ML14097A145

Nuclear Generating Plant - Issuance of Amendment No. 244, Revise Technical Specifications to Remove Certain Requirements from Section 5, Administrative Controls

ML14097A145
Person / Time
Site:	Crystal River
Issue date:	07/11/2014
From:	Gratton C Plant Licensing Branch IV
To:	Hobbs T Duke Energy Florida
Gratton C, NRR/DORL/LPLIV-2
References
TAC MF1504
Download: ML14097A145 (35)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 July 11, 2014 Mr. Terry D. Hobbs Decommissioning Director Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, FL 34428-6708

SUBJECT:

CRYSTAL RIVER UNIT 3- ISSUANCE OF AMENDMENT TO THE FACILITY OPERATING LICENSE REGARDING CHANGES TO THE ADMINISTRATIVE CONTROLS SECTION OF THE TECHNICAL SPECIFICATIONS (TAC NO.

MF1504)

Dear Mr. Hobbs:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 244 to Facility Operating License No. DPR-72 for Crystal River Unit 3 Nuclear Generating Plant (CR-3) in response to your letter dated April 25, 2013, as supplemented by letters dated September 4, 2013, and February 26, 2014. The amendment revises and removes certain requirements from the Section 5.0, "Administrative Controls," portions of the CR-3 Technical Specifications that are no longer applicable to the facility in its permanently defueled condition.

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

Sincerely, Christopher Gratton, Senior Project Manager Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-302

Enclosures:

1. Amendment No. 244 to DPR-72

2. Safety Evaluation cc w/enclosures: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY FLORIDA, INC.

CITY OF ALACHUA CITY OF BUSHNELL CITY OF GAINESVILLE CITY OF KISSIMMEE CITY OF LEESBURG CITY OF NEW SMYRNA BEACH AND UTILITIES COMMISSION CITY OF NEW SMYRNA BEACH CITY OF OCALA ORLANDO UTILITIES COMMISSION AND CITY OF ORLANDO SEMINOLE ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 244 License No. DPR-72

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

A. The application for amendment by Duke Energy Florida, Inc., et al. (the licensees), dated April 25, 2013, as supplemented by letters dated September 4, 2013, and February 26, 2014, 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 Facility Operating License and Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-72 is hereby amended to read as follows:

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 244 are hereby incorporated in the license. Duke Energy Florida, Inc. shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Douglas A. Broaddus, Chief Plant Licensing IV-2 and Decommissioning Transition Branch Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: July 11, 2014

ATTACHMENT TO LICENSE AMENDMENT NO. 244 FACILITY OPERATING LICENSE NO. DPR-72 DOCKET NO. 50-302 Replace the following page of Facility Operating License No. DPR-72 with the attached revised page. The revised page is identified by amendment number and contains a vertical line indicating the areas of change.

Remove 4 4 Replace the following pages of the Appendix "A" Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.

Remove Insert Remove 5.0-1 5.0-1 5.0-22 5.0-2 5.0-2 5.0-23 5.0-3 5.0-3 5.0-23A 5.0-4 5.0-4 5.0-238 5.0-7 5.0-7 5.0-24 5.0-9 5.0-9 5.0-25 5.0-10 5.0-10 5.0-26 5.0-11 5.0-11 5.0-27 5.0-12 5.0-12 5.0-28 5.0-13 5.0-13 5.0-29 5.0-14 5.0-14 5.0-30 5.0-15 5.0-15 5.0-31 5.0-16 5.0-16 5.0-17 5.0-17 5.0-18 5.0-18 5.0-19 5.0-19 5.0-20 5.0-21

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:

2.C.(1) Maximum Power Level Duke Energy Florida, Inc. is authorized to operate the facility at a steady state reactor core power level not in excess of 2609 Megawatts (1 00 percent of rated core power level).

2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 244, are hereby incorporated in the license. Duke Energy Florida, Inc. shall operate the facility in accordance with the Technical Specifications.

The Surveillance Requirements contained in the Appendix A Technical Specifications and listed below are not required to be performed immediately upon implementation of Amendment 149. The Surveillance Requirements shall be successfully demonstrated prior to the time and condition specified below for each.

a) SR 3.3.8.2.b shall be successfully demonstrated prior to entering MODE 4 on the first plant start-up following Refuel Outage 9.

b) SR 3.3.11.2, Function 2, shall be successfully demonstrated no later than 31 days following the implementation date of the ITS.

c) SR 3.3.17.1, Functions 1, 2, 6, 10, 14, & 17 shall be successfully demonstrated no later than 31 days following the implementation date of the ITS.

d) SR 3.3.17 .2, Function 10 shall be successfully demonstrated prior to entering MODE 3 on the first plant start-up following Refuel Outage 9.

e) SR 3.6.1.2 shall be successfully demonstrated prior to entering MODE 2 on the first plant start-up following Refuel Outage 9.

f) SR 3. 7 .12.2 shall be successfully demonstrated prior to entering MODE 2 on the first plant start-up following Refuel Outage 9.

g) SR 3.8.1.1 0 shall be successfully demonstrated prior to entering MODE 2 on the first plant start-up following Refuel Outage 9.

h) SR 3.8.3.3 shall be successfully demonstrated prior to entering MODE 4 on the first plant start-up following Refuel Outage 9.

