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Nuclear Generating Plant - Issuance of Amendment for Operating License and Technical Specification Based on Permanently Shutdown and Defueled Status
	ML15224B286
Person / Time
Site:	Crystal River
Issue date:	09/04/2015
From:	Michael Orenak; Plant Licensing Branch IV
To:	Hobbs T; Duke Energy Florida
	Orenak M
References
	TAC MF3089
	Download: ML15224B286 (120)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 4, 2015 Mr. Terry D. Hobbs General Manager, Decommissioning Crystal River Nuclear Plant (NA2C) 15760 W. Power Line Street Crystal River, FL 34428-6708

SUBJECT:

CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT- ISSUANCE OF AMENDMENT FOR PERMANENTLY SHUTDOWN AND DEFUELED OPERATING LICENSE AND TECHNICAL SPECIFICATIONS (TAC NO MF3089)

Dear Mr. Hobbs:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 247 to Facility Operating License No. DPR-72 for the Crystal River Unit 3 Nuclear Generating Plant (CR-3). The amendment revises the facility operating license and associated technical specifications (TSs) to conform to the permanent shutdown and defueled status of CR-3. The amendment is in response to your application dated October 29, 2013, as supplemented by letters dated May 7, 2014, June 17, 2014, and March 6, 2015.

The proposed amendment revises the operating license and associated TSs to reflect the permanent cessation of reactor operations and the permanently defueled condition of the reactor vessel at CR-3. In general, the changes eliminate those TSs applicable in operating MODES; MODES where fuel is emplaced in the reactor vessel, and certain TSs required for movement of irradiated fuel assemblies. Changes were also made to the TS definitions, administrative controls, and related to programs and procedures. The proposed amendment also revises the facility operating license to clarify or remove certain conditions no longer relevant and add conditions consistent with other permanently shutdown and defueled reactors.

T. Hobbs A copy of our related Safety Evaluation is also enclosed. The Notice of Issuance will be included in the Commission's next biweekly Federal Register notice.

Michael D. Orenak, 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. 247 to DPR-72

2. Safety Evaluation cc w/encls: 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. 247 License No. DPR-72

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

A. The application for amendment to the Crystal River Unit 3 Nuclear Generating Plant (the facility) Facility Operating License No. DPR-72 filed by Duke Energy Florida, Inc., et al. (the licensees), dated October 29, 2013, as supplemented by letters dated May 7, 2014, June 17, 2014, and March 6, 2015, 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 Enclosure 1

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.

2. Accordingly, the license is amended as indicated in the attachment to this license amendment, and Facility Operating License No. DPR-72 is hereby amended to read as follows.

Paragraph 2.8.(1) of Facility Operating License No. DPR-72 is hereby amended to read:

(1) Duke Energy Florida, Inc., pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess and use the facility; Paragraph 2.8.(3) of Facility Operating License No. DPR-72 is hereby amended to read:

(3) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear material configured as reactor fuel, in accordance with the limitations for storage as described in the Final Safety Analysis Report, as supplemented and amended; Paragraph 2.8.(4) of Facility Operating License No. DPR-72 is hereby amended to read:

(4) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70 to possess at any time any byproduct, source and special nuclear material as sealed neutron sources used previously for reactor startup, as fission detectors, and sealed sources for reactor instrumentation and to possess and use at any time any byproduct, source, and special nuclear material as sealed sources for radiation monitoring equipment calibration in amounts as required; Paragraph 2.C.(1) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(1) Deleted per Amendment No. 247 Paragraph 2.C.(2) of Facility Operating License No. DPR-72 is hereby amended to read:

The Technical Specifications contained in Appendix A, as revised through Amendment No. 247, are hereby replaced with the Permanently Defueled Technical Specifications (POTS). Duke Energy Florida, Inc. shall maintain the facility in accordance with the Permanently Defueled Technical Specifications.

Paragraph 2.C.(3) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(3) Deleted per Amendment No. 247

Paragraph 2.C.(5) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(5) Deleted per Amendment No. 247 Paragraph 2.C.(7) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(7) Deleted per Amendment No. 247 Paragraph 2.C.(8) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(8) Deleted per Amendment No. 247 Paragraph 2.C.(9) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(9) Deleted per Amendment No. 247 Paragraph 2.C.(10) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(10) Deleted per Amendment No. 247 Paragraph 2.C.(11) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(11) Deleted per Amendment No. 247 Paragraph 2.C.(14) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(14) Mitigation Strategies License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(1.) Fire fighting responses strategy with the following elements:

a. Pre-defined coordinated fire response strategy and guidance

b. Assessment of mutual aid fire fighting assets

c. Designated staging areas for equipment and materials

d. Command and control

e. Training of response personnel (2.) Operations to mitigate fuel damage considering the following:

a. Protection and use of personnel assets

b. Communications

c. Minimizing fire spread

d. Procedures for implementing integrated fire response strategy

e. Identification of readily-available pre-staged equipment

f. Training on integrated fire response strategy

g. Spent fuel pool mitigation measures (3.) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders Paragraph 2.C.(15) of Facility Operating License No. DPR-72 is hereby amended to read:

2.C.(15) Deleted per Amendment No. 247 Paragraph 2.E of Facility Operating License No. DPR-72 is hereby amended to read:

(1) Deleted per Amendment No. 247

3. This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance. Implementation of the amendment shall also include revision of the facility's Updated Final Safety Analysis Report to incorporate the commitment contained in Attachment E to the licensee's letter dated March 6, 2015.

FOR THE NUCLEAR REGULA TORY COMMISSION Meena K. Khanna, 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: September 4, 2o1 s

ATTACHMENT TO LICENSE AMENDMENT NO. 247 FACILITY OPERATING LICENSE NO. DPR-72 DOCKET NO. 50-302 Replace the following pages of Facility Operating License No. DPR-72 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.

Facility Operating License No. DPR-72 REMOVE INSERT

- 4a - - 4a -

- 5c - - 5c -

Technical Specifications REMOVE INSERT All pages All pages

REVISED LICENSE PAGES TO FACILITY OPERATING LICENSE NO. DPR-72

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

{l) Duke Energy Florida, Inc., pursuant to Section l04b of the Act and 10 CFR Part 50, _*Licensing of Production and

Utilization Facilities, 11 *to possess and use the facility;

{2) The licensees to possess *tne facility at tne designated location in Citrus COWlty, Florida, in accordance with the procedures and limitations set forth in this license; (3) Duke Energy Florida, Inc., pursuant to the Act and 10. CFR Part 70, to possess at any time special nuclear material configured as reactor fuel, in accordance with the limitations for storage as described in the Final Safety Analysis Report, as supplemented and amended; (4) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70 to possess at any time any byproduct, source and special nuclear material as sealed neutron sources used previously for reactor startup, as fission detectors, and sealed sources for reactor instrumentation and to possess and use at any time any byproduct, source. and special nuclear material as sealed sources for radiation monitoring equipment calibration in amounts as required; (5) ,Duke Energy Florida, Inc., pursuant to the Act and 10 CTR Parts 30, 40 and 70, to receive, possess and use in am:>Wlts as required any oyproauct, source or special nuclear material without restriction to cher.lical or physical form, for saaple analysis or instrument calibration or associated with raoio-active apparatus or comp:ments; (6) Duke Energy Florida, Inc., pursuant to the Act and 10 CF.R Parts 30 and 7iJ, to possess, out not separate, such byproouct and special nuclear materials as may be produced by the operation of the facility.

Z.B.(7) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30 and 70, to receive and.possess, but not J+iJ.eJ r.> .

separate, that by-product and special nuclear materials re'1-associated with four (4) fuel assemblies (B&H ldentifi':"' f/mli 15' cation Numbers 1A-Ol. 04. 05 and 36 which were previously * /

irradiated in the Oconee Nuclear Statfon, Unit No. 1) 7-.:<1- 7Y acquired by Florida Powe~ Corporatio~from Duke Power

- - Company for use as reactor fuel in th' facility. *

<.:. This license shall be aeemed to contain ana is subject to the

-conditions specified in the following Conv-nission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, section 50.54 ar.d 50.59 of part Su, Section 70.32 of Part 70; and is subject to all applicable provisions

- - on April 29, 2013, the name "Florida Power Corporation" was changed to "Duke Energy Florida, Inc."

Facility Operating License No. DPR-72 Amendment No. 247

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) Deleted per Amendment No. 247 2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 247are hereby replaced with the Permanently Defueled Technical Specifications (PDTS). Duke Energy Florida, Inc. shall maintain the facility in accordance with the Permanently Defueled Technical Specifications.

Facility Operating license No. DPR- 72 Amendment No. 247

- 4a -

2.C.(3) Deleted per Amendment No. 247 2.C.(4) DELETED per Amendment No. 20 dated 7-3-79.

2.C. (5) Deleted per Amendment No. 247

.Facility Operating License No. DPR-72 Amendment No. 247

5-2.C.(6) Deleted per Amendment No. 21. 7-3-79 2.C.(7) Deleted per Amendment No. 247 2.C.(8) Deleted per Amendment No. 247 2.C.(9) Deleted per Amendment No. 247 2.C.(10) Deleted per Amendment No. 247 2.C.(11) Deleted per Amendment No. 247 2.C.(12) Deleted per Amendment No. 237 Facility Operating License No. DPR-72 Amendment No. 247

-Sc-2.C.(14) Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(1.) Fire fighting responses strategy with the following elements:

a. Pre-defined coordinated fire response strategy and guidance

b. Assessment of mutual aid fire fighting assets

c. Designated staging areas for equipment and materials d: Command and control *

e. Training of response personnel (2.) Operations to mitigate fuel damage considering the following:

a. Protection and use of personnel assets

b. Communications

c. Minimizing fire spread .

d. Procedures for implementing integrated fire response strategy

e.

Identification of readUy-available pre-staged equipment

.f. Training on Integrated fire response strategy

g. Spent fuel pool mitigation measures (3.) Actions to minimize release to include consideration of:

a. Water spray scrubbing .

b. Dose to onslte responders I

2.C.(15) Deleted per Amendment No. 247 Amendment No. 247

DO NOT REMOVE

- ti -

E. Deleted per Amendment No. 247 Facility Operating License No. DPR-72 Amendment No. 247

DO NOJ REfviOVE Deleted per Amendment No. 247 Facility Operatinp License No. DPR-72 Amendment No. 247

DO NOJ REt\roVE

- a-Deleted per Amendment No. 247 Facility Operating License No. DPR-72 Amendment No. 247

- !:) -

F. In accordance with the requirement inp:>sed Dy the octooer d, 1976, order of the Unitea States Court Appeals for the District of Colwri:>ia Circuit in i.~atural .t<esources Defense Council v. Nuclear Regulatory: Comission, No. 74-l:i65 and 74-1586, that the '~uclear Regulatory conmiss1on "shall make any licenses granted between July ~l, 1976 and such time when the manciate is issued subject to the outcome of the proceeaings herein," the license issued herein shall be subject to the outcome of such proceedings.

G. This amended lice"se is effective as of the date of issuancejRotl:t:...~

Facility Operating License No. DPR~7?, as amended, shall exr.ire q7, at midnight, Oecember 3, 2016. . MAR 31 1987 FOR 'l'dB NUCLEAR Rro.JIA'IO.RY O:.MMISSIOO

~rlsinlll ~ed br.

Roger s. Boyd, Director Division of Project Management Office' of Nuclear Reactor .Regulation Attachnents:

Appendices A & B - Technical Specifications Date of Issuance: JAN 2 8 fl/7

REVISED APPENDIX A TECHNICAL SPECIFICATIONS TO FACILITY OPERATING LICENSE NO. DPR-72

TABLE OF CONTENTS 1.0 USE AND APPLICATION .................................... 1.1-1 1.1 Definitions ......................................... 1.1-1 1.2 Logical Connectors .................................. 1. 2-1 1.3 Completion Times .........................*.......... 1. 3-1 1.4 Frequency ........................................... 1.4-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY ... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ............ 3.0-2 3.7 PLANT SYSTEMS ....................................... 3. 7-1 3.7.13 Fuel Storage Pool Water Level ................... 3. 7-1 3.7.14 Spent Fuel Pool Boron Concentration ............. 3.7-2 3.7.15 Spent Fuel Assembly Storage ..................... 3.7-4 4.0 DESIGN FEATURES ........................................ 4.0-1 s.o ADMINISTRATIVE CONTROLS ................................ 5 .0-1 Crystal River Unit 3 i Amendment No. 247

TABLE OF CONTENTS B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY . B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY .......... B 3.0-16 B 3.7 PLANT SYSTEMS . . . . . . . . . * . . * . . * . . . . . . . . * . . * . . . . . . . . * . B 3 . 7 -1 B 3.7-13 Fuel Storage Pool Water Level .................. B 3.7-1 B 3.7.14 Spent Fuel Pool Boron Concentration ............ B 3.7-4 B 3.7-15 Spent Fuel Assembly Storage .................... B 3.7-7 Crystal River Unit 3 ii Amendment No. 247

Definitions 1.1

1. 0 USE AND APPLICATION 1.1 Definitions

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

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

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

Crystal River Unit 3 1.1-1 Amendment No. 247

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

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

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

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

(continued)

Crystal River Unit 3 1.2-1 Amendment No. 149

Logical Connectors

1. 2 1~2 Logical Connectors* (continued)

EXAMPLES The~following examples illustrate the use of logical connectors.

EXAMPLE 1. 2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Verify . . .

A.2 Restore . . .

In this example the logical connector AND is used to indicate that both Required Actions A.1 and A.2 must be completed when in Condition A.

(continued)

Crystal River Unit 3 1. 2-2 Amendment No. 149

Logical Connectors

1. 2 1.2 Logical Connectors EXAMPLES EXAMPLE 1. 2-2 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met. A.1 Trip OR A.2.1 Verify AND A.2.2.1 Reduce OR A.2.2.2 Perform OR A. 3 Align .

This example represents a more complicated use of logical connectors. Required Actions A.1, A.2, and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector OR and the left justified placement. Any one of these three Actions may be chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND.

Required Action A.2.2 is met by performing either A.2.2.1 or A.2.2.2. The indented position of the logical connector OR indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed.

Crystal River Unit 3 1. 2-3 Amendment No. 149

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

BACKGROUND Limitfng Conditions for Operation (LCOs) specify m1n1mum requirements for ensuring safe handling and storage of nuclear fuel. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s).

DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the facility is in a specified condition stated in the Applicability of the Specification. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the facility is not within the Specification Applicability.

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

Crystal River Unit 3 1.3-1 Amendment No.247

Frequency 1.4 1.0 USE AND APPLICATION

1. 4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

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

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

(continued)

Crystal River Unit 3 1.4-1 Amendment No.247

Frequency 1.4

1. 4 Frequency EXAMPLES The following example illustrates the type of frequency statement that appears in the Permanently Defueled Technical Specifications (POTS).

EXAMPLE 1. 4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform (activity). 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Example 1.4-1 contains the type of SR encountered in the POTS. The Frequency specifies an interval (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ) during which the associated Surveillance must be performed at least one time. Completion of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for flexibility.

1.4-2 Amendment No. 247

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

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

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

Crystal River Unit 3 3.0-1 Amendment No. 247

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREM.ENT (SR) APPLICABILTTY SR 3.0.1 SRs shall be met during the specified conditions in the Applicability for individual Specifications, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between perfor~ances of the Surveillance, shall be failure to meet the LCO. Failure to perfor~ a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.

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

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

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

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

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

Crystal River Unit 3 3 .0-2 A~endment No. 247

Fuel Storage Pool Water Level

3. 7 .13

3. 7 PLANT SYSTfMS 3.7.13 Fuel Storage Pool Water Level L.CO 3.7.13 The fuel storage pool water level shall be ~ 156 ft Plant Daturr.

APPLICABILITY: During movement of irradiated fuel assemblies in fuel storage pool.

ACTIONS CONDITT ON REQUIRED ACTION COMP! fTION TT~E A. Fuel storage pool A.l Suspend movement of Immediately water level not within irradiated fuel 1 i mi t. assemblies in fuel storage pool.

SURVEILLANCE REQ!:!_LREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify the fuel storage pool water level is 7 days

~ 156 ft Plant Datum.

Crystal River Unit 3 3.7-1 Amendment No. 247 I

Spent Fuel Pool Boron Concentration 3.7.14

3. 7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Boron Concentration LCO. 3.7.14 The spent fuel pool boron concentration shall be ~1925 ppm.

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

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

A.2.1 Initiate action to Im:"'lediately restore spent fuel pool boron concentration to within limit.

