Text

License Amendment Request 313, Revision 1, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions and Response to Requests for Additional Information
	ML13255A056
Person / Time
Site:	Crystal River
Issue date:	09/04/2013
From:	Elnitsky J; Duke Energy Florida
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	3F0913-02
	Download: ML13255A056 (87)
	v • d • e

DUKE Crystal River Nuclear Plant DUEG 15760 W. Power Line Street

ENERGY, Crystal River, FL 34428 Docket 50-302 Operating License No. DPR-72 10 CFR 50.90 September 4, 2013 3F0913-02 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - License Amendment Request #313, Revision 1, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions and Response to Requests for Additional Information

References:

1. CR-3 to NRC letter dated April 25, 2013, "Crystal River Unit 3 - License Amendment Request #313, Revision 0, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions" (ADAMS Accession No. ML13128A286)

2. NRC to FPC letter dated March 13, 2013, "Crystal River Unit 3 Nuclear Generating Plant Certification of Permanent Cessation of Operation and Permanent Removal of Fuel from the Reactor" (ADAMS Accession No. ML13058A380)

3. FPC to NRC letter dated April 15, 2013, "Crystal River Unit 3 - Request for Approval of the Certified Fuel Handler Training and Retraining Program" (ADAMS Accession No. ML1313OA1 25)

4. Email from C. Gratton (NRC) to D. Westcott (CR-3) dated July 17, 2013, "RAI Request to amend Section 5.0 of the CR-3 Technical Specifications (MF1504)" (ADAMS Accession No. ML13198A142)

Dear Sir:

Pursuant to 10 CFR 50.90, Duke Energy Florida, Inc. (DEF), formerly known as Florida Power Corporation (FPC), hereby provides License Amendment Request (LAR), #313, Revision 1, and the responses to the request for additional information (RAI).

The LAR proposes to revise portions of Section 5.0, Administrative Controls, of the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS).

The changes made in Revision 1 are in response to issues raised in the RAIs (Reference 4). LAR #313, Revision 1, supersedes LAR #313, Revision 0, and replaces it in its entirety.

In Reference 2, the NRC acknowledged CR-3's certification of permanent cessation of power operation and permanent removal of fuel from the reactor vessel. Accordingly, pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel. The basis for this LAR is that certain Administrative Controls in the current CR-3 ITS may be revised or removed for permanently defueled conditions.

DEF requests approval of this LAR by October 31, 2013, with a 30 day implementation period.

The CR-3 Plant Nuclear Safety Committee has reviewed this request and recommended it for approval.

U. S Nuclear Regulatory Commission Page 2 of 2 3F0913-02 This correspondence contains a regulatory commitment identified in Attachment E.

If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Licensing Supervisor, at (352) 563-4796.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 4, 2013.

Sincerely, roject Management and Construction JE/

Attachments:

A. Responses to Request for Additional Information B. Description of Proposed License Amendment Request, Background, Justification for the Request, and Regulatory Analysis C. Proposed Technical Specification Page

Changes, Strikeout and Shadowed Text Format D. Proposed Technical Specification Page Changes, Revision Bar Format E. Regulatory Commitment xc:

NRR Project Manager Regional Administrator, Region II

DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #313, REVISION I ATTACHMENT A RESPONSES TO REQUEST FOR ADDITIONAL INFORMATION

U. S. Nuclear Regulatory Attachment A 3F0913-02 Page 1 of 7 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION By letter dated April 25, 2013, Duke Energy Florida, Inc., formerly known as Florida Power Corporation, requested that the Nuclear Regulatory Commission (NRC) review and approve License Amendment Request (LAR) #313, Revision 0, "Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions." On July 17, 2013, via an e-Mail, the NRC provided a request for additional information (RAI) to Crystal River Unit 3 (CR-3).

RAI TS-1 The proposed change to TS 5.1.1 changes the title of Plant General Manager to Plant Manager; however there is no proposed change that addresses stored nuclear fuel instead of nuclear safety. In addition, TS 5.1.1 states that the manager is responsible for overall unit operation, however pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.82(a)(2),

the operating license for CR-3 no longer authorizes operation of the reactor. Please provide a basis that addresses these differences.

Response to RAI TS-1 Revised proposed Technical Specification (TS) pages are provided in LAR #313, Revision 1, to respond to the RAI observation and to be consistent with the change request.

RAI TS-2 FPC stated that as of May 28, 2011, all the fuel from the reactor has been placed into storage in the spent fuel pools. The NRC staff does not agree with FPC's statement that there is a reduced demand to maintain the safety of the fuel stored in the fuel pools. Please explain why FPC believes there is a reduced demand to maintain the safety of the fuel stored in the fuel pools.

Response to RAI TS-2 The justification for the change to TS 5.2.2.a has been revised to better reflect the intent of the statement to focus on the human performance aspect of maintaining safety of the stored fuel.

The justification now states that there is "... reduced challenge to the operating crew to maintain the safety of fuel stored in the fuel pools." CR-3 last operated on September 26, 2009, and therefore, the most recently irradiated fuel in the pools has been cooling and their fission products have been decaying for almost four years. Conservative calculations demonstrate that currently, on a loss of cooling, the pools will take 107 hours0.00124 days 0.0297 hours 1.76918e-4 weeks 4.07135e-5 months to reach 212 0F, and more than 19 days to boil off inventory to a level of 10 feet above the top of the fuel racks. This provides the operating crew an abundance of time to recognize and mitigate the cause of pool temperature increases or pool inventory losses and therein lays the "reduced challenge to the operating crew." With the plant permanently shutdown and defueled there are significantly fewer transient or accident scenarios that require response from the operating crew. This is in contrast to the immediate responses required for the operating crew of a plant in power operation to a transient or accident.

U. S. Nuclear Regulatory Attachment A 3F0913-02 Page 2 of 7 RAI TS-3 By deleting current paragraph d of TS 5.2.2, CR-3 is allowed to perform fuel handling operations and movement of loads over storage racks containing fuel without a radiation protection staff being on site. FPC's basis for deleting this paragraph does not discuss fuel handling operations or movement of loads over storage racks containing fuel, nor does it discuss any accidents and/or events that might occur during these evolutions. The NRC staff does not agree that fuel movements or movement of loads over storage racks containing fuel should be allowed without on-site radiation protection staff support being available. Please provide a technical basis for not needing a radiation protection staff on site during fuel handling operations and movement of loads over storage racks containing fuel.

Response to RAI TS-3 LAR #313, Revision 1, provides a change to the original TS paragraph 5.2.2.d to require a person qualified in Radiation Protection procedures to be onsite during fuel handling operations and during movement of heavy loads over the fuel storage racks.

RAI TS-4 FPC is proposing to delete the reference to 10 CFR 50.54(m)(2)(i) in paragraph b of TS 5.2.2 below:

Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i) and 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.

Paragraph b of TS 5.2.2 allows during such absences of the shift crew composition that fuel movement and/or movement of loads over fuel are permitted and permits a shift crew position to be unmanned during shift change due to lateness or absence of the oncoming member. The NRC staff does not agree that fuel movements, movement of loads over fuel, and unmanned shift positions during shift turnover should be permitted while the shift crew is less than the minimum. Please provide a technical basis for allowing the fuel movements, movement of loads over fuel, and unmanned shift positions during shift turnover when the shift crew is less than the minimum.

Response to RAI TS-4 The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months grace period to establish minimum shift staffing is existent in the current TS, and is consistent with NUREG-1430, "Standard Technical Specifications -

Babcock and Wilcox Plants." This allowance exists now and no change is being requested.

CR-3 will establish administrative controls to ensure that fuel handling activities and heavy load lifts above spent fuel stored in the spent fuel pool will not occur without a Certified Fuel Handler (CFH) providing oversight of the activity. This administrative control will be established on or before November 30, 2013.

RAI TS-5 On February 24, 2012, the NRC's Office of Enforcement released Enforcement Guidance Memorandum (EGM) 12-001 (ADAMS Accession No.

MLI 1258A243), "Dispositioning

U. S. Nuclear Regulatory Attachment A 3F0913-02 Page 3 of 7 Noncompliance with Administrative Controls Technical Specifications Programmatic Requirements that Extend Test Frequencies and Allow Performance of Missed Tests." The EGM explains that the restructuring of TS chapters during the development of improved standard TS (STSs) resulted in unintended consequences when Section 3.0, "Surveillance Requirement Applicability" provisions were made applicable to Section 5.0 TSs. Specifically, applying STS rules of usage would prohibit licensees from using the Surveillance Requirements (SRs) 3.0.2 and 3.0.3 allowances contained in Section 5.0 TSs. Applying the guidance in the memorandum will ensure SR 3.0.2 or SR 3.0.3, or both, are made available for TS programs and will properly apply 10 CFR 50.36 requirements for tests associated with inservice test activities under 10 CFR 50.55a(f).

During the staffs review of CR-3 TSs, it was noticed that TS 5.6.2.9.c and TS 5.6.2.9.d state that the provisions of SR 3.0.2 and SR 3.0.3 are applicable to inservice testing activities. This is not an accurate statement. SR 3.0.2 and SR 3.0.3 only apply to SRs listed in TSs. SR 3.0.2 and SR 3.0.3 does not apply to inservice test activities under 10 CFR 50.55a(f) that are not listed in TS.

CR-3 TS Section 1.0, "Use and Application," and Section 3.0, "Surveillance Requirement (SR) Applicability," explains how SR 3.0.2 and SR 3.0.3 are applied to SRs in TSs.

In addition, CR-3 TS 5.6.2.9.e states, "Nothing in the ASME OM Code [ASME Code for Operation and Maintenance of Nuclear Power Plants] shall be construed to supersede the requirements of any TS." This statement is contrary to 10 CFR 50.55a(f)(5)(i), which states that the inservice test program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (f)(4) of this section. In other words, if there is a conflict between the 10 CFR 50.55a regulation and TSs, then the regulation should be followed and the TS should be changed to comply with the regulation.

CR-3 TS 5.6.2.9 paragraph c, d, and e are in conflict with10 CFR 50.55a(f) and FPC has not proposed any changes to TS 5.6.2.9, "Inservice Testing Program," so that TS 5.6.2.9 will comply with 10 CFR 50.55a(f).

Please provide changes to TS 5.6.2.9 so that it no longer conflicts with 10 CFR 50.55a(f).

Response to RAI TS-5 LAR #313, Revision 1, is proposing to eliminate TS 5.6.2.9, "Inservice Testing Program." The basis for removal is provided within the LAR. This will eliminate the identified conflict.

RAI TS-6 FPC is proposing to delete TS 5.7.2.a. However, FPC has not proposed any changes to TS 3.3.17 Required Actions B.1 and F.1 in this application and TS 3.3.17 required actions B.1 and F.1 require taking action in accordance with TS 5.7.2.a.

Please remove the proposed change to TS 5.7.2.a, or provide changes that address TS 5.7.2.a in TS 3.3.17 required actions B.1 and F.1.

Response to RAI TS-6 LAR #313, Revision 1, removes the proposed change to TS 5.7.2.a.