Facility Operating License No. DPR-72 Amendment No. 244

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

The Plant Manager or his designee shall approve, prior to implementation, each proposed test, experiment or modifications to systems or equipment that affect stored nuclear fuel.

5.1.2 The Shift Supervisor shall be responsible for the shift command function.

Crystal River Unit 3 5.0-1 Amendment No. 244

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively. The onsite and offsite organizations shall include the positions responsible for activities affecting the safe handling and storage of nuclear fuel.

a. Lines of authority, responsibility, and communications shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of department responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These shall be documented in the FSAR;

b. The Decommissioning Director shall have overall responsibility for the safe handling and storage of nuclear fuel and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure the safe handling and storage of nuclear fuel.

The Plant Manager shall be responsible to control those onsite activities necessary for the safe handling and storage of nuclear fuel; and

c. The individuals who train the Certified Fuel Handlers, carry out health physics or perform quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

5.2.2 Unit Staff The unit staff organization shall include the following:

a. Each duty shift shall be composed of at least one Shift Supervisor and one Non-certified Operator.

b. Shift crew composition may be less than the minimum requirement of 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

(continued)

Crystal River Unit 3 5.0-2 Amendment No. 244

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued)

c. At least one person qualified to stand watch in the control room (Non-certified Operator or Certified Fuel Handler) shall be present in the control room when nuclear fuel is stored in the spent fuel pools.

d. An individual qualified in Radiation Protection procedures shall be on site during fuel handling operations and during movement of heavy loads over the fuel storage racks. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

e. Oversight of fuel handling operations shall be provided by a Certified Fuel Handler.

f. The Shift Supervisor shall be a Certified Fuel Handler.

Crystal River Unit 3 5.0-3 Amendment No. 244

Unit Staff Qualifications 5.3 5.0 ADMINISTRATIVE CONTROLS 5.3 Unit Staff Qualifications 5.3.1 Each member of the unit staff shall meet or exceed the m1n1mum qualifications of ANSI N18.1, 1971 for comparable positions, except for the Radiation Protection Manager, who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975.

5.3.2 A training and retraining program for the Certified Fuel Handler positions shall be maintained under the direction of the Plant Manager.

Crystal River Unit 3 5.0-4 Amendment No. 244

Procedures, Programs, and Manuals 5.6 5.0 ADMINISTRATIVE CONTROLS 5.6 Procedures, Programs, and Manuals 5.6.1 Procedures 5.6.1.1 Scope Written procedures shall be established, implemented, and maintained covering the following activities:

a. The procedures applicable to the safe storage of nuclear fuel recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978;

b. Quality assurance for effluent and environmental monitoring;

c. Fire Protection Program implementation; and

d. All programs specified in Specification 5.6.2.

5.6.2 Programs and Manuals The following programs shall be established, implemented, and maintained. Programs and Manuals may be titled as Reports.

5.6.2.1 Not Used 5.6.2.2 Not Used 5.6.2.3 Offsite Dose Calculation Manual (ODCM):

This Manual contains offsite dose calculation methodologies, the radioactive effluent controls program, and radiological environmental monitoring activities. The ODCM shall contain:

1. The methodologies and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents;

2. The methodologies and parameters used in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints;

3. The controls for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable in accordance with 10 CFR 50.36a. These include:

(continued)

Crystal River Unit 3 5.0-7 Amendment No.244

Procedures, Programs and Manuals S.6 S.6 Procedures, Programs and Manuals S.6.2.3 ODCM (continued)

2. For Iodine-131, Iodine-133, tr1t1um, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1SOO mrems/yr to any organ;

h. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR SO, Appendix I;

i. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR SO, Appendix I; and

j. Limitations on the annual dose or dose commitment to any member of the public beyond the site boundary due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40 CFR 190.

Licensee Initiated Changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change(s), and

b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR S0.36a, and Appendix I to 10 CFR Part SO and not adversely impact the accuracy or reliability of effluent dose, or setpoint calculations.

2. Shall become effective after review and acceptance by the on-site review function and the approval of the Plant Manager; and (continued)

Crystal River Unit 3 5.0-9 Amendment No. 244

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.3 ODCM (continued)

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

5.6.2.4 Not Used

5. 6. 2. 5 Not Used 5.6.2.6 Not Used 5.6.2.7 Not Used 5.6.2.8 Not Used 5.6.2.9 Not Used 5.6.2.10 Not Used 5.6.2.11 Not Used (continued)

Crystal River Unit 3 5.0-10 Amendment No.244

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.12 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of the Control Room Emergency Ventilation System (CREVS) per the requirements specified in Regulatory Guide 1.52, Revision 2, 1978, and/or as specified herein, and in accordance with ANSI N510-1975 and ASTM D 3803-89 (Re-approved 1995).

a. Demonstrate for each train of the CREVS that an inplace test of the high efficiency particulate air (HEPA) filters shows a penetration < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, 1978, and in accordance with ANSI N510-1975 at the system flowrate of between 37,800 and 47,850 cfm.

