities necessary for safe storage and maintenance of spent nuclear fuel.

c. A responsible officer shall have corporate responsibility for the safe storage and handling of spent nuclear fuel and shall take measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the facility to ensure safe management of spent nuclear fuel.

d. Individuals who train the CERTIFIED FUEL HANDLERs and those who carry out the health physics and quality assurance functions may report to the appropriate manager on site; however, these individuals shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

6.2.2 Facility Staff The facility organization shall meet the following:

a. Each on duty shift shall include at least the following shift staffing:

One (1) Shift Manager (see f. below)

One (1) NON-CERTIFIED OPERATOR (see g. below)

b. Shift crew composition may be one less than the minimum requirements of 6.2.2.a for a period of time not to exceed two hours, in order to accommodate unexpected absence of on-duty shift crew members. Immediate action must be OYSTER CREEK 6-1 Amendment No. 54, 59, 65, 69, 102, 134, 194, 195, 203, 210, 213, 220, 290

taken to restore the shift crew composition to within the requirements given above.

During such absences, no fuel movement or movement of loads over the spent fuel shall be permitted. This provision does not permit any shift crew position to be unmanned upon shift change due to an incoming shift crew member being late or absent.

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

d. Oversight of fuel handling operations shall be provided by a CERTIFIED FUEL HANDLER.

e. An individual qualified in radiation protection measures shall be on site during movement of fuel and during the movement of loads over the fuel.

f. The Shift Manager shall be a CERTIFIED FUEL HANDLER.

g. The position of NON-CERTIFIED OPERATOR may be filled by a CERTIFIED FUEL HANDLER.

6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1 of 1978 for comparable positions unless otherwise noted in the Technical Specifications. Technicians and maintenance personnel who do not meet ANSI/ANS 3.1 of 1978, Section 4.5, are permitted to perform work for which qualification has been demonstrated.

6.3.2 The management position responsible for radiological controls shall meet or exceed the qualifications of Regulatory Guide 1.8 (Rev. 1-R, 9/75). Each other member of the radiation protection organization for which there is a comparable position described in ANSI N18.1-1971 shall meet or exceed the minimum qualifications specified therein, or in the case of radiation protection technicians, they shall have at least one year's continuous experience in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power stations and shall have been certified by the management position responsible for radiological controls as qualified to perform assigned functions. This certification must be based on an NRC approved, documented program consisting of classroom training with appropriate examinations and documented positive findings by responsible supervision that the individual has demonstrated his ability to perform each specified procedure and assigned function with an understanding of its basis and purpose.

6.3.3 The NRC approved training and retraining program for CERTIFIED FUEL HANDLERS shall be maintained.

6.4 DELETED 6.5 DELETED 6.6 DELETED 6.7 DELETED OYSTER CREEK 6-2 Amendment No.: 92,102,134,161,203,220,274,290

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 QATR.

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

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

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,

10. 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,

2. A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and

3. Participation in an lnterlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

6.8.5 Station Battery Monitoring and Maintenance Program This Program provides for restoration and maintenance, based on the recommendations of IEEE Standard 450, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries For Stationary Applications," of the following:

a. Actions to restore station battery cells with float voltage < 2.13 volts, and

b. Actions to equalize and test station battery cells that have been discovered with electrolyte level below the top of the plates.

OYSTER CREEK 6-5 Amendment No.: 290

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. 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.8.1.

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.8.3, and IV.C of Appendix I to 10 CFR Part 50.

Basis: 6.9.1.e - RELOCATED TO THE ODCM.

f. DELETED 6.9.2 DELETED OYSTER CREEK 6-6 Amendment No.: 290

6.9.3 Unique Reporting Requirements Special reports shall be submitted to the Director of Regulatory Operations Regional Office within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable reference specification.

a. Materials Radiation Surveillance Specimen Reports (4.3A)

b. (Deleted)

c. Results of required leak tests performed on sealed sources if the tests reveal the presence of 0.005 microcuries or more of removable contamination.

d. Deleted e-j. Pursuant to the ODCM.

k. Records of results of analyses required by the Radiological Environmental Monitoring Program.

I. Failures and challenges to Relief and Safety Valves which do not constitute an LER will be the subject of a special report submitted to the Commission within 60 days of the occurrence. A challenge is defined as any automatic actuation (other than during surveillance or testing) of Safety or Relief Valves.

m. Plans for compliance with standby liquid control Specifications 3.2.C.3(b) and 3.2.C.3(e)(1) or plans to obtain enrichment test results per Specification 4.2.E.5.

n. Inoperable high range radioactive noble gas effluent monitor (3.13.H)

OYSTER CREEK 6-7 Amendment No.: 290

6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years:

a. Records and logs of facility operation covering time interval at each power level.

b. Records and logs of principle maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.

c. All Licensee Event Reports.

d. Records of surveillance activities, inspections and calibrations required by these technical specifications.

e. Records of reactor tests and experiments.

f. Records of changes made to operating procedures.

g. Deleted.

h. Records of sealed source leak tests and results.

i. Records of annual physical inventory of all source material of record.

6.10.2 The following records shall be retained for the duration of the Facility Operating License:

a. Record and drawing changes reflecting facility design modifications made to systems and equipment described in the Final Safety Analysis Report.

b. Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.

c. Records of facility radiation and contamination surveys.

d. Records of doses received by all individuals for whom monitoring was required

e. Records of gaseous and liquid radioactive material released to the environs.

f. Records of transient or operational cycles for those facility components designed for a limited number of transients or cycles.

g. Records of training and qualification for current members of the plant staff.

h. Records of inservice inspections performed pursuant to these technical specifications.

i. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59.

j. Deleted.

OYSTER CREEK 6-8 Amendment No.: 290

k. Records of Environmental Qualification which are covered under the provisions for paragraph 6.14.

I. Deleted.

m. Records of results of analyses required by the Radiological Environmental Monitoring Program.

n. Records of reviews performed for changes made to the OFFSITE DOSE CALCULATION MANUAL and the PROCESS CONTROL PLAN.

o. Records of radioactive shipments 6.10.3 Quality Assurance Records shall be retained as specified by the QATR.

6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.

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.

OYSTER CREEK 6-9 Amendment No.: 69,78,84,117,134,203,210,213,251, 290

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 ENVIRONMENTAL QUALIFICATION A. By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of: Division of Operating Reactors "Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or, NUREG-0588 "Interim Staff Position of Environmental Qualification of Safety-Related Electrical Equipment," December 1979. Copies of these documents are attached to Order for Modification of License DPR-16 dated October 24, 1980.

B. By no later than December 1, 1980, complete and auditible records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the DOR Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified.

6.15 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following:

1. Provisions establishing preventative maintenance and periodic visual inspection requirements, and

2. System leak test requirements, to the extent permitted by system design and radiological conditions, for each system at a frequency of once every 24 months. The systems subject to this testing are (1) Core Spray, (2) Containment Spray, (3) Reactor Water Cleanup, (4) Isolation Condenser, and (5) Shutdown Cooling.

6.16 IODINE MONITORING The licensee shall implement a program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas* under accident conditions. This program shall include the following:

a. Training of personnel,

b. Procedures for monitoring, and

c. Provisions for maintenance of sampling and analysis equipment.

• Areas requiring personnel access for establishing hot shutdown condition.

OYSTER CREEK 6-10 Amendment No.: 69,78,84,108,125,134,161, 166,210,213,251,273, 290

6.17 Deleted 6.18 PROCESS CONTROL PLAN

a. Licensee initiated changes to the PCP:

1. 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:

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

b. a determination that the changes did not reduce the overall conformance of the solidified waste product to existing criteria for solid wastes; and

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

2. Shall become effective upon review and approval by licensee management.

6.19 OFFSITE DOSE CALCULATION MANUAL

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.

OYSTER CREEK 6-11 Amendment No. ~ 290

6.20 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 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 NRC approval pursuant to 10 CFR 50.59.

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

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 NRC prior to implementation. Changes to the bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

6.22 CONTROL ROOM ENVELOPE HABITABILITY PROGRAM A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room HVAC System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge.