This change will be included later in the proposed CR-3 Permanently Defueled TS and will eliminate the stated concern since ITS 3.3.17 will also be proposed for removal.

U. S. Nuclear Regulatory Attachment A 3F0913-02 Page 4 of 7 RAI Health Physics and Human Performance Branch (AHPB)-I Prior to their performance, who approves heavy load moves that could affect the safe handling and storage of nuclear fuel? Is this person a Certified Fuel Handler?

Response to RAI-AHPB-1 Per the Final Safety Analysis Report, Section 9.6.4, CR-3 is committed to the safe handling of heavy loads. All cranes and hoists lifting heavy loads over spent fuel comply with the guidelines of NUREG-0612, "Control of Heavy Loads at Nuclear Power Plants: Resolution of Generic Technical Activity A-36," and are consistent with CR-3's responses and commitments related to the handling of heavy loads.

Various plant procedures are in place for lifting and rigging processes, including training and qualification requirements and defined safe load paths.

Additionally, prior to any spent fuel cask movements, the crane used to lift the cask will be upgraded to single failure proof.

All work activities in the plant, including those that could impact the safe handling and storage of nuclear fuel are approved prior to implementation by the plant work controls center, currently staffed with a current or previously licensed operator. This practice will continue using Certified Fuel Handlers in lieu of licensed operators.

RAI AHPB-2 Can any control actions that could affect the safe handling and storage of nuclear fuel be taken from the control room?

Response to RAI-AHPB-2 The CR-3 control room maintains oversight and command/control for all activities performed in the plant. There are no activities that are performed in the control room, other than some electrical distribution system manipulations, that could directly affect the safe handling of nuclear fuel. However, there are activities that could be performed from the control room that have the potential to affect the safe cooling and storage of the nuclear fuel. These include spent fuel cooling, pool make up, and system alignment. All of these activities are procedurally controlled and performed by trained and qualified operators with oversight by a CFH on shift.

Additionally, in the event of a control action that affects spent fuel cooling, there is approximately 107 hours0.00124 days 0.0297 hours 1.76918e-4 weeks 4.07135e-5 months to boil and greater than 19 days to 10 feet above the fuel racks due to boil-off. This will provide sufficient time to mitigate any consequences from an adverse control action.

RAI AHPB-3 Will the control room remain the center of the command function? If not, where will the alternate command center be located?

Response to RAI-AHPB-3 At this time, the control room is the center of the command function and there are no plans to change this arrangement.

U. S. Nuclear Regulatory Attachment A 3F0913-02 Page 5 of 7 RAI AHPB-4 Will procedures, drawings and instructions continue to be controlled in accordance with 10 CFR, Part 50, Appendix B, Criterion VI, "Document Control," requirements?

Response to RAI-AHPB-4 CR-3 is currently in the process of determining the interim and potentially final configuration of the plant, prior to commencing major decommissioning activities.

It is expected, based on accident analysis results, that there will be very limited safety related Structures, Systems, and Components (SSCs). There will, however, be several augmented quality SSCs that will be maintained for spent fuel cooling and radiation monitoring equipment.

Specific procedures, drawings, and instructions will continue to be controlled commensurate with their safety significance.

Non-safety related SSCs remaining in operation will continue to be controlled with procedures, drawings, and instructions under a more relaxed quality program.

However, most SSCs will be processed to a dormancy state which will be considered abandoned, with energy and fluids removed to the extent practicable. The abandoned state will be established to preclude the SSC from regaining any energy by isolating, draining, removing power, etc., as necessary. Configuration control will remain in place until the SSC is in its final state. There is no guarantee that full configuration control will continue on abandoned SSC.

RAI AHPB-5 Describe how a fuel-handling accident with serious injuries would be identified and mitigated.

Focus on the alarms, displays, and other cues that would allow identification of the problem, how the chain of command would communicate, time constraints, and reporting responsibilities.

Response to RAI-AHPB-5 In the event of a fuel handling accident, CR-3 Abnormal Operating Procedure AP-1090, "Spent Fuel Damage," would be entered. This procedure directs the actions once the Control Room is informed of the event.

The primary notification would be through telephone or radio communication from the staff performing the fuel handling activities. The secondary method of notification would be via radiation monitors alarming in the control room.

Communication protocols in the procedure would require the control room operator to notify the CFH/Emergency Coordinator (EC) first, who then would notify the station duty manager and the station management team and activate the Emergency Plan as needed. The EC would be the individual determining the response of the plant to the event, including declaring the accident condition and determining when assessment activities would commence.

Initial reporting requirements would be determined by the EC, as appropriate, for the condition.

The plant has an established process for identifying and assisting injured personnel even during an accident scenario.

CR-3 performs an annual medical emergency drill to maintain proficiency related to handling a contaminated injured individual; from discovery of the injured individual to hospital activities associated with the potentially contaminated individual.

U. S. Nuclear Regulatory Attachment A 3F0913-02 Page 6 of 7 The fuel handling accident analysis assumes that all fuel rods in a fuel assembly are breached and the resultant radioactive cloud is released from the pool and is released via the Auxiliary Building ventilation.

With the age of the fuel, there is no consequential radioactive iodine released and the majority of the dose would be from Krypton-85. The offsite dose is a small fraction of the Environmental Protection Agency's Protective Action Guide limit and as such, there are no time constraints in staffing the Emergency Response Organization or mitigating the event.

Preliminary evaluation results demonstrate the dose rate on the refueling floor after a fuel handling accident to be a small fraction of the 10 CFR Part 20 allowable limits due to the age of the fuel and the nature of the accident. Should there be an injured individual near the pool during this accident, it would not result in a dose significant event for the injured individual or the responders.

RAI AHPB-6 Confirm that no one in the chain-of-command above the Shift Supervisor is required to be a Certified Fuel Handler or to attend equivalent training.

Response to RAI-AHPB-6 The CR-3 management structure will not require any positions above the Shift Supervisor to be a Certified Fuel Handler or attend equivalent training. CR-3 has determined that based on the permanently shutdown and defueled status of the plant, and the time available to mitigate credible events, operational activities/responses to events are less complex or immediate than the responses in an operating plant.

RAI AHPB-7 How many people in the organization will be Certified Fuel Handlers?

Response to RAI-AHPB-7 Each Operations shift will have one Certified Fuel handler who will also be the Shift Supervisor.

There are sufficient shifts to staff the plant 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day, seven days a week. There will be sufficient off-shift Operations Certified Fuel Handlers trained and qualified to step in as a Shift Supervisor or supplement the shift as necessary.

RAI AHPB-8 Clarify if any current Radiation Protection procedures will be impacted as a result of this amendment request.

Response to RAI-AHPB-8 There are no expected impacts to Radiation Protection procedures resulting from implementing this LAR.

U. S. Nuclear Regulatory Attachment A 3F0913-02 Page 7 of 7 RAI AHPB-9 Confirm that the Explosive Gas and Storage Tank will remain operable until all radioactive gas is eliminated from the site Response to RAI-AHPB-9 As of July 8, 2013, all CR-3 Waste Gas Decay Tanks have been vented and purged.

No consequential residual radiation or radioactive material remains in these tanks. Removal of the relief valves allowing the tanks to be vented to the atmosphere was completed on July 8, 2013.

This satisfies the commitment made in LAR #313, Revision 0, and as such, the commitment will not be repeated in this response. The Waste Gas System currently remains in service with low level monitored releases through the Auxiliary Building ventilation.

DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #313, REVISION 1 ATTACHMENT B DESCRIPTION OF PROPOSED LICENSE AMENDMENT REQUEST, BACKGROUND, JUSTIFICATION FOR THE REQUEST, AND REGULATORY ANALYSIS

U. S. Nuclear Regulatory Attachment B 3F0913-02 Page 1 of 20 DESCRIPTION OF PROPOSED LICENSE AMENDMENT REQUEST, BACKGROUND, JUSTIFICATION FOR THE REQUEST, AND REGULATORY ANALYSIS 1.0 Description of Proposed License Amendment Request Pursuant to 10 CFR 50.90, Duke Energy Florida, Inc., formerly known as Florida Power Corporation, proposes to amend the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS). This License Amendment Request (LAR) proposes to revise and remove certain requirements from the Section 5, Administrative Controls, portions of the ITS that are no longer applicable to CR-3 in the permanently defueled condition.

This revision proposes additional changes in response to a request for additional information from the Nuclear Regulatory Commission (NRC).

Therefore, LAR #313, Revision 1, supersedes LAR #313, Revision 0, and replaces it in its entirety.

2.0

Background

CR-3 has been shutdown since September 26, 2009, when the plant entered the Cycle 16 refueling outage. In the process of creating a construction opening for replacement of steam generators during that outage, a delamination of the outer concrete shell of the containment was discovered.

The construction opening and adjacent concrete shell of the containment was repaired during 2010 and 2011.

During tensioning of the containment prestressing tendons following the concrete repair, delaminations occurred in two other sections of the containment shell.

In consideration of performing a second repair of the containment shell, all fuel was removed from the reactor vessel and placed in storage in the Spent Fuel Pools as of May 28, 2011. On February 5, 2013, Progress Energy Florida, a subsidiary of Duke Energy, announced that CR-3 would be retired.

The NRC has acknowledged CR-3's certification of permanent cessation of power operation and permanent removal of fuel from the reactor vessel, and pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel.

3.0 Justification For The Request The decay heat load in the spent fuel pools is low since freshly irradiated fuel was last added to the pools almost four years ago. Therefore, due to the low decay heat load, significant time is available to respond to loss of cooling or loss of inventory events.

The following generic conclusions from NUREG-1738 and NUREG-1275 demonstrate that CR-3 is in a low risk condition that supports the ITS changes proposed in this LAR.

Based on conservative calculation results presented in NUREG-1738, "Technical Study of Spent Fuel Accident Risk at Decommissioning Nuclear Power Plants," (Reference 1) the time to boil off Spent Fuel Pool inventory down to three feet above the top of the fuel would be approximately 14.5 days with no action taken to restore cooling or inventory.

Even this reduced inventory condition would not result in any significant effect to members of the public. In addition, this long decay time has resulted in essentially depleting the radioactive iodine nuclides from the pool inventory, removing the most significant contributor to offsite consequences.

Spent fuel cooling is being provided by the systems normally used during plant operation and is as described in the CR-3 Final Safety Analysis Report. These systems are typical for a pressurized water reactor plant.

Therefore, the conclusions of NUREG-1275, Volume 12, "Operating Experience Feedback Report - Assessment of Spent Fuel Cooling," (Reference 2) are applicable to CR-3. The report concludes that based on 12 years of operating experience, the staff determined that loss of spent fuel pool coolant

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 2 of 20 inventory has occurred at a rate of about 1 event per 100 reactor years and that none of these events resulted in a water level less than 20 feet above the fuel. It also concludes that loss of cooling with a temperature increase of 20°F has occurred at a rate of approximately 3 events per 1000 reactor years.

