ML13128A286

License Amendment Request #313, Revision 0, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions

ML13128A286
Person / Time
Site:
Issue date:	04/25/2013
From:	Elnitsky J Duke Energy Florida, Florida Power Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
3F0413-01
Download: ML13128A286 (81)

Text

PDuke Energy.

Crystal River Nuclear Plant Docket No. 50-302 Operating License No. DPR-72 Ref: 10 CFR 50.90 April 25, 2013 3F0413-01 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 -License Amendment Request #313, Revision 0, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions

References:

1. 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)

2. FPC to NRC letter dated April 15, 2013, "Crystal River Unit 3 - Request for Approval of the Certified Fuel Handler Training and Retraining Program"

Dear Sir:

Pursuant to 10 CFR 50.90, Florida Power Corporation (FPC) hereby provides this License Amendment Request (LAR) to revise portions of Section 5.0, Administrative Controls, of the Crystal River Unit 3 (CR-3) Improved Technical Specifications (ITS).

In Reference 1, 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.

Part of this request proposes changes to the staffimg and training requirements for the operating staff. Reference 2 was submitted proposing a Certified Fuel Handler Training and Retraining Program for NRC approval.

This request also proposes to eliminate the Explosive Gas and Storage Tank Radioactivity Monitoring Program. A commitment to release all gases in the Radioactive Waste Storage System gas decay tanks prior to implementation of the requested License Amendment is contained in Attachment D.

FPC 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.

Crystal River Nuclear Plant 15760 W. Power Line Street Crystal River, FL 34428

U. S Nuclear Regulatory Commission Page 2 of 2 3F0413-01 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 April 25, 2013.

n Elnitsky, Vice President ject Management and Construction JE/scp Attachments: A. Description of Proposed License Amendment Request, Background, Justification for the Request, and Regulatory Analysis B. Proposed Technical Specification Page Changes, Strikeout and Shadowed Text Format C. Proposed Technical Specification Page Changes, Revision Bar Format D. List of Regulatory Commitments xc: NRR Project Manager Regional Administrator, Region II Senior Resident Inspector

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #313, REVISION 0 ATTACHMENT A DESCRIPTION OF PROPOSED LICENSE AMENDMENT REQUEST, BACKGROUND, JUSTIFICATION FOR THE REQUEST, AND REGULATORY ANALYSIS

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 1 of 21 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, Florida Power Corporation (FPC) 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.

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 three and one half 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

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 2 of 21 Cooling," 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 inventory has occurred at a rate of about I 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.

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 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 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

U. S. Nuclear Regulatory Attachment A 3FO413-01 Page 3 of 21 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 demand on 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.e. 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.

This reference is removed since CR-3 will not return to operation in

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 4 of 21 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 upon a total loss of spent fuel pool cooling the temperature in the spent fuel pool will take approximately four days to reach 2000 F.

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 deleted since CR-3 is no longer authorized to have fuel in the reactor. Deletion of this paragraph recognizes the expanded response time available for radiation protection staff, in the event of a loss of cooling or inventory in the spent fuel pools, as compared to power operation considering the stored fuel exposure

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 5 of 21 history.

5.2.2.d (New) A new paragraph is added to establish the requirement for having oversight of fuel handling operations performed by a Certified Fuel Handler.

5.2.2.e (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 N 18.1, 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 established, 5.6.1.1 .a 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

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 6 of 21 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.

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 XI 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 XI 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 No Changes Program 5.6.2.10 Steam Generator The Steam Generator Program established and implemented (OTSG) Program practices to ensure that OTSG tube integrity was maintained.

This program is no longer required since CR-3 is permanently

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 7 of 21 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 Water This program provided controls for monitoring secondary 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. 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.

In Attachment D to this LAR, a Regulatory Commitment is made to vent and remove from service the Radioactive Waste System gas decay tanks by August 30, 2013. Based on that commitment, 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.

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 pressurization.

5.6.2.20 Containment This program was established to implement the leakage rate testing Leakage Rate Testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50,

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 8 of 21 Program 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 section is being eliminated. The Special Reports that currently exist are associated with plant conditions that cannot exist in a permanently defueled plant or for 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 > 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 0, 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. Florida Power Corporation (FPC) 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 Attachment A 3F0413-01 Page 0 of 21 SIGNIFICANT HAZARDS CONSIDERATION FOR PROPOSED CHANGES Does the proposed change Does the proposed Does the proposed involve a significant change create the change involve a increase in the possibility of a new or significant reduction in Identification and Description of Change 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 Plant General Manager to the Plant administrative and cannot cannot create a new or Manager. increase the probability or different kind of consequences of a fuel accident.

handling accident.

