Text

License Amendment Request 323, Revision 0, Permanently Defueled Technical Specifications for the Independent Spent Fuel Storage Installation to Reflect Permanent Removal of Spent Fuel from the Spent Fuel Pools
	ML16243A249
Person / Time
Site:	Crystal River
Issue date:	08/31/2016
From:	Reising R; Duke Energy Florida
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	3F0816-01
	Download: ML16243A249 (113)
	v • d • e

Crystal River Nuclear Plant 15760 W. Power Line Street Crystal River, FL 34428 Docket 72-1035 Docket 50-302 Operating License No. DPR-72 10 CFR 50.90 August 31, 2016 3F0816-01 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - License Amendment Request #323, Revision 0, Permanently Defueled Technical Specifications for the Independent Spent Fuel Storage Installation to Reflect Permanent Removal of Spent Fuel from the Spent Fuel Pools

References:

1. NRC to CR-3 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. CR-3 to NRC Letter dated November 16, 2015, Notification of Schedule Change for the Post-Shutdown Decommissioning Activities Report (ADAMS Accession No. ML15322A117)

Dear Sir:

Pursuant to 10 CFR 50.90, Duke Energy Florida, LLC, previously known as Duke Energy Florida, Inc. (DEF), hereby requests approval of License Amendment Request (LAR) #323, Revision 0, where the CR-3 Facility Operating License and the Permanently Defueled Technical Specifications (PDTS) will be revised to reflect removal of all Crystal River Unit 3 (CR-3) spent nuclear fuel from the spent fuel pools and its transfer to dry cask storage within the Independent Spent Fuel Storage Installation (ISFSI) located within the Crystal River Unit 3 Nuclear Plant (CR-3) Protected Area.

By letter dated March 13, 2013 (Reference 1), the NRC acknowledged the CR-3 certification of permanent removal of fuel from the reactor vessel pursuant to 10 CFR 50.82 (a)(1)(ii).

Therefore, the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. A supplement to the Post Shutdown Decommissioning Activities Report for CR-3 dated November 16, 2015 (Reference 2) documented the DEF expectation that all spent fuel be completely transferred to the ISFSI by the end of February 2018. In support of this condition, the CR-3 license and associated Technical Specifications (TS) are being proposed for revision to comport to facility configuration with all spent nuclear fuel in dry storage within the ISFSI.

AREVA submitted an Application for Amendment 14, to the standardized NUHOMS Certificate of Compliance (CoC) No. 1004 for Spent Fuel Storage Casks, Revision 0, on April 16, 2015.

AREVA submitted a revision to the above referenced Application for Amendment 14 of this CoC on November 11, 2015. This revision requests the deletion of the License Condition to require a return to the spent fuel pool for inspection. AREVA has requested approval of the CoC amendment by October 2016. Once the CoC is revised, there will no longer be a requirement to return spent fuel to the spent fuel pools.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 2 of 2 to this letter contains a description, technical analysis, significant hazards determination, and environmental considerations evaluation for the proposed amendment. contains marked-up Facility Operating License (FOL) and POTS pages. All POTS Bases pages are being deleted in their entirety. provides the FOL and POTS pages that will remain after approval of this LAR in a revision bar format.

As discussed in this submittal, the remaining design basis accidents and transients analyzed in Chapter 14 of the CR-3 Final Safety Analysis Report (FSAR) are no longer applicable for the condition where all spent nuclear fuel is transferred to dry cask storage within an ISFSI.

The CR-3 Plant Nuclear Safety Committee has reviewed the proposed amendment and recommended it for submittal, and a copy of this submittal has been provided to the State of Florida in accordance with 1 O CFR 50.91 (b).

There are no new regulatory commitments made within this submittal.

DEF requests approval of this POTS LAR by August 31, 2017. Once approved, the amendment will be implemented within sixty (60) days following DEF's submittal of written notification to the NRC that all spent nuclear fuel assemblies have been transferred out of the spent fuel pools and placed in dry storage within the ISFSI.

This license amendment will not be implemented if the revision to the AREVA CoC, Amendment 14, is not approved.

If you have any questions regarding this submittal, please contact Mr. Phil Rose, Lead Engineer, Nuclear Regulatory Affairs, at (352) 563-4883.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August 30, 2016.