A CR-3 specific calculation of spent fuel pool heatup was performed using conservative assumptions and inputs. CR-3 last operated on September 26, 2009, and therefore, the most recently irradiated fuel in the pools has been cooling and their fission products have been decaying for almost four years.

The calculations demonstrate that currently, on a loss of cooling, the pools will take 107 hours0.00124 days 0.0297 hours 1.76918e-4 weeks 4.07135e-5 months to reach 212'F, and more than 19 days to boil off inventory to a level of 10 feet above the top of the fuel racks. The level of 10 ft was chosen since NEI 12-02, Revision 1, "Industry Guidance for Compliance with NRC Order EA-12-051,"

established that 10 ft is the minimum that still provides substantial radiation shielding.

The changes proposed herein are consistent with the operating experience and changes made by other permanently defueled plants and the low risk conditions that exist at CR-3.

The following table identifies each section that is being changed, the proposed changes, and the basis for the changes:

CR-3 Specification Proposed Change and Basis 5.1 Responsibility 5.1.1 This section defines the responsible position for overall unit operation and for approval of each proposed test, experiment or modification to systems or equipment that affect stored nuclear fuel.

The position title is changed from Plant General Manager to Plant Manager; and the scope of the position is changed from overall unit operation to overall facility functions and from the effect on nuclear safety to the effect on stored nuclear fuel.

5.1.2 This section identifies the responsibilities for the control room command function associated with Modes of plant operation, and is based on personnel positions and qualifications for an operating plant. It identifies the need for a delegation of authority for command in an operating plant when the principal assignee leaves the control room.

This section is being changed to eliminate the MODE dependency for this function and personnel qualifications associated with an operating plant. The proposed change establishes the Shift Supervisor as having command of the shift. Delegation of command is unnecessary for CR-3 where all fuel is in the spent fuel pools, and no fuel has been in a critical core for over three and a half years. Any event involving loss of pool cooling would evolve slowly enough that no immediate control room response would be required to protect the health and safety of the public or station personnel.

5.2 Organization 5.2.1 Onsite and Offsite The introduction to this section identifies that organizational

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 3 of 20 Organizations positions are established that are responsible for the safety of the nuclear plant.

5.2. 1. a This is changed to require that positions be established that are responsible for the safe handling and storage of nuclear fuel. This change removes the implication that CR-3 can return to operation.

5.2.1.b This section identifies the organizational position responsible for overall nuclear plant safety, for the safe operation of the plant, and for control of activities necessary for the safe operation and maintenance of the plant.

To reflect the reduced safety concerns from an operating plant to a permanently defueled plant, the responsibility for overall nuclear safety is changed to the overall responsibility for safe handling and storage of nuclear fuel. The assignment of this responsibility is changed from the Vice President - Crystal River Nuclear Plant to the Decommissioning Director. The responsibility to control those onsite activities necessary for safe operation and maintenance of the plant is changed to control those onsite activities necessary for safe handling and storage of nuclear fuel and is changed from the Vice President - Crystal River Nuclear Plant to the Plant Manager.

5.2.1.c This paragraph addresses the requirement for organizational independence of the operations, health physics and quality assurance personnel from operating pressures.

This is changed to replace "operating staff" with "Certified Fuel Handlers" and to replace "their independence from operating pressures" to "their ability to perform their assigned functions."

These changes reflect the changed function of the previous operating staff to a focus on safe handling and storage of nuclear fuel, and to remove the implication that CR-3 can return to operation.

5.2.2 Unit Staff This paragraph addresses that one auxiliary nuclear operator must 5.2.2.a be assigned to the operating shift whenever fuel is in the reactor.

Since this can never occur again at CR-3, the minimum requirement is changed to a minimum crew compliment of one Shift Supervisor and one Non-certified Operator. This reflects the reduced challenge to the operating crew to maintain the safety of fuel stored in the fuel pools. The Certified Fuel Handler will be the Shift Supervisor in accordance with new paragraph 5.2.2.f. In this position, he will retain command and control responsibility for operational decisions and will be responsible for the functions required for event reporting and emergency response.

5.2.2.b This paragraph addresses the conditions under which the minimum shift compliment may be reduced. It contains a reference to 10 CFR 50.54(m) which establishes the minimum requirements for a licensed operating staff for facility operation.

U. S. Nuclear Regulatory Attachment B 3F0913-02 Page 4 of 20 This reference is removed since CR-3 will not return to operation in the future, and the requirement for licensed operating personnel will no longer be required to protect public health and safety.

5.2.2.c This paragraph establishes the requirement for one licensed Reactor Operator to be in the control room when fuel is in the reactor, and for one Senior Reactor Operator to be in the control room during operating Modes 1 - 4.

This paragraph is changed to reflect the requirement for having one qualified watch stander (either a Non-certified Operator or Certified Fuel Handler) in the control room when fuel is stored in the spent fuel pools.

This reflects the reduced requirement for control room personnel training and qualification for a plant authorized for nuclear fuel storage only. CR-3 has submitted a Certified Fuel Handler Training and Retraining Program for NRC approval. The training and qualification for the Non-certified Operator will be determined in accordance with the systems approach to training (SAT) as defined in 10 CFR 55.4. This process ensures that the Non-certified Operator will be qualified to perform the functions necessary to monitor and ensure safe fuel storage is maintained.

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

For any conditions, incidents, or events that occur when the Non-certified Operator is in the control room alone and are not within the scope of qualifications that are possessed by the Non-certified Operator, the Certified Fuel Handler will immediately be contacted for direction by phone, radio, and/or plant page system. This philosophy is deemed acceptable because the necessity to render immediate actions to protect the health and safety of the public is not challenged. A conservative engineering calculation indicates that upon a total loss of spent fuel pool cooling the temperature in the spent fuel pool will take approximately 107 hours0.00124 days 0.0297 hours 1.76918e-4 weeks 4.07135e-5 months to reach 212 0F.

5.2.2.d This paragraph established the requirement for a person qualified in Radiation Protection procedures to be onsite when fuel is in the reactor.

This paragraph is revised to require a person qualified in Radiation Protection procedures to be onsite during fuel handling operations and during movement of heavy loads over the fuel storage racks.

5.2.2.e (New)

A new paragraph is added to establish the requirement for having oversight of fuel handling operations performed by a Certified Fuel

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 5 of 20 Handler.

5.2.2.f (New)

A new paragraph is added to establish that the Shift Supervisor must be a Certified Fuel Handler.

In the permanently defueled plant, the Certified Fuel Handler is the senior position on the operating crew. It is not necessary for the Shift Supervisor to hold a Senior Reactor Operator license if the plant cannot operate to generate power.

5.3 Unit Staff Qualifications 5.3.1 This paragraph establishes that the unit staff must meet or exceed the minimum qualifications of ANSI N18.1, 1971 and for the Radiation Protection Manager to meet the qualifications of NRC Regulatory Guide 1.8, September 1975. The paragraph also establishes the requirements for the Shift Technical Advisor.

This paragraph is changed to remove the requirements for the Shift Technical Advisor since that position is only required for a plant authorized for power operations.

5.3.2 (New)

This new paragraph is added to identify that responsibility for training and retraining of Certified Fuel Handlers is assigned to the Plant Manager.

Sections 5.4 and 5.5 are not currently used.

5.6 Procedures. Programs, and Manuals 5.6.1 Procedures 5.6.1.1 Scope This section states the requirement for procedures to be 5.6.1.1.a established, implemented and maintained covering various plant activities. Subparagraph (a) establishes a requirement to have applicable procedures recommended in NRC Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

This requirement is changed to reduce the scope to procedures applicable to the safe storage of nuclear fuel recommended in NRC Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

This recognizes the reduced requirements associated with protection of the stored nuclear fuel.

5.6.2 Programs and Manuals 5.6.2.1 Not Used 5.6.2.2 Not Used 5.6.2.3 Offsite Dose In 5.6.2.3.2, the authority for approval of changes to the ODCM is Calculation Manual changed from the Plant General Manager to the Plant Manager, (ODCM) consistent with the position title change in 5.1.1.

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 6 of 20 5.6.2.4 Primary Coolant This program was established to minimize leakage from portions of Sources Outside systems outside containment that could contain highly radioactive Containment fluids during a serious transient or accident.

The program is being eliminated since these conditions can no longer exist for a permanently defueled plant 5.6.2.5 Component Cyclic This program provided controls to track cyclic and transient or Transient Limit occurrences to ensure that components were maintained within their design limits.

The program is being eliminated since serious transient or accident conditions can no longer exist for a permanently defueled plant, and the monitored components are not required to assure spent fuel cooling.

5.6.2.6 Not Used 5.6.2.7 Not Used 5.6.2.8 Inservice This program established the controls for periodic inspection of Inspection Program ASME Code Class 1, 2, 3, MC and CC components including applicable supports in accordance with ASME Section XI.

The Preface to ASME Section Xl states:

"The rules of this section constitute requirements to maintain the nuclear power plant and to return the plant to service, following plant outages, in a safe and expeditious manner. The rules require a mandatory program of examinations, testing, and inspections to evidence adequate safety and to manage aging and deterioration effects."

This program is no longer required since CR-3 is permanently defueled and cannot operate. Therefore, the ASME Section Xl systems and components will not be subjected to the temperature and pressure effects that the Inservice Inspection Program was in place to protect against.

5.6.2.9 Inservice Testing This program established the controls for periodic testing of ASME Program Code Class 1, 2, and 3, components including applicable supports in accordance with the ASME Operations and Maintenance (OM)

Code.

ASME OM Code, ISTA-1100, 'Scope' states:

"These requirements apply to:

(a) pumps and valves that are required to perform a specific function in shutting down a reactor to the safe shutdown condition, in maintaining the safe shutdown condition, or in mitigating the consequences of an accident; (b) pressure relief devices that protect systems or portions of systems that perform one or more of these three functions; and (c) dynamic restraints (snubbers) used in systems that perform one or more of these three functions, or to ensure the integrity of the

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 7 of 20 reactor coolant pressure boundary."

This program is no longer required since CR-3 is permanently defueled and cannot operate. Therefore the functions described in (a) above are no longer required.

5.6.2.10 Steam The Steam Generator Program established and implemented Generator (OTSG) practices to ensure that OTSG tube integrity was maintained.

Program This program is no longer required since CR-3 is permanently defueled and cannot operate. Therefore, the steam generator tubes will not be subjected to the temperature and pressure effects that the Steam Generator Program was in place to protect against.

5.6.2.11 Secondary This program provided controls for monitoring secondary water Water Chemistry Program chemistry to inhibit steam generator tube degradation and low pressure turbine disc stress corrosion cracking.

This program is no longer required since CR-3 is permanently defueled and cannot operate. The majority of the Secondary systems are drained.

5.6.2.12 Ventilation Filter No Changes Testing Program (VFTP) 5.6.2.13 Explosive Gas This program provided controls for potentially explosive gas and Storage Tank mixtures contained in the Radioactive Waste Disposal (WD)

Radioactivity Monitoring System, and the quantity of radioactivity contained in gas storage Program tanks or fed into the offgas treatment system.