b. Demonstrate for each train of the CREVS that an inplace test of the carbon adsorber shows a system bypass < 0.05% when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 at the system flowrate of between 37,800 and 47,850 cfm.

c. Demonstrate for each train of the CREVS that a laboratory test of a sample of the carbon adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, 1978, meets the laboratory testing criteria of ASTM D 3803-89 (Re-approved 1995) at a temperature of 30°( and relative humidity of 95% with methyl iodide penetration of less than 5.0%.

d. Demonstrate for each train of CREVS that the pressure drop across the combined roughing filters, HEPA filters and the carbon adsorbers is ~ AP=4 water gauge when tested in accordance with Regulatory Guide 1.52, Revision 2, 1978, and ANSI N510-1975 at the system flowrate of between 37,800 and 47,850 cfm.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test frequencies.

5.6.2.13 Not Used (continued)

Crystal River Unit 3 5.0-11 Amendment No. 244

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.14 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, in accordance with applicable ASTM Standards.

The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has the following properties within limits of ASTM D 975 for Grade No. 2-D fuel oil:

1. Kinematic Viscosity,

2. Water and Sediment,

3. Flash Point,

4. Specific Gravity API;

b. Other properties of ASTM D 975 for Grade No. 2-D fuel oil are within limits within 92 days following sampling and addition of new fuel to storage tanks.

c. Total particulate contamination of stored fuel oil is < 10 mg/L when tested once per 92 days in accordance with ASTM D 2276-91 (gravimetric method).

5.6.2.15 Not Used (continued)

Crystal River Unit 3 5.0-12 Amendment No. 244

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.16 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.

The SFDP shall contain the following:

a. Provisions for cross train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;

b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;

c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to the system(s) supported by the inoperable support system is also inoperable); or

b. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

c. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable.

(continued)

Crystal River Unit 3 5.0-13 Amendment No. 244

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.16 SFDP (continued)

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

5.6.2.17 Technical Specifications (TS) Bases Control Program Changes to the Bases of the TS shall be made under appropriate administrative controls and reviewed according to the review process specified in the Quality Assurance Plan.

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

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

b. A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.

The Bases Control Program shall contain prov1s1ons to ensure that the Bases are maintained consistent with the FSAR.

Proposed changes that meet the criteria of Specification 5.6.2.17.a or Specification 5.6.2.17.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71.

5.6.2.18 Not Used 5.6.2.19 Not Used 5.6.2.20 Not Used Crystal River Unit 3 5.0-14 Amendment No.244

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.21 Control Complex Habitability Envelope Integrity Program A Control Complex Habitability Envelope Integrity Program shall be established and implemented to ensure that CCHE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CCHE 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 challenge from smoke. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CCHE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements.

1. The definition of the CCHE and the CCHE boundary.

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

3. Requirements for (i) determining the unfiltered air in-leakage past the CCHE boundary into the CCHE 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 CCHE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

4. The Control Complex Habitability Envelope Integrity Program will be used to verify the integrity of the Control Complex boundary. Conditions that are identified to be adverse shall be trended and used as part of the 24 month assessment of the CCHE boundary.

5. The quantitative limits on unfiltered air in-leakage into the CCHE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air in-leakage measured by the testing described in paragraph 3. The unfiltered air in-leakage limit for radiological challenges is the in-leakage flow rate assumed in the licensing basis analyses of DBA consequences.

Unfiltered air in-leakage limits for hazardous chemicals and smoke must ensure that exposure of CCHE occupants to these hazards will be within the assumptions in the licensing basis.

6. The provisions of SR 3.0.2 are applicable to the Frequencies for assessing CCHE habitability, determining CCHE unfiltered in-leakage as required by paragraph 3.

Crystal River Unit 3 5.0-15 Amendment No. 244

Reporting Requirements 5.7 5.0 ADMINISTRATIVE CONTROLS 5.7 Reporting Requirements 5.7.1 Routine Reports 5.7.1.1 Reports required on an annual basis include:

a. Not Used

b. Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year.

The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM).

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

c. Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted prior to May 1 of each year, and in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. 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 50, Appendix I, Section IV 8.1.

(continued)

Crystal River Unit 3 5.0-16 Amendment No.244

Reporting Requirements 5.7 5.7 Reporting Requirements 5.7.1.2 Not Used 5.7.2 Special Reports Special Reports shall be submitted in accordance with 10 CFR 50.4 within the time period specified for each report.

The following Special Reports shall be submitted:

a. When a Special Report is required by Condition B or F of LCO 3.3.17, "Post Accident Monitoring (PAM)

Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

(continued)

Crystal River Unit 3 5.0-17 Amendment No. 244

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), alternative methods are used to control access to high radiation areas. Each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation (measured at 30 em) is > 100 mrem/hr but < 1000 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).

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

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

b. A radiation monitoring device that continuously integrates the radiation dose in the area and alarms when a preset 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 are aware of them.

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance.