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

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.

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 1 of 21 operating plant. The proposed change establishes the Shift Supervisor as having command of the shift. 4 4 -4 5.2.1 .a The introduction to this section No. This change in the No. This change in the No. This change does identifies that organizational positions are description of the description of the not directly involve any established that are responsible for the safety of functional responsibility of functional responsibility physical limits or the nuclear plant. organizational positions of organizational parameters and therefore places emphasis on the safe positions cannot create a cannot affect any margin This is changed to require that positions be storage and handling of new or different kind of of safety.

established that are responsible for the safe nuclear fuel. This focus on accident since they do storage and handling of nuclear fuel. This their principal not change the function change removes the implication that CR-3 can responsibility cannot of any plant structures, return to operation. increase the probability or systems, or components.

consequences of a fuel handling accident.

5.1.2.b This section identifies the No. This change in the No. This change in the No. This change does organizational position responsible for overall description of the description of the not directly involve any nuclear plant safety, for the safe operation of the functional responsibility of functional responsibility physical limits or plant, and for control of activities necessary for organizational positions of organizational parameters and therefore the safe operation and maintenance of the plant. places emphasis on the safe positions cannot create a cannot affect any margin handling and storage of new or different kind of of safety.

This section is being changed to recognize that nuclear fuel. This focus on accident since they do the safety concerns for a permanently defueled their principal not change the function plant are for the safe storage and handling of responsibility cannot of any plant structures, nuclear fuel. It changes responsibility for increase the probability or systems, or components.

overall safety for storage and handling of consequences of a fuel nuclear fuel to the Decommissioning Director. handling accident.

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

5.2.1 .c This paragraph addresses the No. This change continues No. This change does No. This change does requirement for organizational independence of to ensure that personnel in j not introduce any Jnot directly involve any

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 2 of 21 the operations, health physics and quality specifically identified changes to the function limits or parameters and assurance personnel from operating pressures. 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 independence from operating pressures" to probability or occurrence accident.

"their ability to perform their assigned of 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.e, 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 conditions No. This change continues No. This change does No. This change does under which the minimum shift compliment to ensure that the minimum not introduce any not directly involve any may be reduced. It contains a reference to 10 shift compliment of changes to the function limits or parameters and CFR 50.54(m) which establishes the minimum qualified personnel will not of any plant structures, therefore cannot affect requirements for a licensed operating staff for be decreased for more than systems, or components any margin of safety.

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

return to operation in the future, and the who are capable of requirement for licensed operating personnel responding to operating will no longer be required to protect public plant transients and health and safety. accidents. This does not involve an increase in the

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 3 of 21 probability or consequences of a fuel handling accident.

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

1-4. control room. This cannot therefore it cannot create increase the probability or a new or different kind of The change establishes the requirements for consequences of a fuel accident.

either a Non-certified operator or Certified Fuel handling accident.

Handler to be in the control room when fuel is stored in the pools. -I. I-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 deleted since CR-3 is no consequences of a fuel therefore it cannot create longer authorized to have fuel in the reactor. handling accident are a new or different kind of governed by the accident.

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

5.2.2.d (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 a trained and qualified to changes to the function limits or parameters and Certified Fuel Handler. safely handle irradiated of any plant structures, therefore cannot affect

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 4 of 21 fuel. Applying these systems, or components any margin of safety.

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

consequences of a fuel handling accident are not increased.

5.2.2.e (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 safely handle irradiated of any plant structures, therefore cannot affect In the permanently defueled plant, the Certified fuel. Applying these systems, or components any margin of safety.

Fuel Handler is the senior position on the qualifications to the therefore it cannot create operating crew. It is not necessary for the Shift supervision of fuel a new or different kind of Supervisor to hold a Senior Reactor Operator movement ensures that the accident.

license if the plant cannot operate to generate probability or power. 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 requirements cause and advise on the therefore it cannot create any margin of safety.

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

N 18.1, 1971, and the Radiation Protection The absence of a staff Manager meet or exceed the qualifications of member with those Regulatory Guide 1.8, September 1975. qualifications does not change the probability or consequences of a fuel handling accident.

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 5 of 21 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 consequences of a fuel handling accident. t +/-

+

5.6.1 .1.a This section states the requirement for No. The procedures No. The applicable No. This change does procedures to be established, implemented and necessary for the safe procedures for the safe not directly involve any maintained covering various plant activities. handling of nuclear fuel are storage of nuclear fuel limits or parameters and included in the group of will direct the correct use therefore cannot affect The scope is reduced to procedures applicable procedures applicable to of fuel handling any margin of safety.

to the safe handling and storage of nuclear fuel. the safe storage of nuclear equipment. These fuel. With these procedures are currently procedures in effect for in place and have been fuel handling, the used effectively for the probability or safe handling of fuel.