The program shall ensure that adequate radiation protection is provided to permit access and occup~mcy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of a 30-day integrated dose of 5 rem TEDE. The program shall include the following elements:

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

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

OYSTER CREEK 6-12 Amendment No. 69,73,84,108,125,134,166,205,245, 290 Corrected by letter of 10/15/2004

The following are exceptions to Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0:

1. The Oyster Creek CRE boundary operability is not dependent on a measured unfiltered air inleakage value (Reference Oyster Creek letter to NRC dated November 17, 2005, Letter No. 2130-05-20218). No inleakage testing for determining the unfiltered air inleakage past the CRE boundary into the CRE is required at the Oyster Creek site.

d. Measurement, at designated locations, of the CRE pressure relative to areas adjacent to the CRE boundary during the pressurization mode of operation by one subsystem (train) of the Control Room Ventilation System operating at the design flow rate, at a Frequency of 24 months. The results shall be trended and used as part of the 24 month assessment of the CRE boundary.

e. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of OBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f. The provisions of Section 1.24 are applicable to the frequencies for assessing CRE habitability measuring CRE pressure and assessing the CRE boundary as required by paragraphs d and c, respectively.

6.23 REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

a. RCS pressure and temperature limits for heat up, cooldown, low temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

i) Limiting Conditions for Operation Section 3.3, "Reactor Coolant" ii) Surveillance Requirements Section 4.3, "Reactor Coolant"

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following document:

i) SIR-05-044-A, "Pressure-Temperature Limits Report Methodology for Boiling Water Reactors"

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

OYSTER CREEK 6-13 Amendment No. 69,78,84,108,125,134,166,183, 191,254, 290

6.24 SURVEILLANCE FREQUENCY CONTROL PROGRAM This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Definition 1.24 and Surveillance Requirement 4.0.2 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

6.25 SNUBBER INSPECTION PROGRAM This program conforms to the examination, testing, and service life monitoring for dynamic restraints (snubbers) in accordance with 10 CFR 50.55a inservice inspection (ISi) requirements for supports. The program shall be in accordance with the following:

a. This program shall meet 10 CFR 50.55a(g) ISi requirements for supports.

b. The program shall meet the requirements for ISi of supports set forth in subsequent editions of the Code of Record and addenda of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code and the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) that are incorporated by reference in 10 CFR 50.55a(a), 50.55a(a)(1 ),

50.55a(a)(1 )(i), and 50.55a(a)(1 )(iv), subject to its limitations and modifications, and subject to Commission approval.

c. The program shall, as allowed by 10 CFR 50.55a(b)(3)(v)(B), meet Subsection ISTA, "General Requirements," and Subsection ISTD, "Preservice and lnservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants," in lieu of Section XI of the ASME B&PV Code ISi requirements for snubbers, or meet authorized alternatives pursuant to 10 CFR 50.55a(z).

d. The 120-month program updates shall be made in accordance with 10 CFR 50.55a (including 10 CFR 50.55a(g)(4)(ii)) subject to the conditions listed therein.

e. Records of the service life of all snubbers, including the date which the service life commences, and associated installation and maintenance records shall be maintained for the duration of the Facility Operating License.

OYSTER CREEK 6-14 Amendment No. 69,108,134,147,166,248,264, 290

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 290 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 letter dated January 7, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML110070507), Exelon Generation Company, LLC (Exelon or the licensee) informed the U.S. Nuclear Regulatory Commission (NRG) that the Oyster Creek Nuclear Generating Station (OCNGS) will permanently cease operations no later than December 31, 2019. Upon docketing of the certifications for permanent cessation of operations (paragraph 82(a)(1 )(i) to Part 50, "Domestic Licensing of Production and Utilization Facilities, of Title 10 of the Code of Federal Regulations (10 CFR)) and permanent removal of fuel from the reactor vessel (10 CFR 50.82(a)(1 )(ii)), pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for OCNGS will no longer authorize operation of the reactor or the emplacement or retention of fuel into the reactor vessel. The irradiated fuel will be stored in the OCNGS spent fuel pool (SFP) and in dry cask storage at the on-site independent spent fuel storage installation (ISFSI) until it is shipped off-site.

By application dated May 17, 2016 (ADAMS Accession No. ML16138A129), as supplemented by letters dated November 2, 2016, and March 1, 2017 (ADAMS Accession Nos. ML16308A029 and ML17060A436, respectively}, Exelon requested changes to the Technical Specifications (TSs) for OCNGS.

On July 19, 2016, the NRG staff published a proposed no significant hazards consideration (NSHC) determination regarding the amendment request in the Federal Register (81 FR 46963). Subsequently, by letter dated November 2, 2016, the licensee provided additional information that expanded the scope of the amendment request as originally noticed in the Federal Register. Accordingly, the NRG staff published a second proposed NSHC determination regarding the amendment request in the Federal Register on November 22, 2016 (81 FR 83876), which superseded the original Federal Register notice in its entirety. The supplemental letter dated March 1, 2017, provided additional information that clarified the application, did not expand the scope of the application as noticed, and did not change NRG staff's second proposed NSHC determination.

Enclosure 2

The proposed changes would revise and remove certain requirements from the Section 6, "Administrative Controls," portions of the OCNGS TSs that are not applicable to the facility in a permanently defueled condition. Specifically, the amendment would revise TS Section 6.1, "Responsibility"; TS Section 6.2, "Organization"; TS Section 6.3, "Facility Staff Qualifications";

TS Section 6.6, "Reportable Event Action"; TS Section 6.7, "Safety Limit Violation";

TS Section 6.8, "Procedures and Programs"; and TS Section 6.9, "Reporting Requirements" to reflect the staffing and training requirements for operating staff when the facility is permanently defueled. In addition, the amendment would add definitions to TS 1, "Definitions." Also, the amendment would make additions to, deletions from, and conforming administrative changes to the TSs.

By letter dated September 6, 2016 (ADAMS Accession No. ML16222A787), the NRC approved the Certified Fuel Handler (CFH) Training and Retraining Program for OCNGS.

2.0 REGULATORY EVALUATION

The regulatory requirements and guidance that the NRC staff considered in its review of the license amendment request are:

• The regulation under 10 CFR 50.120, "Training and qualification of nuclear power plant personnel," which requires the use of a Systems Approach to Training (SAT) for personnel positions, including Certified Fuel Handlers.

• The regulation under 10 CFR 50.36, "Technical specifications," paragraph (c)(5),

which provides requirements for the content of the TSs in the category "Administrative Controls."

• NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition," Chapter 13, "Conduct of Operations,"

Section 13.2.2, Revision 4, "Non-Licensed Plant Staff Training," August 2016 (ADAMS Accession No. ML15006A129), which provides guidance for the review of training programs for non-licensed plant staff.

• NUREG-1433, Revision 4, "Standard Technical Specifications General Electric BWR/4 Plants," April 2012 (Volume 1, "Specifications," and Volume 2, "Bases," at ADAMS Accession Nos. ML12104A192 and ML12104A193, respectively).

3.0 TECHNICAL EVALUATION

3.1 TS 6.1 - Responsibility 3.1 .1 TS 6.1.1 The current TS 6.1.1 states:

The Vice President - Oyster Creek shall be responsible for overall facility operation. Those responsibilities delegated to the Vice President as stated in the Oyster Creek Technical Specifications may also be fulfilled by the Plant Manager. The Vice President shall delegate in writing the succession to this responsibility during his and/or the Plant Manager absence.

The proposed TS 6.1.1 would state:

The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsibility during the Plant Manager's absence.

The Plant Manager or the designee shall approve, prior to implementation, each proposed test, experiment, or modification to systems or equipment that affect safe storage and maintenance of spent nuclear fuel.

3.1.1.1 NRC Staff Evaluation of TS 6.1.1 The proposed TS 6.1.1 describes the managerial position with responsibility for overall plant operations following permanent shutdown. The assignment of the position for this responsibility is changed from the OCNGS Site Vice President to the Plant Manager. The proposed change adds to the Plant Manager or the designee the responsibility of approving, prior to its implementation, proposed tests, experiments, and modifications to systems or equipment that affect the safe storage and maintenance of spent nuclear fuel once the unit is in a permanently defueled condition.