5.8.2 In addition to the requirements of Specification 5.8.1, areas with radiation levels ~ 1000 mrem/hr at 30 em shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor or health physics superv1s1on. Doors shall remain locked except during periods of access by personnel.

Direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

(continued)

Crystal River Unit 3 5.0-18 Amendment No. 244

High Radiation Area 5.8 5.8 High Radiation Area (continued) 5.8.3 For individual high radiation areas with radiation levels of

> 1000 mrem/hr at 30 em, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that are not be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

Crystal River Unit 3 5.0-19 Amendment No. 244

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 RELATED TO AMENDMENT NO. 244 TO FACILITY OPERATING LICENSE NO. DPR-72 DUKE ENERGY FLORIDA, INC., ET AL.

CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT DOCKET NO. 50-302

1.0 INTRODUCTION

By application dated April 25, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13128A286), as supplemented by letters dated September 4, 2013, and February 26, 2014 (ADAMS Accession Nos. ML13255A056 and ML14064A343, respectively), Duke Energy Florida, Inc., et al. (the licensee), requested changes to the Technical Specifications (TSs) for Crystal River Unit 3 Nuclear Generating Plant (CR-3).

The supplemental letter dated September 4, 2013, expanded the scope of the application as originally noticed. Therefore, the U.S. Nuclear Regulatory Commission (NRC) staff re-noticed the application, and included a revised proposed no significant hazards consideration determination in the Federal Register on November 12, 2013 (78 FR 67406). The supplemental letter dated February 26, 2014, provided additional information that clarified the supplement dated September 4, 2013, did not expand the scope of the application as noticed on November 12, 2013, and did not change the NRC staff's proposed no significant hazards consideration determination as published in the Federal Register on November 12, 2013.

On February 20, 2013 (ADAMS Accession No. ML13056A005), the licensee submitted certifications under Title 10 of the Code of Federal Regulations (1 0 CFR), Section 50.82(a) that it had permanently ceased operation of the CR-3 reactor and that it had permanently moved all fuel to the spent fuel pool (SFP). The changes proposed in the September 4, 2014, application would revise and remove certain requirements from the Section 5.0, "Administrative Controls,"

portions of the CR-3 Technical Specifications that are no longer applicable to the facility in its permanently shut down and defueled condition.

2.0 REGULATORY EVALUATION

The following regulatory requirements and guidance were considered by the staff in its review of the license amendment request:

2.1 Title 10 of the Code of Federal Regulations (1 0 CFR), Section 50.120, "Training and qualification of nuclear power plant personnel." This regulation requires the use of a Enclosure 2

Systems Approach to Training (SAT) for personnel positions, including Certified Fuel Handlers (CFHs).

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

the specific sub-chapter considered in this review was Chapter 13.2.2, "Non-Licensed Plant Staff Training," Revision 3, March 2007.

2.3 NUREG-1625, "Proposed Standard Technical Specifications for Permanently Defueled Westinghouse Plants," Draft Report for Comment, March 1998. This NUREG provides examples of TSs for defueled plants.

2.4 Section 55.4 of 10 CFR, "Definitions," defines a SAT to mean a training program that includes the elements of job analysis, learning objectives derived from job analysis, training design and implementation based on the learning objectives, evaluation of trainee mastery of the learning objectives, and program revision based the performance of trained personnel in the job setting.

2.5 Inspection Procedure 88010, "Operator Training/Retraining," which contains the NRC inspection guidance to determine whether the licensee is complying with the regulations and license requirements related to the training of licensee employees and other personnel and is implementing an adequate training program.

3.0 TECHNICAL EVALUATION

The NRC staff reviewed the proposed changes to Section 5.0 of the CR-3 TSs, using the guidance in NUREG-1625, "Proposed Standard Technical Specifications for Permanently Defueled Westinghouse Plants," as appropriate and applicable, in the following areas:

Responsibility; Organization; Procedures, Programs, and Manuals; Reporting Requirements; and High Radiation Area, and provides its evaluation of each change in the following sections.

3.1 TS 5.1: Responsibility

3. 1.1 TS 5.1.1 The current TS states that the Plant General Manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

In addition, the Plant General Manager or his designee shall approve, prior to implementation, each proposed test, experiment or modifications to systems or equipment that affect nuclear safety.

The licensee proposed the following changes to TS 5.1.1 to address the permanently defueled and decommissioning status of the facility:

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

The Plant Manager or his designee shall approve, prior to implementation, each proposed test, experiment or modifications to systems or equipment that affect stored nuclear fuel.

The NRC staff reviewed the proposed changes to TS 5.1.1 and found that the changes modified the position title from "Plant General Manager" to "Plant Manager," and the scope of the position responsibility from "unit operation" to "facility operation," and from "the effect on nuclear safety" to "the effect on stored nuclear fuel." The staff finds that these changes are acceptable because the position title change is editorial in nature, the overall management responsibilities are unchanged, and the description of the plant and the related responsibilities of the plant staff discussed in this technical specification have been updated to reflect that the reactor has been permanently shut down.