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

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 6 of 21 5.6.2.3 In this section, the authority for No. This is a change to the No. The change No. The changes approval of changes to the Offsite Dose requirements for the proposed here, proposed here for ODCM Calculation Manual (ODCM) is changed from position responsible for identifying a different approval do not directly the Plant General Manager to the Plant Manager approving ODCM changes. position responsible for involve any limits or consistent with the position title change in 5.1.1. In a permanently defueled ODCM change approval, parameters for operating plant, the fuel handling cannot create a new or systems and therefore accident is the only different kind of accident cannot affect any margin credible accident since this does not of safety.

previously evaluated. This change the function of action cannot increase the any plant structures, probability or systems, or components.

consequences of a fuel handling accident.

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 This program was established to minimize permanently defueled of any plant structures, therefore cannot affect leakage from portions of systems outside plant. This change systems, or components any margin of safety.

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

consequences of operating transients and accidents.

The program is being eliminated.

This change cannot 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.

handling accident.

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 7 of 21 design limits.

This orogram is being eliminated. I. +

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 not apply to nuclear fuel or changes to the function Inspection Program This program required periodic inspections, fuel handling equipment. of any plant structures, provided confidence that examinations, and tests of plant pressure Therefore eliminating this systems, or components plant systems that were boundary components to ensure their continued 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 any margin of safety.

+/-

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 In this state, the any margin of safety.

within the spent fuel pools. condition of the steam Therefore, eliminating the generator tubes is This program is being eliminated.

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

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 8 of 21 This program provided controls for monitoring systems do not not introduce any program was intended to secondary water chemistry to inhibit steam interconnect with the fuel changes to the function protect will no longer generator tube degradation and low pressure cooling or fuel handling of any plant structures, function for power turbine disc stress corrosion cracking. systems. Therefore, systems, or components production. Therefore, eliminating the Secondary therefore it cannot create eliminating this program This program is being eliminated. Water Chemistry Program a new or different kind of cannot affect any margin cannot increase the accident. of safety.

probability or occurrence of 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 are therefore cannot affect explosive gas mixtures contained in the created and must be 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 fuel this control program this control program storage after fuel has been cannot create a new or cannot affect any margin moved into the spent fuel different kind of of safety.

This program is being eliminated, pools. Therefore, accident.

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 9 of 21 eliminating this program cannot increase the probability or occurrence of a fuel handling accident. 4 4

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, temperature limits, including heatup and program cannot increase reactor contains no fuel. eliminating this program cooldown rates, criticality, and hydrostatic and the probability or Therefore, eliminating cannot affect any margin leak test limits, be established and documented occurrence of a fuel this control program of safety.

in the PTLR. handling accident. cannot create a new or different kind of This program is being eliminated. accident.

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 therefore it cannot create This program is being eliminated in accordance of 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 No. Eliminating requirements for programs reporting requirements reporting requirements This section is being eliminated, that are no longer required that are no longer that are no longer or conditions that cannot required cannot create a required cannot affect exist in a permanently new or different kind of any margin of safety.

defueled plant cannot accident.

increase the probability or occurrence of a fuel handling accident.

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 10 of 21 5.8.2 High Radiation Area Controls No. This is a change to the No. The change No. The changes 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 A 3F0413-01 Page 0 of 21 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.

Florida Power Corporation (FPC) 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 Re2ulatory 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.

U. S. Nuclear Regulatory Attachment A 3F0413-01 Page 1 of 21 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 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(g)(1) requires that a plant whose construction permit was issued before January 1, 1971 must meet the requirements of paragraph (g)(4). The construction permit for CR-3 was issued September 25, 1968. 50.55a(g)(4) requires implementation of ASME Section XI, "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 Improved Technical Specification (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 (MLO 10430066)

2. NUREG-- 1275, Volume 12, "Operating Experience Feedback Report - Assessment of Spent Fuel Cooling," February 1997 (ML010670175)

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #313, REVISION 0 ATTACHMENT B PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES, STRIKEOUT AND SHADOWED TEXT FORMAT

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

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

5.1.2 The Contrel Room Supervisor shall be responsible for the eentrel room eommand function. Durimg any absence of the Control Room Supervisor fr-om the eontrol room while the unit is im MOD[ 1, 2k, 3, or 4, am individual with an aetive Semier Reactotr Operator (SRO) lieense shall be desigmated to assume the comtrol room command function. During any absence of the Comtrol Roo~m.