As of July 8, 2013, all CR-3 Waste Gas Decay Tanks have been vented and purged.

No consequential residual radiation or radioactive material remains in these tanks. Removal of the relief valves allowing the tanks to be vented to the atmosphere was completed on July 8, 2013. Based on these actions this program is being eliminated.

5.6.2.14 Diesel Fuel Oil No Changes Testing Program 5.6.2.15 Not Used 5.6.2.16 Safety Function No Changes Determination Program (SFDP) 5.6.2.17 Technical No Changes Specification (TS) Bases Control Program 5.6.2.18 Core Operating This program established that core operating limits be established Limits Report (COLR) prior to each reload cycle.

This program is being eliminated since no reactor core can be reloaded into the CR-3 reactor.

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 8 of 20 5.6.2.19 Reactor Coolant This program ensured that RCS pressure and temperature limits, System (RCS) Pressure including heatup and cooldown rates, criticality, and hydrostatic And Temperature Limits and leak test limits, be established and documented in the PTLR.

Report (PTLR)

Per NRC Regulatory Guide 1.184, this program is being eliminated. The reactor coolant piping has been drained and is not subject to pressurization. The reactor vessel contains water, but is vented and not subject to over pressurization.

5.6.2.20 Containment This program was established to implement the leakage rate Leakage Rate Testing testing of the containment as required by 10 CFR 50.54(o) and 10 Program CFR 50, Appendix J, Option B, as modified by approved exemptions.

Per NRC Regulatory Guide 1.184, this program is being eliminated.

5.6.2.21 Control Complex No Changes Habitability Envelope Integrity Program 5.7 Reporting Requirements 5.7.1 Routine Reports No Changes 5.7.2 Special Reports The Special Reports 5.7.2 b and 5.7.2.c are being eliminated.

These Special Reports are associated with programs that are being eliminated.

5.8 High Radiation Area 5.8.2 The first paragraph contains the requirements for control of keys to areas with radiation levels a 1000 mrem/hr at 30 cm. It identifies that one of the personnel responsible for control of the keys is the Control Room Supervisor. This is being changed to the Shift Supervisor consistent with 5.1.2.

4.0 Regulatory Analysis 4.1 No Significant Hazards Consideration Determination License Amendment Request (LAR) #313, Revision 1, seeks NRC approval to change certain requirements from the Section 5, Administrative Controls, portions of the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS) that are no longer applicable as CR-3 is in a permanently defueled condition. Duke Energy Florida, Inc. (DEF) has evaluated whether or not the proposed changes would result in a significant hazards consideration by application of the three standards set forth in 10 CFR 50.92(c). No significant hazards will exist if the proposed changes would not:

1. Involve a significant increase in the probability or consequences of an accident previously evaluated; or

2. Create the possibility of a new or different kind of accident from any accident previously evaluated; or

3.

Involve a significant reduction in a margin of safety.

The following table provides the evaluation of each of the proposed changes that provides the basis for the determination that no significant hazards consideration is involved.

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 9 of 20 SIGNIFICANT HAZARDS CONSIDERATION FOR PROPOSED CHANGES Does the proposed Does the proposed Does the proposed change involve a change create the change involve a significant increase in possibility of a new or significant reduction in Identification and Description of Change the probability or different kind of a margin of safety?

consequences of an accident from any accident previously accident previously evaluated?

evaluated?

5.1.1 This section defines the responsible No. The change reflects No. This change reflects No. The position title position for overall unit operation and for that the remaining credible an organizational change proposed here does not approval of each proposed test, experiment or accident is a fuel handling to transition from an involve any physical modification to systems or equipment that affect accident or loss of spent operating plant to a plant limits or parameters stored nuclear fuel and fuel handling, fuel cooling. The change permanently defueled and therefore cannot in the position title of the plant. Such an affect any margin of The responsible position title is changed from responsible person is administrative change safety.

the rlantGespnsible post title icange f administrative and cannot cannot create a new or Manager.

increase the probability or different kind of accident.

consequences of a fuel handling accident.

5.1.2 This section identifies the responsibilities No. This is a change to No. The changes No. The changes for the control room command function the requirements for proposed here for control proposed here for control associated with Modes of plant operation, and control room staffing. In a room staffing cannot room staffing do not is based on personnel positions and permanently defueled create a new or different directly involve any limits qualifications for an operating plant. It identifies plant, the fuel handling kind of accident since or parameters and the need for a delegation of authority for accident is the only they do not change the therefore cannot affect command in an operating plant when the credible accident function of any plant any margin of safety.

principal assignee leaves the control room.

previously evaluated. This structures, systems, or action cannot increase the components.

This section is being changed to eliminate the probability or MODE dependency for this function and consequences of a fuel personnel qualifications associated with an handling accident.

operating plant. The proposed change establishes the Shift Supervisor as having command of the shift.

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 10 of 20 5.2.1.a The introduction to this section identifies that organizational positions are established that are responsible for the safety of the nuclear plant.

This is changed to require that positions be established that are responsible for the safe storage and handling of nuclear fuel. This change removes the implication that CR-3 can return to operation.

No. This change in the description of the functional responsibility of organizational positions places emphasis on the safe storage and handling of nuclear fuel. This focus on their principal responsibility cannot increase the probability or consequences of a fuel handling accident.

No. This change in the description of the functional responsibility of organizational positions cannot create a new or different kind of accident since they do not change the function of any plant structures, systems, or components.

No. This change does not directly involve any physical limits or parameters and therefore cannot affect any margin of safety.

5.1.2.b This section identifies the organizational position responsible for overall nuclear plant safety, for the safe operation of the plant, and for control of activities necessary for the safe operation and maintenance of the plant.

This section is being changed to recognize that the safety concerns for a permanently defueled plant are for the safe storage and handling of nuclear fuel. It changes responsibility for overall safety for storage and handling of nuclear fuel to the Decommissioning Director.

It changes responsibility for control over onsite activities necessary for safe handling and storaae of nuclear fuel to the Plant Manaaer.

No. This change in the description of the functional responsibility of organizational positions places emphasis on the safe handling and storage of nuclear fuel. This focus on their principal responsibility cannot increase the probability or consequences of a fuel handling accident.

No. This change in the description of the functional responsibility of organizational positions cannot create a new or different kind of accident since they do not change the function of any plant structures, systems, or components.

No. This change does not directly involve any physical limits or parameters and therefore cannot affect any margin of safety.

5.2.1.c This paragraph addresses the No. This change No. This change does No. This change does requirement for organizational independence of continues to ensure that not introduce any not directly involve any the operations, health physics and quality personnel in specifically changes to the function limits or parameters and assurance personnel from operating pressures.

identified positions retain of any plant structures, therefore cannot affect independence from systems, or components any margin of safety.

This is changed to replace "operating staff" with organizational pressures therefore it cannot create "Certified Fuel Handlers," and to replace "their and will not increase the a new or different kind of

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 11 of 20 independence from operating pressures" to probability or occurrence of accident.

"their ability to perform their assigned a fuel handling accident.

functions."

5.2.2.a This paragraph addresses that one No. This change, in No. This change does No. This change does auxiliary nuclear operator must be assigned to conjunction with new not introduce any not directly involve any the operating shift whenever fuel is in the paragraph 5.2.2.f, changes to the function limits or parameters and reactor.

continues to ensure that of any plant structures, therefore cannot affect personnel trained and systems, or components any margin of safety.

Since this can never occur again at CR-3, the qualified for the safe therefore it cannot create minimum requirement is changed to a minimum handling and storage of a new or different kind of crew compliment of one Shift Supervisor and nuclear fuel are onsite.

accident.

one Non-certified Operator.

This cannot increase the probability or consequences of a fuel handling accident.

5.2.2.b This paragraph addresses the No. This change No. This change does No. This change does conditions under which the minimum shift continues to ensure that not introduce any not directly involve any compliment may be reduced. It contains a the minimum shift changes to the function limits or parameters and reference to 10 CFR 50.54(m) which compliment of qualified of any plant structures, therefore cannot affect establishes the minimum requirements for a personnel will not be systems, or components any margin of safety.

licensed operating staff for facility operation.

decreased for more than a therefore it cannot create limited period. It removes a new or different kind of This reference is removed since CR-3 will not the qualification accident.

return to operation in the future, and the requirements for personnel requirement for licensed operating personnel who are capable of will no longer be required to protect public responding to operating health and safety.

plant transients and accidents. This does not involve an increase in the probability or consequences of a fuel handling accident.

5.2.2.c This paragraph establishes the No. This change No. This change does No. This change does requirement for one licensed Reactor Operator continues to ensure that not introduce any not directly involve any to be in the control room when fuel is in the personnel trained and changes to the function limits or parameters and reactor and for one Senior Reactor Operator to qualified for the handling of any plant structures, therefore cannot affect

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 12 of 20 be in the control room during operating Modes and storage of nuclear fuel systems, or components any margin of safety.

1-4.

man the control room.

therefore it cannot create This cannot increase the a new or different kind of The change establishes the requirements for probability or accident.

either a Non-certified operator or Certified Fuel consequences of a fuel Handler to be in the control room when fuel is handling accident.

stored in the pools.

5.2.2.d This paragraph established the No. This is an No. This change does No. This change does requirement for a person qualified in Radiation administrative change that not introduce any not directly involve any Protection procedures to be onsite when fuel is cannot affect the changes to the function limits or parameters and in the reactor.

probability of a fuel of any plant structures, therefore cannot affect handling accident. The systems, or components any margin of safety.

This paragraph is revised to require a person consequences of a fuel therefore it cannot create qualified in Radiation Protection procedures to handling accident are a new or different kind of be onsite during fuel handling operations and governed by the accident.

during movement of heavy loads over the fuel characteristics of the fuel storage racks.

element and are not affected by the presence or absence of radiation protection trained personnel.

5.2.2.e (New) A new paragraph is added to No. Certified Fuel No. This change does No. This change does establish the requirement for having oversight Handlers are specifically not introduce any not directly involve any of fuel handling operations to be performed by trained and qualified to changes to the function limits or parameters and a Certified Fuel Handler.

safely handle irradiated of any plant structures, therefore cannot affect fuel. Applying these systems, or components any margin of safety.

qualifications to fuel therefore it cannot create movement ensures that a new or different kind of the probability or accident.

consequences of a fuel handling accident are not increased.

5.2.2.f (New) A new paragraph is added to No. Certified Fuel No. This change does No. This change does establish that the Shift Supervisor must be a Handlers are specifically not introduce any not directly involve any Certified Fuel Handler.

trained and qualified to changes to the function limits or parameters and

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 13 of 20 In the permanently defueled plant, the Certified safely handle irradiated of any plant structures, therefore cannot affect Fuel Handler is the senior position on the fuel. Applying these systems, or components any margin of safety.

operating crew. It is not necessary for the Shift qualifications to the therefore it cannot create Supervisor to hold a Senior Reactor Operator supervision of fuel a new or different kind of license if the plant cannot operate to generate movement ensures that accident.

power.

the probability or consequences of a fuel handling accident are not increased.