The Plant Manager will be the senior position at OCNGS and will be duly authorized by Exelon with sufficient authority to carry out the specific responsibilities as described in the proposed TS 6.1.1 and TS 6.2.1.b (see Section 3.2.1 of this safety evaluation). Although not an Exelon corporate officer, the Plant Manager will be vested by Exelon with similar authority over activities subject to the safe storage and handling of spent nuclear fuel. The Plant Manager will be duly authorized by Exelon with sufficient authority and organizational freedom to identify, initiate, recommend, and provide solutions at a defueled facility.

The Plant Manager will report to a "responsible officer" (Vice President or higher) within the Exelon corporate organization who will have the necessary authority and full responsibility for the safe management, storage, and handling of spent nuclear fuel as stated in the licensee's letter dated November 2, 2016.

3.1 .2 TS 6.1 .2 Currently, there is no TS 6.1.2.

The proposed TS 6.1.2 would state:

The Shift Manager shall be responsible for the shift command function.

3.1.2.1 NRC Staff Evaluation of TS 6.1.2 TS 6.1.2 is proposed to be added to identify the Shift Manager as having the command function of the shift. The NUREG-1433, Revision 4, Volume 1 Standard TSs {STSs) Section 5.1.2 states, in part, that the "shift supervisor" shall be responsible for the main control room (MCR) command function. TS 6.1.2 is being modified from the STS, because safe operation in the permanently defueled condition consists primarily of ensuring safe management of the spent irradiated fuel that is stored on site. Associated activities (e.g., fuel handling) do not necessarily rely on the MCR. The MCR will remain the physical center of the command function; however,

since control of activities may be performed either remotely from the MCA or locally in the plant, the location of the command center is functionally where the Shift Manager is located. The proposed OCNGS TS changes recognize that the delegation of authority for command and control aspects is different for a permanently shut down and defueled facility from that for an operating plant when the Shift Manager leaves the MCA as stated in the STSs. Once OCNGS is in a permanently defueled condition with fuel in the SFP, the number of relevant controls located in the MCA and the gradual nature of abnormal or accident situations (i.e., fuel handing accident, postulated liquid waste system leak) would not warrant that the command function remain in the MCA. Adequate communications capability is provided to allow operators and plant personnel to safely manage storage and handling of irradiated fuel without reliance on the MCA for the command function.

3.

1.3 NRC Staff Conclusion

for TS 6.1 The NRC staff finds that the proposed changes to TS 6.1.1 and TS 6.1.2 are acceptable because, once the reactor is permanently defueled, any potential events are expected to evolve slowly and, therefore, the proposed changes involve sufficient management oversight to maintain the facility in a safe manner and protect the environment and the health and safety of the public 3.2 TS 6.2 - Organization 3.2.1 TS 6.2.1 - Corporate The current TS 6.2.1 states:

6.2.1 Corporate 6.2.1.1 An onsite and offsite organization shall be established for unit operation and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safety of the nuclear power plant.

6.2.1.2 Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to and including operating organization positions. These relationships shall be documented and updated as appropriate, in the form of organizational charts. These organizational charts will be documented in the Updated FSAR [Final Safety Analysis Report] and updated in accordance with 10 CFR 50.71e.

6.2.1.3 The Chief Nuclear Officer shall have corporate responsibility for overall plant nuclear safety and shall take measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support in the plant so that continued nuclear safety is assured.

The proposed TS 6.2.1 would state:

6.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for facility staff and corporate management. The onsite and offsite organization shall include the positions for activities affecting the safe storage and handling of spent nuclear fuel.

a. Lines of authority, responsibility and communication shall be established and defined from the highest management levels through intermediate levels to and including facility organization positions. These relationships shall be documented and updated as appropriate, in the form of organizational descriptions. These organizational descriptions will be documented in the Updated FSAR and updated in accordance with 10 CFR 50.71 e.

b. The Plant Manager shall be responsible for overall facility safe operation and shall have control over those onsite activities necessary for safe storage and maintenance of the spent nuclear fuel.

c. A responsible officer shall have corporate responsibility for the safe storage and handling of spent nuclear fuel and shall take measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the facility to ensure safe management of spent nuclear fuel.

d. Individuals who train the CERTIFIED FUEL HANDLERs and those who carry out the health physics and quality assurance functions may report to the appropriate manager on site; however, these individuals shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

3.2.1.1 NRC Staff Evaluation of TS 6.2.1 The proposed TS 6.2.1 changes the section title from "Corporate" to "Onsite and Offsite Organizations." The reformatted TS section discusses both corporate and onsite organizations.

This change is considered administrative in nature. TS 6.2.1.1 identifies the onsite and offsite organizational positions that are responsible for safe operation of the nuclear plant. The reformatting, renumbering, and rewording process is administrative and involves no technical changes to the existing TS. The paragraphs following this introductory statement are renumbered as subparts a through d. The proposed changes reflect a permanently defueled facility and require that positions be established that are responsible for the safe management of the spent fuel. The terms "unit, "unit operation, "power plant, and "plant" are typically associated with an operating reactor. The proposed changes replace these terms, where applicable, with terms such as facility," facility staff," and "spent nuclear fuel, which are more appropriate in representing the permanently shut down and defueled condition.

The terms "safe storage and maintenance of spent nuclear fuel" and "safe management of spent nuclear fuel" are considered analogous to "nuclear safety" for a facility that is in a

permanently defueled condition. Following the permanent cessation of operations and defueling, nuclear safety will focus predominately on ensuring the safe control and management of spent fuel. The assignment of this responsibility is changed from the Chief Nuclear Officer (CNO) to a responsible officer. This proposed change provides Exelon the flexibility to assign overall responsibility to a responsible officer position other than CNO. The CNO is considered a corporate officer position. This position has no qualification requirements beyond the applicable requirements established in the American National Standards Institute/American Nuclear Society (ANSI/ANS) 3.1-1978, "Selection and Training of Nuclear Power Plant Personnel." The current TS 6.2.2.3 is essentially being moved to TS 6.2.1.d with some minor editorial changes and the addition of the CFH.

3.2.1.2 NRG Staff Conclusion for TS 6.2.1 The proposed changes appropriately reflect the changed function of the previous operating staff to a focus on the safe handling and storage of spent nuclear fuel and remove the implication that OCNGS can return to operation once it has been permanently defueled and the final certification required by 10 CFR 50.82(a)(1 )(ii) has been submitted to the NRG. Therefore, the NRG staff finds the proposed changes to TS 6.2.1 acceptable.

3.2.2 TS 6.2.2 - Facility Staff The current TS 6.2.2 states:

6.2.2 FACILITY STAFF 6.2.2.1 The Vice President - Oyster Creek shall be responsible for overall unit safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

6.2.2.2 The facility organization shall meet the following:

a. Each on duty shift shall include at least the following shift staffing:

• One (1) Shift Manager (see h. below)

• Two (2) licensed Nuclear Plant Operators

• Three (3) licensed or non-licensed Nuclear Plant Operators

• One (1) Shift Technical Adviser [(STA)] (see h.

below)

Except for the Shift Manager, shift crew composition may be one less than the minimum requirements, for a period of time not to exceed two hours, in order to accommodate unexpected absence of on-duty shift crew members.