Supervisor from the control room while the umit is im MODE 5 or 6, an . dividual with an active SRO Theense or Reactor 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. 201

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 safety of the.u.lear power pla-t. the safe handling and storage and-handling 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 Plant Decommissionini Director shall have corporate responsibility for overall responsibility for the safe handling and storage 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 fuelsafety. The Vice President -Crystal River Nuclear Plant Manager shall be responsible for the overall safe operation of the plant and shall have to control over those onsite activities necessary for the safe handling and storage of nuclear fueloperation and maintenance of the plant; 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 independem.e from operat.. g 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 assig.ed to the operatimg shift any time there is fuel imn the reator and (continued)

Crystal River Unit 3 5.0-2 Amendment No. 201

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued) an additional auxiliary nuclear 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 CFR 50.-4,m)-2)-i) and 5.2.2.a for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

e. At least one licensed Reactor Operator ERG) shall be present i" the control room whe, fuel is in the reaeter. i.

addition, while the unit is in MOD[ 1, 2, 3, or 4, at leasyt em.e licensed Senior Reactor Operator (SRO) shall be presen iI the eontrIl roomIftI.

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. Amn idividal qualified im Radiation ProteItion procedures shall be on site when fuel is in the reactor. The position may be vacant for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, in order to providE for unexpected absence, provided immediate a-tion is tIken to fill the required position.

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

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

Crystal River Unit 3 5.0-3 Amendment No. 233

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.

and the Shift Technical Adviser who shall have a Bahe lr's" degree, or the equivalent, in a scientific or engineering diseipline wi.th specffic training in plant design and response and analysis of the plant trasiets and accident.

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. 49

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. 201 1

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 Prim.ary Coolant Soures Outside Containment Not Used This program provides controls to minimize leakage from these portions of systems outside containment that culd contain high4-y radioative fluids during a seus transient or accident tO levels as low as practi-able. The systems include Low Pressure injection, Reactor Building Spray and Makeup and Purification.

The program shall include the fell-wing:

a. Preventive maintena1ee and peridic visual inspection requremets, and

b. Integrated leak test requirements for each system at refueling cycle intervals or less5.

5.6.2.5 Component Cyclic or Transient Limit Not Used This program provides controls to track the FAR Table 4.8, cyclic 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. 2-1-3

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.7 Not Used 5.6.2.8 Inservice Inspection Program Not Used This program provides eentrels for ner imeispection of ASME Code Class 1, 2, 3, ME amd CC compnenets, includimg applicable supports. The program shalil include the followimg.

a. rrov*io* that imservice inspection of ASME Code Class 1, 2, 3, ME amd EE compoenets shall be performed in accordacee with Section XI of the ASME Boiler and Pressure Vessel Cd and applicable Addenda as required by 10 CFR 50.55a,

b. The provisions of SR 3.0.2 are applicable to the frequencie for perforin nsrvice inspection activities;

e. nerieinspeetion of each reaete, eeeant pump flywheel shall be performed at least once every twentyyears. Te

,ser 11 imeispection shall be either an ultrasonic examination of the volume from the inner bore of the flywheel to the circle of one half the outer radius or -a.

surface examination for exposed surfaces of the disassembled flywheels. The recommendations delineated in Regulatory Guide 1.14, Positions 3, 4, and 5 of Section C.4.b shall

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

(continued)

Crystal River Unit 3 5.0-11 Amendment No. 218

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals (continued) 5.6.2.9 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components, including applicable supports. The program shall include the following:

a. Provisions that inservice testing of ASME Code Class 1, 2, and 3 pumps, valves, and snubbers shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a;

b. Testing frequencies 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 accelerated Frequencies specified as two years or less in the Inservice Testing Program for performing inservice testing activities;

d. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

e. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

(continued)

Crystal River Unit 3 5.0-12 Amendment No. 232

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 Steam Ge.erat. r E.. TSG) Program Not Used A .teamGenerator Program shall be established and implemented to emsure that OTSG tube integrity is maintained. in addition, the.

Steam Generator Program shall include the fellow~ing proisio-:-

a. Provisions for condition monitoring assessments. Eomditi 1 monitoring assessment means an evaluation of the "as found" eemditien of the tubing with respect to the performacee criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an T-"G inspetiomn outage, as determined from thee inservie inspection results or by other means, prior to the plugging of tubes.ioiditio. .V V itein assessments shall be-eondueted during each outage during which the OTSG tubers we inspeted or.plugged to confirm that the performan.e are being met.