Sincerely, (2_

Ro~e::niob President Operations Support ARR/par

Enclosures:

1. Discussion of

Change, Technical

Analysis, Significant Hazards Determination and Environmental Considerations

2. Marked-up Facility Operating License and Technical Specification Pages (Red Line)

3. Clean Facility Operating License and Technical Specification Pages (Rev Bar) xc:

NMSS Project Manager Regional Administrator, Region I State of Florida

DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / 72 - 1035 LICENSE NUMBER DPR - 72 LICENSE AMENDMENT REQUEST #323, REVISION 0, PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS FOR THE INDEPENDENT SPENT FUEL STORAGE INSTALLATION TO REFLECT PERMANENT REMOVAL OF SPENT FUEL FROM THE SPENT FUEL POOLS ENCLOSURE 1 DISCUSSION OF CHANGE, TECHNICAL ANALYSIS, SIGNIFICANT HAZARDS DETERMINATION, AND ENVIRONMENTAL CONSIDERATIONS

U.S. Nuclear Regulatory Commission 3F0816-01 Page 1 of 38 LICENSE AMENDMENT REQUEST #323, REVISION 0- PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS FOR THE INDEPENDENT SPENT FUEL STORAGE INSTALLATION TO REFLECT PERMANENT REMOVAL OF SPENT FUEL FROM THE SPENT FUEL POOLS Discussion of Change, Technical Analysis, Significant Hazards Determination and Environmental Considerations

1.0 DESCRIPTION

Pursuant to 10 CFR 50.90, Duke Energy Florida, LLC, previously known as Duke Energy Florida, Inc. (DEF), requests an amendment to Facility Operating License Number DPR-72 for Crystal River Unit 3 (CR-3). The proposed amendment would revise the Operating License and associated Permanently Defueled Technical Specifications (PDTS) to reflect removal of all CR-3 spent nuclear fuel from the spent fuel pools (SFP) and its transfer to dry cask storage within the onsite Independent Spent Fuel Storage Installation (ISFSI).

By letter dated February 20, 2013, DEF submitted a certification of permanent removal of fuel from the reactor vessel pursuant to 10 CFR 50.82(a)(1)(ii) and on March 13, 2013, the NRC officially acknowledged this letter (Reference 1). Therefore, the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. An update to the Post-Shutdown Decommissioning Activities Report (PSDAR) for CR-3 dated November 16, 2015 (Reference 2), documented that DEF expects to have all spent fuel transferred to the ISFSI by the end of February 2018. Transfer of fuel out of the SFP supports decommissioning of CR-3, which involves the eventual dismantlement of the SFP. In support of this condition, the CR-3 license and associated PDTS are being proposed for revision, in accordance with 10 CFR 50.36(c)(6), to comport to a facility configuration with all spent nuclear fuel in dry storage within the onsite ISFSI at CR-3 using casks certified for use under a general 10 CFR 72 license.

AREVA submitted an Application for Amendment 14, to the standardized NUHOMS Certificate of Compliance (CoC) No. 1004 for Spent Fuel Storage Casks, Revision 0, on April 16, 2015 (Reference 5). AREVA submitted a revision to the above referenced Application for Amendment 14 of this CoC on November 11, 2015 (Reference 6). This revision requests the deletion of the License Condition to require returning a dropped transfer cask or storage cask (DSC) to the spent fuel pool for inspection. AREVA has requested approval of the CoC amendment by October 2016. Once the CoC is revised, there will no longer be a requirement to return spent fuel to the spent fuel pools.

The existing CR-3 PDTS contain Limiting Conditions for Operation (LCOs) that provide for appropriate functional capability of equipment required for safe storage and management of irradiated fuel with fuel stored in a spent fuel pool (SFP). As such, the existing PDTS provide a level of control in excess of that needed for safe storage and management of irradiated fuel with all fuel stored in an ISFSI. The majority of the existing PDTS are only applicable when irradiated fuel assemblies are within the SFP. Once all spent fuel assemblies have been transferred to the ISFSI, all remaining LCOs (and associated Surveillance Requirements (SRs))

will no longer be applicable and are being proposed for deletion. The PDTS being proposed reflect the removal of all spent fuel from the SFP. The proposed changes will result in a PDTS that will be applicable to CR-3 after the last spent fuel assembly has been removed from the SFP and placed within the ISFSI.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 2 of 38 There is one other pending license amendment request associated with the CR-3 Facility Operating License or PDTS currently docketed for CR-3. This change requested deletion of the Cyber Security License Condition after all fuel is removed from the SFP and transported to the ISFSI (Reference 7). Final disposition of this change will be performed closer to approval of this amendment request.

2.0 PROPOSED CHANGE

The proposed amendment would modify the CR-3 Facility Operating License to comport to the condition of all irradiated fuel in dry storage within the onsite Independent Spent Fuel Storage Installation (ISFSI) at CR-3 using casks certified for use under a general 10 CFR 72 license.