Immediate action must be taken to restore the shift crew composition to within requirements given above. This provision does not permit any shift crew position to be unmanned upon shift change due to an incoming shift crew member being late or absent.

b. At all times when there is fuel in the vessel, at least one licensed senior reactor operator [(SRO)] shall be on site and one licensed reactor operator should be at the controls.

c. At all times when there is fuel in the vessel, except when the reactor is in COLD SHUTDOWN or REFUEL modes, two licensed senior reactor operators and two licensed reactor operators shall be on site, with at least one licensed senior reactor operator in the control room and one licensed reactor operator at the controls.

d. At least two licensed reactor operators shall be in the control room during all reactor startups, shutdowns, and other periods involving planned control rod manipulations.

e. All CORE ALTERATIONS shall be directly supervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f. An individual qualified in radiation protection measures shall be on site when fuel is in the reactor.

g. (deleted)

h. Each on duty shift shall include a Shift Technical Advisor except that the Shift Technical Advisors position need not be filled if the reactor is in the refuel or shutdown mode and the reactor is less than 212 F. The Shift Technical Advisor position may be filled by an on-shift Senior Reactor Operator (dual-role SRO/STA) provided the individual meets the requirements of 6.3.3.

i. (deleted)

j. The Senior Manager - Operations or an Operations Manager, and the Shift Manager require Senior Reactor Operators licenses. The licensed Nuclear Plant Operators require a Reactor Operators license.

6.2.2.3 Individuals who train the operating staff and those who carry out the health physics and quality assurance function shall have sufficient organizational freedom to be independent of operational pressures, however, they may report to the appropriate manager on site.

The proposed TS 6.2.2 would state:

6.2.2 Facility Staff The facility organization shall meet the following:

a. Each on duty shift shall include at least the following shift staffing:

• One (1) Shift Manager (see f. below)

• One (1) NON-CERTIFIED OPERATOR (see g. below)

b. Shift crew composition may be one less than the minimum requirements of 6.2.2.a for a period of time not to exceed two hours, in order to accommodate unexpected absence of on-duty shift crew members. Immediate action must be taken to restore the shift crew composition to within requirements given above.

During such absences, no fuel movement or movement of loads over the spent fuel shall be permitted. This provision does not permit any shift crew position to be unmanned upon shift change due to an incoming shift crew member being late or absent.

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

d. Oversight of fuel handling operations shall be provided by a CERTIFIED FUEL HANDLER.

e. An individual qualified in radiation protection measures shall be on site during movement of fuel and during the movement of loads over the fuel.

f. The Shift Manager shall be a CERTIFIED FUEL HANDLER.

g. The position of NON-CERTIFIED OPERATOR may be filled by a CERTIFIED FUEL HANDLER.

3.2.3 NRC Staff Evaluation of TS 6.2.2 The proposed TS 6.2.2 changes the case of the title from upper case letters, "FACILITY STAFF," to title case letters, "Facility Staff." This change is considered to be editorial in nature.

The proposed TS 6.2.2 also relocates current TS 6.2.2.1 paragraph to proposed TS 6.2.1.b as part of the reformatting of this TS section. The proposed TS 6.2.2 reformats TS 6.2.2.2 to reflect the introductory statement of TS 5.2.2 in the STS. The reformatting, renumbering, and rewording process is administrative and involves no technical changes to the existing TS. The paragraphs following this introductory statement are renumbered as subparts a through g. The remaining subparts are removed as part of the TS reformatting outlined in the submittal.

3.2.3.1 NRC Staff Evaluation of TS 6.2.2.2.a The existing TS 6.2.2.2.a is proposed to be renumbered as TS 6.2.2.a, which change is considered to be editorial in nature. This section describes the minimum shift staffing for plant operations. Since plant operations cannot occur at OCNGS once the certifications required by 10 CFR 50.82(a)(1) have been submitted to the NRC, the minimum staffing requirement is changed to a minimum crew compliment of one shift manager and one non-certified operator.

The number and complexity of the systems of an operating plant will be reduced to the systems required to provide and support SFP cooling. This crew compliment is sufficient to monitor SFP parameters, such as pool level and temperature, while maintaining the ability to ensure spent fuel handling operations are carried out in a safe manner. Moreover, the spectrum of credible accidents and operational events, and the quantity and complexity of activities required for safety are greatly reduced from those at an operating plant. The shift manager will be qualified as a CFH in accordance with the proposed TS 6.2.2.f. In this position, this individual will retain command and control responsibility for operational decisions and will be responsible for the functions required for event reporting and emergency response. The non-certified operator position can be filled by either a non-certified operator or by a CFH in accordance with the proposed TS 6.2.2.g. The shift manager cannot also be counted as the required non-certified operator.

The training and qualification for the NON-CERTIFIED OPERATOR will be determined in accordance with the SAT as defined in 10 CFR 55.4. This process ensures that the NON-CERTIFIED OPERATOR will be qualified to perform the functions necessary to monitor and ensure safe storage of fuel. The SAT process requires:

(1) Systematic analysis of the jobs to be performed, (2) Learning objectives derived from the analysis which describe desired performance after training, (3) Training design and implementation based on the learning objectives, (4) Evaluation of trainee mastery of the objectives during training, and (5) Evaluation and revision of the training based on the performance of trained personnel in the job setting.

There will be a sufficient number of individuals qualified as CFHs to staff the facility 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day, 7 days per week. The CFHs will be qualified in accordance with the CFH Training and Retraining Program approved by the NRC by letter dated September 6, 2016. Additional on-shift staffing will be provided to satisfy applicable security, fire protection, and emergency preparedness requirements. The MCR will remain the physical center of the command function.

However, since control of activities may be performed either remotely from the control room or locally in the plant, the location of the command center is functionally where the shift manager is located, in accordance with proposed TS 6.1.2. Activities that could be performed from the MCR that have the potential to affect forced cooling of spent nuclear fuel include changing the electrical power distribution system alignment. All spent fuel handling activities including starting and stopping cooling water pumps are performed locally at the SFP. Indications and/or alarms are also received in the MCR that would be indicative of SFP abnormalities. The shift manager is responsible for directing the response to those abnormalities, from either the MCR or local to the SFP, in accordance with applicable response procedures. For any condition,

incident, or event that occurs when the non-certified operator is in the MCR alone and is not within the scope of qualifications that are possessed by the non-certified operator, the shift manager will be immediately contacted for direction by phone, radio, and/or plant page system.

This philosophy is deemed acceptable because the necessity to render immediate actions to protect the health and safety of the public is not challenged.

3.2.3.2 NRC Staff Evaluation of TS 6.2.2.2.b The existing TS 6.2.2.2.b is proposed to be renumbered as TS 6.2.2.b, which change is considered to be .editorial in nature. The existing TS 6.2.2.2.b establishes the requirement for at least one licensed SRO to be on-site and at least one licensed Reactor Operator to be at the controls when fuel is in the reactor. Following the submission to the NRC of the certifications required by 10 CFR 50.82(a)(1), OCNGS will no longer be authorized to operate the reactor or load fuel into the reactor vessel, and the requirements of 10 CFR 50.54(m) requiring a licensed operator staffing will no longer apply. Therefore, the proposed changes that remove requirements for licensed operators are acceptable. The proposed TS 6.2.2.b addresses the conditions under which the minimum shift compliment may be reduced, due to unforeseen circumstances. It allows for shift crew composition to be less than the minimum requirement of the proposed TS 6.2.2.a for a period of time, not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to accommodate unexpected absence of on-duty shift crew members, provided immediate action is taken to restore the shift crew composition to within the minimum requirements. It also ensures that no fuel movement or movement of loads over the spent fuel occur during an absence. The NRC staff reviewed the proposed change and determined that this provision does not permit any shift crew position to be unmanned during movements of fuel or load over fuel, as well as during shift change, due to absence or tardiness of an oncoming shift crew member. Therefore, the proposed change is acceptable.

3.2.3.3 NRC Staff Evaluation of TS 6.2.2.2.c The existing TS 6.2.2.2.c is proposed to be renumbered as TS 6.2.2.c, which change is considered to be editorial in nature. The existing TS 6.2.2.2.c establishes the requirement for two licensed SROs and two licensed Reactor Operators to be onsite with at least one SRO in the control room and one Reactor Operator at the controls when fuel is in the reactor vessel.

Following the submission to the NRC of the certifications required by 10 CFR 50.82(a)(1 ),

OCNGS will not be required to have operators licensed pursuant to 10 CFR Part 55; therefore, TS 6.2.2.2.c will not apply and this paragraph is proposed to be deleted.