3.1.2 TS 5.1.2 TS 5.1.2 identifies the responsibilities for the control room command function for the various operating MODES, and describes delegation of authority requirements for the position. The current TS states that the Control Room Supervisor shall be responsible for the control room command function. During any absence of the Control Room Supervisor from the control room while the unit is in MODE 1, 2, 3, or 4, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. During any absence of the Control Room Supervisor from the control room while the unit is in MODE 5 or 6, an individual with an active SRO license or Reactor Operator license shall be designated to assume the control room command function.

The licensee proposes to change TS 5.1.2 to state:

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

As part of the amendment request, the licensee proposed to eliminate the MODE dependency of the shift command function delegation requirements currently in the CR-3 TSs due to the permanently shutdown and defueled condition of the reactor. The NRC staff found that the CR-3 TSs define "MODE" as corresponding to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1. Table 1.1-1 only applies with fuel in the reactor vessel, therefore, due to the permanent defueled state of the reactor, operating modes are no longer applicable to CR-3, and should not be used to determine delegation of authority decisions.

The licensee also proposed to change the title of the person responsible for the command of the shift from "Control Room Supervisor" to "Shift Supervisor" and the command function from "control room command function" to "shift command function." These changes are editorial and are acceptable.

The proposed change also eliminates the requirements regarding delegation of authority of the shift command function when the Shift Supervisor leaves the control room. Due to the decision to permanently cease operation of the unit and to permanently transfer the spent fuel to the SFP, all of the stored fuel will reside in the SFP. Because of the robust design of the SFP and

the low decay heat load of the stored fuel, events involving the SFP are expected to evolve slowly and, while the control room would continue to be manned with qualified staff consistent with proposed TS 5.2.2, continuous manning of the control room by the Shift Supervisor would not be necessary to protect the environment and the health and safety of the public. Consistent with this shift command philosophy, delegation of the command function during absence of the Shift Supervisor from the control room would not be required.

Based on the reasons stated above, the NRC staff finds the changes toTS 5.1.2 acceptable.

3.2 TS 5.2: Organization 3.2.1 TS 5.2.1 - Onsite and Offsite Organizations The current TS states, in part, in the introduction to TS 5.2.1 that "[T)he onsite and offsite organizations shall include the positions responsible for activities affecting safety of the nuclear power plant."

The licensee proposed the following changes to the TS 5.2.1 opening paragraph:

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

The NRC staff reviewed the introduction toTS 5.2.1 and found that it has been changed to reflect that the plant is permanently defueled. As such, the introduction has been modified to be appropriate for activities associated with a decommissioning reactor. The NRC staff has reviewed the changes and finds them to be acceptable based on the plant's permanently defueled status.

3.2.2 TS 5.2.1.b The current TS states that the "Vice President - Crystal River Nuclear Plant shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure nuclear safety. The Vice President- Crystal River Nuclear Plant shall be responsible for the overall safe operation of the plant and shall have control over those on site activities necessary for the safe operation and maintenance of the plant."

The licensee proposed the following changes toTS 5.2.1.b:

The Decommissioning Director shall have overall responsibility for the safe handling and storage of nuclear fuel and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure the safe handling and storage of nuclear fuel. The Plant Manager shall be responsible to control those onsite activities necessary for the safe handling and storage of nuclear fuel

The NRC staff reviewed the proposed changes toTS Section 5.2.1.b and finds that the changes for the responsible staff continue to define the responsibility for safety of the facility, both for overall management and for onsite activities. In addition, the proposed changes appropriately reflect the decommissioning status of the plant. The proposed changes, in combination with the changes proposed in TS 5.2.2.a- c, are consistent with NUREG-1625 regarding the designation of staff responsible for spent fuel safety and the control of onsite activities. Therefore, the staff finds that these changes are acceptable.

3.2.3 TS 5.2.1.c The current TS states that the individuals who train the operating staff carry out health physics or perform quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

The licensee proposed the following changes to TS 5.2.1 c:

The individuals who train the Certified Fuel Handlers, carry out health physics or perform quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their ability to perform their assigned functions.

The purpose of TS 5.2.1.c is to ensure that staff who train critical staff members, carry out health physics or perform quality assurance functions, can do so without being subject to pressure from schedule or budget. Prior to decommissioning, this included operations staff trainers. The licensee proposed to change the TS from "operating staff' to "Certified Fuel Handlers," and from "independence from operating pressures" to "perform their assigned functions," to reflect the decommissioning status of the plant. The NRC staff finds that this is acceptable because the changes more accurately describe the requirements as a result of the plant's decommissioning status.

3.2.4 TS 5.2.2.a through c- Unit Staff The current TSs 5.2.2.a - c discuss requirements for unit staff organization. Sub-paragraph a.

requires that one auxiliary nuclear operator be assigned to the operating shift any time there is fuel in the reactor and an additional auxiliary nuclear operator be assigned in MODES 1, 2, 3, and 4. Sub-paragraph b. states that shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and TS 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. Finally, sub-paragraph c. states that at least one licensed reactor operator shall be present in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed SRO shall be present in the control room. The proposed changes modify these requirements to reflect the decommissioning status of the unit.