The amendment would also revise the CR-3 PDTS to eliminate operational requirements and certain design requirements involving storage of spent fuel that will no longer be applicable following the transfer of the last spent fuel assembly from the spent fuel pool (SFP) to the ISFSI.

A new PDTS design requirement is being added that prohibits storage of spent fuel in the SFP to comport to the ISFSI-Only Emergency Plan and the ISFSI-Only Security Plan which do not address the spent fuel pools or items stored within them. This will also comport to the AREVA COC 1004, Amendment 14 License Condition that fuel is no longer required to be returned to the spent fuel pools.

The proposed changes to the PDTS also involve relocating administrative requirements from Section 5, Administrative Controls, to the CR-3 Quality Assurance Topical Report contained in the CR-3 Final Safety Analysis Report (FSAR), Section 1.7, or to the CR-3 procedure that is the equivalent to a Technical Requirements Manual, (Compliance Procedure CP-0500, SPECIAL ACTIONS AND REPORTING REQUIREMENTS), and subsequently controlling them in accordance with 10 CFR 50.54(a) and 10 CFR 50.59, respectively. This relocation is being proposed pursuant to the criteria contained in 10 CFR 50.36 and in accordance with recommendations contained in NRC Administrative Letter 95-06.

The proposed changes to the Facility Operating License are as follows:

Revise the wording in 1.E to reflect that Duke Energy Florida, LLC completely owns CR-3, and all previous co-owners have been removed from the license.

Revise the wording in 2.A to reflect that Duke Energy Florida, LLC owns the facility.

Eliminate License Condition 2.C.(14) related to mitigation strategy.

Add a statement in 2.G pertaining to how the license remains effective until final license termination.

Reformat to move text forward and delete empty spaces.

Remove statement on Page 2 that identified the page was revised and submitted on February 24, 1977.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 3 of 38 General Analysis Applicable to Proposed Change DEF is in the process of decommissioning CR-3. In support of this activity, the spent fuel is being transferred from the spent fuel pools to the ISFSI. The proposed changes to the CR-3 PDTS reflect the removal of all the spent fuel from the spent fuel pools. With no spent fuel in either spent fuel pool, the design bases for spent fuel storage in the pools and the design basis accident for fuel handling are no longer applicable.

The CR-3 Final Safety Analysis Report (FSAR), Section 14, Safety Analysis, currently addresses the design basis accidents (DBA) and transient scenarios applicable to CR-3 in the permanently defueled condition with irradiated fuel stored in the spent fuel pools (SFP). These postulated accidents are predicated on spent fuel being stored in the spent fuel pools.

However, upon transfer of all irradiated fuel to storage in the ISFSI, the accident scenarios postulated in the FSAR are no longer possible. The ISFSI is a passive system that does not rely on electrical power for heat transfer. With removal of the spent fuel from the spent fuel pools, there are no remaining spent fuel assemblies to be monitored and there are no credible fuel related accidents that require actions of a Certified Fuel Handler, Shift Supervisor, or a Non-certified Operator to prevent occurrence or mitigate the consequences.

DEF plans to continue using a decommissioning method called SAFSTOR, in which most fluid systems are drained and the plant is left in a stable condition until final dismantlement.

Administrative controls that are required to be in place when decontamination or dismantling activities of radioactive systems, structures, and components are being performed are designed to minimize the likelihood of an off-normal or accident event, and thereby the consequences of such an event. The proposed changes do not have an adverse impact on the remaining decommissioning activities or any of their postulated consequences.

The spent fuel will be stored in the ISFSI until it is shipped off site in accordance with the schedules described in the PSDAR and updated Irradiated Fuel Management Plan.

During decommissioning (with all spent fuel in dry storage within an ISFSI), no plant systems are relied upon for spent fuel storage. In this condition there are no credible accidents whose prevention or mitigation would need to be addressed by plant PDTS. The spent fuel storage canisters used in the ISFSI are subject to their own CoC and associated storage canister Technical Specification.

A list of the FSAR Chapter 14 DBAs is provided in Section 5.2, Applicable Regulatory Requirements/Criteria, of this submittal. There are no accident scenarios that apply to the condition with all spent fuel stored in dry casks within an ISFSI. Therefore, no analyzed accidents remain applicable to CR3 in the condition with all spent fuel stored in dry casks within an ISFSI.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 4 of 38 The definition of safety-related structures, systems, and components (SSCs) in 10 CFR 50.2, Definitions, states that safety-related SSCs are those relied on to remain functional during and following design basis events to assure:

1. The integrity of the reactor coolant boundary;

2. The capability to shut down the reactor and maintain it in a safe shutdown condition; or,

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.67(b)(2) or 100.11.