Proposed TS 6.2.2.c reflects the requirement for having one qualified watch stander (either a NON-CERTIFIED OPERATOR or CFH) in the MCR when fuel is stored in the SFP. This reflects the reduced requirement for control room personnel training and qualification for a plant authorized for nuclear fuel storage only.

3.2.3.4 NRC Staff Evaluation of TS 6.2.2.2.d The existing TS 6.2.2.2.d establishes the requirement for two licensed Reactor Operators to be at the controls during startups, shutdowns, and other periods involving planned control rod manipulations. Once the certifications for permanent cessation of operations and removal of fuel from the reactor vessel have been submitted to the NRC pursuant to 10 CFR 50.82(a)(1),

the OCNGS license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel, and OCNGS will no longer be required to have

operators licensed pursuant to 10 CFR Part 55; therefore, TS 6.2.2.2.d will not apply and this paragraph is proposed to be deleted.

3.2.3.5 NRC Staff Evaluation of TS 6.2.2.2.e The existing TS 6.2.2.2.e is proposed to be renumbered as TS 6.2.2.d, which change is considered to be editorial in nature. This section establishes that all core alterations be directly supervised by a licensed SRO or SRO limited to fuel handling. Once the certifications for permanent cessation of operations and removal of fuel from the reactor vessel have been submitted to the NRC pursuant to 10 CFR 50.82(a)(1 ), OCNGS license will no longer authorize the emplacement or retention of fuel in the reactor vessel. This paragraph is proposed to be changed to establish the requirement for oversight of fuel handling operations in the SFP to be performed by a CFH. Fuel moves and heavy load moves that could affect the safe handling and storage of spent nuclear fuel would be approved by the shift manager. Proposed TS 6.2.2.f requires the shift manager to be a CFH.

3.2.3.6 NRC Staff Evaluation of TS 6.2.2.2.f The existing TS 6.2.2.2.f is proposed to be renumbered as TS 6.2.2.e, which change is considered to be editorial in nature. This section establishes the requirement for a person qualified in radiation protection measures to be onsite when fuel is in the reactor. This requirement is being modified to require an individual qualified in radiation protection measures to be onsite during the movement of fuel and during the movement of loads over fuel. Following the submission to the NRC of the certification of permanent removal of fuel from the reactor vessel, fuel will no longer be permitted to be emplaced or retained in the vessel. The proposed TS reflects those remaining activities where individuals qualified in radiation protection measures are required to be present.

The proposed TS 6.2.2.f adds the requirement that the Shift Manager be a CFH. This requirement ensures that the senior individual on shift is appropriately trained and qualified, in accordance with the NRG-approved CFH Training and Retraining Program, to supervise shift activities. The OCNGS management structure will not require positions above the Shift Manager to be qualified as a CFH or attend equivalent training. Exelon has stated that once the plant is permanently shut down and defueled, the time available (i.e., greater than 1O hours) to mitigate credible events is expected to be greater than that for current design basis events for OCNGS. As such, management oversight of the facility can be performed by individuals meeting the applicable requirements of ANSI/ANS 3.1-1978 (as required by the proposed TS 6.3.1) and need not be qualified as CFHs (see Section 3.3.1 of this safety evaluation).

3.2.3.7 NRC Staff Evaluation of TS 6.2.2.2.g TS 6.2.2.2.g was deleted previously. The proposed change adds the provision that the NON-CERTIFIED OPERATOR position required in TS 6.2.2.a may be filled by either a NON-CERTIFIED OPERATOR or by a CFH. This minimum shift crew composition is appropriate for the safe management of spent irradiated nuclear fuel at a permanently defueled facility. This paragraph is proposed to be renumbered as TS 6.2.2.g, which change is considered to be editorial in nature.

3.2.3.8 NRC Staff Evaluation of TS 6.2.2.2.h TS 6.2.2.2.h establishes the requirements for the STA position. This paragraph is proposed to be deleted to remove the requirements for the STA position since that position is only required for a plant authorized for power operations. Once the certifications required by 10 CFR 50.82(a)(1) have been submitted to the NRC, the requirements of this paragraph will no longer be applicable because the OCNGS license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel. This paragraph is proposed to be removed as part of the TS reformatting, which change is considered to be editorial in nature.

3.2.3.9 NRC Staff Evaluation of TS 6.2.2.2.i TS 6.2.2.2.i was previously deleted and is proposed to be removed as part of the TS reformatting, which change is considered to be editorial in nature.

3.2.3.10 NRC Staff Evaluation of TS 6.2.2.2.j TS 6.2.2.2.j establishes the requirement for the Senior Manager- Operations and the Shift Manager to hold an SRO license and the Nuclear Plant Operator to hold a Reactor Operator license. This paragraph is proposed to be deleted. Once the certifications required by 10 CFR 50.82(a)(1) have been submitted to the NRC, the requirements of 10 CFR 50.54(m) will no longer be applicable, because the OCNGS license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel. These certifications also obviate the need for the operators' licenses specified in 10 CFR Part 55. Therefore, there is no longer a need for operations management staff to hold an SRO license. This paragraph is proposed to be removed as part of the TS reformatting, which change is considered to be editorial in nature.

3.2.3.11 NRC Staff Evaluation of TS 6.2.2.3 The existing TS 6.2.2.3 is proposed to be relocated to TS 6.2.1.d as part of the reformatting of the TS section (see Section 3.2.1.1 of this Safety Evaluation).

3.

2.4 NRC Staff Conclusion

for TS 6.2.2 Once the certifications required by 10 CFR 50.82(a)(1) have been submitted to the NRC, certain requirements will no longer be applicable because the OCNGS 10 CFR Part 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel. The NRC staff finds that the proposed changes to TS 6.2.2 reflect the scope of the activities that would result from permanent cessation of operations and, therefore, are acceptable.

3.3 TS 6.3 - Facility Staff Qualifications 3.3.1 TS 6.3 The current TS 6.3 states:

6.3 Facility Staff Qualifications 6.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1 of 1978 for comparable positions unless otherwise noted in the Technical Specifications, with the following exceptions: 1) the licensed operators who shall comply only with the requirements of 10 CFR 55, and (2) technicians and maintenance personnel who do not meet ANSI/ANS 3.1 of 1978, Section 4.5, are permitted to perform work for which qualification has been demonstrated.

6.3.2 The management position responsible for radiological controls shall meet or exceed the qualifications of Regulatory Guide 1.8 (Rev. 1-R, 9/75). Each other member of the radiation protection organization for which there is a comparable position described in ANSI N18.1-1971 shall meet or exceed the minimum qualifications specified therein, or in the case of radiation protection technicians, they shall have at least one year's continuous experience in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power stations and shall have been certified by the management position responsible for radiological controls as qualified to perform assigned functions. This certification must be based on an NRC approved, documented program consisting of classroom training with appropriate examinations and documented positive findings by responsible supervision that the individual has demonstrated his ability to perform each specified procedure and assigned function with an understanding of its basis and purpose.

6.3.3 The Shift Technical Advisors shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, response and analysis of the plant for transients and accidents.

The proposed TS 6.3 would state:

6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum qualifications of ANSI/ANS 3.1 of 1978 for comparable positions unless otherwise noted in the Technical Specifications.

Technicians and maintenance personnel who do not meet ANSI/ANS 3.1 of 1978, Section 4.5, are permitted to perform work for which qualification has been demonstrated.

6.3.2 The management position responsible for radiological controls shall meet or exceed the qualifications of Regulatory Guide 1.8 (Rev. 1-R, 9/75). Each other member of the radiation protection organization for which there is a comparable position described in ANSI N18.1-1971 shall meet or exceed the minimum qualifications specified therein, or in the case of radiation protection technicians, they shall have at least one year's continuous experience in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power stations and shall have been certified by the management position responsible for radiological controls as qualified to perform assigned functions. This certification must be based on an NRC approved, documented program consisting of classroom training with appropriate examinations and documented positive findings by responsible supervision that the individual has demonstrated his ability to perform each specified procedure and assigned function with an understanding of its basis and purpose.