5.3.1 This paragraph is changed to remove the No. The Shift Technical No. This change does No. This change does requirements for the Shift Technical Advisor Advisor position was not introduce any not directly involve any since that position is only required for a plant established to assist the changes to the function physical equipment limits authorized for power operations.

control room operating of any plant structures, or parameters and personnel to diagnose the systems, or components therefore cannot affect The paragraph retains the previous cause and advise on the therefore it cannot create any margin of safety.

requirements for the personnel filling unit staff response to operating a new or different kind of positions meet or exceed the minimum transients and accidents.

accident.

qualifications of ANSI N18.1, 1971, and the The absence of a staff Radiation Protection Manager meet or exceed member with those the qualifications of Regulatory Guide 1.8, qualifications does not September 1975.

change the probability or consequences of a fuel handling accident.

5.3.2 This new paragraph is added to identify No. This section No. This change does No. This change does that responsibility for the training and retraining recognizes the importance not introduce any not directly involve any of Certified Fuel Handlers is assigned to the of establishing and changes to the function physical equipment limits Plant Manager.

maintaining Certified Fuel of any plant structures, or parameters and Handler qualifications and systems, or components therefore cannot affect assigns a manager therefore it cannot create any margin of safety.

responsibility for this a new or different kind of program. Training and accident.

retraining Certified Fuel Handlers specifically trained to safely handle nuclear fuel will not increase the probability or

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 14 of 20 consequences of a fuel handling accident.

5.6.1. 1.a This section states the requirement for procedures to be established, implemented and maintained covering various plant activities.

The scope is reduced to procedures applicable to the safe handling and storage of nuclear fuel.

No. The procedures necessary for the safe handling of nuclear fuel are included in the group of procedures applicable to the safe storage of nuclear fuel. With these procedures in effect for fuel handling, the probability or consequences of a fuel handling accident will not be increased.

No. The applicable procedures for the safe storage of nuclear fuel will direct the correct use of fuel handling equipment. These procedures are currently in place and have been used effectively for the safe handling of fuel.

These procedures will not direct the use of plant structures, systems, or components in a different manner, therefore, they cannot create a new or different kind of accident.

No. This change does not directly involve any limits or parameters and therefore cannot affect any margin of safety.

+

I.

5.6.2.3 In this section, the authority for approval of changes to the Offsite Dose Calculation Manual (ODCM) is changed from the Plant General Manager to the Plant Manager consistent with the position title change in 5.1.1.

No. This is a change to the requirements for the position responsible for approving ODCM changes.

In a permanently defueled plant, the fuel handling accident is the only credible accident previously evaluated. This action cannot increase the probability or consequences of a fuel handling accident.

No. The change proposed here, identifying a different position responsible for ODCM change approval, cannot create a new or different kind of accident since this does not change the function of any plant structures, systems, or components.

No. The changes proposed here for ODCM approval do not directly involve any limits or parameters for operating systems and therefore cannot affect any margin of safety.

5.6.2.4 Primary Coolant Sources Outside No. The fuel handling No. This change does No. This change does Containment accident is the only not introduce any not directly involve any credible accident for a changes to the function limits or parameters and permanently defueled of any plant structures, therefore cannot affect

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 15 of 20 This program was established to minimize plant. This change systems, or components any margin of safety.

leakage from portions of systems outside eliminates an inspection therefore it cannot create containment that could contain highly program that is no longer a new or different kind of radioactive fluids during a serious transient or necessary to limit the accident.

accident.

consequences of operating transients and accidents.

This change cannot The program is being eliminated.

i s

th e pro t

increase the probability or consequences of the fuel handling accident.

5.6.2.5 Component Cyclic or Transient Limit No. Eliminating an No. Eliminating an No. This change does administrative event administrative event not directly involve any This program provided controls to track cyclic tracking program cannot tracking program cannot limits or parameters and and transient occurrences to ensure that increase the probability or create a new or different therefore cannot affect components were maintained within their consequences of a fuel kind of accident.

any margin of safety.

design limits.

handling accident.

This program is being eliminated.

5.6.2.8 Inservice Inspection Program No. The Inservice No. This change does No. For an operating Inspection Program does not introduce any plant the Inservice This program required periodic inspections, not apply to nuclear fuel or changes to the function Inspection Program fuel handling equipment.

of any plant structures, provided confidence that boundary components to ensure their continued Therefore eliminating this systems, or components plant systems that were program cannot increase therefore it cannot create either a potential source integrity for power operation.

the probability or a new or different kind of of an accident or occurrence of a fuel accident.

transient or served to This program is being eliminated, handling accident.

mitigate events continued to meet their physical design requirements. For a permanently shutdown plant, no transient or accident can occur, so ending this inspection program cannot affect

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 16 of 20 any marcin of safety.

I 5.6.2.9 Inservice Testing Program This program required periodic testing of ASME Code Class 1, 2, and 3, components including applicable supports in accordance with the ASME Operations and Maintenance (OM)

Code.

No. The Inservice Testing Program does not apply to nuclear fuel or fuel handling equipment.

Therefore eliminating this program cannot increase the probability or occurrence of a fuel handling accident.

No. This change does not introduce any changes to the function of any plant structures, systems, or components therefore it cannot create a new or different kind of accident.

No. For an operating plant, the Inservice Testing Program provided confidence that plant components that were required for safe shutdown would perform as expected. For a permanently shutdown plant, the transients or accidents that would require safe shutdown equipment cannot occur, so ending this testing program cannot affect anv marain of safetv.

5.6.2.10 Steam Generator (OTSG) Program No. The condition of the No. The CR-3 steam No. This change does steam generator tubes generators will remain not directly involve any The Steam Generator Program established and inside the containment has out of service until limits or parameters and implemented practices to ensure that OTSG no effect on fuel handing in removed from the plant.

therefore cannot affect tube integrity was maintained, the auxiliary building within In this state, the any margin of safety.

the spent fuel pools.

condition of the steam This program is being eliminated.

Therefore, eliminating the generator tubes is program cannot increase immaterial and cannot the probability or create a new or different occurrence of a fuel kind of accident.

handling accident.

5.6.2.11 Secondary Water Chemistry Program No. The secondary piping No. This change does No. The components This program provided controls for monitoring systems do not not introduce any this program was secondary water chemistry to inhibit steam interconnect with the fuel changes to the function intended to protect will generator tube degradation and low pressure cooling or fuel handling of any plant structures, no longer function for turbine disc stress corrosion cracking.

systems. Therefore, systems, or components power production.

This program is being eliminated, eliminating the Secondary therefore it cannot create Therefore, eliminating Water Chemistry Program a new or different kind of this program cannot cannot increase the accident.

affect any margin of

U. S. Nuclear Regulatory 3F0913-02 Attachment B Page 17 of 20 probability or occurrence of safety.

a fuel handling accident.

5.6.2.13 Explosive Gas and Storage Tank No. This program is No. This program is No. This change does Radioactivity Monitoring Program required for an operating required for an operating not directly involve any plant where hydrogen and plant where hydrogen limits or parameters and This program provided controls for potentially radioactive gases are and radioactive gases therefore cannot affect explosive gas mixtures contained in the created and must be are created and must be any margin of safety.

Radioactive Waste Disposal (WD) System, and controlled. Controlled controlled. Controlled the quantity of radioactivity contained in gas release of any gases release of any gases storage tanks or fed into the offgas treatment currently in the tanks, in currently in the tanks, in system.

accordance with existing accordance with existing procedures, will ensure procedures, will ensure there will be no hazard to there will be no hazard to This program is being eliminated, public heath and safety.

public heath and safety.

Therefore, elimination of Therefore, elimination of this program cannot this program cannot increase the probability or create a new or different consequences of a fuel kind of accident.

handling accident.

5.6.2.18 Core Operating Limits Report (COLR)

No. This program for No. Since CR-3 can No. Since CR-3 can controlling the design and never load a core into the never load a core into the This program established that core operating operation of the reactor reactor again, eliminating reactor again, eliminating limits be established prior to each reload cycle, core has no bearing on this control program this control program fuel storage after fuel has cannot create a new or cannot affect any margin been moved into the spent different kind of accident.

of safety.

This program is being eliminated, fuel pools. Therefore, eliminating this program cannot increase the probability or occurrence of a fuel handling accident.

5.6.2.19 Reactor Coolant System (RCS)

No. This program contains No. This report is no No. The limits Pressure And Temperature Limits Report no actions or limits that longer needed since the established in this report (PTLR) affect the storage or reactor coolant system is do not apply to nuclear handling of nuclear fuel.

not subject to fuel stored in the spent This program ensured that RCS pressure and Therefore, eliminating this pressurization and the fuel pools. Therefore, program cannot increase reactor contains no fuel.

eliminating this program

U, S. Nuclear Regulatory 3F0913-02 Attachment B Page 18 of 20 temperature limits, including heatup and the probability or Therefore, eliminating cannot affect any margin cooldown rates, criticality, and hydrostatic and occurrence of a fuel this control program of safety.

leak test limits, be established and documented handling accident.

cannot create a new or in the PTLR.

different kind of accident.

This program is being eliminated.

5.6.2.20 Containment Leakage Rate Testing No. Since fuel can never No. This change does No. This change does Program be returned to the CR-3 not introduce any not directly involve any containment, ending changes to the function limits or parameters and This program was established to implement the containment leakage rate of any plant structures, therefore cannot affect leakage rate testing of the containment, testing cannot increase the systems, or components any margin of safety.

probability or occurrence of therefore it cannot create This program is being eliminated in accordance a fuel handling accident.

a new or different kind of with Regulatory Guide 1.184.

accident.

5.7.2 Special Reports No. Eliminating reporting No. Eliminating reporting No. Eliminating reporting requirements for programs requirements that are no requirements that are no This section is being revised to eliminate that are no longer required longer required cannot longer required cannot reporting requirements associated with in a permanently defueled create a new or different affect any margin of programs that are being eliminatedt plant cannot increase the kind of accident.

safety.

probability or occurrence of a fuel handling accident.

5.8.2 High Radiation Area Controls No. This is a change to No. The change No. The changes the requirements for the proposed here, proposed here for key Changes one of the personnel responsible for position title responsible identifying a different control do not directly locked high radiation area key control from the for key control. In a position title responsible involve any limits or Control Room Supervisor to the Shift permanently defueled for key control, cannot parameters and therefore Supervisor.

plant, the fuel handling create a new or different cannot affect any margin accident is the only kind of accident since of safety.

credible accident they do not change the previously evaluated. This function of any plant action cannot increase the structures, systems, or probability or components.

consequences of a fuel handling accident.

U. S. Nuclear Regulatory Attachment B 3F0913-02 Page 19 of 20 4.2 Environmental Impact Evaluation 10 CFR 51.22(c)(9) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if the amendment changes a requirement with respect to use of a facility component within the restricted area provided that (i) the amendment involves no significant hazards consideration, (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (iii) there is no significant increase in individual or cumulative occupational radiation exposure.