The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after all spent fuel assemblies have been transferred to the ISFSI, there are no longer any SSCs at CR-3 that are required to be relied upon for accident mitigation. Therefore, with no fuel stored in the spent fuel pools, none of the SSCs at CR-3 meet the definition of a safety-related SSC as stated in 10 CFR 50.2.

10 CFR 50.36, Technical Specifications, promulgates the regulatory requirements related to the content of Technical Specifications. As detailed in subsequent sections of this proposed amendment, this regulation lists four criteria to define the scope of equipment and parameters that must be included in a plants Technical Specifications. A discussion of the applicability of these four criteria in the permanently defueled condition with all fuel removed from the spent fuel pools is provided in Section 5.2, Applicable Regulatory Requirements/Criteria, of this submittal. In a permanently defueled condition with all spent fuel in dry storage within an ISFSI, the scope of equipment and parameters that need be included in the CR-3 PDTS is limited to a description of the design features and high radiation area administrative controls.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 5 of 38 2.1 Technical Specifications The following table provides a summary of which portions of the PDTS are being deleted in their entirety and which are being revised consistent with a plant configuration where all spent fuel is located within the onsite ISFSI. The details of, and justification for, the proposed changes follow in subsequent sections (arranged by PDTS Section).

PDTS Being Deleted PDTS Being Revised/Maintained 1.0 USE AND APPLICATION 1.1 Definitions 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY 3.7 PLANT SYSTEMS 3.7.13 Spent Fuel Pool Water Level 3.7.14 Spent Fuel Pool Boron Concentration 3.7.15 Spent Fuel Pool Storage 4.0 DESIGN FEATURES 4.1 Site Location 4.3 Fuel Storage 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Unit Staff Qualifications 5.6 Procedures, Programs, and Manuals 5.7 Reporting Requirements 5.8 High Radiation Area The PDTS Table of Contents is being revised accordingly.

The corresponding PDTS Bases are also being deleted to reflect the proposed changes.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 6 of 38 2.2 Facility Operating License This section describes the proposed changes to the CR-3 Facility Operating License and the justification for each change.

License Section 1, Commission Findings Section 1 of the license contains historical conclusions determined by the Commission during past licensing activities that are informational in nature. Revised the wording in 1.E to reflect that Duke Energy Florida, LLC completely owns CR-3, and all previous co-owners have been removed from the license.

1.E.

The licensee is financially qualified and Duke Energy Florida, LLC is technically qualified and financially qualified to engage in the activities authorized by this operating license in accordance with the rules and regulations of the commission.

License Section 2., Amendments to the License

Section 2 discusses the license and what is permitted by the licensee with regards to the site and byproduct, source and/or special nuclear material (SNM). Revised the wording in 2.A to reflect that Duke Energy Florida, LLC is the sole owner and licensee of the Crystal River Unit 3 Nuclear Generating Plant.

2.A This amended license applies to the Crystal River Unit 3 Nuclear Generating Plant, a pressurized water nuclear reactor and associated equipment (the facility), owned by the licensee and operated by Duke Energy Florida, LLC. The facility is located and amended.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 7 of 38 License Section 2.C, License Conditions

Section 2 also addresses the conditions of the license which includes the conditions imposed by regulation and the conditions identified below. License Condition 2.C.(14) addresses requirements added to the license to assure the licensee would be capable of mitigating large fires and explosions.

This section is proposed for deletion in its entirety. After all spent fuel is stored within the ISFSI, the mitigation strategy license condition is no longer required.

2.C.(14)

Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions and that include the following key areas:

(1)

Fire fighting responses strategy with the following elements:

a.

Pre-defined coordinated fire response strategy and guidance

b.

Assessment of mutual aid fire fighting assets

c.

Designated staging areas for equipment and materials

d.

Command and control

e.

Training of response personnel (2)

Operations to mitigate fuel damage considering the following:

a.

Protection and use of personnel assets

b.

Communications

c.

Minimizing fire spread

d.

Procedures for implementing integrated fire response strategy

e.

Identification of readily-available pre-staged equipment

f.

Training on integrated fire response strategy

g.

Spent fuel pool mitigation measures (3)

Actions to minimize release to include consideration of:

a.

Water spray scrubbing

b.