6.3.3 The NRC approved training and retraining program for CERTIFIED FUEL HANDLERs shall be maintained.

3.3.1.1 NRC Staff Evaluation of TS 6.3.1 TS 6.3.1 establishes that each member of the unit staff meet or exceed the minimum qualification specified in ANSI/ANS 3.1 of 1978, with two exceptions. Following the submission to the NRC of the certifications required by 10 CFR 50.82(a)(1 ), OCNGS will no longer be required to have operators licensed pursuant to 10 CFR Part 55; therefore, exception (1) of the existing TS 6.3.1 would no longer apply and the requirement is proposed to be removed. The term "unit" is proposed to be changed to "facility" to be more appropriately represent the permanently defueled condition.

3.3.1.2 NRC Staff Evaluation of TS 6.3.2 TS 6.3.2 describes the qualifications for the radiological protection organization. There are no changes proposed to this section.

3.3.1.3 NRC Staff Evaluation of TS 6.3.3 TS 6.3.3 requires the STA to have a bachelor's degree or equivalent in a scientific or engineering discipline and specific, listed training. The STA position is required for a plant authorized for power operations; the STA position will no longer be required once OCNGS is permanently shut down and defueled and, thus, no longer authorized for power operations.

TS 6.3.3 is proposed to be changed to require that an NRG-approved training and retraining program for the CFHs be maintained. As approved by the NRC by letter dated September 6, 2016, the CFH Training and Retraining Program ensures that the qualifications of CFHs are commensurate with the tasks to be performed and the conditions requiring response. The OCNGS CFH Training and Retraining Program provides adequate confidence that appropriate training of personnel who will perform the duties of CFHs is conducted to ensure that the facility is maintained in a safe and stable condition.

3.3.1.4 NRC Staff Evaluation of Editorial Changes to TS 6.3 The licensee proposes to change "Facility Staff Qualifications" to "FACILITY STAFF QUALIFICATIONS." This change to all capital letters is editorial in nature. The TS 6.3 is currently located on TS page 6-2a and the licensee proposes to move it to TS page 6-2. This change is also editorial in nature.

3.

3.2 NRC Staff Conclusion

for TS 6.3 Once the certifications required by 10 CFR 50.82(a)(1) have been submitted to the NRC, certain requirements will no longer be applicable because the OCNGS 10 CFR Part 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel. The NRC staff finds that the proposed changes to TS 6.3 reflect the scope of the activities that would result from permanent cessation of operations and, therefore, are acceptable.

3.4 TS 6.4 3.4.1 TS 6.4 The current TS 6.4 is marked as deleted.

The proposed TS 6.4 would be marked as deleted.

3.4.2 NRC Staff Evaluation of TS 6.4 The existing TS 6.4 is proposed to be relocated for a change in pagination from TS page 6-3 to TS page 6-2.

3.

4.3 NRC Staff Conclusion

for TS 6.4 The proposed change is editorial in nature; therefore, the NRC staff finds the change to TS 6.4 acceptable.

3.5 TS 6.5 3.5.1 TS 6.5 The current TS 6.5 is marked as deleted.

The proposed TS 6.5 would be marked as deleted.

3.5.2 NRC Staff Evaluation of TS 6.5 The existing TS 6.5 is proposed to be relocated for a change in pagination from TS page 6-3 to TS page 6-2.

3.

5.3 NRC Staff Conclusion

for TS 6.5 The proposed change is editorial in nature; therefore, the NRC staff finds the change to TS 6.5 acceptable.

3.6 TS 6.6 - Reportable Event Action 3.6.1 TS 6.6 The current TS 6.6 states:

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

a. The Commission shall be notified and a report submitted pursuant to the requirements of Section 50. 73 to 10 CFR Part 50; and

b. Each REPORTABLE EVENT shall be reported to the cognizant manager and the cognizant department director and the Vice President - Oyster Creek. The functionally cognizant department staff shall prepare a Licensee Event Report (LER) in accordance with the guidance outlined in 10 CFR 50.73(b). Copies of all such reports shall be submitted to the functionally cognizant department director and the Vice President - Oyster Creek.

The proposed TS 6.6 would be marked as deleted.

3.6.2 NRC Staff Evaluation of TS 6.6 TS 6.6 specifies that LERs shall be submitted pursuant to the requirements of 10 CFR 50.73, "Licensee event report system." The actions of TS 6.6 are required by regulation and, therefore, it is not necessary to restate the requirements in the TSs. Existing licensee administrative procedures control the LER process and detail the required actions currently specified by TS 6.6. Therefore, TS 6.6 is proposed to be deleted and moved from TS page 6-9 to TS page 6-2. These proposed changes are considered to be administrative and editorial changes.

3.

6.3 NRC Staff Conclusion

for TS 6.6 Since the licensee is required to comply with 10 CFR 50. 73, the proposed changes to TS 6.6 are administrative and editorial in nature. Therefore, the NRC staff finds the changes to TS 6.6 acceptable.

3. 7 TS 6. 7 - Safety Limit Violation 3.7.1 TS 6.7 The current TS 6. 7 states:

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated:

a. If any Safety Limit is exceeded, the reactor shall be shut down immediately until the Commission authorizes the resumption of operation.

b. The Safety Limit violation shall be reported to the Commission and the Vice President-Oyster Creek.

c. A Safety Limit Violation Report shall be prepared. The report shall be submitted to the Vice President-Oyster Creek. This report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility components systems or structures, (3) corrective action taken to prevent recurrence.

d. The Safety Limit Violation Report shall be submitted to the Commission within ten days of the violation.

The proposed TS 6.7 would be marked as deleted.

3.7.2 NRC Staff Evaluation of TS 6.7 TS 6. 7 defines the requirements to immediately shut down the operating reactor if a safety limit is exceeded. Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been submitted to the NRC pursuant to 10 CFR 50.82(a)(1)(i) and (ii), the OCNGS will no longer authorize operation. Therefore, TS 6.7 is proposed to be deleted and moved from TS page 6-9 to TS page 6-2.

3.

7.3 NRC Staff Conclusion

for TS 6.7 Once the certifications required by 10 CFR 50.82(a)(1) have been submitted to the NRC, certain requirements will no longer be applicable because the OCNGS 10 CFR Part 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel. The NRC staff finds that the proposed changes to TS 6. 7 reflect the scope of the activities that would result from the permanent cessation of operations and, therefore, are acceptable.

3.8 TS 6.8 - Procedures and Programs 3.8.1 TS 6.8.1 The current TS 6.8.1 states:

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

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33 as referenced in the QATR.

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

c. Refueling 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.

j. Plant Staff Overtime pursuant to Technical Specification 6.2.2.2(i),

above.

The proposed TS 6.8.1 would state:

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 QATR.

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.

3.8.2 NRC Staff Evaluation of TS 6.8.1 The licensee proposed changes to modify the requirement for written procedures to be established, implemented, and maintained, to apply to procedures for the safe storage of nuclear fuel. The proposed reduction in scope of TS 6.8.1.a recognizes the reduced requirements associated with the protection of stored nuclear fuel, as opposed to the operation of the nuclear power plant, and appropriately reflects the permanently shut down and defueled status of the facility.

The proposed change to TS 6.8.1.c is to specify that procedures are required for fuel handling operations, rather than refueling operations, because refueling of the reactor will not be authorized by the 10 CFR Part 50 license once the certifications required by 10 CFR 50.82(a)(1) have been submitted. Procedures governing fuel handling operations provide the guidance necessary to ensure safe handling of spent fuel in the SFP and transfer from the SFP to dry fuel storage casks. Procedures governing responses to fuel handling accidents, personnel injuries, spent fuel pool events, and external events provide the necessary guidance to mitigate the consequences of such events.

The proposed change to TS 6.8.1.j would delete this specification that refers to TS 6.2.2.2(i),

which was deleted in license Amendment No. 27 4, dated March 23, 2009 (ADAMS Accession No. ML083500217). TS 6.8.1.j will be removed completely since it is the last item listed. This change is considered to be editorial in nature.

3.8.3 TS 6.8.3.b The current TS 6.8.3.b states:

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 unit systems or equipment, at least one of these individuals shall be a member of unit management or supervision holding a Senior Reactor Operator's License on the unit.