.riteria

b. ,erformance criteria for OTG tube integrity. OTcG tub-e integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion: All in sericesteam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retainin a safety factor of 3.0 against burst under noralsed state full power operation rmr to secondary pressure differential and a safety fF 1.4 agaist burst applied to the design basis accident primary to secondary pressure differentials. Apart from the above reqi ements, additional loading conditions associ ated itthe design basis accidents, or combination of accidents in accordance with the design and licensinrg 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 withth loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary leas (conti nued)

Crystal River Unit 3 5.0-13 Amendment No. 234

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 UIJI Program I continued)

I*

e. --- II-- jeaKde I -. /-- I r I --

ler1ormane I-__erion:, lile = P primary to sec a aci.IeIt iIIucea Iea*,,e rate for GLilyY I - 11I UIC 01iGL IJc  %% I %A II I U I-III  %-Il CII l I %J U I JJ I-UW%

rupture, shall not exceed the leakatge rate assumed in the accident analysis in terms of total leakage rate for all OTh5Gs and leakage rate for am imdividual O):FS.

Leakage is mot to exceed one gallon Ie mrte per OEFG J. The operational LLAKAGL pertormance criterion is c.n.-a 4f -i A 4 Ir-A :1 A :12 "Ol~fr- riti. 4Ani Ir-AI(Arr I

c. Provisier5 for OISG tube repair eriteria. A tube feund by III ~VIL ~ "pI "

tI3JLL~I 1 j L V"IEILII I IVV3 I YIW I L

0. I II I

.g I I-.

or eAe~aII1*g 4/ UI tIe nomIIaI TUUe Wa; I tni- eS SInaIi b1e

d. ,rovisions for 0TSG tube inspections. ,eriodic O TSG tube inspections shall be performed. The number and portionso the tubes inspected and methods of inspection shall be performed with the objective of detectimg flaws of any type (e.g., volumetric flaws, axial and circumferential cracks)ý that may be present along the lemgth of the tube, fromth

• tube to tubesheet II it. it-U ~iJ.weld at the tube inlet to the tube tor 1i3 1 T *Ll~i tubesheet weld at the tube outlet, and that may satisfy th-e applicable tube repair criteria. The tube to tubesheet weld is not pa~rt of the tube. -In addition to meeting the requremnts of d.1. d.2. an-d d.-3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that O)TSG tube integrity is maintained WA*41 t -i- nT~cr- c4no t--0% An. -

'"n~n-

- A - - - __ __ - - - - - - -I

%A%: C IV 1 Ul 11E I I UV- 1J I I UV IIIU L-U U-LVI IIIIIIIC Li C Lylie UtIIU kcaiOvn oF flaws to which the tubes may be susceptible and, based on this assessment, to determine whieh inspectio~n methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each BTSG during the first refueling outage following OTSG replacement-.

2. Inspect 100% of the tubes at sequential periods of 14 Ilull L Ia A IPL I IE11OI AA f-11 PIIa I I

~%M#kc. T-k- fiwa* Qý "ý"*i pq - w conidere- d to bIn a f ter I I I*e fi* rs rice t e firtntinuedei iset.iom of the~ gin

.'ere toater b utib. +/-fl addition, insmect.t tu elf (conti nued)

Crystal River Unit 3 5.0-14 Amendment No. 234

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.10 of the9 period and the remainima 50% by the refuelin-three refueling outages (whiehever is less) without being inspected.-

3. if craIk indications are fluld i any OTIIG tube, theI the next inspection for each )T-SG for the degradation mechanism that caused the eraek imdication shall mot exceed 24 effective full power momths or one refueli-ng outage Ewhiehever is less). if def-initive imformatiei-,,

sueh as from examination of a pulled tube, diagnostice mom destructive testing, or engineering evatluartion indicates that a crack like indication is not asocated with a crack~s), then the indication need not be treated as a eraek.

e. Provisions for monitoring operational primary to secondar-y (continued)

Crystal River Unit 3 5.0-15 Amendment No. 234

Procedures, Programs and Manuals 5.6 TIITS RAGE iNTENTIONALLY LEFT BLANK Crystal River Unit 3 5.0-16 Amendment No. 234

Procedures, Programs and Manuals 5.6 Til-rS PAG[ INTTENTIONALLY LEFT BLANK Crystal River Unit 3 5.0-17 Amendment No. 234

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.11 Seendary Water Ehemistry Program Not Used This program provides controls for monitoring secondary water l Ic - I Ilt I

InIb I steIM generato- tlube.-IdelalatII* 1I dI1IloW pressud tubn isc stres coroo Hrcing.