A.2.2 Verify by Immediately administrative means a Storage Pool A and Storage Pool B spent fuel poo 1 verification has been I performed since the last movement of fuel J assemblies in

~~~~~~~~~~~~~* *~~~t-h_e~s-p_e~n~t~f-u-el__p_o_o_i_._

Crystal River Unit 3 3.7-2 Amendment No. 247

Spent Fuel Pool Boron Concentration 3.7.14 SURVEILLANCE FREQUENCY

~**-**-----* --~------

SR 3.7.14.1 Verify the spent fuel pool boron days concentration is ~ 1925 ppm.

Crystal River Unit 3 3.7-3 Amendment No. 247 l

Spent Fuel Assembly Storage 3.7.15

3. 7 PLAr\T SYSTCVlS 3.7.15 Spent Fuel Assembly Storage LCO 3.7.15 The combination of initial enrichment and burnup of each spent fuel assembly stored in Storage Pool A and StorJge Pool B, shall be within the acceptable region of Figure 3.7.15-1 or Figure 3.7.15-2.

APPLICABIL!TY: ~henever any fuel assembly is stored in Storage Pool A or Storage Pool B of the spent fuel pool .

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the A.1 Initiate action to Immediately LCO not met. move the noncomplying fuel assembly to an acceptable configuration.

Crys~al River Unit 3 3.7-4 Amendment No. 247

Spent Fuel Assembly Storage 3.7.15

~!JRVEI U:_~N<:.E~ REgyr REMENTS --~-~- -~-----**- ***'"~~----~*-~"~--**----* ...*.. **-*-**"--*******~--~~"* ...

SURVEIL ... ANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the initial Prior to enrichment and burnup of the fuel assembly storing the is in accordance with Figure 3.7.15-1 or fuel assembly Figure 3.7.15-2. in Storage Pool A or Storage Pool B.

Crystal River Unit 3 3.7-5 Amendment No. 247

Spent Fuel Assembly Storage 3.7.15 45 40 35 30 en ll::

025 3r:

E

~o E

I m

15 10 5

0 2 2.5 3 3.5 4 4.5 5 Initial Enrichment, Weight Percent U235

1. Category B: Fuel from this category can be stored with no restrictions except as noted below.

2. Category A: Fuel from this category can be stored with fuel from Categories A or B.

3. Category F: Fuel from this category must be stored in a one-out-of-two checkerboard configuration with fuel from Category B or empty water cells. Category F fuel stored in a checkerboard pattern with either Category B fuel or empty water cells must be separated from Category A fuel by a transition row of Category B fuel.

Figure 3.7.15-1 Burnup versus Enrichment Curve for Spent Fuel Storage Pool A Crystal River Unit 3 3.7-6 Amendment No. 2 4 7 I

Spent Fuel Assembly Storage 3.7.15 45 40 35 30

J Cl

~

(;25 3:

E

§20 c

=

ID 15 10 5

0 2 2.5 3 3.5 4 4.5 5 Initial Enrichment, Weight Percent U235

1. Category B: Fuel from this category can be stored with no restrictions except as noted below.

2. Category BP: Fuel from this category (between lower and upper curves) can be stored in the peripheral cells of the pool.

3. Category BE: Unacceptable for storage unless surrounded by eight empty water cells.

4. Fuel of any enrichment and burnup including fresh, unburned fuel may be stored in Pool B if surrounded by eight empty water cells. Category BE fuel assemblies must be separated by two adjacent empty cells in Pool B.

Figure 3.7.15-2 Burnup versus Enrichment Curve for Spent Fuel Storage Pool B Crystal River Unit 3 3.7-7 Amendment No. 24 7 I

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site The 4,738 acre site is characterized by a 4,400 foot m1n1mum exclusion radius centered on the Reactor Building; isolation from nearby population centers; sound foundation for structures; an abundant supply of cooling water; an ample supply of power; and favorable conditions of hydrology, geology, seismology, and meteorology.

4. 2 Not Used (continued)

Crystal River Unit 3 4.0-1 Amendment No.247

Design Features 4.0

4. 0 DESIGN FEATURES 4.3 Fuel Storage

4. 3 .1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a. Fue*1 assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b. k.n s; 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.6 of the FSAR;

c. A nominal 9.11 inch center to center distance between fuel assemblies placed in the B pool;

d. A nominal 10.5 inch center to center distance between fuel assemblies placed in the A pool.

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent;

b. k..,, s 0.9S is fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.6 of the FSAR;

c. k..,, ~ 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.6 of the FSAR; and

d. A nominal 21.125 inch center to center distance between fuel assemblies placed in the storage racks.

(continued)

Crystal River Unit 3 4.0-2 Amendment No.247

Design Features 4.0 4.0 DESIGN FEATURES 4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 138 feet 4 inches.

4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1474 fuel assemblies and six failed fuel containers.

Crystal River Unit 3 4.0-3 Amendment No.247

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 facility staff 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 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. 247

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 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

Not Used 5 .4

5. 0 ADMINISTRATIVE CONTROLS

5. 4 Not Used Crystal River Unit 3 5.0-5 Amendment No. 149

Not Used

5. 5

5. 0 ADMINISTRATIVE CONTROLS

5. 5 Not Used Crystal River Unit 3 5.0-6 Amendment No. 149

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

a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and setpoint determination;

b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to ten times the concentration values of 10 CFR 20.1001 - 20.2401, Appendix B, Table II, Column 2;

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302;

d. Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10 CFR 50, Appendix I;

e. Determination of cumulative and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year at least every 31 days;

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

g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents from the site to areas at or beyond the site boundary shall be limited to the following:

1. 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 (continued)

Crystal River Unit 3 5.0-8 Amendment No. 149

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

2. For 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;

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 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 50, Appendix I; and

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

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 50.36a, and Appendix I to 10 CFR Part 50 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. 247

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 ~anuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.12 Not Used 5.6.2.13 Not Used 5.6.2.14 Not Used 5.6.2.15 Not Used 5.6.2.16 Not Used 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 maintai1ed 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 5.6.2.21 Not Used Crystal River Unit 3 5.0-11 Amendment No.24~

Reporting Requirements 5.7 5.0 ADMINISTRATIVE CONTROLS

5. 7 Reporting Requirements 5.7.1 Routtne Repor:t~

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 CODCM).

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental sanples 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 l, 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. !he 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. lhe 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 B.l.

____ _______(continued)

Crystal River Unit 3 5.0-12 Amendment No.247

Reporting Require~ents 5.7

5. 7 Reporting Requirements 5.7.1.2 Not Used 5.7.2 Not Used Crystal River Unit 3 5.0-13 Amendment No. 247

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Pursuant to 10 CfR 20, paragraph 20.160l(c), alternative mPthods 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 cm) 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 groJp of individuals pernitted to enter such areas shall be provid~d 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 perfor~ 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 cm from the radiation source or from any surface penetrated by the radiation but less than 500 rads/hr at 1 meter from the radiation source or from any surface penetrated by the radiation shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administr0tive control of the Shift Supervisor or health physics supervision. 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 perfor~ed within the area.

(continued)

Crystal River Unit 3 5.0-14 Amendment No.247

High Radiation Are<l

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 cm from the radiation source or from any surface penetrated by the radiation but less than 500 rads/hr at 1 meter from the radiation source or from any surface penetratE>d by the radiation, accessible to personnel, that are located within large areas such as reactor contain~ent, where no enclosure exists for purposes of locking, or that are not 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 5.0-15 Amen.dment No. 247 I

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 247 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

On September 26, 2009, Crystal River Unit 3 Nuclear Generating Plant (CR-3) shut down for a refueling outage during which its steam generators were to be replaced. As a result of containment damage occurring during the replacement, Duke Energy Florida, Inc. (DEF, the licensee) decided to retire CR-3. As of May 28, 2011, all fuel assemblies have been removed from the reactor vessel and placed in the spent fuel pool (SFP). By letter dated February 20, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13056A005}, DEF submitted a certification to the U.S. Nuclear Regulatory Commission (NRC) of permanent cessation of power operations and removal of fuel from the reactor vessel, pursuant to Title 10 of the Code of Federal Regulations (10 CFR}, paragraphs 50.82(a)(1 )(i) and (ii).

By letter dated October 29, 2013 (ADAMS Accession No. ML13316C083}, as supplemented by letters dated May 7, and June 17, 2014, and March 6, 2015 (ADAMS Accession Nos. ML14139A006, ML14178B284, and ML15076A035, respectively}, DEF requested an amendment to Facility Operating License DPR-72 for CR-3. The proposed amendment would revise the facility operating license and revise the associated technical specifications (TSs) to reflect the permanent cessation of operations of CR-3. The supplemental letter dated March 6, 2015, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staffs original proposed no significant hazards consideration determination as published in the Federal Register on October 28, 2014 (79 FR 64222).

In the October 29, 2013, license amendment request (LAR), DEF proposed numerous changes to conform the CR-3 TSs to the permanently shutdown and defueled condition of CR-3. In general, the October 29, 2013, LAR proposed to eliminate most of the previous operating TSs (that applied to CR-3 when it was authorized to operate) because these TSs are only applicable in operating modes or modes where fuel is emplaced within the reactor vessel. Operation or emplacement of fuel into the reactor vessel is no longer authorized at CR-3 and, therefore, these TSs are no longer applicable or relevant based on the mode restrictions. The LAR also proposed changes to TS definitions and various organizational and program specifications.

Enclosure 2

Additionally, the LAR proposed changes to several conditions of the CR-3 operating license that DEF stated were either clarifications, redundant with other requirements, or no longer applicable based on the permanently shutdown and defueled status of CR-3. These license condition changes, for the most part, are not directly related to the TS changes.

By letter dated September 26, 2013, "Permanently Defueled Emergency Plan and Emergency Action Level Scheme, and Request for Exemption to Certain Radiological Emergency Response Plan Requirements Defined by 10 CFR 50" (ADAMS Accession No. ML13274A584), DEF requested exemption from specific emergency planning (EP) planning standards of 10 CFR 50.47, "Emergency plans," and specific requirements of Appendix E to Part 50, "Emergency Planning and Preparedness for Production and Utilization Facilities," for CR-3.

Contained within the September 26, 2013, submittal were analyses of the possible accidents at CR-3 in its permanently defueled condition. The analyses demonstrate that for all design-basis accidents (DBAs), radiation exposure levels at the CR-3 exclusion area boundary (EAB) would be less than the Environmental Protection Agency's (EPA) Protective Action Guidelines (PAGs).

The NRG staff approved the exemption request on March 30, 2015 (ADAMS Accession No. ML15058A906) and approved a conforming amendment request on March 31, 2015 (ADAMS Accession No. ML15027A209).

By letter dated July 11, 2014 (ADAMS Accession No. ML14097A145), the NRG issued Amendment No. 244 to Facility Operating License No. DPR-72 for CR-3. The amendment revised and removed certain requirements from the Section 5.0, "Administrative Controls,"

portions of the CR-3 TSs that are no longer applicable to the facility in its permanently defueled condition.

2.0 REGULATORY EVALUATION

2.1 Technical Specifications Section 182a of the Atomic Energy Act of 1954, as amended (AEA), requires applicants for nuclear power plant operating licenses to include TSs as part of their application. The NRC's regulatory requirements related to the content of the TSs are contained in 10 CFR 50.36, "Technical specifications." Pursuant to 10 CFR 50.36, each operating license issued by the Commission must include TSs which must include items in the following categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. According to 10 CFR 50.36(c)(6), for nuclear power reactor facilities that have submitted the certifications required by 10 CFR 50.82(a)(1 ), technical specifications involving safety limits, limiting safety system settings, and limiting control system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls will be developed on a case-by-case basis.

Section 50.36 of 10 CFR provides four criteria to define the scope of equipment and parameters to be included in the TS LCOs. These criteria were developed for licenses authorizing operation (i.e., for operating reactors). These criteria are not applicable to permanently shutdown and defueled facilities. For instance, Criterion 1 of 1o CFR 50.36(c)(2)(ii)(A) states that TS LCOs must be established for "installed instrumentation that is used to detect, and indicate in the

control room [CR], a significant abnormal degradation of the reactor coolant pressure boundary

[RCPB]." Since no fuel is present in the reactor or reactor coolant system (RCS) at CR-3, this criterion regarding the integrity of the RCPB is not applicable.

Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) states that TS LCOs must be established for a "process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The purpose of this criterion is to capture those process variables that have initial values assumed in the DBA and transient analyses, and which are monitored and controlled during power operation. The scope of DBAs applicable to a reactor that is permanently shut down and defueled is reduced from those postulated for an operating reactor, and most TSs satisfying Criterion 2 are no longer applicable. However, there is one existing CR-3 TS that defines the initial condition of the DBA associated with irradiated fuel movement that is discussed in Section 3.4 of this safety evaluation (SE).

Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) states that TS LCOs for operation must be established for a structure, system, or component (SSC) "that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The intent of this criterion is to capture into TSs those SSCs that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion), so that the plant response to DBAs and transients limits the consequences of these events to within the appropriate acceptance criteria. There are no transients that continue to apply to permanently shutdown and defueled reactors. The scope of DBAs that apply to CR-3 in its permanently shutdown and defueled condition is discussed in more detail in Section 3.0 of this SE.

Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) states that TS LCOs must be established for an SSC "which operating experience or probabilistic risk assessment has shown to be significant to public health and safety." The intent of this criterion is to ensure that risk insights and operating experience are factored into the establishment of TS LCOs. All of the accident sequences that dominated risk at CR-3 when it was authorized to operate are no longer applicable with the reactor in its permanently shutdown and defueled condition.

The regulations in 10 CFR 50.51 (b) require licensees that have provided certifications for permanent cessation of power operations and permanent removal of fuel, in accordance with 10 CFR 50.82(a)(1 )(i) and 10 CFR 50.82(a)(1 )(ii), to take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition.

2.2 Radiological Consequences Pursuant to the change process in 10 CFR 50.59, "Changes, tests, and experiments," DEF has revised Chapter 14, "Safety Analysis," of the CR-3 Final Safety Analysis Report (FSAR).

Chapter 14 of the FSAR describes the DBA and transient scenarios that could apply to CR-3.

The licensee states that there are no transients that continue to apply to CR-3 in its permanently shutdown and defueled condition and that the only accident scenarios that could potentially

apply to CR-3 would be a fuel handling accident (FHA) and a radioactive waste handling accident. A release of waste gas accident is no longer applicable because all waste gas has been released from the waste gas decay tanks and the tank relief valves have been removed.

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

By letter dated September 17, 2001, the NRG approved the implementation of the AST methodology for FHA dose consequence analysis at CR-3 by License Amendment No. 199 (ADAMS Accession No. ML012430210). The submittal also included changes to the CR-3 TSs to reflect implementation of AST assumptions in accordance with 10 CFR 50.67.

The FHA-specific dose acceptance criteria are specified in NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR [Light-Water Reactor] Edition," (SRP), Section 15.0.1, "Radiological Consequence Analyses Using Alternative Source Terms," Revision 0, July 2000 (ADAMS Accession No. ML003734190). The dose acceptance criteria for the FHA are a total effective dose equivalent (TEDE) of 6.3 roentgen equivalent man (rem) at the EAB for the worst 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 6.3 rem at the outer boundary of the low population zone (LPZ), and 5 rem in the CR for the duration of the accident.

3.0 TECHNICAL EVALUATION

3.1 Accident Analysis During normal power reactor operations, the forced flow of water through the RCS removes the heat generated by the reactor. The RCS, operating at high temperatures and pressures, transfers this heat through the steam generator tubes to the secondary system. The most severe postulated accidents for nuclear power plants involve damage to the nuclear reactor core and the release of large quantities of fission products to the RCS. Many of the accident scenarios postulated in the FSAR involve failures or malfunctions of systems that could affect the reactor core. With the termination of reactor operations at CR-3 and the permanent removal of the fuel from the reactor core, such accidents are no longer possible. The irradiated fuel is now stored in the CR-3 spent fuel pool (SFP). The reactor, RCS, and secondary system are no longer in operation and have no function related to the storage of the irradiated fuel. Therefore, the postulated accidents involving failure or malfunction of the reactor are no longer applicable.

Chapter 14 of the CR-3 FSAR describes the transients and DBAs that are applicable to CR-3 in its permanently shutdown and defueled condition. Since CR-3 is permanently shut down and defueled, there are no applicable transients. Additionally, the only DBAs that could potentially apply at this time to the permanently shutdown and defueled CR-3 would be the FHA and a radioactive waste handling accident.

3.2 Fuel Handling Accident In the AST evaluation supporting CR-3 power operation, the radiological consequence analysis evaluated the consequences of an FHA in the containment, with no credit taken for containment isolation. Since the assumptions and parameters used for an FHA inside containment with no isolation are identical to those for an FHA in the auxiliary building, the resulting radiological consequences are the same regardless of the location of the accident. After permanent cessation of operations, an FHA onto the top of the core (or elsewhere within containment) is no longer possible, and therefore, has been removed from the CR-3 licensing basis. However, an FHA in the SFP (which is located in the auxiliary building) is still possible at CR-3, as long as spent fuel is stored in the SFP.

The licensee defines the FHA in the SFP as the dropping of a spent fuel assembly onto the SFP floor or the racks that hold the spent fuel such that the cladding of all the fuel rods in one assembly ruptures. The gap activity in the damaged rods is instantaneously released into the SFP. The activity is assumed to pass through 23 feet of the required minimum water level over the top of the irradiated fuel assemblies in the SFP. It is postulated that the activity released from the SFP then mixes with the auxiliary building atmosphere before being released directly to the environment. The FHA analysis postulates that the release from the SFP to the auxiliary building atmosphere is not mitigated in route to the environment. This assumption is consistent with the CR-3 current licensing basis FHA analysis, which does not credit the SFP ventilation system for accident mitigation. The activity is assumed to be exhausted from the auxiliary building at a rate established to complete the release in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (consistent with RG 1.183).