The proposed TS 6.8.3.b would state:

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.

3.8.4 NRC Staff Evaluation of TS 6.8.3.b The proposed change to TS 6.8.3.b changes one of the two approvers of temporary procedure changes that may affect the operational status of the facility systems or equipment to be a member of operations management or supervision who is a CERTIFIED FUEL HANDLER. The term "unit" is proposed to be changed to "facility" to more appropriately represent the permanently defueled condition. The requirement for the manager to hold a Senior Reactor Operator License is proposed to be removed and replaced by the requirement to be a CERTIFIED FUEL HANDLER. Once the certifications required by 10 CFR 50.82(a)(1) have been submitted to the NRC, the requirements of 10 CFR 50.54(m) will no longer be applicable, because the OCNGS Part 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor vessel. The submission of these certifications would also obviate the need for the operators' licenses specified in 10 CFR Part 55. Therefore, there is no longer a need for operations management staff to hold an SRO license. By letter dated September 6, 2016, the NRC approved the CFH Training and Retraining Program for OCNGS.

3.8.5 TS 6.8.4.a.9 The current TS 6.8.4.a.9 states:

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, The proposed TS 6.8.4.a.9 would state:

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, 3.8.6 NRC Staff Evaluation of TS 6.8.4.a.9 The proposed change to TS 6.8.4.a.9 would correct a pre-existing typographical error in that Iodine 131 (1-131) and Iodine 133 (1-133) were stated as "1-131, 1-133." This change is considered to be editorial in nature.

3.8.7 NRC Staff Evaluation of Editorial Changes to TS 6.8 TS page 6-10 is proposed to be changed to TS page 6-3. TS page 6-11 is proposed to be changed to TS page 6-4. TS page 6-12 is proposed to be changed to TS page 6-5. These changes are considered to be editorial in nature.

3.

8.9 NRC Staff Conclusion

for TS 6.8 Once the certifications required by 10 CFR 50.82(a)(1) have been submitted to the NRC, certain requirements will no longer be applicable because the OCNGS 10 CFR Part 50 license will no longer authorize operation of the reactor or emplacement or retention of fuel in the reactor

vessel. The NRC staff finds that the proposed changes to TS 6.8 reflect the scope of the activities that would result from the permanent cessation of operations and, therefore, are acceptable.

3.9 TS 6.9.1 - Routine Reports 3.9.1 TS 6.9.1 - Routine Reports The current heading for TS 6.9.1 is "ROUTINE REPORTS." The licensee proposed to change the heading of TS 6.9.1 to "Routine Reports." This change is considered editorial in nature.

3.9.1.1 NRC Staff Evaluation The heading is proposed to be changed from upper case letters to title case letters. This is an editorial change.

3.9.

1.2 NRC Staff Conclusion

for TS 6.9.1 Changing the upper case letters to title case letters is an editorial change; therefore, the NRC staff finds this change to TS 6.9.1 acceptable since it involves no technical changes.

3.9.2 TS 6.9.1.a - Startup Report The current TS 6.9.1.a states:

a. Startup Report. A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. The report shall address each of the tests identified in the FSAR and shall in general include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specified details required in license conditions based on other commitments shall be included in this report.

Startup reports shall be submitted within ( 1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of startup test program, and resumption or commencement of commercial power operation),

supplementary reports shall be submitted at least every three months until all three events have been completed.

The proposed TS 6.9.1.a would be marked as deleted.

3.9.2.1 NRC Staff Evaluation of TS 6.9.1.a TS 6.9.1.a describes the start-up report as a summary of plant startup and power escalation testing following the receipt of the operating license, increase in licensed power level, installation of nuclear fuel with a different design or manufacturer, and modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the unit. The proposed change is to mark this paragraph as deleted. Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel are submitted to the NRC, none of the criteria for a startup report will apply.

3.9.

2.2 NRC Staff Conclusion

for TS 6.9.1.a Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been submitted to the NRC pursuant to 10 CFR 50.82(a)(1 )(i) and (ii), the OCNGS license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. Therefore, the startup report criteria will no longer apply, and the NRC staff finds the proposed deletion of TS 6.9.1.a regarding startup reports acceptable.

3.9.3 TS 6.9.1.d - Radioactive Effluent Release Report The current TS 6.9.1.d states:

The Radioactive Effluent Release Report covering the operation of the unit 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 unit. The material provided shall be consistent with the objectives outlined in the ODCM [Offsite Dose Calculation Manual] and Process Control Program and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix I, Section IV.B.1.

The proposed TS 6.9.1.d would state:

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.

3.9.3.1 NRC Staff Evaluation of TS 6.9.1.d The term "unit" is proposed to be changed to "facility," to more appropriately represent the permanently shut down and defueled condition. This is an editorial change.

3.9.

3.2 NRC Staff Conclusion

for TS 6.9.1.d Changing the term "unit" to "facility" is an editorial change; therefore, the NRC staff finds this change to TS 6.9.1.d acceptable since it involves no technical changes.

3.9.4 TS 6.9.1.e -Annual Radiological Environmental Operating Report The current TS 6.9.1.e states:

The Annual Radiological Environmental Operating Report covering the operation of the unit 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.

The proposed TS 6.9.1.e would state:

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.

3.9.4.1 NRC Staff Evaluation of TS 6.9.1.e The term "unit" is proposed to be changed to "facility," to more appropriately represent the permanently shut down and defueled condition. This is an editorial change.

3.9.4.2. NRC Staff Conclusion for Technical Specification 6.9.1.e Changing the term "unit" to "facility" is an editorial change; therefore, the NRC staff finds this change to TS 6.9.1.e acceptable since it involves no technical changes.

3. 9. 5 TS 6. 9.1. f - Core Operating Limits Report (COLR)

The current TS 6.9.1.f states:

1. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle for the following:

a. The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.1 O.A

b. The K1 core flow adjustment factor for Specification 3.1 O.C.

c. The MINIMUM CRITICAL POWER RATIO (MCPR) for Specification 3.10. C.

d. The LOCAL LINEAR HEAT GENERATION RATE (LLHGR) for Specification 3.10. B.

e. The Average Power Range Monitor (APRM) stability protection settings for Specifications 2.3.A.1 and 2.3.B.

and shall be documented in the COLR.

2. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRG, specifically those described in the following documents.

a. GPU Nuclear (GPUN) Topical Report (TR) 020, Methods for the Analysis of Boiling Water Reactors Lattice Physics, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)

b. GPUN TR 021, Methods for the Analysis of Boiling Water Reactors Steady State Physics, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)

c. GPUN TR 033, Methods for the Generation of Core Kinetics Data for RETRAN-02, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)

d. GPUN TR 040, Steady-State and Quasi-Steady-State Methods Used in the Analysis of Accidents and Transients, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)

e. GPUN TR 045, BWR-2 Transient Analysis Model Using the Retran Code, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)

f. NEDE-31462P and NEDE-31462, Oyster Creek Nuclear Generating Station SAFER/CORECOOL/GESTR-LOCA Loss-of-Coolant Accident Analysis, (The approved revision at the time reload analyses are performed shall be identified in the COLR.)

g. NEDE-24011-P-A, General Electric Standard Application for Reactor Fuel, (GESTAR II) (The approved revision at the time reload analyses are performed shall be identified in the COLR.)

h. DELETED

i. XN-75-55-(A); XN-75-55, Supplement 1-(A); XN-75-55, Supplement 2-(A), Revision 2, "Exxon Nuclear Company

WREM-Based NJP-BWR ECCS Evaluation Model and Application to the Oyster Creek Plant," April 1977

j. XN-75-36(NP)-(A); XN-75-36(NP), Supplement 1-(A), "Spray Cooling Heat Transfer Phase Test Results, ENC- 8x8 BWR Fuel 60 and 63 Active Rods, Interim Report," October 1975

k. NEDC-33065P, Rev. 0, "Application of Stability Long-Term Solution Option II for Oyster Creek," April 2002.

3. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met.

4. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements shall be provided, upon issuance for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

Basis: 6.9.1.e - RELOCATED TO THE ODCM.