Teprogram sa J-L nlud*e I

a. identification of a samIlplingj schedule for the critical variables aid controLl po I s for these variables;- 1

b. identiflcation of theJl procedures used to-meaurt h vl vl of the cri-tical variables; ue t eaue

e. Identification of proess sampling2points, which shall include mo.itoring th deeondensate pumps for evidence of= condense in le 1age

d. Procedures for the recording and management of data;

e. rrocedures defi.ning corrective actions for all off cointrol SII- c heist*ry I on I L itins;I Iand

f. Aprocedtire identi fyq t- ~hiyu-o,'~ fo, the InterpretatIon of thIe t a theslll quenIcl timing of administrO~a-tive eventsY which is, required to initia-te eorective action.

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 anoin ace test of the carbon adsorber shows a system bypass < 0.0 % 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 300 C 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. 1-99

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 explosive gas mixtures contained in the Radioactive Waste Disposal WD) Syst.em, the quantity of radioactivity contained in gas storage tanks or fed into the offgas treatment system. The gaseous radioactivity quantities shall be determined following the methodology in Branch Technical Position (BTr) [TSD 11 5, "rostulated 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 Radioactive--

Release due to Tank Failures".

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Radioactive Waste Disposal (WD) System and a surveillance program to ensure the limits are maintained. Such limitsý shall be appropriate to the system's design criteria, (i.e.,

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

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

The provision f SR 3.0.2 and SR 3.0.3 are applicable to t Explosive Gas and Storage Tank Radioactivity Monitoring Progr sureilancefequnis (continued)

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 COR OPERATING LIMITS , ,CL"R Not Used

a. Core operating limits shall be established prior to each reload cyle, or prior to any rema. . g portion of a reload cycle, and shall be do.umented i the CO*R for the SL 2.1.1.1 "rote.tive API L"m'-t LCO 3.1.1 SHIUTDOWN MARGIN SR 3.1.7.1 A,'/RP" Position ind,.at.on Agreement LC 3 1.3 Moderator Temnperature Coeffl- Aient MTC).

LC0 3.2.1 Regulating Rod insertion Limt LEO 3.2.2 AXIAL POW[R SHArING ROD (APSR) insertion Limit (continued)

Crystal River Unit 3 5.0-22 Amendment No. 2-H

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

--

LCEO 3.2.3 AXIAL POWER IMBALANCE Operating Limit LCO I.2.A QUADRANT POWER TILT LC 325 Power reaking factors r*E;I LCO 3.3.1 ,r,,Reacter, Cl%* , , r*

~.....~

J/k I

• ~ oetinSse UiLl

"... ~ :'" 2~

J~'lri-k1. -

~ (RP5) G.

M Instrumentatio I I lnJ A..

I III .

• R 3.4.1.1 Reatotr C-,-l-- SYstem Pressure BNB Limits 5R 3.4 .1 .3 ....... .. .. '-t*

. .. . . .. . . .. . . . -

LC0 3.9.1 B.r....e .

tration ..

b. The analyZ!tical methods used to determine the core operatin limts hal be those peiuly reviewed amd approv.ed buy BAW 10179P A, "aeyCriteria and Methodologyfo Aee~ptable Cycle Relead AnaFFls (h a revison at the me the elad nalyss ae performed)-

and- approved The Lee mdemet 144, SER dated -- 25, 192 rev*s on number forI A 11. 91 A . shall be identifed in te COLER.

• -.I I I III. Aý e.- -- u- e,-A1,t h "r . 1+_., 1. . JL ,P 1

e. T c operatigiimits shall be determined sueh that all applicable9 liis(~. ultermal mechanmical limits, corethemalhydauli liits Emrgency Core Cooling 5ystem

([CCS) li-its, nula lmt uc 5M, transien analyis lmits, and acc ident analysis lim~its) of- the safety analysis are met.

d. The COLR, incjludingay midcycle revision or suppl~emensL shall be provided upon issuamee for eacreloa cyle toh 4 _ 3..=:

_ - -Z.. e.F I -"

5.6.2.19 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT Apeni(PTLR) G,-+*:"" Not Lne- Used- 1986. The anltclmtonsdt ZPL.1....." I be I I1  :) Il?!e IVoS0I IFIII Iu .e1.1 JNRCl M  !? I*LI I I the and docaumetd ,l nL flAW. ,241,Ma1,, 97.a The analytical%.3 3.4.11 Low Temperature Overpressure rrotection b-.---CSp!!ssure and temperature limits, incluin etup and

-rl rates.., c.riticlt.y, and* .i hyd ro ad l test liit, a*.----Oth.?%har halbe estabise ITS an-oueted belcal in the PTLR.