The NRC staff finds that the assumptions in the licensee's analysis are consistent with the CR-3 current licensing basis FHA analysis, which does not credit the auxiliary building ventilation system for accident mitigation, but conservatively assumes that it continues to operate, expelling accident-related activity to the environment and the CR ventilation intake.

Regulatory Guide 1.183 provides for an iodine reduction factor of 200 with at least 23 feet of water above the damaged fuel. The licensee states that this would be the case for fuel in the spent fuel racks which are damaged by a heavy load drop, but for a damaged assembly which lies horizontally across the top of the spent fuel racks, the water depth could be slightly less than 23 feet. Therefore, the licensee used an iodine removal factor of 100 for the FHA analysis that is consistent with the previously approved licensing basis analysis.

A fission product decay period of 4 years is assumed in the FHA analysis. No credit is taken for CR isolation or recirculation filtration in the FHA analysis. Specifically, the Control Room Emergency Ventilation System (CREVS), support systems, or automatic actuation instrumentation (from the radiation monitor in the CR ventilation ductwork) are not credited in the DEF analysis. For the calculation of the CR, EAB and LPZ doses, the licensee used current licensing basis atmospheric dispersion factors.

The licensee's analysis of radiological consequences resulting from the postulated FHA at CR-3 in its permanently shutdown and defueled condition concluded that the radiological consequences at the EAB, LPZ, and in the CR are within the dose criteria specified in 10 CFR 50.67 and the accident-specific dose criteria described in SAP, Section 15.0.1. The NRC staff has evaluated the licensee's analysis. In performing this evaluation, the NRC staff relied upon information provided by the licensee in its September 26, 2013, submittal, as well as NRC staff experience in performing similar evaluations. The NRC staff reviewed the methods,

parameters, and assumptions that the licensee used in its radiological dose consequence analyses and finds that they are consistent with the conservative guidance provided in RG 1.183. Based on the above, the NRC staff concludes that the radiological consequences of an FHA at the EAB, LPZ, and in the CR for CR-3 in its permanently shutdown and defueled condition continue to comply with the applicable regulations and guidance.

3.3 Radioactive Waste Handling Accident According to the CR-3 FSAR, postulated accidents that could result in the release of radioactive liquids are those that involve a drop (spill) of a high integrity container (HIC) on the south berm, adjacent to the auxiliary building, such that its entire contents of radioactive dewatered demineralizer resin escapes. The licensee determined that the consequences of a dropped spent resin HIC bounds all postulated radioactive waste handing accidents as described in its September 26, 2013, submittal.

The licensee determined that the dose to the most limiting individual at the EAB from the radioactive waste handling accident is 40 millirem (mrem) TEDE and that the thyroid dose is negligible due to the irradiated fuel having undergone four years of radioactive decay. The NRC staff verified that both the thyroid dose and the TEDE dose from this bounding radioactive waste handling accident would not trigger the EPA PAGs at the EAB and are below the dose criteria specified in 10 CFR 50.67. In addition, the NRC staff finds that the licensee has adequately demonstrated that the maximum consequences of the radioactive waste handling accident are below the 100 mrem acceptance criteria established in Nuclear Energy Institute (NEI) 99-01, "Development of Emergency Action Levels for Non-Passive Reactors, Revision 6, dated November 2012 (ADAMS Accession No. ML12326A805) for the declaration of a Site Area Emergency. Based on the above, the NRC staff concludes that the radiological consequences of a radioactive waste handling accident at the EAB for CR-3 in its permanently shutdown and defueled condition continue to comply with the applicable regulations and guidance.

3.4 Spent Fuel Pool Water Level Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) states that TS LCOs must be established for a "process variable, design feature, or operating restriction that is an initial condition of a [OBA] or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The purpose of this criterion is to capture those process variables that have initial values assumed in the OBA. CR-3 TS 3.7.13, "Fuel Storage Pool Water Level," specifies the TS required LCOs and SRs that ensure that the minimum water level in the SFP meets the assumptions of iodine decontamination factors following an FHA.

DEF's analysis of the postulated FHA assumes that there is 23 feet of water between the top of the damaged fuel bundle and the fuel pool surface during the FHA. In the analysis, FHA is defined as the dropping of a spent fuel assembly onto the SFP floor or racks, such that the cladding of all the fuel rods in one assembly rupture. The gap activity in the damaged rods is instantaneously released into the SFP. The release occurs under 23 feet of water, which acts as a filter. The activity exhaust rate from the auxiliary building is established to complete the release in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , as required by RG 1.183, but does not credit the auxiliary building ventilation for any mitigation of the release.

Since the 23-foot water level of the SFP above the damaged fuel bundle is an initial condition of the FHA OBA, it satisfies Criterion 2 for inclusion in TSs and, therefore, is being retained as a TS for CR-3 in its permanently shutdown and defueled condition.

Additionally, alarms are maintained to detect a reduction in the CR-3 SFP water level and alert the operators. The CR-3 FSAR Section 9.3.2, "System Description," states that the SFP level transmitters provide visual indication on the main control board to alert operators to a possible pool overflow or a decrease in pool level. Additionally, the CR-3 FSAR Section 9.3.2.7, "Operational Limits, states that the alarms for the SFP level are provided on the Main Control Board annunciators.

The LAR submitted by DEF does not propose any change to TS 3.7.13. Therefore, the discussion in this SE of the CR-3 SFP water level TS is provided only for completeness since the SFP water level is an important initial condition in the FHA analysis and, pursuant to 10 CFR 50.36(c)(2)(ii)(B), must continue to be part of the CR-3 TSs.

3.5 Accident Analysis Summary As described above, the NRC staff reviewed the assumptions, inputs, and methods used by the licensee to assess the radiological consequences of DBAs based upon the permanently shutdown and defueled condition at CR-3. The NRC staff finds that the licensee used analysis methods and assumptions consistent with the conservative regulatory requirements and guidance identified in Section 2.2 of this SE. The NRC staff compared the doses estimated by the licensee to the applicable criteria identified in Section 2.2. The NRC staff finds that the licensee has demonstrated with reasonable assurance that the estimates of the EAB, LPZ, and CR doses will comply with these criteria. The NRC staff further finds that there is reasonable assurance that the CR-3 TSs, as modified by this license amendment, will continue to provide sufficient safety margins with adequate defense-in-depth to address unanticipated events and to compensate for uncertainties in accident progression and analysis assumptions and parameters. Therefore, the NRC staff concludes that the proposed license amendment is acceptable with respect to the radiological consequences of DBAs at CR-3 in its permanently shutdown and defueled condition.

3.6 Proposed TS Changes 3.6.1 Section 1.1, "Definitions" The licensee proposed deleting the following definitions because they pertain to an operating reactor. Since CR-3 is permanently shut down and defueled, the definitions have no relevance and no longer apply:

ALLOWABLE THERMAL POWER ALLOWABLE THERMAL POWER shall be the maximum reactor core heat transfer rate to the reactor coolant permitted by consideration of the number and configuration of reactor coolant pumps (RCPs) in operation.

AXIAL POWER IMBALANCE AXIAL POWER IMBALANCE shall be the power in the top half of the core expressed as a percentage of RATED THERMAL POWER {RTP} minus the power in the bottom half of the core expressed as a percentage of RTP.

AXIAL POWER SHAPING RODS (APSRS)

APSRs shall be the part length control components used to control the axial power distribution of the reactor core. The APSRs are positioned manually by the operator and are not trippable.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds within the necessary range and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel, including the required sensor, alarm, display, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an inplace qualitative assessment of sensor behavior and normal calibration of the remaining adjustable devices in the channel. Whenever a sensing element is replaced, the next required inplace assessment consists of comparing the other sensing elements with the recently installed sensing element. The CHANNEL CALIBRATION may be performed by means of any series of sequential, overlapping, or total channel steps so that the entire channel is calibrated.

The CHANNEL CALIBRATION shall also include testing of safety related Reactor Protection System (RPS), Engineered Safeguards Actuation System (ESAS), and Emergency Feedwater Initiation and Control (EFIC) bypass functions for each channel affected by the bypass operation.

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

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarms, interlocks, display, and trip functions.

b. Bistable channels (e.g., pressure switches and switch contacts) - the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify OPERABILITY, including required alarm and trip functions.

c. The ESAS CHANNEL FUNCTIONAL TEST shall include testing of ESAS safety related functions for each channel affected by bypass operation.

CONTROL RODS CONTROL RODS shall be all full length safety and regulating rods that are used to shut down the reactor and control power level during maneuvering operations.

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

CORE OPERATING LIMITS REPORT (COLR)

The COLR is the unit specific document that provides cycle specific parameter limits for the current reload cycle. These cycle specific limits shall be determined for each reload cycle in accordance with Specification 5.6.2.18. Plant operation within these limits is addressed in individual Specifications.

DOSE EQUIVALENT 1-131 DOSE EQUIVALENT 1-131 shall be that concentration of 1-131 (microcuries/gram) that alone would produce the same thyroid dose as the quantity and isotopic mixture of 1-131, 1-132, 1-133, 1-134, and 1-135 actually present. The thyroid dose conversion factors used for his calculation shall be those listed in International Committee on Radiation Protection (ICRP) 30, Supplement to Part 1, page 192-212, Table titled, "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity.:

E-AVERAGE DISINTEGRATION ENERGY E shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half live_s > 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.

EFFECTIVE FULL POWER DAY (EFPD)

EFPD shall be the ratio of the number of hours of production of a given THERMAL POWER to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , multiplied by the ratio of the given THERMAL

POWER to the RTP. One EFPD is equivalent to the thermal energy produced by operating the reactor core at RTP for one full day. (One EFPD is 2609 MWt

[megawatt thermal] times 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or 62616 MWhr.)

EMERGENCY FEEDWATER INITIATION AND CONTROL (EFIC) RESPONSE TIME The EFIC RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its EFIC actuation setpoint at the channel sensor until the emergency feedwater equipment is capable of performing its safety function (i.e.,

valves travel to their required positions, pump discharge pressures reach their required values, etc.) Times shall include diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

ENGINEERED SAFETY FEATURE (ESF) RESPONSE TIME The ESF RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (I.E., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing, that is captured and conducted to collection systems or a sump or collecting tank; or

2. LEAKAGE into the containment atmosphere from sources that are both specifically located and quantified and know not to interfere with the operation of leakage detection systems and not to be pressure boundary LEAKAGE; or

3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the secondary system (primary to secondary LEAKAGE).

b. Unidentified LEAKAGE All LEAKAGE that is not identified LEAKAGE.

c. Pressure Boundary LEAKAGE LEAKAGE except primary to secondary LEAKAGE) through a non-isolable fault in an RCS component body, pipe wall, or vessel wall.

MODE A MODE shall correspond 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.

NUCLEAR HEAT FLUX HOT CHANNEL FACTOR (Fa(Z))

Fa(Z) shall be the maximum local linear power density in the core divided by the core average fuel rod linear power density, assuming nominal fuel pellet and fuel rod dimensions.

NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR shall be the ratio of the integral of linear power along the fuel rod on which minimum departure from nucleate boiling ratio occurs to the average fuel rod power.

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

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

These tests are:

a. Described in Chapter 13, "Initial Tests and Operation" of the FSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

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 5.6.2.19.

Plant operation within these operating limits is addressed in LCO 3.4.3, "RCS Pressure and Temperature Limits."

QUADRANT POWER TILT (QPT)

QPT shall be defined by the following equation and is expressed as a percentage.

Power In Any Core Quadrant Q PT = 100 (----------------------------------------------- -1 )

Average Power of all Quadrants RATED THERMAL POWER (RTP)

RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2609 MWt.

REACTOR PROTECTION SYSTEM (RPS) RESPONSE TIME The RPS RESPONSE, TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel _sensor until electrical power is interrupted at the control rod drive trip breakers. The response time may be measured by means of any series of sequential overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SOM)

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

a. All CONTROL RODS (safety and regulating) are fully inserted except for the single CONTROL ROD of highest reactivity worth, which is assumed to be fully withdrawn; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the post-trip RCS average temperature.

With any CONTROL RODS not capable of being fully inserted, the reactivity worth of these CONTROL RODS must be accounted for in the determination of SOM.

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

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

Section 1.1 of the CR-3 TSs provides definitions for selected terms that are used throughout the TSs. Since these terms will not be used in the permanently defueled technical specifications (PDTSs), their definitions are proposed to be deleted from the PDTSs. The NRC staff examined the TS definitions proposed to be deleted and found that all of the terms listed above, as well as Table 1.1-1 of the CR-3 TSs, "MODES," are only meaningful to a reactor authorized to operate and, therefore, are not applicable to PDTSs. Since CR-3 is permanently shut down and defueled, the NRC staff concludes that the deletion of these definitions and Table 1.1-1 will have no impact on the continued safe use of the CR-3 facility and, therefore, is acceptable.

3.6.2 Section 1.2, "Logical Connectors" The licensee is not proposing any changes to Section 1.2 of the CR-3 TS.

3.6.3 Section 1.3, "Completion Times" Section 1.3 of the CR-3 TSs establishes the completion time convention used throughout the TSs and provides guidance for its use. The licensee proposed to replace each reference to "operation of the unit" and "unit" with the new terminology, "handling and storage of nuclear fuel" and facility," respectively, since operation of the unit is no longer permitted and safe management of irradiated fuel is the primary objective of the proposed PDTSs. In addition, the licensee proposed to delete references to "MODE," and "THERMAL POWER," to be consistent with the proposed removal of these definitions from the TSs and because these terms are no longer used in the Required Actions of the remaining LCOs in the proposed POTS.

Below are the proposed changes with deletions in strikethrough and additions underlined.

BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe operation of the unit handling and storage of nuclear fuel. The ACTIONS associated with an LCO state Conditions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated Condition are Required Action(s) and Completion Time(s).

DESCRIPTION The Completion Time is the amount of time allowed for completing a Required Action. It is referenced to the time of discovery of a situation (e.g., inoperable equipment or variable not within limits) that requires entering an ACTIONS Condition unless otherwise specified, providing the tlflit facility is in a MODE or specified condition stated in the Applicability of the Specification. Required Actions must be completed prior to the expiration of the specified Completion Time. An ACTIONS Condition remains in effect and the Required Actions apply until the Condition no longer exists or the tlflit facility is not within the Specification Applicability.

If situations are discovered that require entry into more than one Condition at a time within a single Specification (multiple Conditions), the Required Actions for each Condition must be performed 'Nithin the associated Completion Time.

When in multiple Conditions, separate Completion Times are tracked for each Condition starting from the time of discovery of the situation that required entry into tho Condition.

Once a Condition has boon entered, subsequent trains, subsystems, components, or variables expressed in the Condition discovered to be inoperable or not within limits, will not result in separate entry into the Condition, unless specifically stated. Tho Required Actions of tho Condition continue to apply to each additional failure, 'Nith Completion Times based on initial entry into tho Condition.

HO't\'Over, 'Nhen a subsequent train, subsystem, component, or variable, expressed in tho Condition, is discovered to be inoperable or not 'Nithin limits, the Completion Time(s) may be extended. To apply this Completion Time extension ti.*10 criteria must first be mot. Tho subsequent inoperability:

a. Must exist concurrent 'Nith tho ill§! inoporability; aAG

b. Must remain inoperable or not within limits after the first inoperability is resolved.

Tho total Completion Time allO'tved for completing a Required Action to address the subsequent inoporability shall be limited to tho more restrictive of either:

a. Tho stated Completion Time, as measured from the initial entry into tho Condition, plus an additional 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ; or

b. Tho stated Completion Time as measured from discovery of the subsequent inoperability.

The above Completion Time extensions do not apply to those Specifications that have exceptions that allow completely separate re entry into the Condition (for

each train, subsystem, component or variable expressed in the Condition) and separate tracking of Completion Times based on this re entry. These exceptions are stated in individual Specifications.

The above Completion Time extension does not apply to a Completion Time with a modified "time zero." This modified "time zero" may be expressed as a repetitive time (i.e., "once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ," where the Completion Time is referenced from a previous completion of the Required Action versus the time of Condition entry) or as a time modified by the phrase from discovery ..."

EXAMPLES The following examples illustrate the use of Completion Times 1Nith different types of Conditions and changing Conditions.

EXAMPLE 1.3 1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required B.1 Be in MODE 3 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action and associated Completion Time not met. B.2 Be in MODE 5 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Condition B has two Required Actions. Each Required Action has its own separate Completion Time. Each Completion Time is referenced to the time that Condition B is entered.

The Required Actions of Condition B are to be in MODE 3 in 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months s~

in MODE 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . A total of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is allowed for roaching MODE 3 and total of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months (not 42 hours4.861111e-4 days 0.0117 hours 6.944444e-5 weeks 1.5981e-5 months ) is allowed for reaching MODE 5 from the time that Condition B *1,ias entered. If MODE 3 is reached in 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months , the time allov.*ed for reaching MODE 5 is the next 33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months because the total time allowed for reaching MODE 5 is 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

If Condition B is entered while in MODE 3, the time allO\\'ed for reaching MODE 5 is the next 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

EXAMPLE 1.3 2 ACTIONS CONQITION REQbJIREQ ACTION COMPLETION

+iM&;

.,,A 0Re pl:lmp A.1 Restore p1:1mp to 7 days iRoperable. OPERABLE staWs B. ReEt1:1iFed B.1 Be iR MOQE 3 6 hOl:IFS AstioR aRd assosiated ANQ CompletioR Time Rot B.2 Be iR MOQE a 36 hOl:IFS

~

\lheR a p1:1mp is deslared iRoperable, CoRditioR A is eRtered. If the p1:1mp is Rot restored to OPERABLE stat1:1s withiR 7 days, CoRditioR B is also eRtered aRd the CompletioR Time slosks foF ReEj1:1ired AstioRs B.1 aRd B.2 staFt. If the iRopeFable p1:1mp is Festmed to OPERABLE stat1:1s aftm CoRditioR B is eRtered, CoRditioRs A aRd B are exited, aRd therefme, the ReEj1:1ired AstioRs of CoRditioR B may be teFmiRated.