It is proposed that TS 6.9.1.f be deleted, except that the basis of 6.9.1.e, which was inadvertently placed under 6.9.1.f, will be relocated under TS 6.9.1.e.

3.9.5.1 NRC Staff Evaluation of TS 6.9.1.f TS 6.9.1.f describes the COLR that establishes prior to each reload cycle the reactor core operating limits. It is proposed that the COLR be deleted, because the COLR pertains only to an activity that does not apply in a permanently defueled condition (i.e., operations).

By letter dated December 13, 1993 (ADAMS Accession No. ML011200256), the NRC issued Amendment No. 166 that deleted portions of Radiological Effluent TSs and relocated them to controlled programs. Amendment No. 166 inadvertently placed the language "Basis: 6.9.1.e -

Relocated to the ODCM" under TS 6.9.1.f instead of under TS 6.9.1.e. This was an administrative error.

3.9.

5.2 NRC Staff Conclusion

for TS 6.9.1.f Once the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel have been submitted to the NRC pursuant to 10 CFR 50.82(a)(1 )(i) and (ii), the OCNGS license will no longer authorize operation of the reactor or emplacement or retention of fuel into the reactor vessel. Therefore, the COLR will no longer apply. Additionally, it is an editorial change to move "Basis: 6.9.1.e - Relocated to the ODCM" from TS 6.9.1.f to TS 6.9.1.e. The NRC staff finds these changes to TS 6.9.1.f acceptable.

3.10. TS 6.9.2 - Reportable Events 3.10.1 TS 6.9.2 The current TS 6.9.2 states:

The submittal of Licensee Event Reports shall be accomplished in accordance with the requirements set forth in 10 CFR 50. 73.

It is proposed that TS 6.9.2 be marked as deleted.

3.10.1.1 NRC Staff Evaluation of TS 6.9.2 TS 6.9.2 specifies that LERs shall be submitted pursuant to the requirements of 10 CFR 50.73.

The actions of TS 6.9.2 are required by regulation and it is not necessary to restate the requirements in TS 6.9.2.

3.10.

1.2 NRC Staff Conclusion

for TS 6.9.2 The actions of TS 6.9.2 are required by regulation and it is not necessary to restate the requirements in TS 6.9.2. Deleting TS 6.9.2 is an administrative change and the NRC staff finds the change acceptable since the licensee is required to follow 10 CFR 50. 73.

3.11 TS 6.9.3 - Unique Reporting Requirements The current heading for TS 6.9.3 is "UNIQUE REPORTING REQUIREMENTS." The proposed change for TS 6.9.3 is "Unique Reporting Requirements."

3.9.11.1 NRC Staff Evaluation of TS 6.9.3 The heading is proposed to be changed from upper case letters to title case letters. This is an editorial change.

3.9.1

1.2 NRC Staff Conclusion

for TS 6.9.3 Changing the upper case letters to title case letters is an editorial change; therefore, the NRC staff finds this change to TS 6.9.3 acceptable since it involves no technical changes.

3.12 TS 6.21 - Technical Specifications (TS) Bases Control Program The current heading for TS 6.21 is 'Technical Specifications (TS) Bases Control Program." The proposed change for TS 6.21 heading is "TECHNICAL SPECIFICATIONS (TS) BASES CONTROL PROGRAM."

3.12.1 NRC Staff Evaluation of 6.21 The heading is proposed to be changed from title case letters to upper case letters. This is an editorial change.

3.1

2.2 NRC Staff Conclusion

for TS 6.21 Changing the title case letters to upper case letters is an editorial change; therefore, the NRC staff finds this change to TS 6.21 acceptable since it involves no technical changes.

3.13 TS 6.22 - Control Room Envelope Habitability Program The current heading for TS 6.22 is "Control Room Envelope Habitability Program." The proposed change for TS 6.22 heading is "CONTROL ROOM ENVELOPE HABITABILITY PROGRAM."

3.13.1 NRC Staff Evaluation of TS 6.22 The heading is proposed to be changed from title case letters to upper case letters. This is an editorial change.

3.1

3.2 NRC Staff Conclusion

for TS 6.22 Changing the title case letters to upper case letters is an editorial change; therefore, the NRC staff finds this change to TS 6.22 acceptable since it involves no technical changes.

3.14 TS 6.23- Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

The current heading for TS 6.23 is "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)." The proposed change for TS 6.23 heading is "REACTOR COOLANT SYSTEM (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)."

3.14.1 NRC Staff Evaluation of TS 6.23 A portion of the heading is proposed to be changed from title case letters to upper case letters.

This is an editorial change.

3.1

4.2 NRC Staff Conclusion

for TS 6.23 Changing the title case letters to upper case letters is an editorial change; therefore, the NRC staff finds this change to TS 6.23 acceptable since it involves no technical changes.

3.15 TS 1 - Definitions The licensee proposes to add to new definitions: 1.52 (Certified Fuel Handler) and 1.53 (Non-Certified Operator). The licensee proposes the following definition 1.52:

1.52 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.

The licensee proposes the following definition 1.53:

1.53 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.

3.15.1 NRC Staff Evaluation of TS 1 The licensee proposed to modify TS 1, to include new definitions for a CFH and a Non-Certified Operator. By letter dated September 6, 2016, the NRC approved the CFH Training and Retraining Program for OCNGS. The NRC staff reviewed the proposed definition for a CFH and finds that it is acceptable because it conforms to the usage contained in the Administrative Controls section of the OCNGS TS.

The NRC staff reviewed the proposed definition for a Non-Certified Operator and finds that it is acceptable, because it conforms to the usage contained in the Administrative Controls section of the OCNGS TS.

3.1

5.2 NRC Staff Conclusion

for TS 1 Based on its evaluation, the NRC staff finds the addition of definitions 1.52 and 1.53 acceptable.

3.16 TS Table of Contents Pages ii and iii The licensee proposed to add the TS definitions 1.52 and 1.53 to TS table of contents page ii.

The licensee proposed conforming administrative changes to TS table of contents page iii for the proposed deletions and condensing of pages in TS Section 6.

3.16.1 NRC Staff Evaluation of TS Table of Contents Pages ii and iii The proposed changes to TS table of contents pages ii and iii are editorial in nature.

3.1

6.2 NRC Staff Conclusion

for TS Table of Contents Pages ii and iii The changes to TS table of contents pages ii and iii are editorial in nature; therefore, the NRC staff finds these changes to TS table of contents pages ii and iii acceptable since they involve no technical changes.

3.17 TS Page Numbering Due to the above TS changes, the licensee proposed to change the TS page numbering.

TS page 6-13 will become TS page 6-6. TS pages 6-14, 6-14a, and 6-15 will become TS page 6-6. TS page 6-16 will become TS page 6-7. TS page 6-17 will become TS page 6-8.

TS page 6-18 will become TS page 6-9. TS page 6-19 will become TS page 6-10.

TS page 6-20 will become TS page 6-11. TS page 6-21 will become 6-12. TS page 6-22 will become TS page 6-13.

3.17.1 NRC Staff Evaluation of TS Page Numbering The proposed renumbering and deletion of TS page numbers is an editorial change.

3.1

7.2 NRC Staff Conclusion

for TS Page Numbering Renumbering and deleting TS page numbers is an editorial change; therefore, the NRC staff finds these changes to TS page numbering acceptable since they involve no technical changes.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the New Jersey State official was notified of the proposed issuance of the amendment on September 19, 2016. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendment changes requirements with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20. 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 as published in the Federal Register on November 22, 2016 (81 FR 83876). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9).

Pursuant to 10 CFR 51.22(b}, no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

6.0 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 Contributor: J. Lamb Date: March 7, 2017

ML16235A413 *via email OFFICE NRR/DORL/LSPB/PM NRR/DORL/LSPB/LAiT NRR/DORL/LSPB/LA NAME JLamb IBetts JBurkhardt DATE 11/18/16 02/16/17 02/21/17 OFFICE NRR/DRA/AHPB/BC NRR/DSS/STSB/BC NSIR/DPR/RLB/BC*

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