Theanaytial methods used to determine the pr-essure and meth mehe used tro deI use ITP to deemm - LTmilimt A... shal I .. be thos J+,-It JLLL~i I ~ ,;II*; -ýJ.*I '51 IVZ` .e

I rd I me % i.w;*U "I 5.2AW*.---RCS~.4 prsur1, Reactor ndt Realem 2 erature livitse including SurvEiAnc" aatp E Program Prgrm.

N o!e dmctndiun u 1 K. .- -.E; .- E--L 3

-IR E P OR

-+-L *J- I . 1 - 1(conItinued)I;I

-+ P R- -- J +J . .. * . 3, .+

3% 1* E;__

• i n *.I;

+I_ l-llll *

• Crystal River Unit 3 5.0-23 Amendment No. 204

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals

-" - - -. -.--- ,nrr' - r - I il A LIM -&.A -r -1 IMl r I -A -Tr 5.6.2.19 - - - -

R[OT .1mTLRInk ~ ine e.-The reacter ves~sel -- res-ure and temperature iMmit I'.AOL

.. I..#. .I JI*'I A .. lI J J.. L L.

F. 1Id 1 Ie ...I II U,..,,1.

1 / .I I CLI .-. * , 4I\L&I . J I Mlii I I1 I,.. J,I IUIM I I

• aI.iL4 I 14IM ,.I~V J I I IaI WLl I , II L Ii I UL I testing limits) of t~he analysis are metý._

d %T.4 d--The . rTLR,

.P LR-, . .. in .. lu in .. revisions including ...p-Th ..n. ,r or _Q a^r- .. ....

su-plements Ithereto,

F1 .. . .. -hall . P_be F, "Fw.,I ,' *, ,-'I 5.6.2.20 Containment Leakage Rate Testing Program Not Used A prga hHbe established to impement tý!.!Takqe -rate

ý,.s'i*,,., e e mt a,,m11em a, r e* q u.* I. I Iy 10 C4Ee) and 10 CFR50 crrl Apenix J, Option B, as A '--

.... .. J . . oIfie

-I- 4* I bI aprved exemtos

-.. . -1m-t&4-

. .. mmeJ. . . Leak*---

This prora halbei aordac 1 ith the gudifidelintes contawied T t dIIII, daL;LICU .. p* iiiLIJi £!J- , dSI.* ISIIjp 1i 1-I*4 LI II '.e, IiVIIIy 4 *irt o n---- . . - i - _ _ _"_ _ - - - -- L _ J _ ._ _ -r. . .

• __

i~I~~L ~tJ.Li~I~J ~ .)~i..L MLIII ZJ.L. J. I~~I iL iy~~ I L~L l IILl i~I mc lI LICI LIIC IIVVI:IIIIUJ l Ip 1/ LCIL -11(1a I /I performed no later than November 6, 20066.

Thep peak calculated containment interalpressure for the design baislosof coolant acdnai U(.,*

1J* .I . %1  %.*I

• 1I- ( , ,I -,, I 1* , , I.*

!C 1 42psig. The

- [ J,.  % 1 containment design pressure is 5 sg The maximum allowable primary containment le.aage ratE, L,---t--P shall be 0.25% of prIIIary cotinment air eigh per day-.

Leakage Rate acceptance criteria are:

1. Containment alealkaIge rate acceptance.cri

.

..... . .... 4--.V terion is..< 1.05!! , -

~ :F4~ rst unt q~' iqR~ F t d:LLLUI

• I .C.

! "

V I LII LII I

• 1 IlJI b  :] II l LII ,

I~ &*IOI*JE (LLE a.L LJ G am ILE ceriteria are 0 L fI _ Ty e nt

-. 7 L. for Type A Tests.

2. Air lock testing acceptance criteria are-:-

a. Overall air lock leakage rag s<0.05 L. when teste

b. For eQacIh* door, leakage rate is < 0.01 L. whem tested at The provisions of SR 3.0.2 do not a to the test freqencis specified in the Containment Leakage Rate Testng rogram.

T provions of SR 3.0.3 are applicable to the Containment Leakagel IRate esI*ng Program.

(continued)

Crystal River Unit 3 5.0-2 3A 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 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 bv oaraaranh 3.