WheR a sesoRd p1:1mp is deslared iRoperable while the fiFst p1:1mp is still iRopmable, CoRditioR A is Rot re eRtered foF the sesoRd p1:1mp. LCO 3.0.3 is eRtered, siRse the ACTIONS do Rot iRsl1:1de a CoRditioR for more thaR oRe iRoperable p1:1mp. The CompletioR Time slosk fm CoRditioR A does Rot stop afteF LCO 3.0.3 is eRtered, b1:1t soRtiR1:1es to be trasked fmm the time CoRditioR A was iRitially eRtemd.

While iR LCO 3.0.3, if oRe of the iRopeFable p1:1mps is restored to OPERABLE stat1:1s aRd the CompletioR Time fOF CoRditioR A has Rot expired, LCO 3.0.3 may be exited aRd operatioR soRtiR1:1ed iR assmdaRse 1Nith CoRditioR A.

VIJhile iR LCO 3.0.3, if oRe of the iRopeFable p1:1mps is restored to OPERABLE stat1:1s aRd the CompletioR Time foF CoRditioR A has expired, LCO 3.0.3 may be exited aRd operatioR soRtiR1:1ed iR assOFdaRse with CoRditioR B. The CompletioR Time fOF CoRditioR B is tFasked from the time the CoRditioR A CompletioR Time expired.

CONDITION REQUIRED ACTION COMPLETION

+lME A 1 Restore 7 days Function X train to OPERABLE stat\::ls B. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months C. One Function C.1 Restore 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months X train Function X train inoperable. to OPERABLE staWs GR C.2 Restore Function Y train to OPERABLE stat\::ls

established for Condition C and tracked from the time the second train was declared inoperable (i.e., the time the situation described in Condition C was discovered).

If Required Action C.2 is completed within the specified Completion Time, Conditions B and C are exited. If the Completion Time for Required Action A.1 has not expired, operation may continue in accordance with Condition A. The remaining Completion Time in Condition A is measured from the time the affected train *.vas declared inoperable (i.e.,

initial entry into Condition A).

It is possible to alternate between Condition A, B, and C in such a manner that operation could continue indefinitely

.vithout ever restoring systems to meet the LCO. However, doing so would be inconsistent with the basis of the Completion Times. Therefore, there shall be administrative controls to limit the maximum time allowed for any combination of Conditions that result in a single contiguous occurrence of failing to meet the LCO. These administrative controls shall ensure that the Completion Times for those Conditions are not inappropriately extended.

EXAMPLE 1.3 4 ACTIONS CONQITION REQl:JIREQ ACTION COMPLETION

~

A Vi. One or more A.1 Restore valve(s) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1,1alve(s) to OPERABLE inoperable. status.

B. Required B.1 Be in MOQE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action and associated ANG Completion Time not B.2 Be in MOQE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months meb A single Completion Time is used for any number of 1.ialves inoperable at tho same time. Tho Completion Time associated with Condition A is based on the initial entry into Condition A and

is not traoked on a per valve basis. Deolaring subsequent valves inoperable, while Condition A is still in effeot, does not trigger the traoking of separate Completion Times.

Onoe one of the valves has been restored to OPERABLE status, the Condition A Completion Time is not reset, but oontinues from the time the first valve was deolared inoperable. The Completion Time may be extended if the valve restored to OPERABLE status Yt'as the first inoperable valve. The Condition A Completion Time may be extended for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months provided this does not result in any subsequent valve being inoperable for > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If the Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (inoluding the extension) expires

'Nhile one or more valves are still inoperable, Condition B is entered.

EXAMPLE 1.3 5 ACTIONS NOTE Separate Condition entry is allowed for eaoh inoperable valve CONDITION REQ611RED ACTION COMPLETION T-lME I\

One or more A1 Restore 1.ial1.ie(s) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 1.ialve(s) to OPERABLE inoperable. status.

Q. ~e~uiFed EU Qe iR MQQli 3. 6 hOUFS Aotion and assooiated ANQ Completion Time not met. B.a Be in MODE 4. 1a hours The Note above the ACTIONS table is a method of modifying hov.' the Completion Time is traoked. If this method of modifying how the Completion Time is traoked was applioable only to a speoifio Condition, the Note would appear in that Condition, rather than at the top of the ACTIONS Table.

The Note allows Condition A to be entered separately for eaoh inoperable valve, and Completion Times traoked on a per valve basis. When a valve is deolared inoperable,

Condition A is entered and its Completion Time starts. If subsequent 1alves are deolared inoperable, Condition A is 1

entered for eaoh valve and separate Completion Times start and are traoked for eaoh valve.

If the Completion Time assooiated with a valve in Condition A expires, Condition B is entered for that *.ialve.

If the Completion Times assooiated *.vith subsequent valves in Condition A expire, Condition B is entered separately for eaoh valve and separate Completion Times start and are traoked for eaoh valve. If a 1alve that oaused entry into 1

Condition B is restored to OPERABLE status, Condition B is exited for that valve.

Sinoe the Note in this example allows multiple Condition entry and traoking of separate Completion Times, Completion Time extensions do not apply.

EXAMPLE 1.3 6 ACTIONS CONQITION REQbllREQ ACTION COMPLETION

+lM A One ohannel A.1 Perform SR Onoe per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

j. '

inoperable. 3.x.x.x.

GR A.2 Reduoe 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Tl=4ERMAL POWER to <E aG% RTP.

B. Required B.1 Be in MOQE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Aotion and assooiated Completion Time not meb Entry into Condition A offers a ohoioe between Required Aotion A 1 or A.2. Required Aotion A.1 has a "onoe per" Completion

Time, which qualifies for the 25%'a extension, per SR 3.0.2, to each performance after the initial performance. If Required Action A.1 is follm*.*ed and the Required Action is not met within the Completion Time (including the 25% extension allowed by SR 3.0.2), Condition Bis entered. If Required Action A.2 is followed and the Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is not met, Condition B is entered.

If after entry into Condition B, Required Action A.1 or A2 is met, Condition B is exited and operation may then continue in Condition A.

EXAMPLE 1.3 7 ACTIONS CQNQITIQN REQbJIREQ ACTION CQMPLETIQN

~

I\

One A1 Verify affected +Aetff:

subsystem susbsytem inoperable. isolated ANG Qnce per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter ANG A.2 Restore 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months subsystem to OPERABLE status ..

B. Required B.1 Be in MOQE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months

.A.ction and associated ANG Completion Time not B.2 Be in MQQE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months meb Required Action A.1 has t*No Completion Times. The 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months completion Timo begins at tho time the Condition is entered and each "Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter interval begins upon completion of Required Action A 1.

If after Condition A is entered, Required Action A.1 is not met within either the initial 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or any subsequent B hour interval from the previous performanoe (including the 25% extension allowed by SR 3.0.2), Condition Bis entered. The Completion Time cloak for Condition A does not stop after Condition B is entered, but continues from the time Condition A *.vas initially entered. If Required Action A.1 is met after Condition B is entered, Condition B is exited and operation may continue in accordance *nith Condition /\, provided the Completion Time for Required Action A.2 has not expired.

The NRC staff has reviewed the proposed terminology changes to the CR-3 TS 1.3 and has determined that they are consistent with the transition from an operating reactor to a permanently shutdown and defueled facility with a primary safety focus of managing irradiated fuel. The proposed changes also remove references to operating modes that are no longer permitted following the certifications of permanent cessation of operations and permanent removal of fuel under the provisions of 10 CFR 50.82(a)(2).

The CR-3 TS Examples 1.3-1 through 1.3-7 provide an explanation of the time requirements for transitioning into a MODE in which the requirements are not applicable, or provide the time requirements for entry into TS 3.0.3, which is also proposed to be deleted (as discussed in Section 3.6.6 of this SE). Additionally, all POTS LCOs have a completion time of "Immediately

and contain no actions linked by logical connectors, negating the need for the explanation of more complex arrangements of Required Actions and Completion Times provided in the examples. Therefore, the NRC staff finds that the CR-3 TS Examples 1.3-1 through 1.3-7 are no longer necessary to understand and properly implement the Required Actions and Completion Times of the POTS.

Because the proposed terminology changes and the proposed deletion of examples in CR-3 TS 1.3 conform to the permanently shutdown and defueled condition of CR-3, the NRC staff concludes that they will have no impact on the continued safe use of the CR-3 facility and, therefore, are acceptable.

3.6.4 Section 1.4, "Frequency" Section 1.4 of the CR-3 TSs defines the proper use and application of Frequency requirements throughout the TSs. The licensee proposed to delete the final paragraph in the description section and Examples 1.4-2 and 1.4-3. The licensee proposed to modify Example 1.4-1 to replace the "Channel Check" term with "activity," in order to illustrate the type of frequency statement that would appear in the POTS.

Below are the proposed changes with deletions in strikethrough and additions underlined.

DESCRIPTION:

Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An

understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

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

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

EXAMPLES The following examples illustrates the type of frequency statement that appears in the Permanently Defueled Technical Specifications (POTS). various way-s that Frequencies are specified. In these examples, the l\pplicability of the Specification (not shown) is MODES 1, 2, and 3.

EXAMPLE 1 .4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform CHA~JNEL CHECK (activity). 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (PDTS1. The Frequency specifies an interval (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ) during which the associated Surveillance must be performed at least one time.

Completion of the Surveillance initiates the subsequent interval. Although the Frequency is stated as 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility. +Re measurement of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment is inoperable, a variable is outside specified limits, or the unit is outside the Applicability of the Specification). If the interval specified by SR 3.0.2 is exceeded while the unit is in a MODE or other specified condition in the Applioability of the Specification, and the performance of the Surveillanoe is not otherwise modified (refer to Example 1.4 3), then SR 3.0.3 becomes applicable.

If the interval as specified by SR 3.0.2 is exoeeded while the unit is not in a MODE or other specified condition in the Applioability of the Speoification for

which performance of the SR is required, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the MODE or other specified condition. Failure to do so 'Nould result in a violation of SR 3.0.4.

EXAMPLE 1.4 2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENC¥ Verify flow is within limits. ~

Once 'Nithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after> 25%

ANG 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter Example 1.4 2 has two Frequencies. The first is a one time performance Frequency, and the second is of the type shown in Example 1.4 1. The logical connector "ANQ" indicates that both Frequency requirements must be met. Each time reactor power is increased from a power level < 25% RTP to > 25% RTP, the Surveillance must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The use of "once" indicates a single performance will satisfy the specified Frequency (assuming no other Frequencies are connected by "8.NQ"). This type of Frequency does not qualify for the 25% extension allowed by SR 3.0.2.

"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If tho reactor power decreases to< 25% RTP, the measurement of both intervals stops. New intervals start upon reactor power reaching 25% RTP.

EXAMPLE 1.4 3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENC¥ NOTE Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after > 25% RTP.

The interval continues whether or not the unit operation is < 25% RTP between performances.

As the Note modifies the required performance of the Surveillance, it is construed to be part of the "specified Frequency." Should tho 7 day interval be exceeded while operation is < 25% RTP, this Note allows 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after power reaches ~

25% RTP to perform the Sl::lrveillanoe. The Sl::lrveillanoe is still oonsidered to be within the "speoified freq1::Jency." Therefore, if the Sl::lrveillance was not performed within the 7 day (pills 25% per SR 3.0.2) interval, bl::lt operation v,ias <

25% RTP, it wol::lld not constit1::Jte a fail1::Jre of the SR or faill::lre to meet the LCO.

Also, no violation of SR 3.0.4 occms when ohanging MODES, even with the 7 day freq1::Jency not met, provided operation does not exceed 12 ho1::Jrs with pmver > 25% RTP.

Onoe the llnit reaohes 25% RTP, 12 hol::lrs 'NOl::lld be allowed for oompleting the Sl::lrveillance. If the Sl::lrveillance was not performed within this 12 hol::lr interval, there wo1::Jld then be a fail1::Jre to perform a Sl::lrveillance 'Nithin the specified freq1::Jency, and the provisions of SR 3.0.3 wollld apply.

The NRC staff has reviewed the proposed changes to Section 1.4 and has determined that they are appropriate for a permanently defueled facility. The revised CR-3 TS Example 1.4-1 illustrates the type of SR contained in the POTS and has been modified to describe the permanently shutdown and defueled status of the facility. The proposed deletion of Examples 1.4-2 and 1.4-3 is appropriate since none of the surveillances in the proposed POTS contain notes that modify the frequency of performance or the conditions during which the acceptance criteria must be satisfied. Because these changes conform to the permanently shutdown and defueled condition of CR-3, the NRC staff concludes that they will have no impact on the continued safe use of the CR-3 facility and, therefore, are acceptable.

3.6.5 Section 2.0, "Safety Limits (Sls)"

Section 2.0 of the CR-3 TSs establishes safety limits (Sls), which preclude violation of the fuel design criteria and RCS design pressure. The licensee proposes to delete the safety limits specified in Section 2.0 and Figure 2.1.1-1 because they are not applicable to the permanently shutdown and defueled status of the facility. Below are the proposed changes with deletions in strikethrollgh and additions underlined.

2.1 "SLs" 2.1.1 "Reaotor Coro SLs" 2.1.1.1 In MODES 1 and 2, the maximl::lm local fllel pin centerline temperatl::lre shall be < 5080 (6.5 E 3) X (Bllrnllp MVVD/MTU) 0 f.

Operation within this limit is ens1::Jred by compliance with the AXIAL PO\AJER IMBALANCE protective limits preserved by the Reactor Protection System setpoints in LCO 3.3.1, "Reactor Protection System (RPS) lnstrl::lmentation," as speoified in the COLR.

2.1.1.2 In MODES 1 and 2, the departl::lre from n1::Jcleate boiling ratio (DNBR) shall be maintained greater than the limits of 1.3 for the BA'!'/ 2 correlation, 1.18 for the BWC correlation and 1.132 for the BHTP correlation. Operation within this limit is ens1::Jred by oompliance with SL 2.1.1.3 and with the AXIAL POWER

IMBALANCE protective limits preserved by the RPS setpoints in LCO 3.3.1, as specified in the COLR.

2.1.1.3 In MODES 1 and 2, Reactor Coolant System (RCS) core outlet temperature and pressure shall be maintained above and to the left of the SL shown in Figure 2.1.1 1.

2.1.2 "RCS Pressure SL" In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained

< 2750 psig.

2.2 SL Violations The following actions shall be completed:

2.2.1 In MODE 1 or 2, if SL 2.1.1.1, SL 2.1.1.2 or SL 2.1.1.3 is violated, be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 In MODE 1 or 2, if SL 2.1.2 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.3 In MODES 3, 4, and 5, if SL 2.1.2 is violated, restore compliance within 5 minutes.

The restrictions of the Sls in CR-3 TS 2.1.1 prevent overheating of the fuel and cladding, and possible cladding perforation, which could result in the release of fission products to the reactor coolant. TS 2.1.1 is applicable in MODES 1 and 2. The restrictions of the Sls in the CR-3 TS 2.1.2 protect the integrity of the RCS against overpressure. TS 2.1.2 is applicable in MODES 1, 2, 3, 4, and 5.

The licensee proposed to delete the Sls specified in Section 2.0 and Figure 2.1.1-1 of the CR-3 TSs because they only address specific process variables that are no longer applicable to the facility in its permanently shutdown and defueled condition. Because CR-3 has permanently shut down and defueled and submitted certifications under the provisions of 10 CFR 50.82(a)(2), placing fuel in the reactor vessel and resuming power operations are no longer authorized. In this condition, there is no departure from nucleate boiling ratio (DNBR) or peak fuel centerline temperature to be monitored and there are no challenges to RCS integrity.

Because these changes conform to the permanently shutdown and defueled condition of CR-3, the NRC staff concludes that they will have no impact on the continued safe use of the CR-3 facility and, therefore, are acceptable.

3.6.6 Section 3.0, "Limiting Condition for Operation (LCO) Applicability" Section 3.0 of the CR-3 TSs contains the general requirements applicable to all LCOs and SRs and applies at all times unless otherwise stated in TSs.

LCO 3.0.1 establishes the applicability statement within each individual TS as the requirement for when the LCO is required to be met. In its October 29, 2013, application, the licensee

proposed to delete the reference to "modes or other," as well as references to LCOs 3.0.2, 3.0.7, and 3.0.8. However, the licensee modified its request in its June 17, 2014, supplemental letter to retain the reference to LCO 3.0.2 in LCO 3.0.1.

The proposed changes to LCO 3.0.2 are shown below with deletions in strikethrough.

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

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

In its October 29, 2013, application, the licensee proposed to delete the references to LCOs 3.0.5 and 3.0.6, as well as the second sentence. However, the licensee modified its request in its June 17, 2014, supplemental letter to retain the second sentence, resulting in the proposed change only deleting the references to LCOs 3.0.5 and 3.0.6.

LCO 3.0.3 establishes the actions that must be implemented when an LCO is not met and (1) an associated required action and completion time is not met and no other condition applies or (2) the condition of the unit is not specifically addressed by the associated actions. The licensee proposes to delete LCO 3.0.3 in its entirety since LCO 3.0.3 does not apply with the reactor defueled.

LCO 3.0.4 prohibits entering an operational mode or specified applicability condition unless the LCOs are met. The licensee states that LCO 3.0.4 is not proposed for inclusion in the CR-3 POTS since all actions in the POTS have a completion time of "Immediately," making LCO 3.0.4 unnecessary.

LCO 3.0.5 allows equipment removed from service or declared inoperable to be returned to service for the purpose of testing. The licensee states that LCO 3.0.5 is not proposed for inclusion in the CR-3 POTS since there are no LCOs for equipment to be operable or in operation in the POTS.

LCO 3.0.6 establishes an exception to LCO 3.0.2 for supported systems that have a support system LCO specified in the TS. The licensee proposes to delete LCO 3.0.6 in its entirety since there are no support or supported systems that will remain in the proposed POTS.

LCO 3.0.7 pertains to certain special tests and operations required to be performed at various times over the life of the unit. LCO 3.0.7 is associated with LCO 3.1.8, "PHYSICS TESTS Exceptions-MOOE1 ,"which is described in TS 3.1, "Reactivity Control Systems" (see Section 3.6.7 of this SE). The licensee proposes to delete LCO 3.0.7 in its entirety.

LCO 3.0.8 establishes conditions under which systems are considered to remain capable of performing their intended safety function when associated snubbers are not capable of providing their associated support function(s). The licensee proposes to delete LCO 3.0.8 in its entirety.

SR 3.0.1 establishes the requirement that SRs must be met during the modes or other specified conditions in the applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. The licensee proposes to delete the reference to "modes," as well as the requirement which states, "Surveillances do not have to be performed on inoperable equipment or variables outside specified limits."

SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per ..." interval. The SR provides an allowance for extending the frequency for performance of a SR to 1.25 times the nominal frequency. The proposed changes to SR 3.0.2 are shown below with deletions in strikethrough.

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

For Frequenoies speoified as "onoe," the above interval extension does not

~