Duke Energy Florida, Inc. (DEF) has reviewed this LAR and has determined that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22, no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the proposed license amendment. The following is the basis for this determination:

(i)

The proposed license amendment does not involve a significant hazards consideration, as described in the significant hazards evaluation.

(ii)

As discussed in the Justification for the Request and the No Significant Hazards Consideration, this change does not result in a significant change or significant increase in the release associated with any Design Basis Accident. There will be no significant change in the types or a significant increase in the amounts of any effluents released offsite during normal operation.

There will be no significant change in the types or increase in the amounts of any effluents that may be released offsite and does not involve irreversible environmental consequences beyond those already associated with the CR-3 Final Environmental Statement.

(iii)

The proposed LAR does not result in a significant increase to the individual or cumulative occupational radiation exposure because this is a change to plant equipment that does not interface with radiologically contaminated systems and does not require operator or other actions that could increase occupational radiation exposure.

Therefore, the proposed LAR does not result in a significant increase to the individual or cumulative occupational radiation exposure.

4.3 Applicable Regulatory Requirements/Criteria 10 CFR 50.82(a)(1) requires that when a licensee has determined to permanently cease operations the licensee shall, within 30 days, submit a written certification to the NRC, consistent with the requirements of § 50.4(b)(8), and once fuel has been permanently removed from the reactor vessel, the licensee shall submit a written certification to the NRC that meets the requirements of § 50.4(b)(9).

CR-3 submitted the required certifications by letter dated February 20, 2013. The NRC acknowledged receipt of the required certifications by letter dated March 13, 2013.

10 CFR 50.36 establishes the requirements for Technical Specifications.

50.36(c)(5),

Administrative Controls, identifies that an Administrative Controls section shall be included in the Technical Specifications and shall include provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure

U. S. Nuclear Regulatory Attachment B 3F0913-02 Page 20 of 20 operation of the facility in a safe manner. This LAR is proposing changes to the Administrative Controls section consistent with the decommissioning status of the plant. This LAR applies the principles identified in 50.36(c)(6), Decommissioning, for a facility which has submitted certification required by 50.82(a)(1) and proposes changes to the Administrative Controls appropriate for the CR-3 permanently defueled condition. As 50.36(c)(6) states, this type of change should be considered on a case-by-case basis.

10 CFR 50.54(m) establishes the requirements for having Reactor Operators and Senior Reactor Operators licensed in accordance with Part 55 based on plant conditions. Based on the permanent cessation of operation for CR-3, the requirements of this section no longer apply and it is permissible to remove those positions from the Technical Specifications.

10 CFR 50.55a establishes that each operating license for a boiling or pressurized water-cooled nuclear power facility is subject to the conditions in paragraphs (f) and (g) of this section.

50.55a(f)(1) and (g)(1) require that a plant whose construction permit was issued before January 1, 1971, must meet the requirements of paragraphs (f)(4) and (g)(4). The construction permit for CR-3 was issued September 25, 1968.

50.55a(f)(4) requires implementation of ASME OM Code, "Code for Operation and Maintenance of Nuclear Power Plants," throughout its service life.

Since CR-3 has permanently ceased operation, its service life has ended and the CR-3 Improved Technical Specifications (ITS)

Program for Inservice Testing can be eliminated from the ITS.

50.55a(g)(4) requires implementation of ASME Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," throughout its service life. Since CR-3 has permanently ceased operation, its service life has ended and the CR-3 ITS Program for Inservice Inspection can be eliminated from the ITS.

5.0 References

1. NUREG-1738, "Technical Study of Spent Fuel Accident Risk at Decommissioning Nuclear Power Plants," February 2001 (ADAMS Accession No. ML010430066)

2. NUREG--1275, Volume 12, "Operating Experience Feedback Report - Assessment of Spent Fuel Cooling," February 1997 (ADAMS Accession No. ML010670175)

DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #313, REVISION 1 ATTACHMENT C PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES, STRIKEOUT AND SHADOWED TEXT FORMAT

Responsibility 5.1 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 5.1.2 The Plant General Manager shall be responsible for overall t

nfacilitv functions and shall delegate in writing the succession to this responsibility during his absence.

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

The Control Room Supervisor shall be responsible for the c.ntr. l room command funetiom. During any absence of the Control Room~

Supervisor from the control room while the unit is in MOD.E 1, 2, 3, or 4, an individual with an aetive Senior Reactor Operator

.SRO), llicese shall be designated to assume the I.

trol room command function. During any absence of the Control Room Supervisor from the control room while the unit is in MODE 5 or-6-,

an inividual with am active 5RO) license or Reacter Operator license shall be designated to assume the control room command The Shift Supervisor shall be responsible for the shift command function.

Crystal River Unit 3 5.0-1 Amendment No. ft

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.

The onsite and offsite organizations shall include the positions responsible for activities affecting safet, of the

.u.lear power plant. the safe hnn,,a andhandling-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 Vice President Crystal River Nuclear Plan-t _pe0j pj n Diretor shall have eorporate responsibility for overall responsibility for the safe handling and storaae of nuclear fuel plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining, and providing technical support to the plant to ensure the safe handling and storage of nuclear fuels-afety. The VI-ee President -,rystal River Nulea*r*

Plant Mngeshall be responsible for the overall safe operation of the plant and shall have, tQ control orver-those onsite activities necessary for the safe handling and storaue of nuclear fueloperation and maintenance of the*p*.ant; and

c.

The individuals who train the operating staff=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 independence from operating pressures their ability to perform their assigned functions.

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

a.--One auxiliary nuclear operator shall be assigned to t-e operating shift any time there is fuel in the reactor and (conti nued)

Crystal River Unit 3 5.0-2 Amendment No. 2-0-1

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) an additional auxiliary nu-lear operator shall be assigned in MODES 1, 2, 3 and 4--

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 10 C[R 50.54(m)(2)0) and 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.

e.

At least one licensed Reaete, Operator (RE))

shall be present in the control room whe. fuel is i-the reator.

I, addition, while the umit is in MODE 1, 2, 3, or 4, at leas-t one licensed Senior Reactor Operator

,SR,) shall be present in the control room.

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 when fuel is in the r* ator 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.

d-.-eOversight of fuel handling operations shall be provided by a Certified Fuel Handler.

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

Crystal River Unit 3 5.0-3 Ampndment No. "

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 minimum 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 19 7 5T and the Shift Technical Adviser who shall have a Ba..el.r' degree, or the equivalemt, in a scientific or e.ghrimg diseipline wi.th speeifie traimimg in plant desigm and response and analysis of the plant transients and aee-dents.

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. -149

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 applieable 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. 1-49

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 Iodine-131, Iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days:

Less than or equal to a dose rate of 1500 mrems/yr to any organ;

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 50, Appendix I;

i.

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

Crystal River Unit 3 5.0-9 Amendment No. 2-Hi

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 Primary Coolant

,.Ann 1'r-'Q% Outside Conta inmemt Not Used This program provides controls to mimimize leakage from these portions of systems outside containment that could contain high-ly radioa.tive fluids during a rious trasient or accident to levels as low as' pra--t-cable.

The systems include Low Pressur-e IIjection, Reaeter Buildig Spray and Makeup aId rurificatiol.

The program shall include the followi _~

a.

Preventive maintenance and periodic visual i"spection reqirmetsl and

b. Integrated leak test requirements for each system at refueling cycle intervals or less.

5.6.2.5 Component Cyclic or Trasient Limit Not Used This program provides controls to track the FSAR Table 4.8, cyclice and transient occurrences to ensure that components are maintained within the design limits.

5.6.2.6 Not Used (continued)

Crystal River Unit 3 5.0-10 Amendment No. 2-13

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.7 Not Used 5.6.2.8 i,,servie Inspetion Program Not Used This program provides contrIls for i--e-vice inspection of AlM[

Code Class 1, 2, 3, MC and CC c.mpone..ts, includi-g appli.able supports.

The program shall i"clude the f

.llowi..-

a.

Provisions that inservice i,,pecon of ASM, Code Class 1--,

2, 3, ME and CC components shall be performed in a.crdance with Seetiam X! of the ASME BDoler ad Pressure Vessel Ce and app-1eable Adde-da as required by 10 CFR 5Q.55a; b.

The provisioms of SR 3.0.2 are applicable to the frequencie for perforin isrv.,e inspection aetivities;

e. inserviee inspection of each reactor coolant pump flywheel shall be performed at least once every IV-atyears.

Thte

,.evi inspeecdon shall be either an ultraseonic examination of the volume from the inner bore of the flywheel to the circle of one half the outer radiuis or a surface examinatinn for exposed surfaces of the disassemble flywheels.

The recomemedations delnmeated in Regulatory Guide 1.14, Positions 3, 4, and 5 of Seetion C.4.b shall

d. Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any TS.

(continued)

Crystal River Unit 3 5.0-11 Amendment No. 2-18

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals (continued) 5.6.2.9 inservice Testing ProgramNot tUsed This program provides cotrols for testing of ASM[ C.dE Class 1, 2, and 3 components, incldi,,g appliable supports.

The program shtil ifclude the foll-w mg.

a. Provisions that i

.service testimg of ASME Code Class 1, 2, and 3 pumps, valves, and snubbers shall be perfome in.

accordance with the ASM[ Code for Operatian anrd Maintenanc Addenda as required by 10 CFR 505a

b. Testing freunce specified in the ASME OM Code-and applicable Addenda;

c. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and gacelerated Testing Program for performin,.

inservice testing activitie

d. The provisions of SR 3.0.3 are applicable to in--.

testing activities; and

e. Nothing in the ASM[ OM ode= shall be construed to supersede the requirements of any TS.

(continued) 5.0-12 Amendment No. 23-Crystal River Unit 3

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 Steam Generator (OTSG) rrogra, Not Used A Steam Generator Program shall be established and implemente t

ensure that OTSG tube integrity is maintained.

in addition, the Steam Generator Program shall include the folle..

pro visis:-

a.

rrovisions for condition monitoring assessments.

Condit.

monitoring assessment means an evaluation of the "as found" endition of the tubing with respect to the performance eiteria for structural integrity and accident induced leakage.

The "as found" e.d~t un refers to the conditio of the tubing during an OTSG inspection outage, as determined from the isrceinspection results or by other means, prior to the plugging-of tubes. Conditio,, me, toin assessments shall be conducted during each outage during whiceh the O)TSG tubes ar in-spected

=pl ugged to confi rm that the performance criteria are boing met.

b. Performance criteria for OTSG tube integrity. OTSG tube integrity shall be maintained by meeting the performance.

criteria for tube structural integrity, acciIent in1uce.

leakage, and operational LEAKAGE.

1. Structural integrity perfo1mance criterion:

All in servie steam generator tudbes shall retain structural integrity over the full range of normal operating conditions (inchuding startup, operation in the powe-r range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.

This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primar to secondary pressure differential and a safety-flactor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials.