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

Procedures, Programs and Manuals 5.6 THIS PAGE INTENTIONALLY LEFT BLANK Crystal River Unit 3 5.0-24 Amendment No. 223

Procedures, Programs and Manuals 5.6 THIS PAGE INTENTIONALLY LEFT BLANK Crystal River Unit 3 5.0-25 Amendment No. 223

Procedures, Programs and Manuals 5.6 TIIT, PAGF TNTENTTONAtl Y LEFT RI ANK 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 5.0-27 Amendment No. 222

Reporting Requirements 5.7 5.7 Reporting Requirements 5.7.1.2 Not Used 5.7.2 Special ReportsNot Used 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 submi-tted 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.

b.-Any abnormal degradation of -emtainme.t the stre"t.re fou-d during the inspeetion performed in accordance with IT5 5.6.2.8 shall be reported to the NRC within 30 days of th eurremt surveillamee e, 1 1 letiom.. The abmermal degradati~.

shall be defined as f-indimgs sueH ats delamiinatiem of the dome concrete, widespread corrosiom of the liner plate,

• IIIA.-~~rosonof nI- rA r,~ ir l prestressing l.;',% 1 /ll,'

]L.,1

- iii% elements *rnI-4 ,i (wires,

-L.* I LI strands, 1,*: ll4*l. -*/I A*.) bars)

- .ll; or anchorage components extending to more than two tendonsan group tendons force trends not meeting the requirementso L1'..iI'J'J J.Q.)J ~J , A)ISI __I I _VJ1JI C& IILUl "1U%. I

__ - 0 V vj" ký -d ý " W" , v4F-- GIJ I -L -1 111 "II CAl W"JI, 0 JJ L.J P A PRAP#-r QIq-11 140 Qc~ri **.-Atj Wli th-ir nI AAr, 04 =A- P- .jk imiiafmr it MODE 4-- fellwi\ eeoeto of am--

imneto ve------. im.... wihth Leodae neilet 5...10 Sta Gee-te -- :SG Pro.rm Th VI I avL a i;I cIL j I ýJ j J

%I1 - - I V I ý I IIWA a..VI. I II%A I %_C L I%JI I.; .

(continued)

Crystal River Unit 3 5.0-28 Amendment No. 223

Reporting Requirements 5.7 5.7 Reporting Requirements 5.7.2 Special Reports (-cntinued)

5. Number of tubes plugged-during the inspection outage for eaTh active degradation me-hanism,

6. Total number and percentage of tubes plugged-to date,

7. The results of cndition monitoring, including the results of tube pulls and in situ 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 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 5.0-30 Amendment No. 201

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

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #313, REVISION 0 ATTACHMENT C 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 unit operation 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 nuclear safety.

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.

(continued)

Crystal River Unit 3 5.0-2 Amendment No.

Organization 5.2 5.2 Organization 5.2.2 Unit Staff (continued)

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

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

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

e. 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. 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 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 (continued)

Crystal River Unit 3 5.0-10 Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.9 Inservice Testing Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components, including applicable supports. The program shall include the following:

a. Provisions that inservice testing of ASME Code Class 1, 2, and 3 pumps, valves, and snubbers shall be performed in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as required by 10 CFR 50.55a;

b. Testing frequencies 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 accelerated Frequencies specified as two years or less in the Inservice Testing Program for performing inservice testing activities;

d. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

e. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

5.6.2.10 Not Used 5.6.2.11 Not Used (continued)

Crystal River Unit 3 5.0-11 Amendment No.

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

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

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

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

c. Demonstrate for each train of the CREVS that a laboratory test of a sample of the carbon adsorber, when obtained as described in Regulatory Guide 1.52, Revision 2, 1978, meets the laboratory testing criteria of ASTM D 3803-89 (Re-approved 1995) at a temperature of 30 0 C 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 5.0-12 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 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 5.0-13 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 5.0-14 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 (continued)

Crystal River Unit 3 5.0-15 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 .0-16 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 5.0-17 Amendment No.

Reporting Requirements 5.7 5.6 Reporting Requirements 5.7.1.2 Not Used 5.7.2 Not Used Crystal River Unit 3 5 .0-18 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 Ž 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 5.0-19 Amendment No.

High Radiation Area 5.8 5.8 High Radiation Area 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 5.0-20 Amendment No.

FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #313, REVISION 0 ATTACHMENT D LIST OF REGULATORY COMMITMENTS

U. S. Nuclear Regulatory Attachment D 3F0413-01 Page 1 of I LIST OF REGULATORY COMMITMENTS The following table identifies those actions committed to by Florida Power Corporation (FPC) in this document. Other statements in this correspondence 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 Commitment Due date/event CR-3 will vent and remove firom service the August 30, 2013 Radioactive Waste System gas decay tanks.