If a Required Aotion requires performanoe of a Surveillance or its Completion Time requires periodio performanoe on a "onoe per ... " basis, the above Frequenoy extension applies to eaoh performanoe after the initial performanoe.

E>meptions to this Speoifioation are stated in the individual Speoifioations.

SR 3.0.3 establishes the requirements regarding the delay of the declaration of an LCO not being met. The SR provides an allowance of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for performance of the Surveillance and requires a risk evaluation for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . In the June 17, 2014, supplemental letter, the licensee provided additional information regarding the risk evaluation methodology approach used in SR 3.0.3. The licensee proposed no changes for SR 3.0.3.

SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a MODE or other specified condition in the Applicability of an LCO. The licensee proposes to delete references to operating modes and the discussion about shutting down the unit.

The NRC staff has reviewed the proposed changes to Section 3.0 of the CR-3 TSs and has determined that they are consistent with the permanently shutdown and defueled condition of CR-3 and, therefore, will have no impact on the continued safe use of the CR-3 facility.

Specifically, the NRC staff found that the licensee's proposed changes to the following LCOs are acceptable:

The proposed changes to LCO 3.0.1 deleting references to "MODES," and LCOs 3.0.7 and 3.0.8 (as explained below); to LCO 3.0.2 deleting references to LCOs 3.0.5 and 3.0.6 (as explained below); and to LCO 3.0.3, deleting it in its entirety, reflect the change in plant status from operating to permanently shut down and defueled and, therefore, the NRC staff concludes that they are acceptable.

Regarding the proposed deletion of LCO 3.0.4, which prohibits entering an operational mode or specified applicability condition unless the LCOs are met, the NRC staff concludes that it is appropriate since all actions in the POTS will have a completion time of "Immediately."

Because there are no LCOs for equipment to be operable or in operation in the POTS, the NRC staff concludes that it is appropriate to delete LCO 3.0.5. Additionally, the NRC staff concludes that the conforming deletion of the reference to LCO 3.0.5 in LCO 3.0.2 is appropriate.

CR-3 has no support or supported systems that will remain within the proposed POTS and, as such, the requirements of LCO 3.0.6 do not apply to any system in the TSs. Because the requirements of LCO 3.0.6 do not apply to any system in the TSs, the NRC staff concludes that the deletion of LCO 3.0.6 is appropriate. Additionally, the NRC staff concludes that the conforming deletion of the reference to LCO 3.0.6 in LCO 3.0.2 is appropriate.

Because LCO 3.1.8 is being deleted (as discussed in Section 3.6.7 of this SE), LCO 3.0.7 is no longer applicable, and, therefore, the NRC staff concludes that the proposed deletion of LCO 3.0.7 is acceptable. Additionally, the NRC staff concludes that the conforming deletion of the reference to LCO 3.0.7 in LCO 3.0.1 is appropriate.

Regarding the proposed deletion of LCO 3.0.8, all plant systems associated with snubbers are no longer required to be operable as a result of the permanent shut down and defueling of CR-3 and are proposed for deletion from the TSs. As such, LCO 3.0.8 no longer applies to any systems remaining in the TSs. Because LCO 3.0.8 no longer applies to any systems remaining in the TS, the NRC staff concludes that the proposed deletion of LCO 3.0.8 is acceptable.

Additionally, the NRC staff concludes that the conforming deletion of the reference to LCO 3.0.8 in LCO 3.0.1 is appropriate.

The NRC staff reviewed the proposed changes to SR 3.0.1 and SR 3.0.4 and determined that the changes are consistent with the transition to a permanently shutdown and defueled facility.

Since 10 CFR 50.82{a)(2) prohibits the licensee from operating the plant or placing fuel in the reactor vessel, the references to "MODE," and the discussions about shutting down the unit, are no longer applicable. Therefore, the NRC staff concludes that deleting these references is appropriate.

The NRC staff reviewed the proposed changes to SR 3.0.2. The NRC staff finds that the statements to be deleted are no longer necessary because the proposed POTS do not contain Frequencies and Completion Times of the type described in the portions proposed to be deleted. Therefore, the NRC staff concludes that the proposed changes are acceptable.

SR 3.0.3 requires that a risk evaluation be performed for any Surveillance delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and that the risk impact be managed. The NRC staff reviewed the licensee's risk evaluation methodology discussed in the June 17, 2014, supplemental letter. The licensee stated that the proposed evaluations evolved from the shutdown safety risk management.process that was used for refueling outages during the time that the reactor operated, and that the evaluations were developed from NRC Generic Letter 88-17, "Loss of Decay Heat Removal," dated October 17, 1988 (ADAMS Accession No. ML031200496). The licensee stated that the proposed risk evaluation addresses the key safety functions associated with spent fuel cooling and was developed from a previously acceptable risk evaluation used for refueling outages during

the period of reactor operation. The NRC staff concludes that the proposed risk evaluation methodology is acceptable because it addresses functions required to control the source of risk (loss of decay heat removal from reactor fuel) and is developed from a previously acceptable risk evaluation methodology that also addressed loss of decay heat removal from reactor fuel.

Based on the above, the NRC staff concludes that the proposed changes to LCO 3.0.1, LCO 3.0.2, LCO 3.0.3, LCO 3.0.4, LCO 3.0.5, LCO 3.0.6, LCO 3.0.7, LCO 3.0.8, SR 3.0.1, SR 3.0.2, and SR 3.0.4 will have no impact on the continued safe use of the CR-3 facility and are, therefore, acceptable.

3.6.7 Section 3.1, "Reactivity Control Systems" Section 3.1 of the CR-3 TSs contains the following LCOs, actions, and SRs that provide for appropriate control of the reactivity of the reactor and, in turn, protect the integrity of a fission product barrier.

TS 3.1.1, "Shutdown Margin (SOM)"

o Applicability - MODES 3, 4, and 5

TS 3.1.2, "Reactivity Balance" o Applicability - MODES 1 and 2

TS 3.1.3, "Moderator Temperature Coefficient (MTC)"

o Applicability - MODES 1 and 2

TS 3.1.4, "Control Rod Group Alignment Limits" o Applicability - MODES 1 and 2

TS 3.1.5, "Safety Rod Insertion Limits" o Applicability - MODES 1 and 2

TS 3.1.6, "Axial Power Shaping Rod (APSR) Alignment Limits" o Applicability - MODES 1 and 2, when the APSRs are not fully withdrawn

TS 3.1.7, "Position Indicator Channels" o Applicability - MODES 1 and 2

TS 3.1.8, "Physics Tests Exceptions - Mode 1" o Applicability - MODE 1 during PHYSICS TESTS

TS 3.1.9, "Physics Tests Exceptions - Mode 2" o Applicability - MODE 2 during PHYSICS TESTS The licensee proposed to delete all of the above Section 3.1 TSs LCOs since they do not apply with the reactor permanently shut down and defueled.

Section 3.1 of the CR-3 TSs assure the appropriate functional capability of plant equipment required for safe operation of the facility when the reactor is in MODES 1 through 6. However, due to the licensee's certification that it has permanently ceased power operations of the reactor and permanently removed the fuel from the reactor, 10 CFR 50.82(a)(2) prohibits operation of the reactor or placing fuel in the reactor vessel. Since the Section 3.1 TSs only address the associated specific plant equipment, control of process variables, design features, or operating restrictions for an operating reactor, the NRC staff finds that they are no longer applicable.

The NRC staff also reviewed Section 3.1 of the CR-3 TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this SE. These TSs indicate MODES for which the TSs are applicable. MODES, as defined in the TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to MODES for permanently shutdown and defueled reactors, such as CR-3, has no meaning and is not relevant.

Based on the above, the NRC staff concludes that the licensee's proposed change to delete Section 3.1, "Reactivity Control Systems," is acceptable.

3.6.8 Section 3.2, "Power Distribution Limits" Section 3.2 of the CR-3 TSs contains the following LCOs, actions, and SRs that provide for appropriate control of power distribution in the reactor and, in turn, protect the integrity of a fission product barrier.

TS 3.2.1, "Regulating Rod Insertion Limits" o Applicability - MODES 1 and 2

TS 3.2.2, "AXIAL POWER SHAPING ROD (APSR) Insertion Limits" o Applicability - MODES 1 and 2

TS 3.2.3, "AXIAL POWER IMBALANCE Operating Limits" o Applicability - MODE 1 with THERMAL POWER > 40% RTP

TS 3.2.4, "QUADRANT POWER TILT (OPT)"

o Applicability - MODE 1 with THERMAL POWER> 20% RTP

TS 3.2.5, "Power Peaking Limits" o Applicability - MODE 1 The licensee proposed to delete all of the above Section 3.2 TSs LCOs, since they do not apply with the reactor permanently shut down and defueled.

Section 3.2 of the CR-3 TSs assure the appropriate functional capability of plant equipment required for safe operation of the facility when the reactor is in MODES 1 and 2; however, 10 CFR 50.82(a)(2) prohibits operation of the reactor or placing fuel in the reactor vessel. Since the Section 3.2 TSs only address the associated specific plant equipment, control of process

variables, design features, or operating restrictions for an operating reactor, the NRC staff finds that they are no longer applicable.

The NRC staff also reviewed Section 3.2 of the CR-3 TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this SE. These TSs indicate MODES for which the TSs are applicable. MODES, as defined in the TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to MODES for permanently shutdown and defueled reactors, such as CR-3, has no meaning and is not relevant.

Based on the above, the NRC staff concludes that the licensee's proposed change to delete Section 3.2, "Power Distribution Limits," is acceptable.

3.6.9 Section 3.3, "Instrumentation" Section 3.3 of the CR-3 TSs contains the following LCOs, actions, and SRs for instrumentation required for safe operation of the facility.

TS 3.3.1, "Reactor Protection System (RPS) Instrumentation" o Applicability- According to Table 3.3.1-1

TS 3.3.2, "Reactor Protection System (RPS) Manual Reactor Trip" o Applicability- MODES 1 and 2 (MODES 3,4, and 5 as specified in the LCO)

TS 3.3.3, "Reactor Protection System (RPS) - Reactor Trip Module (RTM)"

o Applicability - MODES 1 and 2 (MODES 3,4, and 5 as specified in the LCO)

TS 3.3.4, "Control Rod Drive (CRD) Trip Devices" o Applicability - MODES 1 and 2 (MODES 3,4, and 5 as specified in the LCO)

TS 3.3.5, "Engineered Safeguards Actuation System (ESAS) Instrumentation" o Applicability- According to Table 3.3.5-1

TS 3.3.6, "Engineered Safeguards Actuation System (ESAS) Manual Initiation" o Applicability - MODES 1, 2, and 3 (MODE 4 as specified in the LCO)

TS 3.3.7, "Engineered Safeguards Actuation System (ESAS) Automatic Actuation Logic" o Applicability - MODES 1, 2, and 3 (MODE 4 as specified in the LCO)

TS 3.3.8, "Emergency Diesel Generator (EOG) Loss of Power Start (LOPS)"

o Applicability - MODES 1, 2, 3, and 4.

TS 3.3.9, "Source Range Neutron Flux" o Applicability - MODE 2 ( as specified in the LCO) and MODES 3, 4, and 5

TS 3.3.10, "Intermediate Range Neutron Flux" o Applicability - MODE 2 (MODES 3, 4, and 5 as specified in the LCO)

TS 3.3.11, "Emergency Feedwater Initiation and Control (EFIC) System Instrumentation" o Applicability - According to Table 3.3.11-1

TS 3.3.12, "Emergency Feedwater Initiation and Control (EFIC) Manual Initiation" o Applicability - MODES 1, 2, and 3

TS 3.3.13, "Emergency Feedwater Initiation and Control (EFIC) Automatic Actuation Logic" o Applicability - MODES 1, 2, and 3

TS 3.3.14, "Emergency Feedwater Initiation and Control (EFIC) - Emergency Feedwater (EFW) - Vector Valve Logic" o Applicability - MODES 1, 2, and 3

TS 3.3.15, "Reactor Building (RB) Purge Isolation - High Radiation" o Applicability - During movement of recently irradiated fuel assemblies within containment

TS 3.3.16, "Control Room Isolation - High Radiation" (Deleted by Amendment 199)

TS 3.3.17, "Post Accident Monitoring (PAM) Instrumentation" o Applicability - MODES 1, 2, and 3

TS 3.3.18, "Remote Shutdown System" o Applicability - MODES 1, 2 and 3 The licensee proposed to delete all of the above Section 3.3 TSs LCOs, since they do not apply with the reactor permanently shut down and defueled.

The NRC staff reviewed the proposed change to CR-3 TS 3.3.1 and determined that TS 3.3.1 is necessary to maintain the ability of the RPS to automatically initiate a reactor scram to preserve the integrity of the fuel cladding, preserve the integrity of the primary system barrier, and minimize the energy which must be absorbed, and prevent criticality following a loss of coolant accident. However, 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, therefore, the NRC staff finds that this system is no longer applicable.

The NRC staff reviewed the proposed changes to CR-3 TSs 3.3.2, 3.3.3, and 3.3.4 and determined that this instrumentation was designed to mitigate accidents related to reactor operation. Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, the NRC staff finds that this instrumentation is no longer applicable.

The NRC staff reviewed the proposed changes to CR-3 TSs 3.3.5, 3.3.6, and 3.3.7 and determined that this instrumentation was designed to monitor safety-related power and assure that adequate power was available. This instrumentation monitored safety division buses to ensure that adequate power was available to operate emergency safeguards equipment. The licensee has analyzed the two remaining DBAs, a fuel handling accident in the auxiliary building where the spent fuel pool is located and a radioactive waste handling accident. Emergency safeguards equipment is not relied upon to prevent the occurrence of either OBA, nor to mitigate the consequences of such accidents. Since safeguards equipment is not required for the permanently shutdown and defueled condition, instrumentation to monitor power to safeguards equipment is not required. Therefore, since 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, the NRC staff finds that this instrumentation is no longer applicable.

The NRC staff reviewed the proposed changes to CR-3 TS 3.3.8, which has to do with emergency diesel generators (EDGs). Since CR-3 is permanently shut down and defueled, 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel. The remaining DBAs for CR-3 in its permanently shutdown and defueled condition do not rely on EDGs for accident prevention or mitigation; therefore, the NRC staff finds that the EDGs are no longer applicable.

The NRC staff reviewed the proposed changes to CR-3 TSs 3.3.9, 3.3.10, 3.3.11, 3.3.12, 3.3.13, and 3.3.14 and determined that this instrumentation ensures that there is sufficient information available on selected unit parameters to monitor and assess unit status following an accident, and allows operators to take manual actions specified in the emergency operating procedures. Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, these systems are no longer applicable. Furthermore, the accidents that these systems and components were designed to mitigate are no longer possible; therefore, the NRC staff finds that these systems are no longer applicable.

The NRC staff reviewed the proposed changes to CR-3 TS 3.3.15 and determined that this instrumentation was designed to ensure that safety analysis assumptions regarding reactor building isolation are bounded during spent fuel movement within containment. Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, spent fuel is not allowed in containment and this instrumentation will never be used again. Therefore, the NRC staff finds that TS 3.3.15 is no longer applicable.

The NRC staff reviewed the proposed changes to CR-3 TS 3.3.17 and determined that this instrumentation was designed to ensure that post-accident radiation levels are monitored. The TS is applicable in MODES 1, 2, and 3. Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, CR-3 will never be in any MODE that is applicable to TS 3.3.17. Therefore, the NRC staff finds that TS 3.3.17 will no longer be applicable.

The NRC staff reviewed the proposed changes to CR-3 TS 3.3.18 and determined that this instrumentation was designed to ensure that the remote shutdown panel receives the appropriate inputs for remote shutdown capability. The TS is applicable in MODES 1, 2, and 3.

Since 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, CR-3 will remain in a permanently shutdown status and the remote shutdown

system instrumentation will never be used again. Therefore, the NRC staff finds that TS 3.3.18 is no longer applicable.

The NRC staff also reviewed Section 3.3 of the CR-3 TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this SE. These TSs indicate MODES for which the TSs are applicable. MODES, as defined in the TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to MODES for permanently shutdown and defueled reactors, such as CR-3, has no meaning and is not relevant. Because DEF has submitted certifications of the permanent shut down and defueling of CR-3, it is prohibited, pursuant to 10 CFR 50.82(a)(2), from operating the reactor or placing fuel in the reactor vessel and, therefore, CR-3 is no longer in a configuration or a condition under which the TS MODES apply.

Based on the above, the NRC staff concludes that the licensee's proposed deletion of TS 3.3.1, TS 3.3.2, TS 3.3.3, TS 3.3.4, TS 3.3.5, TS 3.3.6, TS 3.3.7, TS 3.3.8, TS 3.3.9, TS 3.3.10, TS 3.3.11, TS 3.3.12, TS 3.3.13, TS 3.3.14, TS 3.3.15, TS 3.3.17, and TS 3.3.18 is acceptable.

3.6.10 Section 3.4, "Reactor Coolant System (RCS)"

Section 3.4 of the CR-3 TSs contains the following LCOs, actions, and SRs related to the RCS boundary.

TS 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits" o Applicability - MODE 1

TS 3.4.2, "RCS Minimum Temperature for Criticality" o Applicability - MODES 1 and 2

TS 3.4.3, "RCS Pressure and Temperature (P!T) Limits" o Applicability - At all times

TS 3.4.4, "RCS Loops - Mode 3" o Applicability - MODE 3

TS 3.4.5, "RCS Loops - Mode 4" o Applicability - MODE 4

TS 3.4.6, "RCS Loops - MODE 5, Loops Filled" o Applicability - MODE 5 with RCS loops filled

TS 3.4.7, "RCS Loops- MODE 5, Loops Not Filled" o Applicability - MODE 5 with loops not filled

TS 3.4.8, "Pressurizer" o Applicability - MODES 1, 2, and 3

TS 3.4.9, "Pressurizer Safety Valves" o Applicability - MODES 1, 2, and 3

TS 3.4.10, "Pressurizer Power Operated Relief Valve (PORV)"

o Applicability - MODES 1, 2, and 3

TS 3.4.11, "Low Temperature Overpressure Protection (LTOP) System" o Applicability - (MODE 4, as specified in the LCO), MODE 5 and (MODE 6, as specified in the LCO)

TS 3.4.12, "RCS Operational LEAKAGE" o Applicability - MODES 1, 2, 3, and 4

TS 3.4.13, "RCS Pressure Isolation Valve (PIV) Leakage" o Applicability - MODES 1, 2, 3, and (MODE 4, as specified in the LCO)

TS 3.4.14, "RCS Leakage Detection Instrumentation" o Applicability - MODES 1, 2, 3, and 4

TS 3.4.15, "RCS Specific Activity" o Applicability - MODES 1, 2, and (MODE 3 as specified in the LCO)

TS 3.4.16, "Steam Generator (OTSG) Tube Integrity" o Applicability - MODES 1, 2, 3, and 4 The licensee proposed to delete all of the above Section 3.4 TSs LCOs since they do not apply with the reactor permanently shut down and defueled.

Section 3.4 of the CR-3 TSs assure the appropriate functional capability of plant equipment required for safe operation of the facility; however, 10 CFR 50.82(a)(2) prohibits operation of the reactor or placing fuel in the reactor vessel. Since the Section 3.4 TSs only address the associated specific plant equipment, control of process variables, design features, or operating restrictions for an operating reactor, the NRC staff finds that they are no longer applicable.

The NRC staff also reviewed Section 3.4 of the CR-3 TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this SE. The NRC staff notes that these TSs indicate MODES for which the TSs are applicable. MODES, as defined in TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to MODES for permanently shutdown and defueled reactors, such as CR-3, has no meaning and is not relevant. Because DEF has submitted certifications of the permanent shut down and defueling of CR-3, it is prohibited, pursuant to 10 CFR 50.82(a)(2), from operating the reactor or placing fuel in the reactor vessel and, therefore, CR-3 is no longer in a configuration or a condition under which the TS MODES apply. Furthermore, because fuel has been permanently removed from the reactor, the RCS is no longer relevant as a fission product barrier.

Based on the above, the NRG staff concludes that the licensee's proposed change to delete Section 3.4, "Reactor Coolant System (RCS), is acceptable.

3.6.11 Section 3.5, "Emergency Core Cooling Systems (ECCS)"

Section 3.5 of the CR-3 TSs contains the following LCOs, actions, and SRs related to the ECCS.

TS 3.5.1, "Core Flood Tanks (CFTs) "

o Applicability - MODES 1, 2, and (MODE 3 as specified in the LCO)

TS 3.5.2, "ECCS - Operating" o Applicability - MODES 1, 2, and 3

TS 3.5.3, "ECCS - Shutdown" o Applicability - MODE 4

TS 3.5.4, "Borated Water Storage Tank (BWST)"

o Applicability - MODES 1, 2, 3, and 4 The licensee proposed to delete all of the above Section 3.5 TSs LCOs, since they do not apply with the reactor permanently shut down and defueled.

Section 3.5 of the CR-3 TSs assure the appropriate functional capability of the ECCS required for mitigation of DBAs; however, 10 CFR 50.82(a)(2) prohibits operation of the reactor or placing fuel in the reactor vessel. Since the Section 3.5 TSs only address the associated specific plant equipment, control of process variables, design features, or operating restrictions for an operating reactor, the NRG staff finds that they are no longer applicable. The individual Section 3.5 TSs are not required for a decommissioning plant where the reactor is no longer authorized to operate and is permanently defueled.