Apart from the aboe requiements, additional loadim eed~itins associated with the design basis accidents, or combination of accidents in accordance with the design and licensinfg basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

in the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with th loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

{contllnued}ý Crystal River Unit 3 5.0-13 Amendment No. 2-34

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 OTSG Program (continued)

2. Accident induced leakage performance criterion: Th-e piay to secondary accident induced leakage rate for aydsign basis accident, other than an O)TSG tube rupture, shall not: exceed the leakage rate assumed in the accident analysis in terms of total leakage rate fo~r all O)TSGs and leakage rate for an individual O)TSG.

Leakage is not to exceed one galloper minte per OTSG

3. The operational LEAKAGE performance criterion is specified in LCO) 3.4.12, "RCS Operational LEAKAGE."

c. rrovisions for O)TSG tube repair criteria. A~ube found by inservice inspection to contain flaw~s with a depth equal t or exceeding 40% of the nominal tube wall thickness shall be plugged.

d. Provisions for OTSG tube inspections.

Periodic O)TSG tube

,ispections shall be performed. The number and portions-of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks)ý that may be present along the length of the tube, fromth tube to tubesheet weld at the tube inlet to the tube to tubesheet weld at the tube outlet, and that may satisfy th-e applicable tube repair criteria. The tube to tubesheet wl is not pa~rt of the tube. in addition to meeting the requirmets of d.1. d.2. and d.3 below, the inspection scope, inspection methods, and inspection intervals shall b-e such as to ensure that O)ThG tube integrity is maintained until the next O)TSG inspection.

An assessment -of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locationis.

1. Inspect 100% of the tubes in each OTSG during the first, refueling outage following O)TSG replacement.

2. Inspect 100%6 of the tubes at sequential periods of 144.

108. 72. and. thereafter.- 60 effective full power months.

The first sequential period shall be considered to begin after the first inservice insection of the OTSGs.

In addition. intspect50 (0I H- -bes bv the refuelima outage nearest the mid

' t econti nueo Crystal River Unit 3 5.0-14 Amendment No. 2-34

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 oTcSG Program (continued) of the period and the remainin. 5096 b. the refuelin" operate fI., ora~ that, 72 effective ful..

w-rmath

.L

~

k.

n-i-cr

.on.hs or three refueling outages (whichever is less) without being inspected.

3. if crack indications are found in any OTSG tube, the the next inspection for each O)T-G for the degradationm mechanism that caused the cr-ack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less).

If definitive informatio-n, such as from examination of at pulled tube, diagnostice non destructive testing, or engineering evaluation indicates that a crack like indication is no~t associated with a crack~s), then the indication need not be treated as a cractK

e. Provisions for monitoring operational primary to secondary LrEAKAGE.

(eontinued)-

Crystal River Unit 3 5.0-15 Amendment No. 2-34

Procedures, Programs and Manuals 5.6 IITI r PArG T

EiTrkl-rl Al I g I

Brr nl Akl/

Crystal River Unit 3 5.0-16 Amendment No. 2-34

Procedures, Programs and Manuals 5.6

.1 I

I I Crystal River Unit 3 5.0-17 Amendment No. 2-34

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.11 Sec,,dary Water Chemistry Program Not Used This

.r

.vides ontrol for me.dr a

eIs try to-# IIIII t

I steam generator t

-ub -,_ I wit anid lwvv pb "

i"i.-

d stres s orrosion ra The-program,,

a.

identifflcaton of-a sMpling schedule for the eritieal variables and controlpont for these variables,

b.

Identification of the proceduires used to measure the values of the er~tileall vaaria1 s

e.

Identific~ation of process sampling points, whieh shall inelude aoiorn the discarge 1

of tecndemsate pumpsfo evidenee of codeseqpqlakge

d.

Procedures for the recording and management of data;

e.

Prroedures defihh orrective aetions for all off control point ehe-i-tr c-Olnd't'ens; and

f.

A prcdr dent ifyn th auhryrepnilfrte

..terpretation

~

t of tedtanth seucendtMng of administr&'--

events, which is reurd to initit

~re-t-ve acdion.

5.6.2.12 Ventilation Filter Testing Program (VFTP)

A program shall be established to implement the following required testing of the Control Room Emergency Ventilation System (CREVS) per the requirements specified in Regulatory Guide 1.52, Revision 2, 1978, and/or as speified herein, and in accordance with ANSI N510-1975 and ASTM D3803-89 (Re-approved 1995).

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

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

C. Demonstrate for each train of the CREVS that a laboratory test of a sampe of the carbon adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, 1978, meets the laboratory testing criteria of ASTM D 3803-89 (Re-approved 1995) at a temperature of 300C and relative humidity of 95% with methyl iodide penetration of less than 5.0%.

(conti nued)

Crystal River Unit 3 5.0-18 Amendment No. -199

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.12 VFTP (continued)

d.

Demonstrate for each train of CREVS that the pressure drop across the combined roughing filters, HEPA filters and the carbon adsorbers is

< AP=4" water gauge when tested in accordance with Regulatory Guide 1.52, Revision 2, 1978, and ANSI N510-1975 at the system flowrate of between 37,800 and 47,850 cfm.

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

5.6.2.13 Explosive Gas and Storage Tank Radioactivity Monitoring Program Not Used This program provides controls for potentially explosivega mixtures contained in the Radioactive Waste Disposal (WO)

Syst-em, the quantity of radioactivity contained in gas storage tanks fed into the offgas treatment system.

The gaseous radioactivi-ty quantities shall be determined following the methodology in Branch Technical rosition (BTr)

[TSB 1:1 5, 'Postulated Radioactive Release due to Waste Gas System Leak or Failure".

The liquid radwaste quantities shall be determined in accordance with Standard Review Plan, Section 15.7.3, 'Postulated Radioactiv~e Release due to Tank Failures".

The program shall include:

a. The limits for concentrations of hydrogen and oxgn nthe Radioactive Waste Disposal (WD) System and a srelac program to ensure the limits are mainta&ined.

Such limits shall be app'roriate to the system's design criteria, (i-e.,

whether or not the system is designed to withstand a hydrogen explosion)-.

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tan-k-and fed into the offgas treatment system is less than the amount that would result in a whole body exposure of L!0.5 rem to any individuial in an unrestricted area, in the event of a uncontrolled release of the tanks' contents.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage T-ank Radioactivity Monitoring Program surveillance frequencie.

(conti nued)

Crystal River Unit 3 5.0-19 Amendment No. +99

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.14 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.

The program shall include sampling and testing requirements, and acceptance criteria, in accordance with applicable ASTM Standards.

The purpose of the program is to establish the following:

a.

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

2-D fuel oil:

1.

Kinematic Viscosity,

2.

Water and Sediment,

3.

Flash Point,

4.

Specific Gravity API;

b.

Other properties of ASTM D 975 for Grade No.

2-D fuel oil are within limits within 92 days following sampling and addition of new fuel to storage tanks.

c.

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

5.6.2.15 Not Used (continued)

Crystal River Unit 3 5.0-20 Amendment No. 199

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

This program ensures loss of safety function is detected and appropriate actions taken.

Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists.

Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions.

This program implements the requirements of LCO 3.0.6.

The SFDP shall contain the following:

a.

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

b.

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

c.

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

d.

Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed.

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a.

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

b.

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

c.

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

(continued)

Crystal River Unit 3 5.0-21 Amendment No. 185

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

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

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

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

a.

A change in the TS incorporated in the license; or

b.

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

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

Proposed changes that meet the criteria of Specification 5.6.2.17.a or Specification 5.6.2.17.b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71.

5.6.2.18 CORE OPERATIN LI..MIS REPORT

+EC,)- Not Used

a. Core operating l-mits shall be established prior to eaIh reload cycle, or prior to any

.em

.~g portion of a reload cyle, ad shall be-documented im the COLR for the IL 2.1.1.1 API Protective Lmt, LCO 3.1.1 SH.UTDOWN MARGIN S-R 3.1.7.1 An/Rln Position Indieation Agreement LCO 3.1.3 Moderator Temperature C-eff"-e-'t (MTC)

LCO) 3.

2.13 Regulating Rod insertion Limits LCO 3.2.2 AXIAL POWER SHAPING ROD I

A1SR) Ilsertion f1its.

(continued)

Crystal River Unit 3 S.0-22 Amendment No. H1

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.18 COLR (continued)

LCO 3.2.3 AXIAL POWER IMBALANCE Operating Limitsr LCO 3.2.4 QUADRANT POWER TILT LCO) 3. 2. 5 rower reaking Factors LCO 3.

3.1:

Reactor Prroecio System (RrS) instrumentation SR 3.4.1.1 Reagctr Coolat System r e

Lits "n

A I

nfl r

4

r.

lflkIklrfI l

I4 LCO 3.9.1 Boron Concentratio-b*. The analytical~methods used to determine the core operating l i.s shall be those-pir.eviy reviewe*d and approved by the-NREt BAW 101:79P A, "Safety-Cri*Itteria and Methodology for Ae~ptable Cycleý R-elad Analy!ses (the approve reviionat he-time the reodaalyses are performed) adLicense Amendment 144, SER dated June 25, 1992.

The appred r ber for BAW 1017P A shall be identifid in the COLR.

wIUEII I I 1*

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c. The core operat ng limits shall be determined such that all applicable imt (ýegfe thermal mechanical limits, L~e thermlj hydraulic limits, Emrgency Core Cooling Systemt (ECCS) li+mit, ucea limits sc s5M rnin analysi lmtad accident analysis limits) of the safety an alys are mt.I

d.

The COLR, includin any midcycle revisions or supplements, shall bepo e

pon issuance for each eloa cyl t he 5.6.2.19 Reactor Coolant System (RCS) PRESSURE AND TEMPERA~RE+MT REPORT (PTIR)-Not Used 3.4.5 tabRf+&-F' 3.4.11 Low Temperature Overpressure rrotectio k-r-RCS pressure and m p ea tur l Si I InIcIuI n heatup and L ýldwn rates, cri ýtiaty an!rd hyrostatic. and leak test limits, shall be est-ablishe and documented, in -the PTLR.

The analytical methods used to determine-th rssure and

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C(IIoni t~i nued)I Crystal River Unit 3 5.0-23 Amendment No. 2-G4

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2. 19 Reactor C--l -n t"

PR,-SUR S

REPORT (PTLR)4/

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IsIhall deerind othat allapial limits(eghau lImits cooldown limits and isrcelaanhyottIc testing limits) of the anlss are met.

a.-The PTLR, includ',,

revisions or su-plements thereto, shall b-provided upon issuane for each reco essel flec perod 5.6.2.20 Containment Leakage Rate Testing rrogram Not Used A pogamshllbe established to 1._.p,,ement the leakage rate tetig fth cnainment as re8ie 1y FR5.4o)ad1 CFR..O,= Ap J, Op t11oMn W

, as" modifie by appo exemptionsI ThI IPro IFgramI shal Ibe inr, dean uI tl theI guIdl I I onta I-neI in ReguVlatory, Guide 1.163, INPe rfo rmac B-ased. Containment Leak TestProram datd Spteber1995, as modified by the follewiftg

1. NEI 94 01 195,Setion 9.2.3; The first Type A test performed afte the November 7,191.

Type A test shall be performed no later than November 6,20.