The licensee proposed to modify TS 5.2.2.a - cas follows:

a. Each duty shift shall be composed of at least one Shift Supervisor and one Non-certified Operator.

b. Shift crew composition may be less than the minimum requirement of 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

c. At least one person qualified to stand watch in the control room (Non-certified Operator or Certified Fuel Handler) shall be present in the control room when nuclear fuel is stored in the spent fuel pools.

The NRC staff reviewed the proposed changes toTS 5.2.2.a, b, and c that reflect the scope of activities resulting from the permanent cessation of operations. With the certifications submitted in accordance with 10 CFR 50.82(a), the licensee is no longer authorized to operate the reactor or load fuel into the reactor vessel. As discussed in NRC letter dated May 21, 2014 (ADAMS Accession No. ML14127A340), the requirements of 10 CFR 50.54(m) requiring licensed operator staffing under various MODES of operation no longer apply to facilities that have submitted certifications in accordance with 10 CFR 50.82(a). Therefore, proposed changes that remove requirements for licensed operators are acceptable, In addition, as discussed in Section 3.1.2 of this safety evaluation, operational MODES are also no longer applicable to reactors that have permanently removed fuel from the reactor vessel.

The NRC staff, therefore, finds that proposed changes removing MODE dependent actions are also acceptable.

The NRC staff found that the proposed minimum crew compliment is consistent with shift manning requirements for permanently shutdown sites with single SFPs. For the reasons discussed above, the NRC staff finds that the proposed changes toTS 5.2.2.a- c. are acceptable.

3.2.5 TS 5.2.2.d TS 5.2.2.d currently states that an individual qualified in Radiation Protection procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

The licensee proposed the following changes toTS 5.2.2.d:

An individual qualified in Radiation Protection procedures shall be on site during fuel handling operations and during movement of heavy loads over the fuel storage racks. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to provide for

unexpected absence, provided immediate action is taken to fill the required position.

The licensee modified TS 5.2.2.d to reflect those activities requiring Radiation Protection staffing following the permanent shutdown of the reactor. Following certification of permanent removal of fuel to the SFP, fuel can no longer reside in the reactor. The TS has been modified to reflect those remaining activities where individuals qualified in Radiation Protection procedures are required to be present. The NRC staff reviewed these changes and found them to be acceptable.

3.2.6 TS 5.2.2.e- f:

The licensee proposed two new TSs under TS 5.2.2, "Unit Staff':

TS 5.2.2.e. Oversight of fuel handling operations shall be provided by a Certified Fuel Handler TS 5.2.2.f. The Shift Supervisor shall be a Certified Fuel Handler The new TSs proposed by the licensee establish the qualification requirements for unit staff having oversight of fuel handling operations and shift management responsibilities. Unit staff qualified as CFHs have completed training in the safe conduct of decommissioning activities, safe handling and storage of spent fuel, and the appropriate response to plant emergencies. The NRC staff reviewed the proposed TSs and found that TS 5.2.2.e- f establish appropriate minimum qualification requirements for the Shift Supervisor, and for the unit staff position that is responsible for overseeing fuel handling operations. Therefore, the NRC staff finds that these changes are acceptable.

3.3 TS 5.3: Unit Staff Qualifications 3.3.1 TS 5.3.1 The current requirements in TS 5.3.1 state that each member of the unit staff shall meet or exceed the minimum qualifications of American National Standards Institute document ANSI N18.1, 1971, for comparable positions, except for the Radiation Protection Manager, who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and the Shift Technical Advisor (STA) who shall have a Bachelor's degree, or the equivalent, in a scientific or engineering discipline with specific training in plant design and response and analysis of the plant transients and accidents.

The licensee proposed to delete the requirements related to the STA position due to the permanent shutdown of the reactor.

The NRC staff reviewed the proposed changes toTS 5.3.1 and finds that the STA position is only needed for power operations. Because the licensee has certified that it will no longer operate the unit, this position is no longer required and the qualification requirements for the STA position can be deleted. Therefore, the staff finds that this change is acceptable.

3.3.2 TS 5.3.2 The licensee proposed adding a newTS 5.3.2 requiring that a training and retraining program for the CFH positions be maintained under the direction of the Plant Manager.

As defined in 10 CFR 50.2, a CFH is a non-licensed operator that has been qualified in accordance with an NRC-approved CFH training program. The CR-3 CFH training and retraining program was previously approved by the NRC staff on June 26, 2014 (ADAMS Accession No. ML14155A181). By establishing this newTS, the NRC staff finds that the new TS 5.3.2 appropriately requires the establishment and maintenance of a program to train and retrain CFHs for positions requiring such qualification that are specified in the TSs. Therefore, the NRC staff finds that TS 5.3.2 is acceptable.

3.4 TS 5.6: Procedures, Programs, and Manuals 3.4.1 TS 5.6.1: Procedures Current TS 5.6.1.1.a requires that written procedures be established, implemented, and maintained covering certain activities. One of the activities requiring written procedures, covered by sub-section a., is the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

The licensee proposed to modify this requirement of TS 5.6.1.1.a as follows: "The procedures applicable to the safe storage of nuclear fuel recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978." The proposed change reduces the scope of the TS to requiring only the establishment, implementation, and maintenance of written procedures applicable to the safe storage of nuclear fuel. This change recognizes the reduced requirements associated with the protection of stored nuclear fuel as opposed to the operating of the nuclear power plant.