The NRC staff also reviewed Section 3.5 of the CR-3 TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this SE. The NRG staff notes that these TSs indicate MODES for which the TSs are applicable. MODES, as defined in the TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to MODES for a permanently shutdown and defueled reactor, such as CR-3, has no meaning and is not relevant.

Because DEF has submitted certifications of the permanent shut down and defueling of CR-3, it is prohibited, pursuant to 10 CFR 50.82(a)(2), from operating the reactor or placing fuel in the reactor vessel and, therefore, CR-3 is no longer in a configuration or a condition under which the TS MODES apply.

Based on the above, the NRG staff concludes that the licensee's proposed change to delete Section 3.5, "Emergency Core Cooling Systems (ECCS), is acceptable.

3.6.12 Section 3.6, "Containment Systems" Section 3.6 of the CR-3 TSs contains the following LCOs, actions, and SRs related to containment systems.

TS 3.6.1, "Containment" o Applicability - MODES 1, 2, 3, and 4

TS 3.6.2, "Containment Air Locks" o Applicability - MODES 1, 2, 3, and 4

TS 3.6.3, "Containment Isolation Valves" o Applicability - MODES 1, 2, 3, and 4

TS 3.6.4, "Containment Pressure" o Applicability - MODES 1, 2, 3, and 4

TS 3.6.5, "Containment Air Temperature" o Applicability - MODES 1, 2, 3, and 4

TS 3.6.6, "Reactor Building Spray and Containment Cooling Systems" o Applicability - MODES 1, 2, 3, and 4

TS 3.6.7, "Containment Emergency Sump pH Control System (CPCS)"

o Applicability - MODES 1, 2, 3, and 4 The licensee proposed to delete all of the above Section 3.6 TSs LCOs since they do not apply with the reactor permanently shut down and defueled.

Section 3.6 of the CR-3 TSs assure the operability of the containment and SSCs necessary to contain radioactive material that may be released from the reactor core following a design-basis loss-of-coolant accident (LOCA) in containment. Due to the permanently defueled and shutdown condition of CR-3, 10 CFR 50.82(a)(2} prohibits the licensee from operating the reactor or placing fuel in the reactor vessel that is inside of containment. The CR-3 irradiated fuel is stored within the SFP in the auxiliary building, which is outside of the containment, and the fuel is no longer allowed to be located within containment. Consequently, the radioactive releases that the containment and its associated SSCs were designed to protect against are no longer possible and the Section 3.6 TSs do not provide protection for the cladding of fuel stored in the SFP. Therefore, the NRC staff finds that the individual Section 3.6 TSs are not needed for a permanently shutdown and defueled condition and are no longer applicable.

The NRC staff also reviewed Section 3.6 of the CR-3 TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this SE. The NRC staff notes that these TSs indicate MODES for which the TSs are applicable. MODES, as defined in the TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to MODES for a permanently shutdown and defueled reactor, such as CR-3, has no meaning and is not relevant.

Because DEF has submitted certifications of the permanent shut down and defueling of CR-3, it is prohibited, pursuant to 10 CFR 50.82(a)(2), from operating the reactor or placing fuel in the reactor vessel, and, therefore, CR-3 is no longer in a configuration or a condition under which the TS MODES apply.

Based on the above, the NRC staff concludes that the licensee's proposed change to delete TS Section 3.6, "Containment Systems, is acceptable.

3.6.13 Section 3.7, "Plant Systems" Section 3.7 of the CR-3 TSs contains the following LCOs, actions, and SRs that provide for appropriate functional capability of plant equipment required for safe operation of the facility.

TS 3.7.1, "Main Steam Safety Valves (MSSVs)"

o Applicability - MODES 1, 2, and 3

TS 3.7.2, "Main Steam Isolation Valves (MSIVs)"

o Applicability - MODES 1, 2, and 3

TS 3.7.3, "Main Feedwater Isolation Valves (MFIVs)"

o Applicability - MODES 1, 2, and 3

TS 3.7.4, "Turbine Bypass Valves (TBVs)"

o Applicability - MODES 1, 2, and 3

TS 3.7.5, "Emergency Feedwater (EFW) System" o Applicability - MODES 1, 2, and 3

TS 3.7.6, "Emergency Feedwater (EFW) Tank" o Applicability - MODES 1, 2, and 3

TS 3.7.7, "Nuclear Services Closed Cycle Cooling Water (SW) System" 0 Applicability - MODES 1' 2, 3, and 4

TS 3.7.8, "Decay Heat Closed Cycle Cooling Water (DC) System o Applicability - MODES 1, 2, 3, and 4

TS 3.7.9, "Nuclear Services Seawater System" o Applicability - MODES 1, 2, 3, and 4

TS 3.7.10, "Decay Heat Seawater System" o Applicability - MODES 1, 2, 3, and 4

TS 3.7.11, "Ultimate Heat Sink (UHS)"

o Applicability - MODES 1, 2, 3, and 4

TS 3. 7 .12, "Control Room Emergency Ventilation System (CREVS)"

o Applicability - MODES 1, 2, 3, and 4

TS 3. 7 .13, "Fuel Storage Pool Water Level" o Applicability - During movement of irradiated fuel assemblies in fuel storage pool

TS 3.7.14, "Spent Fuel Pool Boron Concentration" o Applicability - When fuel assemblies are stored in the SFP and a SFP verification has not been performed since the last movement of fuel assemblies in the SFP

TS 3.7.15, "Spent Fuel Assembly Storage" o Applicability - Whenever any fuel assembly is stored in Storage Pool A or Storage Pool B of the SFP

TS 3. 7 .16, "Secondary Specific Activity" o Applicability - MODES 1, 2, 3, and 4

TS 3.7.17, "Steam Generator Level" o Applicability - MODES 1, and 2

TS 3.7.18, "Control Complex Cooling System" o Applicability - MODES 1, 2, 3, and 4

TS 3.7.19, "Diesel Driven EFW (DD-EFW) Pump Fuel Oil, Lube Oil and Starting Air" o Applicability - When the associated DD-EFW Pump is required to be OPERABLE The licensee proposed to delete TSs 3.7.1through3.7.12 and TSs 3.7.16 through 3.7.19. The licensee proposed to retain TS 3.7.13, TS 3.7.14, and TS 3.7.15 and to revise these TSs to delete the REQUIRED ACTIONS note that states that LCO 3.0.3 is not applicable. LCO 3.0.3 is requested to be deleted from the CR-3 TSs and, therefore, the removal of these references to TS 3.0.3 is a conforming change.

The NRG staff reviewed the proposed deletions of TSs 3.7.1 through 3.7.12 and TSs 3.7.16 through 3.7.19. These TSs exist assure the appropriate SSCs will be available to protect the operating reactor core from DBAs and transients, but these TS do not apply to the permanently shutdown and defueled condition of CR-3. Because DEF has submitted certifications of the permanent shut down and defueling of CR-3, it is prohibited, pursuant to 10 CFR 50.82(a)(2),

from operating the reactor or placing fuel in the reactor vessel. Consequently, the operating, transient, and accident conditions that these SSCs were designed to protect against are no longer possible. Therefore, the NRG staff finds that TSs 3.7.1 through 3.7.12 and TSs 3.7.16 through 3. 7.16 are not needed for the permanently shutdown and defueled condition and are no longer applicable.

The NRG staff also reviewed TSs 3.7.1through3.7.12 and TSs 3.7.16 through 3.7.19 proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this SE. The NRG staff notes that these TSs indicate MODES for which the TSs are applicable. MODES, as defined in the TSs, correspond to any

one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel.

The reference to MODES for permanently shutdown and defueled reactors, such as CR-3, has no meaning and is not relevant. Because DEF has submitted certifications of the permanent shut down and defueling of CR-3, it is prohibited, pursuant to 10 CFR 50.82(a)(2), from operating the reactor or placing fuel in the reactor vessel and, therefore, CR-3 is no longer in a configuration or a condition under which the TS MODES apply.

Therefore, based on the evaluation above, the NRG staff concludes that the licensee's proposed change to delete TSs 3.7.1through3.7.12 and TSs 3.7.16 through 3.7.19 is acceptable.

The NRG staff reviewed the proposed retention of TS 3.7.13, "Fuel Storage Pool Water Level,"

and finds that the LCO and corresponding SRs are sufficient to ensure adequate water supply in the SFP to satisfy 10 CFR 50.36(c)(2)(ii)(B). The NRG staff finds that retaining the SR to "Verify the fuel storage pool water level is ~ 156 ft Plant Datum" sufficiently satisfies:

10 CFR 50.36(c)(2)(ii)(B) by preserving the current requirements for safe storage of irradiated fuel.

The initial conditions assumed for the FHA in the FSAR for iodine removal efficiency.

Additional protection for plant personnel and shielding to reduce general area radiation dose during handling and storage.

See Section 3.4 of this SE for further information on the SFP water level.

The NRG staff reviewed the proposed deletion of the reference to LCO 3.0.3 in the Required Actions note in TS 3.7.13, TS 3.7.14, and TS 3.7.15 and determined that this deletion is consistent with the transition to a permanently shutdown and defueled facility. The proposed deletion does not affect the design or use of the existing fuel racks, and, therefore, no criticality analysis was made in association with the deletion. The proposed deletion also keeps intact the systems for the SFP needed to keep the fuel in a subcritical condition The deletion of the Required Actions note that states, "LCO 3.0.3 is not applicable," is also appropriate since LCO 3.0.3 is no longer applicable and is being deleted as part of this license amendment (see Section 3.6.6 of this SE). Therefore, the NRG staff concludes that the proposed deletion of the reference to LCO 3.0.3 in TS 3.7.13, TS 3.7.14, and TS 3.7.15 is acceptable.

3.6.14 Section 3.8, "Electrical Power Systems" Section 3.8 of the CR-3 TSs, "Electrical Power Systems," contains the following LCOs, actions, and SRs related to the availability of alternating current (AC) and direct current (DC) electrical power supplies and distribution systems.

TS 3.8.1, "AC Sources - Operating" o Applicability - MODES 1, 2, 3, and 4

TS 3.8.2, "AC Sources - Shutdown" o Applicability - MODES 5 and 6

TS 3.8.3, "Diesel Fuel Oil, Lube Oil, and Starting Air" o Applicability - When associated EOG is required to be OPERABLE

TS 3.8.4, "DC Sources-Operating" o Applicability - MODES 1, 2, 3, and 4

TS 3.8.5, "DC Sources-Shutdown" o Applicability - MODES 5 and 6

TS 3.8.6, "Battery Cell Parameters" o Applicability - When associated DC electrical power subsystems are required to be OPERABLE

TS 3.8.7, "Inverters - Operating" o Applicability - MODES 1, 2, 3, and 4

TS 3.8.8, "Inverters - Shutdown" o Applicability - MODES 5 and 6

TS 3.8.9, "Distribution Systems - Operating" o Applicability - MODES 1, 2, 3, and 4

TS 3.8.10, "Distribution Systems - Shutdown" o Applicability - MODES 5 and 6 The licensee proposed to delete all of the above Section 3.8 TSs LCOs since they do not apply with the reactor permanently shut down and defueled.

The NRC staff reviewed the proposed deletions of TSs 3.8.1, 3.8.4, 3.8.7, and 3.8.9. These TSs assure the appropriate functional capability of plant equipment required for safe operation of the facility only when the reactor is in MODES 1 through 4. Since 10 CFR 50.82(a)(2) no longer authorizes the licensee to operate CR-3 or place fuel in the CR-3 reactor vessel, the licensee is not authorized to be in MODES 1 through 4 (see Section 3.6.1 of this SE for the elimination of MODES in the TS definitions). The NRC staff finds that TSs 3.8.1, 3.8.4, 3.8.7, and 3.8.9 do not apply with the reactor in a permanently shutdown and defueled condition.

Therefore, the NRC staff finds that TSs 3.8.1, 3.8.4, 3.8.7, and 3.8.9 are not needed for the permanently shutdown and defueled condition and are no longer applicable.

The NRC staff reviewed Section 3.8 of the CR-3 TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this SE. The NRC staff notes that these TSs indicate MODES for which the TSs are applicable. MODES, as defined in the TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to MODES for permanently shutdown and defueled reactors, such as CR-3, has no meaning and is not

relevant. Because DEF has submitted certifications of the permanent shut down and defueling of CR-3, it is prohibited, pursuant to 10 CFR 50.82(a)(2), from operating the reactor or placing fuel in the reactor vessel. Therefore, CR-3 is no longer in a configuration or a condition under which the TS MODES apply.

Since CR-3 has permanently ceased operations, the generation of fission products has ceased and the remaining source term will decay. The SFP water level (TS 3.7.13), SFP boron concentration (TS 3.7.14), and SFP assembly storage (TS 3.7.15) LCOs are retained to preserve the current requirements for safe storage of irradiated fuel in the SFP. The licensee stated that SFP cooling and makeup related equipment and support equipment (e.g., electrical power systems) are not required to be continuously available since, due to the large volume of water in the SFP and the low decay heat load, there is sufficient time to effect repairs, establish alternate sources of makeup flow, or establish alternate sources of cooling in the event of a loss of cooling and makeup flow to the SFP. The NRC staff reviewed the licensee's Permanently Defueled Emergency Plan (ADAMS Accession No. ML13274A584), which cited several sources of makeup to the pools that are available. One source is the Fire Service system, which uses the Diesel Driven Fire Service pump in the event there is a concurrent loss of electrical power.

Because the remaining DBAs or transients do not rely upon the electrical SSCs for accident mitigation and the 10 CFR 50.36 criteria are not met, the NRC staff concludes that the licensee's proposed change to delete the TSs in Section 3.8, Electrical Power Systems, is acceptable.

3.6.15 Section 3.9, "Refueling Operations" Section 3.9 of the CR-3 TSs, "Refueling Operations," contains the following LCOs, actions, and SRs related to refueling operations.

TS 3.9.1, "Boron Concentration" o Applicability - Mode 6

TS 3.9.2, "Nuclear Instrumentation" o Applicability - Mode 6

TS 3.9.3, "Containment Penetrations" o Applicability - During movement of recently irradiated fuel assemblies within containment.

TS 3.9.4, "Decay Heat Removal (OHR) and Coolant Circulation - High Water Level" o Applicability - Mode 6 with the refueling canal water level ;::: 156 ft Plant Datum

TS 3.9.5, "Decay Heat Removal (OHR) and Coolant Circulation - Low Water Level" o Applicability - Mode 6 with the refueling canal water level < 156 ft Plant Datum

TS 3.9.6, "Refueling Canal Water Level" o Applicability - During movement of irradiated fuel assemblies in containment

The licensee proposed to delete all of the above Section 3.9 TSs LCOs since they are only applicable to an operating reactor and, therefore, do not apply to the permanently shutdown and defueled condition of CR-3.

The NRC staff reviewed these proposed changes and determined that the Section 3.9 TSs provide the LCOs and SRs necessary to maintain functionality of plant systems required for refueling operations. Due to the permanently shutdown and defueled status of the plant, 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, thus refueling operations are no longer permitted at CR-3. Therefore, the NRC staff finds that the Section 3.9 TSs are not needed for a permanently shutdown and defueled condition and are no longer applicable.

The NRC staff also reviewed Section 3.9 of the CR-3 TSs proposed for deletion to ensure that these LCOs no longer satisfy the 10 CFR 50.36 criteria for inclusion in TSs, as described in Section 2.1 of this SE. The NRC staff notes that these TSs indicate MODES for which each TS is applicable. MODES, as defined in the TSs, correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning with fuel in the reactor vessel. The reference to MODES for a permanently shutdown and defueled reactor has no meaning and is not relevant. Because DEF has submitted certifications of the permanent shut down and defueling of CR-3, it is prohibited, pursuant to 10 CFR 50.82(a)(2), from operating the reactor or placing fuel in the reactor vessel and, therefore, CR-3 is no longer in a configuration or a condition under which the TS MODES apply.

Based on the above, the NRC staff concludes that the licensee's proposed change to delete the TSs in Section 3.9, "Refueling Operations," is acceptable.

3.6.16 Section 4.0, "Design Features" This section contains a description of the design features of the facility.

TS 4.1 , "Site" The proposed change to TS 4.1, "Site," is shown below with the deletion indicated in strikethrough:

The 4,738 acre site is characterized by a 4,400 foot minimum exclusion radius centered on the Reactor Building; isolation from nearby population centers; sound foundation for structures; an abundant supply of cooling water; an ample supply of emergency-power; and favorable conditions of hydrology, geology, seismology, and meteorology.

Based on the fact that no applicable DBAs or transients rely upon the electrical SSCs for accident mitigation (see Section 3.6.14 of this SE regarding the electrical power systems TSs),

"emergency" power supplies are not needed in the permanently shutdown and defueled condition. Therefore, the NRC staff concludes that this change is acceptable.

TS 4.2, "Reactor Core" TS 4.2, "Reactor Core," provides a general description of the number of and design material requirements for the fuel and control rod assemblies used in the reactor core. The licensee proposed to delete this design features description of the fuel and control rod assemblies since it is only applicable to an operating reactor and does not apply to the permanently shutdown and defueled condition of CR-3.

The proposed changes to TS 4.2 are shown below with deletions in strikethrough and additions underlined:

4.2 Reactor Core Not Used 4.2.1 Fuel Assemblies The reactor shall contain 177 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy 4 or M5 fuel rods with an initial composition of natural or slightly enriched uranium dioxide (UO.?f-as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance Nith approved applications of 1

fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRG Staff approved codes and methods and shovm by tests or analyses to comply with all fuel safety design bases.

4.2.2 Control Rods The reactor core shall contain 60 safety and regulating CONTROL ROD assemblies and 8 AXIAL POWER SHAPING ROD (APSR) assemblies. The material shall be silver indium cadmium or lnconel as approved by the NRG.

The NRC staff reviewed the proposal to delete the reactor core fuel and control rod assemblies design features description from the TSs. Due to the permanently shutdown and defueled status of CR-3, 10 CFR 50.82{a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, therefore, the design features related to the reactor core fuel assemblies and control rods are no longer relevant at CR-3. Therefore, the NRC staff concludes that the proposed deletion is acceptable.

TS 4.3, "Fuel Storage" TS 4.3, "Fuel Storage," provides a description and the requirements regarding prevention of criticality of spent fuel, prevention of SFP drainage, and spent fuel capacity limitations. TS 4.3 is proposed to be retained in the permanently defueled TS.

The proposed changes do not affect the design or use of the existing fuel racks, and, therefore, no criticality analysis was made in association with the changes. The proposed changes also keep intact the systems for the SFP needed to keep the fuel in a subcritical condition. Further, TS 4.3.1, "Criticality," which provides a description and the requirements regarding the

prevention of criticality of spent fuel, is being retained in the permanently defueled TSs for CR-3.

Therefore, the NRC staff concludes that the proposed retention of TS 4.3 is acceptable.

3.6.17 Section 5.0, "Administrative Controls" By letter dated July 11, 2014 (ADAMS Accession No. ML14097A145), the NRC approved CR-3 License Amendment No. 244. The amendment revised and/or removed subsections from TS 5.1, "Responsibility, TS 5.2, "Organization, TS 5.3, "Unit Staff Qualifications," TS 5.6, "Procedures, Programs, and Manuals," TS 5.7.2, "Special Reports", and TS 5.8, "High Radiation Area, that were no longer applicable to the facility in its permanently shutdown and defueled condition. In its October 29, 2013, submittal, the licensee submitted the following additional changes to Section 5.0.

TS 5.2.1, "Onsite and Offsite Organizations" The first proposed change to TS 5.2.1 is shown below with deletions in strikethrough and additions underlined:

Onsite and offsite organizations shall be established for unit operation facility staff and corporate management, respectively.

Due to the permanently shutdown and defueled status of CR-3, 10 CFR 50.82(a)(2) prohibits the licensee from operating the reactor or placing fuel in the reactor vessel, therefore, referencing "unit operation" is no longer relevant at CR-3. However, the licensee will continue to maintain facility staff to use and maintain the SFP and related SSCs. Therefore, the NRC staff concludes that the proposed change is acceptable.

Additionally, the licensee stated that commas were added to sub item (c) to differentiate between the three classifications of personnel who are assured of organizational freedom to perform their assigned functions. This change is considered editorial only and does not change the meaning of the item. The revised paragraph, with the underlined additions, states the following:

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 NRC staff finds that the proposed change (i.e., the addition a comma) is editorial in nature and is acceptable.

3.6.18 TS 5.6.2.3, "Offsite Dose Calculation Manual (ODCM)"

The ODCM contains offsite dose calculation methodologies, the radioactive effluent controls program, and the radiological environmental monitoring activities. The licensee proposes to remove the dose rate limitations on releases of lodine-131 (1-131) and lodine-133 (1-133) because calculations determined that, due to the time period since the reactor was last operated (i.e., greater than 4 years), the spent fuel no longer contains these two isotopes.