The Peak calculated containment inteernal pressure for the design basis]loss of Icoolat accident, P psig.

The containment des~ign pressure is 55 psig.

The maximum allowabl primary containment "-'le rat e*,

at-* P shall be 0.25%6 of primary containment air w-eight per day Leakage Rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is < 1.0 L_-

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n criteria are < 0.60 L f t

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,EO-r.75-Le 0 for Type A Tests.

2.

Air lock testing acceptance criteria are:

a0.

Overall air lock leakage n

i 0.05 L when teste at-> a-P.

b.

For each door, leakage rate is < 0.01: L. when tested at

> 8. 0 1-i 1 t The provisions of SR 3.0.2 do not ap to t test frequ!enies specified in the Cointlainment Leakge Rate Testin rrogram.

Thep provisions of SR 3.0.3 are applicable to the Containmenft LeaageRat Tetig Program.

(conti nued)

Crystal River Unit 3 5.0-23A Amendment No. 199

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.21 Control Complex Habitability Envelope Integrity Program A Control Complex Habitability Envelope Integrity Program shall be established and implemented to ensure that CCHE habitability is maintained such that. with an OPERABLE Control Room Emergency Ventilation System (CREVS),

CCHE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a challenge from smoke.

The program shall ensure that adeauate radiation protection is

'rovided to permit access and occupancy of the CCHE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.

The orogram shall include the followina elements

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

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

3. Reauirements for (i) determining the unfiltered air in-leakage past the CCHE boundary into the CCHE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Reaulatorv Guide 1.197, "Demonstrating Control Room Envelooe Integrity at Nuclear Power Reactors," Revision

0. May 2003, and (ii) assessing CCHE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.

4. The Control Complex Habitability Envelope Integrity Proaram will be used to verify the integrity of the Control Complex boundary.

Conditions that are identified to be adverse shall be trended and used as part of the 24 month assessment of the CCHE boundary.

5. The quantitative limits on unfiltered air in-leakage into the CCHE.

These limits shall be stated in a manner to allow direct comparison to the unfiltered air in-leakage measured by the testing described in paragraph 3.

The unfiltered air in-leakage limit for radiological challenges is the in-leakage flow rate assumed in the licensing basis analyses of DBA consequences.

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

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

Crystal River Unit 3 5.0-23B Amendment No. 230

Procedures, Programs and Manuals 5.6 TIIT C rI AGE T

ENlTrklTTr*kI L

FI Crystal River Unit 3 5.0-24 Amendment No. 2-3

Procedures, Programs and Manuals 5.6 T'I Sr -f PAGEr kiNIrTETf~IOAlL LEF BL V

ANK1 Crystal River Unit 3 5.0-25 Amendment No. M

Procedures, Programs and Manuals 5.6 T'IS PAGEfl rr TENTkITT/IALL LEF B.

LUANKi Crystal River Unit 3 5.0-26 Amendment No. -223

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

a.

Not Used

b.

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

The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring for the reporting period.

The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM).

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted in a supplementary report as soon as possible.

c.

Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted prior to May 1 of each year, and in accordance with 10 CFR 50.36a.

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program, and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV B.1.

(continued)

Crystal River Unit 3 S.0-27 Amendment No. 222

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

The following Special Reports shall be submitted:

a.

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

Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

u-.--Any abnormal degradation of the coptainment struture fou-during the imspection performed in accordanee with ITS 5.6.2.8 shall be reported to the NRC within 30 days of the current s.rvei1an. e mpl.et -on.

The ab..rmal degradati shall be defined as findimgs such as delamimation of the dome conrete, widespread corrosion of the *i*er

plate, eorreioon of prestr-essing elements (wires, strands, bars) or anchorage components extending to more than two tendons and group tendons force trends not meeting the requiremnents otr IOCFR5O.5a(b)E2)E'x),B).

The report shall incl'de the deseription of degradation, operability determination, root eause determimatiom and the corrective aetions.

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Eeontined-ý Crystal River Unit 3 5.0-28 Amendment No. 223

Reporting Requirements 5.7 5.7 Reporting Requirements 5.7.2 Spe-ial Reports

(.. ti..ed)"

5. Numiber of tubes plugged~durimg the inspection outage for each active degradation mechanism,

6. Total number amd per.entage of tubes plugged-to date,

7. The resuilts of ecndition moni toring, ihcludi ng the results of tube pulls and im si.tu testing.

Crystal River Unit 3 5.0-29 Amendment No. 2-34

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), alternative methods are used to control access to high radiation areas.

Each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation (measured at 30 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 group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

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

b.

A radiation monitoring device that continuously integrates the radiation dose in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c.

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

5.8.2 In addition to the requirements of Specification 5.8.1, areas with radiation levels > 1000 mrem/hr at 30 cm 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 Control Room-Shif-tSupervisor 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 performed within the area.

(continued)

Crystal River Unit 3 5.0-30 Amendment No. 2-GI

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

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

Crystal River Unit 3 5.0-31 Amendment No. 149

DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #313, REVISION 1 ATTACHMENT D PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES, REVISION BAR FORMAT

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.

Organization 5.2 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for unit operation and corporate management, respectively.

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

a.

Lines of authority, responsibility, and communications shall be established and defined from the highest management levels through intermediate levels to and including all operating organization positions.

These relationships shall be documented and updated, as appropriate, in the form of organization charts, functional descriptions of department responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation.

These shall be documented in the FSAR;

b.

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

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

c.

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

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

a.

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

b.

Shift crew composition may be less than the minimum requirement of 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days 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.

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.

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 minimum 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.

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.

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.

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

2.

For Iodine-131, Iodine-133, tritium, and for all radionuclides in particulate form with half-lives greater than 8 days:

Less than or equal to a dose rate of 1500 mrems/yr to any organ;

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 50, Appendix I;

i.

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

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 (conti nued)

Crystal River Unit 3 5.0-10 Amendment No.

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

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

a.

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

b.

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

c.

Demonstrate for each train of the CREVS that a laboratory test of a sample of the carbon adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, 1978, meets the laboratory testing criteria of ASTM D 3803-89 (Re-approved 1995) at a temperature of 300C and relative humidity of 95% with methyl iodide penetration of less than 5.0%.

d.

Demonstrate for each train of CREVS that the pressure drop across the combined roughing filters, HEPA filters and the carbon adsorbers is

< AP=4" water gauge when tested in accordance with Regulatory Guide 1.52, Revision 2, 1978, and ANSI N510-1975 at the system flowrate of between 37,800 and 47,850 cfm.

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

5.6.2.13 Not Used (continued)

Crystal River Unit 3 5.0-11 Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.14 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established.

The program shall include sampling and testing requirements, and acceptance criteria, in accordance with applicable ASTM Standards.

The purpose of the program is to establish the following:

a.

Acceptability of new fuel oil for use storage tanks by determining that the following properties within limits of No. 2-D fuel oil:

prior to addition to fuel oil has the ASTM D 975 for Grade

1.

Kinematic Viscosity,

2.

Water and Sediment,

3.

Flash Point,

4.

Specific Gravity API;

b.

Other properties of ASTM D 975 for Grade No.

2-D fuel oil are within limits within 92 days following sampling and addition of new fuel to storage tanks.

c.

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

5.6.2.15 Not Used (conti nued)

Crystal River Unit 3 5.0-12 Amendment No.

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

This program ensures loss of safety function is detected and appropriate actions taken.

Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists.

Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions.

This program implements the requirements of LCO 3.0.6.

The SFDP shall contain the following:

a.

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

b.

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

c.

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

d.

Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed.

For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a.

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

b.

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

c.

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

(continued)

Crystal River Unit 3 5.0-13 Amendment No.

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

The SFDP identifies where a loss of safety function exists. If a

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

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

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

a.

A change in the TS incorporated in the license; or

b.

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

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

Proposed changes that meet the criteria of Specification 5.6.2.17.a or Specification 5.6.2.17.b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71.

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

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.21 Control Complex Habitability Envelope Integrity Program A Control Complex Habitability Envelope Integrity Program shall be established and implemented to ensure that CCHE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS),

CCHE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a challenge from smoke.

The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CCHE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident.

The program shall include the following elements.

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

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

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

4. The Control Complex Habitability Envelope Integrity Program will be used to verify the integrity of the Control Complex boundary.

Conditions that are identified to be adverse shall be trended and used as part of the 24 month assessment of the CCHE boundary.

5. The quantitative limits on unfiltered air in-leakage into the CCHE.

These limits shall be stated in a manner to allow direct comparison to the unfiltered air in-leakage measured by the testing described in paragraph 3.

The unfiltered air in-leakage limit for radiological challenges is the in-leakage flow rate assumed in the licensing basis analyses of DBA consequences.

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

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

Crystal River Unit 3 5.0-15 Amendment No.

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

a.

Not Used

b.

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

The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring for the reporting period.

The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM).

The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979.

In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted in a supplementary report as soon as possible.

c.

Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted prior to May 1 of each year, and in accordance with 10 CFR 50.36a.

The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program, and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV B.1.

(continued)

Crystal River Unit 3 5.0-16 Amendment No.

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

The following Special Reports shall be submitted:

a.

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

Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

(conti nued)

Crystal River Unit 3 5.0-17 Amendment No.

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), alternative methods are used to control access to high radiation areas.

Each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation (measured at 30 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 group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

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

b.

A radiation monitoring device that continuously integrates the radiation dose in the area and alarms when a preset integrated dose is received.

Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c.

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

5.8.2 In addition to the requirements of Specification 5.8.1, areas with radiation levels > 1000 mrem/hr at 30 cm 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.

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

(continued)

Crystal River Unit 3 5.0-18 Amendment No.

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

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

Crystal River Unit 3 5.0-19 Amendment No.

DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #313, REVISION I ATTACHMENT E REGULATORY COMMITMENT

U. S. Nuclear Regulatory 3F0913-02 Attachment E Page 1 of 1 REGULATORY COMMITMENT The following table identifies the actions committed to by Duke Energy Florida, Inc. in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments. Please notify the Crystal River Unit 3 (CR-3)

Licensing Supervisor of any questions regarding this document or any associated regulatory commitments.

Regulatory Commitments Due Date/Event CR-3 will establish administrative controls to ensure that fuel November 30, 2013 handling activities and heavy load lifts above spent fuel stored in the spent fuel pool will not occur without a Certified Fuel Handler providing oversight of the activity.

v t e Letter Sequence Response to RAI
TAC:MF3089, Control Room Habitability (Approved, Closed)
Initiation Request, Request, Request, Request Acceptance... Supplement, Supplement, Supplement Administration Questions 10 April 2014 8 January 2015 Answers 4 September 2013 Results Approval Other:
MONTHYEAR2014-04-10 [Table View]

3F0913-02, License Amendment Request 313, Revision 1, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions and Response to Requests for Additional Information

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3F0913-02, License Amendment Request 313, Revision 1, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions and Response to Requests for Additional Information

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