The NRC staff reviewed the proposed change and found that it appropriately revises the scope of the requirement to those procedures that are applicable to the permanently shutdown status of the reactor. The change is also consistent with NUREG-1625. Therefore, the staff finds that this change is acceptable.

3.4.2 TS 5.6.2: Programs and Manuals This section defines the programs and manuals that are applicable to CR-3.

TS 5.6.2.4, "Primary Coolant Sources Outside Containment Program," was established to minimize leakage from portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident. The licensee proposes to delete this program since serious transient or accident conditions no longer exist for a permanently defueled plant.

TS 5.6.2.5, "Component Cyclic or Transient Limit Program," provides controls to track cyclic and transient occurrences to ensure that components are maintained within their design limits. The licensee proposes to delete this program since serious transient or accident conditions no longer exist for a permanently defueled plant, and the monitored components are not required to assure spent fuel cooling.

TS 5.6.2.1 0, "Steam Generator (OTSG [once-through steam generator]) Program," ensures that the OTSG tube integrity is maintained. The licensee proposes to delete this program since CR-3 is permanently defueled and cannot operate; therefore, the steam generator tubes will not be subjected to the temperature and pressure effects that the steam generator program was put in place to protect against.

TS 5.6.2.11, "Secondary Water Chemistry Program," provides controls for monitoring secondary water chemistry to inhibit steam generator tube degradation and low pressure turbine disc stress corrosion cracking. The licensee proposes to delete this program because the components that the program was established to protect using water chemistry control are associated with reactor operation. With the licensee's decision to cease reactor operations, these components are no longer in operation and do not need protection from degradation or stress corrosion cracking.

TS 5.6.2.13, "Explosive Gas and Storage Tank Radioactivity Monitoring Program," provides controls for potentially explosive gas mixtures contained in the Radioactive Waste Disposal System, and the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system. The licensee stated that as of July 8, 2013, all CR-3 waste gas decay tanks have been vented and purged, and that no consequential residual radiation or radioactive material remains in these tanks. In addition, removal of the relief valves allowing the tanks to be vented to the atmosphere was completed on July 8, 2013. The licensee proposes to delete this program as a result.

TS 5.6.2.18, "Core Operating Limits Report," establishes the core operating limits prior to each reload cycle. The licensee proposes to delete this program since it is prohibited from reloading fuel into the CR-3 reactor core.

TS 5.6.2.19, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR),"

ensures that RCS pressure and temperature limits, including heatup and cooldown rates, criticality, and hydrostatic and leak test limits, were established and documented in the PTLR.

The licensee proposes to delete this program since the reactor coolant piping has been drained and is not subject to pressurization and the reactor vessel is vented and not subject to pressurization.

The NRC staff has reviewed the proposed changes to the following:

• Primary Coolant Sources Outside Containment Program,

• Component Cyclic or Transient Limit Program,

• Steam Generator Program,

• Secondary Water Chemistry Program,

• Explosive Gas and Storage Tank Radioactivity Monitoring Program,

• Core Operating Limits Report, and

• RCS Pressure and Temperature Limits Report.

The NRC staff has determined that the proposed deletions discussed above would appropriately reflect the condition of a permanently shutdown and defueled reactor. Since the licensee has permanently defueled the reactor, 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel. The above programs and reports are, therefore, no longer necessary. Thus, the proposed changes toTS Sections 5.6.2.4, 5.6.2.5, 5.6.2.1 0, 5.6.2.11, 5.6.2.13, 5.6.2.18, and 5.6.2.19 appropriately reflect the change in plant status and the NRC staff finds them to be acceptable.

TS 5.6.2.8, "lnservice Inspection Program," establishes the controls for periodic inspection of American Society of Mechanical Engineers (ASME) Code Class 1, 2, 3, MC and CC components including applicable supports in accordance with Section XI of the ASME Boiler and Pressure Vessel Code (BPVC). The licensee proposes to delete this program since CR-3 is permanently defueled and cannot operate; therefore, the ASME Section XI BPVC systems and components will not be subjected to the temperature and pressure effects that the inservice inspection program was put in place to protect against.

TS 5.6.2.9, "lnservice Testing Program," establishes the controls for periodic testing of ASME Code Class 1, 2, and 3, components including applicable supports in accordance with the ASME Operation and Maintenance (OM) Code. The licensee proposes to delete this program since CR-3 is permanently defueled and cannot operate; therefore, the functions described in the ASME OM Code are no longer required.

The NRC staff has reviewed the proposed changes to the lnservice Inspection and the lnservice Testing Programs and has determined that the programs do not apply to nuclear fuel or fuel handling equipment and, therefore, their deletion would appropriately reflect the condition of a permanently defueled facility. Since the licensee has permanently defueled the facility, 10 CFR 50.82(a)(2) prohibits the licensee from operating the plant or placing fuel in the reactor vessel.

Therefore, the NRC staff finds that the proposed changes are acceptable.