The NRC staff evaluated the acceptability of revising the radionuclides for which controls are maintained in the ODCM for limiting the dose rate resulting from the release of radioactive gaseous effluents at the site. When CR-3 was authorized to operate, TS 5.6.2.3 satisfied 10 CFR 50.36(c)(5) relating to administrative controls for procedures, recordkeeping, and reporting requirements for ensuring safe operation of the facility. With CR-3 permanently shutdown and defueled, the licensee states that TS 5.6.2.3 can be revised.

When CR-3 was authorized to operate, 1-131 and 1-133 were being produced in the reactor as radioactive fission products. Since CR-3 is permanently shutdown and defueled, 1-131 and 1-133 are no longer being produced onsite. 1-131 has a half-life of approximately 8 days and 1-133 has a half-life of approximately 1 day. Since CR-3 has been shutdown for at least 4 years and is not authorized to restart per 10 CFR 50.82(a)(2), any 1-131 or 1-133 produced by the reactor has sufficiently decayed. Additionally, no new 1-131 or 1-133 will be produced such that there is no offsite exposure concern from these fission products. Therefore, the NRC staff finds that the administrative controls for 1-131 and 1-133 are no longer applicable and thus concludes that the proposed revision is acceptable.

3.6.19 TS 5.6.2.12, "Ventilation Filter Testing Program (VFTP)"

The TS 5.6.2.12 requires that a program shall be established to implement the required testing of the CREVS. The licensee proposed to remove the Ventilation Filter Testing Program (VFTP) because neither the Auxiliary Building, nor the Control Complex ventilation filters, are credited to provide any decontamination factor for the remaining postulated accidents in the permanently shutdown and defueled condition.

The NRC staff determined that the VFTP is only applicable to systems that have been evaluated by the NRC staff and approved for deletion in the preceding sections of this SE. The NRC staff also confirmed that the accident analysis applicable to the permanently shutdown and defueled condition (i.e., the FHA) does not rely on ventilation filters for accident mitigation. In accordance with 10 CFR 50.82(a)(2), the licensee is prohibited from operating the plant or placing fuel in the reactor vessel. Since there are no DBAs for CR-3 that require operable filters, the NRC staff concludes that the proposed deletion of TS 5.6.2.12 is acceptable.

3.6.20 TS 5.6.2.14, Diesel Fuel Oil Testing Program TS 5.6.2.14, "Diesel Fuel Oil Testing Program," regards the testing of both new and stored fuel oil used to supply the EDGs and diesel driven emergency feedwater pump. The licensee states that the program is proposed for removal because neither the emergency diesels nor the diesel driven emergency feedwater pump perform any safety function in the permanently shutdown and defueled condition of CR-3. The EDG (TS 3.8.1) and EFW system (TS 3.7.5) are both also proposed for removal from the POTS.

The NRC staff determined that, with the irradiated fuel having decayed for a period of greater than 90 days, the EOG fuel oil and lube oil systems are not needed. The requirement for EDGs and the diesel driven emergency feedwater pump, which are supported by the fuel oil being tested per this program, have been proposed for deletion from the TSs. The OBA analyses applicable to the permanently shutdown and defueled condition of CR-3 do not rely on EDGs for

accident mitigation. Therefore, the NRC staff concludes that the proposed deletion of TS 5.6.2.14 is acceptable.

3.6.21 TS 5.6.2.16, "Safety Function Determination Program (SFDP)"

The SFDP ensures that a loss of safety function is detected and that appropriate actions taken.

The licensee proposed to eliminate the SFDP since none of the three LCOs (TS 3. 7 .13, TS 3. 7 .14, and TS 3. 7.15) remaining in the POTS rely on the operability of any active equipment or systems to satisfy the LCO.

Upon entry into TS LCO 3.0.6, an evaluation shall be made to determine if a loss of safety function exists. The program implements the requirements of LCO 3.0.6. LCO 3.0.6 directs an evaluation in accordance with the SFDP to determine if a loss of safety function exists based on the status of redundant TS safety systems and associated support systems (i.e., systems that support the functionality of the safety system) to ensure that the appropriate required actions are taken to maintain overall reactor safety.

With the permanent cessation of reactor operations at CR-3 and the permanent removal of the fuel from the reactor vessel, there are no active SSCs at CR-3 that are required for accident mitigation. Therefore, none of the systems in the current SFDP meet the definition of a safety-related SSC in 10 CFR 50.2 (with the exception of the passive SFP structure).

Therefore, the requirements of the SFDP, which directs cross-train checks of multiple and redundant safety systems, no longer apply.

The NRC staff determined that there are no longer any active safety-related SSCs that continue to function at CR-3 due to its permanently shutdown and defueled status. None of the systems that are currently in the SFDP remain in operation in a safety-related capacity. This is consistent with the removal of all active safety-related systems from the CR-3 POTS as proposed by this licensing action. In addition, the SFDP is invoked by LCO 3.0.6, which is proposed to be deleted in its entirety, as discussed in Section 3.6.6 of this SE. Therefore, the NRC staff concludes that the proposed deletion of TS 5.6.2.16 is acceptable.

3.6.22 TS 5.6.2.21, "Control Complex Habitability Envelope Integrity Program" TS 5.6.2.21 requires that a Control Complex Habitability Envelope Integrity Program be established and implemented. This program ensures that Control Complex Habitability Envelope (CCHE) habitability is maintained such that, with an OPERABLE 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 DBA conditions without personnel receiving radiation exposures in excess of 5 rem TEDE for the duration of the accident.

The revised accident analysis for the FHA accounts for radioactive material inventory in the most recently irradiated fuel elements in the pools after 4 years of decay. The licensee's calculation determined that the dose to occupants of the control room following this accident would be less than one mrem TEDE for an extended occupancy period. The calculation did not credit Control Complex ventilation isolation or the CCHE Integrity Program to limit in-leakage.

Therefore, this TS is proposed for elimination.

The NRC staff evaluated the remaining accident analyses at the permanently shutdown and defueled CR-3 and confirmed that no SSC in the CCHE Integrity Program is used to mitigate the CR, EAB, or LPZ dose consequences, as detailed in Sections 3.1 through 3.4 of this SE. Since CR-3 is permanently shutdown and defueled, and since greater than 4 years of decay time has elapsed since permanent shut down, the remaining DBAs analyses applicable to the facility demonstrate that the dose consequences within the CCHE are acceptable without relying on SSCs remaining functional for accident mitigation, including FHAs. Based on the above, the NRC staff concludes that the proposed deletion of TS 5.6.2.21 is acceptable.

3.6.23 TS 5.7.2, "Special Reports" TS 5.7.2 requires that Special Reports be submitted in accordance with 10 CFR 50.4, 'Written communications, within the time period specified for each report. TS 5.7.2 contains only one Special Report, which is related to the inoperability of Post-Accident Monitoring (PAM)

Instrumentation, however, the PAM (TS 3.3.17) LCO is proposed to be deleted from the POTS.

All other required Special Reports were removed from TS 5.7.2 by License Amendment No. 244, issued July 11, 2014.

Since the PAM instrumentation is proposed to be deleted and since the NRC staff found this deletion to be acceptable in section 3.6.9 of this SE, a Special Report on the PAM instrumentation is no longer required. Therefore, the NRC staff concludes that the proposed deletion of TS 5.7.2 is acceptable.

3.6.24 Section 5.8, "High Radiation Area" TS 5.8.1 and TS 5.8.2 provide requirements regarding access to very high radiation areas. The proposed change to TS 5.8.2 is shown below with additions underlined:

In addition to the requirements of Specification 5.8.1, areas with radiation levels

~ 1000 mrem/hr at 30 cm from the radiation source or from any surface penetrated by the radiation but less than 500 rads/hr at 1 meter from the radiation source or from any surface penetrated by the radiation 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 supervision. Doors shall remain locked except during periods of access by personnel.

The proposed change to TS 5.8.3 is show below with a deletion in strikethrough and additions underlined:

For individual high radiation areas with radiation levels of~ 1000 mrem/hr at 30 cm from the radiation source or from any surface penetrated by the radiation but less than 500 rads/hr at 1 meter from the radiation source or from any surface penetrated by the radiation, 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.

By letter dated February 26, 2014 (ADAMS Accession No. ML14064A343), the licensee provided a regulatory commitment to propose additional changes to TS 5.8.2 and TS 5.8.3 in the POTS LAR that would comply with the requirements of 10 CFR 20.1602, "Control of access to very high radiation areas." The licensee submitted those changes in the supplement dated May 7, 2014. The NRC staff compared the language inserted into TS 5.8.2 and TS 5.8.3 to the language of 10 CFR 20.1602 and finds that the inserted language complies with 10 CFR 20.1602. Additionally, an editorial correction was made by removing the word "be" in TS 5.8.3. Based on the fact that the inserted language complies with 10 CFR 20.1602 and the editorial correction improves readability, the NRC staff concludes that the changes to TS 5.8.2 and TS 5.8.3 are acceptable.

3.7 Changes to Facility Operating License DPR-72 3.7.1 Change to License Condition 2.8.(1)

Current License Condition 2.B.(1) states; (1) Duke Energy Florida, Inc., pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess, use and operate the facility; Revised License Condition 2.B.(1) would state:

(1) Duke Energy Florida, Inc., pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess and use the facility; The license no longer authorizes operation of the facility and only authorizes possession of the existing fuel. The licensee must still be allowed to possess the special nuclear material that is still present onsite as reactor fuel and use the systems required to support safe fuel storage (e.g., the SFP) during the decommissioning period, in accordance with the specified limitations for storage. Therefore, the NRG staff concludes that the proposed change to License Condition 2.B.(1) is acceptable.

3.7.2 Changes to License Condition 2.B.(3)

Current License Condition 2.B.(3) states:

(3) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended;

Revised License Condition 2.B.(3) would state:

(3) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear material configured as reactor fuel, in accordance with the limitations for storage, as described in the Final Safety Analysis Report, as supplemented and amended; The proposed changes remove the authorization for receipt and use of special nuclear material (SNM) as reactor fuel and eliminated the reference to use of the SNM for reactor operations.

The proposed change also limits the possession of SNM pursuant to the license condition as being "configured" as reactor fuel. Pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor. As such, the licensee has no need to receive SNM in the form of reactor fuel and cannot use SNM as reactor fuel for reactor operations. The continued authorization to possess SNM "configured" as reactor fuel is necessary as the licensee currently possess the reactor fuel that was used for the past operations of the reactor. Based on the above, the NRG staff concludes that the licensee has no need to receive or use SNM as reactor fuel and that the proposed change to License Condition 2.B.(3) will have no impact on the continued safe use of the CR-3 facility and is, therefore, acceptable.

3.7.3 Changes to License Condition 2.B.(4)

Current License Condition 2.B.(4) states:

(4). Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; Revised License Condition 2.B.(4) would state:

(4). Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30, 40, and 70 to possess at any time any byproduct, source and special nuclear material as sealed neutron sources used previously for reactor startup, as fission detectors, and sealed sources for reactor instrumentation and to possess and use at any time any byproduct, source, and special nuclear material as sealed sources for radiation monitoring equipment calibration in amounts as required; The proposed changes remove the authorization for receipt and use of byproduct, source, and SNM as sealed neutron sources for reactor startup but retains authorization to possess such sources previously used for reactor startup. The deletion of the authorization to receive and use sources for reactor startup is consistent with the fact that CR-3 is no longer authorized to operate and the continued authorization to possess is consistent with the safe storage of byproduct, source, and SNM. The proposed changes also remove the authorization for use of byproduct, source, and SNM as fission detectors. Since fission is a reactor operation process, this change is also consistent with the fact that CR-3 is no longer authorized to operate.

The proposed changes also clarify the wording allowing the possession and use of byproduct, source, and SNM as sealed sources for radiation monitoring equipment calibration. The license condition proposed by the licensee is similar to the license condition currently in place for multiple decommissioning and operating nuclear power facilities. The incorporation of this licensing provision for the possession and use of byproduct, source, and SNM will allow for appropriate uses and provide reasonable assurance that the applicable radiation protection requirements of 10 CFR Parts 20, 30, 40, and 70 will be met. Based on the above, the NRC staff concludes that the proposed changes to License Condition 2.B.(4) will have no impact on the continued safe use of the CR-3 facility and are, therefore, acceptable.

3.7.4 Change to License Condition 2.C.(1)

Current License Condition 2.C.{1) states:

2.C.(1) Duke Energy Florida, Inc. is authorized to operate the facility at a steady state reactor core power level not in excess of 2609 Megawatts (100 percent of rated core power level).

Revised License Condition 2.C.(1) would state:

2.C.{1) Deleted per Amendment No. 247 The proposed change removes the authorization to operate the CR-3 facility as a power reactor.

By letter dated February 20, 2013, DEF submitted to the NRC a certification in accordance with 10 CFR 50.82{a)(1 )(i) indicating it would permanently cease power operations, and 10 CFR 50.82(a)(1 )(ii) that it had permanently defueled the reactor vessel at CR-3. On May 28, 2011, DEF permanently ceased power operation at CR-3. As a permanently shutdown and defueled facility, and in accordance with 10 CFR 50.82(a)(2}, DEF is no longer authorized to operate the reactor or emplace nuclear fuel into the reactor vessel. Based on the above, the NRC staff concludes that the licensee can no longer operate the reactor and, therefore, that the proposed deletion of License Condition 2.C.(1) is appropriate.

3.7.5 Change to License Condition 2.C.(2)

Current License Condition 2.C.{2} states:

2.C.(2) The Technical Specifications contained in Appendix A, as revised through Amendment No. 245, 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 1O 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 O 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.

i) SR 3.8.4.5 shall be successfully demonstrated prior to entering MODE 4 on the first plant start-up following Refuel Outage 9.

j) SR 3.8.7.1 shall be successfully demonstrated no later than 7 days following the implementation date of the ITS.

k) SR 3.8.8.1 shall be successfully demonstrated no later than 7 days following the implementation date of the ITS.

Revised License Condition 2.C.(2) would state:

2.C.(2) The Technical Specifications contained in Appendix A, as revised through Amendment No. 247, are hereby replaced with the Permanently Defueled Technical Specifications (POTS). Duke Energy Florida, Inc. shall maintain the facility in accordance with the Permanently Defueled Technical Specifications.

The proposed change modifies the license condition to account for the permanently shutdown and defueled state of the CR-3 facility and removes the surveillance requirements that were required upon the issuance of license amendment No. 149. By letter dated February 20, 2013, DEF submitted to the NRG a certification in accordance with 10 CFR 50.82(a)(1)(i) indicating that it had permanently ceased power operations, and 10 CFR 50.82(a)(1 )(ii) indicating that it had permanently defueled the reactor vessel at CR-3. In accordance with 10 CFR 50.82(a)(2),

DEF is no longer authorized to operate the reactor or emplace nuclear fuel into the reactor vessel, and, therefore, the NRC staff concludes that the change to the first paragraph of License Condition 2.C.(2) is acceptable.

The surveillance requirements listed in the remainder of License Condition 2.C.(2) have been satisfied by DEF. Because the surveillance requirements have already been satisfied in their entirety, the NRC staff concludes that their removal from License Condition 2.C.(2) will have no impact on the continued safe use of the CR-3 facility, and, therefore, is acceptable.

3.7.6 Change to License Condition 2.C.(3)

Current License Condition 2.C.(3) states:

2.C.(3) Duke Energy Florida, Inc. shall not operate the reactor in operational Modes 1 and 2 with less than three reactor coolant pumps in operation until safety analyses for less than three pump operation have been submitted by the licensees and approval has been granted by the Commission by amendment to this license.

Revised License Condition 2.C.(3) would state:

2.C.(3) Deleted per Amendment No. 247 The proposed change removes the authorization to operate CR-3 in Modes 1 and 2 with less than three reactor coolant pumps in operation once safety analyses for less than three pump operation have been submitted to and granted by the NRC. As a permanently shutdown and defueled facility, and in accordance with 10 CFR 50.S2(a)(2), DEF is no longer authorized to operate the reactor in any mode or emplace nuclear fuel into the reactor vessel. Since the reactor can no longer operate in any mode, the reactor coolant pumps are not relevant to the decommissioning status of CR-3. Based on the above, the NRC staff concludes that the proposed deletion of License Condition 2.C.(3) will have no impact on the continued safe use of the CR-3 facility and, therefore, is acceptable.

3.7.7 Changes to License Condition 2.C.(5)

Current License Condition 2.C.(5) states:

2.C.(5) Within six months of the date of issuance of this license, Florida Power Corporation*** shall complete modifications to the level indication of the borated water storage tank, and installation of dual setpoint pilot-operated relief valve on the pressurizer.

Revised License Condition 2.C.(5) would state:

2.C.(5) Deleted per Amendment No. 247

- - On April 29, 2013, the name "Florida Power Corporation" was changed to "Duke Energy Florida, Inc."

The proposed change removes the requirement to complete level indication modifications of the borated water storage stank, and to install the setpoint pilot operated relief valve on the pressurizer. DEF stated that this one time action license condition has been satisfied. Based on the completion of the one time required action, the NRC staff concludes that the removal of License Condition 2.C.(5) will have no impact on the continued safe use of the CR-3 facility and, therefore, is acceptable.

3.7.8 Changes to License Condition 2.C.(7)

Current License Condition 2.C.(7) states:

2.C.(7) Prior to startup following the first regularly scheduled refueling outage, Florida Power Corporation*** shall modify to the satisfaction of the Commission, the reactor coolant system flow indication to meet the single failure criterion with regard to pressure sensing lines to the flow differential pressure transmitters.

Revised License Condition 2.C.(7) would state:

2.C.(7) Deleted per Amendment No. 247 The proposed change removes the requirement to modify the RCS flow indication to meet the single failure criterion. DEF stated that this license condition is no longer necessary since it has been satisfied. Based on the fact that the one time required action has been completed, the NRC staff concludes that the removal of License Condition 2.C.(7) will have no impact on the continued safe use of the CR-3 facility and is, therefore, acceptable.

3.7.9 Changes to License Condition 2.C.{8)

Current License Condition 2.C.(8) states:

2.C.(8) Within three months of issuance of this license, Florida Power Corporation*** shall submit to the Commission a proposed surveillance program for monitoring the containment for the purpose of determining any future delamination of the dome.

Revised License Condition 2.C.(8) would state:

2.C.(8) Deleted per Amendment No. 247 The proposed change removes the requirement to submit a proposed surveillance program for monitoring for the containment dome delamination. DEF stated that this one time action license condition has been satisfied. Based on the fact that the one time required action has been completed, the NRC staff concludes that the removal of License Condition 2.C.(8) will have no impact on the continued safe use of the CR-3 facility, and, therefore, is acceptable.

3. 7 .10 Changes to License Condition 2.C.(9)

Current License Condition 2.C.(9) states:

2.C.(9) Fire Protection Duke Energy Florida, Inc. shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the Safety Evaluation Reports, dated July 27, 1979, January 22, 1981, January 6, 1983, July 18, 1985, and March 16, 1988, subject to the following provisions:

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. {Arndt. #147, 1-22-93}

Revised License Condition 2.C.(9) would state:

2.C.(9) Deleted per Amendment No. 247 The proposed change removes the requirement to implement and maintain in effect the approved CR-3 fire protection program that provides reasonable assurance of the ability to achieve and maintain safe shutdown in the event of a fire in accordance with 10 CFR 50.48.

The licensee proposed that 10 CFR 50.48(f) is sufficient to ensure that a fire protection program is maintained and, therefore, having a license condition that also requires a fire protection program for a permanently shutdown and defueled plant is redundant.

Since fuel has been removed from the reactor, achieving and maintaining safe shutdown in the event of a fire is no longer applicable to CR-3. However, other elements of the fire protection program continue to be applicable during decommissioning to address fire events that could result in radiological hazards per the requirements in 10 CFR 50.48(f). The NRG staff reviewed the proposed changes to License Condition 2.C.(9) and, due to the permanently shutdown and defueled condition of CR-3, found that that reliance on 10 CFR 50.48(f) is appropriate to ensure that a fire protection program is maintained. Thus, the NRG staff concludes that the removal of License Condition 2.C.(9) will have no impact on the continued safe use of the CR-3 facility and, therefore, is acceptable.

3.7.11 Changes to License Condition 2.C.(10)

Current License Condition 2.C.(10) states:

2.C.(10) The design of the reactor coolant pump supports need not include consideration of the effects of postulated ruptures of the primary reactor coolant loop piping and may be revised in accordance with Florida Power Corporation*** amendment request of April 24, 1986.

{Added per Arndt. #89, 5-23-86}

Revised License Condition 2.C.(10) would state:

2.C.(10) Deleted per Amendment No. 247 The proposed change removes the consideration of the effects of postulated ruptures of the RCS loop piping on the reactor coolant pump supports. DEF stated that this condition has been implemented and that the reactor coolant system is no longer subject to pressurization.

As a permanently shutdown and defueled facility, and in accordance with 10 CFR 50.82{a)(2),

DEF is no longer authorized to operate the reactor in any mode or emplace nuclear fuel into the reactor vessel. Since the reactor can no longer operate in any mode, the reactor coolant pumps or supports are not relevant to the decommissioning status of CR-3. Based on the above, the NRC staff concludes that the proposed deletion of License Condition 2.C.{10) will have no impact on the continued safe use of the CR-3 facility and, therefore, is acceptable.

3.7.12 Changes to License Condition 2.C.(11)

Current License Condition 2.C.(11) states:

2.C.(11) A system of thermocouples added to the decay heat (DH) drop and Auxiliary Pressurizer Spray (APS) lines, capable of detecting flow initiation, shall be operable for Modes 4 through 1. Channel checks of the thermocouples shall be performed on a monthly basis to demonstrate operability. If either the DH or APS system thermocouples become inoperable, operability shall be restored within 30 days or the NRC shall be informed, in a Special Report within the following fourteen (14) days, of the inoperability and the plans to restore operability. {Arndt. #164, 1-27-98}

Revised License Condition 2.C.(11) would state:

2.C.(11) Deleted per Amendment No. 247 The proposed change removes the requirements to have operable DH and APS line thermocouples operable in Modes 4 through 1. DEF stated that this license condition can be deleted because CR-3 is permanently shutdown and defueled in accordance with 10 CFR 50.82(a)(2) and, therefore, the DH and APS thermocouples are no longer needed.

As a permanently shutdown and defueled facility, and in accordance with 10 CFR 50.82(a)(2),

DEF is no longer authorized to operate the reactor in any mode or emplace nuclear fuel into the reactor vessel. Since the reactor can no longer operate in any mode, the DH and APS thermocouples and RCS are not relevant to the decommissioning status of CR-3. Based on the above, the NRC staff concludes that the proposed deletion of License Condition 2.C.{11) will have no impact on the continued safe use of the CR-3 facility, and, therefore, is acceptable.

3.7.13 Changes to License Condition 2.C.(14)

Current License Condition 2.C.(14) states:

2.C.(14) Mitigation Strategies License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(1.) Fire fighting responses strategy with the following elements:

a. Pre-defined coordinated fire response strategy and guidance

b. Assessment of mutual aid fire fighting assets

c. Designated staging areas for equipment and materials

d. Command and control

e. Training of response personnel (2.) Operations to mitigate fuel damager considering the following:

a. Protection and use of personnel assets

b. Communications

c. Minimizing fire spread

d. Procedures for implementing integrated fire response strategy

e. Identification of readily-available pre-staged equipment

f. Training on integrated fire response strategy

g. Spent fuel pool mitigation measures (3.) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders Revised License Condition 2.C.(14) would state:

2.C.(14) Mitigation Strategies License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(1.) Fire fighting responses strategy with the following elements:

a. Pre-defined coordinated fire response strategy and guidance

b. Assessment of mutual aid fire fighting assets

c. Designated staging areas for equipment and materials

d. Command and control

e. Training of response personnel (2.) Operations to mitigate fuel damage considering the following:

a. Protection and use of personnel assets

b. Communications

c. Minimizing fire spread

d. Procedures for implementing integrated fire response strategy

e. Identification of readily-available pre-staged equipment

f. Training on integrated fire response strategy

g. Spent fuel pool mitigation measures (3.) Actions to minimize release to include consideration of:

a. Water spray scrubbing

b. Dose to onsite responders The proposed change corrects a spelling error in part (2.) of this license condition. The word "damager" was corrected to "damage." Based on the fact that this corrects an administrative error, the NRC staff finds the change acceptable.

3.7.14 Changes to License Condition 2.C.(15)

Current License Condition 2.C.(15) states:

2.C.(15) Upon implementation of Amendment No. 230 adopting TSTF-448, Revision 3, the determination of control complex habitability envelope (CCHE) unfiltered air inleakage as required by Surveillance Requirement (SR) 3.7.12.4, in accordance with ITS 5.6.2.21.3(i) and the assessment of CCHE habitability as required by ITS 5.6.2.21.3(ii),

shall be considered met. Following implementation:

a) The first performance of SR 3. 7 .12.4, in accordance with Specification 5.6.2.21.3(i), shall be within the specified Frequency of 6 years, plus the 18-month allowance of SR 3.0.2, as measured from May 18, 2007, the date of the most recent successful inleakage test.

b) The first performance of the periodic assessment of CCHE habitability, ITS 5.6.2.21.3(ii), shall be within 3 years, plus the 9-month allowance of SR 3.0.2, as measured from May 18, 2007, the date of the most recent successful inleakage test.

c) 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. This assessment will be performed within 60 days of implementation of Amendment.

Revised License Condition 2.C.(15) would state:

2.C.(15) Deleted per Amendment No. 247

The proposed change removes the requirements of TSTF-448 that involve surveillance testing of the CCHE unfiltered air inleakage. In its September 26, 2013, submittal, DEF analyzed the fuel handling accident for dose results at the CR, EAB, and LPZ. The calculation accounts for radioactive material inventory in the most recently irradiated fuel elements in the pools after 4 years of decay. For the analysis, DEF took no credit for CR isolation or filtered recirculation of control room air. The licensee calculated that the accident dose at the CR would be 1.3E-04 rem, which is less than the 10 CFR 50.67 CR dose limit of 5 rem. Based on the fact that the dose at the CR is less than the 10 CFR 50.67 dose limit and that the licensee took no credit for CR isolation or filtered recirculation, the NRC staff concludes that the deletion of License Condition 2.C.(15) will have no impact on the continued safe use of the CR-3 facility and, therefore, is acceptable.

3.7.15 Changes to License Condition 2.E Current License Condition 2.E states:

E. This license is subject to the following antitrust conditions and applies only to Duke Energy Florida, Inc. (DEF):

(1) DEF will interconnect with and coordinate reserves by means of the sale and exchange of emergency bulk power with any entity or entities in its service area* engaging in or proposing to engage in electric bulk power supply on terms that will provide for DEF cost (including a reasonable return) in connection therewith and allow the other participant(s) full access to the benefits of reserve coordination.

Explanatory Notes:**

(a) Interconnections will not be limited to low voltages when higher voltages are available from DEF installed facilities in the area where interconnection is desired, when the proposed arrangement is found to be technically and economically feasible.

(b) Emergency service agreements will not be limited to a fixed amount, but emergency service provided under such agreements will be furnished to the fullest extent available and desired where such supply does not impair service to the supplier's customers.

(c) An example of the type of reserve sharing arrangement available to any participant and which would provide "full access to the benefits of reserve coordination" would be one in which the following conditions would obtain:

The use of term "service area" in no way indicates an assignment or allocation of wholesale market areas. It is intended only as a general indication of an area with.in the State of Florida where DEF provides some class of electric service.

- In order to clarify the commitments, certain explanatory notes have been added where necessary.

(1) DEF and each participant(s) shall provide to the other emergency power if and when available from its own generation, or through its transmission from the generation of others to the extent it can do so without disrupting service to its own customers.

(2) The participant(s) to the reserve sharing arrangement shall, jointly with DEF establish from time to time the minimum reserves to be installed and/or purchased as necessary to maintain in total an adequate reliability of power supply on the interconnected system of DEF and participant(s). The reserve responsibility thus determined shall be calculated as a percentage of peak load. No participant(s) to the interconnection shall be required to maintain more than such percentage as a percentage of its peak load; provided that if the reserve requirements of DEF are increased over and above the amount DEF would be required to maintain without such interconnection then the other participant(s) shall be required to carry or provide for as its reserve responsibility the full amount in kilowatts of such increase. Under no circumstances will minimum spinning or operating reserve requirements exceed the installed reserve requirement.

(d) Interconnection and coordination agreements will satisfy this condition if they do not embody restrictive provisions pertaining to inter-system coordination. Industry practice as developed in this area from time to time will satisfy this condition if it is non-restrictive.

(2) DEF will purchase from or sell "bulk power" to any other entity or entities in the aforesaid area engaging in or proposing to engage in the generation of electric power in bulk, at its cost (including a reasonable return) when such transactions would serve to reduce the overall costs of new bulk power supply for itself or the other participant or participants to the transaction. This refers specifically to the opportunity to coordinate in the planning of new generation, transmission and associated facilities.

Explanatory Notes:

(a) It is not to be considered that this condition requires DEF to purchase or sell bulk power if it finds such purchase or sale unfeasible or its costs in connection with such purchase or sale would exceed its benefits therefrom.

(b) If DEF engages in coordinated development of its bulk power supply system with that of any other bulk power supply system, by selling unit power at the cost of its new power supply, or engages in joint ventures with the same result, DEF shall not refuse

proportional participation on a comparable basis from the same unit to any other entity in its service area (see Commitment I, supra) engaging in or proposing to engage in bulk power supply to the extent it is technically feasible to provide such unit power from the unit or units in question.

(3) DEF will facilitate the exchange of bulk power by transmission over its system between or among two or more entities with which it is interconnected on terms which will fully compensate it for the use of its system to the extent that subject arrangements reasonably can be accommodated from a functional and technical standpoint.

Explanatory Notes:

(a) This condition applies to entities with which DEF may be interconnected in the future as well as those to which it is now interconnected.

(b) DEF is obligated under this condition to transmit bulk power for other entities on the terms stated above, and to include in its planning and construction programs sufficient transmission capacity as required therefor, provided that such other entities give DEF sufficient advance notice as may be required to accommodate the arrangement from a functional and technical standpoint and that the other entities will be obligated to compensate DEF fully for the use of its system.

(4) DEF will sell power in bulk to any entity in the aforesaid area now engaging in or proposing to engage in the retail distribution of electric power.

(5) It is recognized that the foregoing conditions are to be implemented in a manner consistent with the provisions of the Federal Power Act and all rates, charges or practices in connection therewith are to be subject to the approval of regulatory agencies having jurisdiction over them.

Revised License Condition 2.E would state:

E Deleted per Amendment No. 247 The proposed change removes the antitrust provision that would prevent anti-competitive behavior by DEF. Because DEF permanently ceased operation and removed fuel from the reactor vessel, and now lacks authorization to operate the plant pursuant to the restrictions of 10 CFR 50.82(a)(2), the antitrust license condition in its facility operating license is no longer necessary to preclude anti-competitive behavior.

Further, activities involved in decommissioning of the nuclear power plant are subject to both the Federal and Florida State antitrust laws such that NRC oversight of any related antitrust concerns would be duplicative.

Based on the fact that CR-3 is no longer authorized to operate and that the plant's decommissioning is being conducted subject to the Federal and Florida State antitrust laws, the NRC staff concludes that the deletion of License Condition 2.E is acceptable.

3.7.15 Withdrawn Requests to Modify the License In its original October 29, 2013, submittal, DEF proposed to modify License Condition 2.D, "Physical and Cyber Security," by deleting the second paragraph regarding cyber security provisions at CR-3, and proposed to delete the entire License Condition 2.C.(14), "Mitigation Strategy License Condition." After discussion with the NRC staff, DEF withdrew the requests to modify License Conditions 2.D and 2.C.(14) by letter dated May 7, 2014. License Condition 2.D will be maintained in the license in its current form. License Condition 2.C.(14) will have one spelling correction, changing the word "damager" to "damage", but otherwise will be maintained in the license in its current form.

3.8 Continued Safe Operations As stated above, 10 CFR 50.51 (b) requires licensees to take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition.

The structures and components required to maintain the safe operation of the facility during the decommissioning period may remain operational beyond the licensed operating period. Under the provisions of 10 CFR 50.82, the licensee must complete decommissioning within 60 years of permanent cessation of operations. Therefore, structures and components, such as those used to maintain the SFP, can remain in operation for up to 60 years after the permanent cessation of operations.

In its supplemental letter dated March 6, 2015, the licensee provided a technical rationale for not needing an aging management program for the neutron absorbing materials in the SFP, the fire service system, and the radiation monitoring system prior to December 31, 2019. With regard to the carborundum neutron absorbing material in the SFP, the licensee extrapolated past coupon tests to show that the neutron poison weight loss would be expected to remain well within acceptable limits in 2023. Additionally, there is no coupon program for the Boral neutron absorbing material, however, the licensee cited industry operating experience that suggests that fuel can be stored safely past 2020.

In its March 6, 2015, supplemental letter, the licensee provided the following regulatory commitment regarding the degradation monitoring of the SFP rack neutron absorbing material:

Regulatory Commitment Due Date/Event If all spent fuel assemblies have not been removed from the spent fuel pool, the December 31, 2019 licensee shall request, prior to that date, an amendment to the license, pursuant to 10 CFR 50.90, to incorporate Boral and Carborundum surveillance programs into the CR-3 Technical Specifications.

The licensee agreed to incorporate this commitment into the CR-3 FSAR. If, by this date, the fuel is not removed from the SFP, the licensee is committed to submitting an aging management program for NRC approval. The scope of the program shall include those long-lived, passive structures and components that are needed to provide reasonable assurance of the safe condition of the spent fuel in the SFP. Once approved, the program shall be described in the CR-3 TSs and shall remain in effect until such time that all spent fuel has been removed from the SFP. Removing all fuel by December 31, 2019, adequately manages the effects of aging in the SFP.

The proposed changes do not affect the design or use of the existing fuel racks, and, therefore, no criticality analysis was made in association with the changes. The license is retaining TS 3.7.14 and TS 3.7.15, which maintain a margin to criticality in the SFP. The proposed changes also keep intact the systems for the SFP needed to keep the fuel in a subcritical condition. Based on the above, the NRC staff concludes that the incorporation of the commitment into the FSAR supports the continued safe operation of the facility and is, therefore, acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the State of Florida official was notified of the proposed issuance of the amendment on August 20, 2015. 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 and changes surveillance requirements. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding which was published in the Federal Register on October 28, 2014 (79 FR 64222). The amendment also relates to changes in recordkeeping, reporting, or administrative procedures or requirements. Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) and 1o CFR 51.22(c)(10).

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

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 amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: D. Frumkin M. Kichline S.Som J. Hickman C. England M. Greenleaf S. Jones J. Raval D. Spaulding-Yeoman J. Parillo H. Jones D. Cunanan R. Grover Date: September 4, 2015

T. Hobbs 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.

Sincerely, IRA/

Michael D. Orenak, 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. 247 to DPR-72

2. Safety Evaluation cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC KWood, NRA LPL4-2 Reading DCunanan, NRR RidsACRS_MailCTR Resource RidsNrrLAPBlechman Resource RidsNrrDraAfpb Resource RidsNrrPMCrystalRiver Resource RidsNrrDraArcb Resource RidsRgn2Mai1Center Resource RidsNrrDssStsb Resource DFrumkin, NRA RidsNrrDeEeeb Resource MKichline, NRA RidsNrrDeEicb Resource SSom,NRR RidsNmssDuwpRDB Resource JHickman, NMSS RidsNrrDssSpbp Resource CEngland, OGG RidsNrrDssScvb Resource MGreenleaf, NRA RidsNrrDssSrxb Resource SJones, NRA RidsNrrDlrRsrg Resource JRaval, NRA RidsNrrDorlDpr Resource DSpaulding-Yeoman, NRA RidsNrrDorllpl2-2 Resource J Parillo, NRR RidsNrrDssStsb Resource RGrover, NRA HJones, NRR ADAMS A ccess1on No.: ML152248286 *SE memo **b1yema1*1 OFFICE NRR/DORULPL4-2/PM NRR/DORULPL4-2/LA NRR/DRA/AFPB/BC** NRR/DRA/ARCB/BC*

NAME MOrenak PBlechman DFrumkin US hoop DATE 8/20/15 8/18/15 12/23/13 12/4/14 OFFICE NRR/DSS/STSB/BC* NRR/DE/EEEB/BC** NMSS/DUWP/RDB/BC* NRR/DSS/SBPB*

NAME AElliott JZimmerman BWatson GCasto DATE 5/28/15 2/13/15 3/4/14 10/27/14 OFFICE NRR/DSS/SCVB/BC* NRR/DLR/RSRG/BC** NRR/DSS/SRXB/BC* NRR/DE/EICB/BC*

NAME ADennig SBloom CJackson JThorp DATE 9/11/14 3/16/15 7/23/15 7/9/15 OGC - NLO subject to OFFICE e-mail comments NRR/DORULPL4-2/BC NRR/DORULPL4-2/PM NAME JWachutka MKhanna MOrenak DATE 8/28/15 9/4/15 9/4/15 OFFICIAL RECORD COPY

v t e Letter Sequence Approval
TAC:MF3089, Control Room Habitability (Approved, Closed)
Initiation Acceptance... Supplement, Supplement Administration Questions 10 April 2014 8 January 2015 Answers Results Approval Other:
MONTHYEAR2015-01-08 [Table View]

ML15224B286: Difference between revisions

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SUBJECT:

Dear Mr. Hobbs:

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1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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ML15224B286: Difference between revisions

Latest revision as of 05:10, 25 February 2020

Text

SUBJECT:

Dear Mr. Hobbs:

Enclosures:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

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