TS 5.6.2.20, "Containment Leakage Rate Testing Program," was established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B, as modified by approved exemptions. The licensee proposes to delete this program consistent with the guidance in NRC Regulatory Guide 1.184, "Decommissioning of Nuclear Power Reactors," July 2000, Section 9.0, "Eliminated Regulatory Requirements."

The NRC staff has reviewed the proposed deletion of the containment leakage rate testing program and has determined that this is acceptable because the program is no longer needed since 10 CFR 50.54(o) excludes permanently defueled units from the requirements of 10 CFR Part 50 Appendix J, and because this deletion appropriately reflects the condition of CR-3 as a permanently defueled facility.

TS 5.6.2.3, "Offsite Dose Calculation Manual," contains offsite dose calculation methodologies, the radioactive effluent controls program, and radiological environmental monitoring activities.

The licensee proposes to change the authority for approval of changes to the offsite dose

calculation manual from the Plant General Manager to the Plant Manager consistent with the changes to the responsibility section that were found to be acceptable in Section 3.1.1 of this safety evaluation.

The NRC staff has reviewed the proposed change to the offsite dose calculation manual and has determined that the change is of an editorial and clarifying nature, such that the current intent of the requirement is unchanged. Furthermore, the proposed change is appropriate given the above changes to the responsibility section. Therefore, the staff finds that the proposed change is acceptable.

Based on the above, the NRC staff finds all of the proposed changes toTS 5.6.2 to be acceptable.

3.5 TS 5.7.2: Special Reports The licensee proposes to eliminate the requirement to submit the special reports described in TS 5.7.2.b and c. These reports include documenting abnormal degradation of the containment structure and documenting steam generator inspection information. The licensee proposes to eliminate these reports because the programs that generate these reports are being eliminated.

The NRC staff reviewed the proposed changes and found that, due to the licensee's submittal of certifications under 10 CFR 50.82(a), the licensee is no longer authorized to operate the reactor.

These reports support programs that are no longer necessary due to the submittal of certifications to permanently cease operation and defuel the reactor, therefore, the NRC staff finds that these deletions are acceptable.

3.6 TS 5.8: High Radiation Area The licensee requested changes toTS 5.8.2 revising the title of the individual responsible for control of High Radiation Area keys from the "Control Room Supervisor" to the "Shift Supervisor," consistent with the CR-3 shutdown organization. The staff finds that this title change is an editorial change that conforms TS 5.8 with the proposed revisions toTS 5.1.2, that are evaluated in Section 3.1.2 of this safety evaluation.

During the review of TS 5.8, the staff identified that the current TS 5.8.2 and TS 5.8.3 are not clear on their applicability (or non-applicability) to Very High Radiation Areas, consistent with the requirements in 10 CFR 20.1602 and the current guidance in the Standard Technical Specifications.

On February 26, 2014, in response to an NRC staff request for additional information dated February 12, 2014, the licensee responded to the staff's request and included a regulatory commitment to submit a revision to Amendment Request No. 316 to the CR-3 TSs, which is currently under review by the NRC staff, that will include wording changes toTS 5.8.2 and TS 5.8.3 that comply with the requirements of 10 CFR 20.1602. The licensee committed to submit the revision by August 15, 2014. The NRC staff finds that this action acceptably resolves its request for additional information.

4.0 REGULATORY COMMITMENTS Regulatory Commitment Due Date/Event CR-3 will submit a revision to License August 15, 2014 Amendment Request No. 316, Revision 0, "Revise and Remove License Conditions and Revision to Improved Technical Specifications to Establish Permanently Defueled Technical Specifications," to revise Technical Specifications 5.8.2 and 5.8.3 proposing wording that complies with the requirements of 10 CFR 20.1602.

5.0 STATE CONSULTATION

Based upon a letter dated May 2, 2003, from Michael N. Stephens of the Florida Department of Health, Bureau of Radiation Control, to Ms. Brenda L. Mozafari, Senior Project Manager, U.S. Nuclear Regulatory Commission, the State of Florida does not desire notification of issuance of license amendments. In an email dated July 25, 2012 (ADAMS Accession No. ML12208A014), from Cynthia Becker, Florida State Bureau of Radiation Control, to Farideh Saba, Senior Project Manager, U.S. Nuclear Regulatory Commission, the State of Florida confirmed that the May 2003 letter continues to reflect the State's position on notification of issuance of license amendments.

6.0 ENVIRONMENTAL CONSIDERATION

S 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 or changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding, which was published in the Federal Register on November 12, 2013 (78 FR 67406). 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.

7.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) there is reasonable assurance that such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of these amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: M. Keefe, NRR K. Bucholtz, NRR R. Pedersen, NRR C. Gratton, NRR Date: July 11, 2014

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DATE 7/7/14 7/7/14 4/10/14 4/14/14 OFFICE NRR/DSS/STSB OGC (NLO w/comment) NRR/DORULPL4-2/BC NRR/DORULPL4-2/PM NAME REIIiott* JWachutka DBroaddus CGratton DATE 4/7/2014 4/22/14 7/10/14 7/11/14