Dose to onsite responders The NRC issued this license condition on August 23, 2007, to incorporate the requirements for the Interim Compensatory Measures (ICM) Order EA-02-026, Section B.5.b mitigation strategies (dated February 25, 2002). Subsequently, 10 CFR 50.54(hh)(2) became effective on May 26, 2009. This section provides mitigation strategies and response procedure requirements for loss of large areas of the plant due to explosions or fire. However, as stated in 10 CFR 50.54(hh)(3),

this section does not apply to a defueled reactor that has submitted the certification for permanent removal of fuel under 10 CFR 50.82(a).

On November 28, 2011, the NRC issued a letter that rescinded Item B.5.b of the ICM Order EA-02-26. Therefore, neither the ICM Order nor 10 CFR 50.54(hh) continue to apply to CR-3.

This paragraph will read as follows.

2.C.(14) Deleted by Amendment No..

U.S. Nuclear Regulatory Commission 3F0816-01 Page 8 of 38 License Condition 2.G

Section 2.G provides the date of the expiration of the license. This date was extended via Amendment #97 to December 3, 2016. A statement is being added to this License Condition to clarify that although the license will expire on this date, per regulation, the license will continue in effect until the commission terminates the license.

2.G This amended license is effective as of the date of issuance. Facility Operating License No. DPR-72, as amended, shall expire at midnight, December 3, 2016.

This License Condition is proposed for revision to document the date of the 10 CFR 50.82(a)(1) permanent shutdown submittal and the requirements of 10 CFR 50.51, Continuation of License.

This was determined to be an appropriate change since there were numerous questions related to the impact of the operating license expiration date.

This paragraph will read as follows:

2.G This amended license is effective as of the date of issuance. Facility Operating License No. DPR-72, as amended, shall expire at midnight, December 3, 2016.

Duke Energy Florida, LLC submitted the 10 CFR 50.82(a)(1) notification to the Nuclear Regulatory Commission on February 20, 2013. Per 10 CFR 50.51(b),

the Facility Operating License No. DPR-72 continues in effect until the Commission notifies the licensee that the License has been terminated.

Editorial Changes to the License Page 2 of the CR-3 Facility Operating License has a statement that this revised page was submitted on 2-24-77. This statement is being proposed for deletion as it does not tie to a specific section of the facility operating license and may confuse rather than clarify. Other than for historical purposes, there is no need to keep this statement as it provides no valuable information.

Additionally, it is being proposed to reformat the pages by moving up text and eliminating blank spaces and pages. No changes to the text are proposed under this change.

3.0 TECHNICAL ANALYSIS

and 4.0

SUMMARY

The following portion of this license amendment request contains a summary and technical justification for the proposed changes to Permanently Defueled Technical Specifications (PDTS)

Sections 1, 3, 4 and 5 (PDTS Section 2 was previously deleted).

U.S. Nuclear Regulatory Commission 3F0816-01 Page 9 of 38 A combined chapter, containing a separate description, the proposed change, technical analysis, and summary of the change is provided separately for each PDTS section. These individual chapters combine to constitute Parts 3.0 and 4.0 of this license amendment request.

For grouping purposes the description, proposed change, technical analysis, and summary of the change for each TS section is labeled as follows.

Txx.1 Description Txx.2 Proposed Change Txx.3 Technical Analysis Txx.4 Summary For the labels above, the numerical suffix 1 corresponds to Description, the suffix 2 corresponds to Proposed Change, the suffix 3 corresponds to Technical Analysis, and the suffix 4 corresponds to Summary.

The xx is a numerical designator that corresponds to the associated PDTS section (e.g., T3.1 would correspond to the Description for PDTS Section 3, whereas T37.3. would correspond to the Technical Analysis for PDTS Section 3.7).

The General Analysis Applicable to Proposed Change documented in Section 2.0, Proposed Change, above, is also applicable to the following proposed PDTS changes.

TS SECTION 1.0, USE AND APPLICATION T

1.1 DESCRIPTION

The existing PDTS Section 1.0, Use and Application, contains the rules of usage for the PDTS. This section is divided into the following four subsections.

1.1 Definitions - Defines terms used and applicable throughout the PDTS and Bases.

1.2 Logical Connectors - An explanation of the logical connectors used to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies.

1.3 Completion Times - Establishes the Completion Time convention and provides guidance for its use.

1.4 Frequency - Defines the proper use and application of Frequency requirements.

Because storage of fuel in an ISFSI does not rely on plant systems or activities addressed by PDTS, the requirements of this section will no longer apply (as discussed below) and are being proposed for deletion in their entirety.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 10 of 38 T

1.2 PROPOSED CHANGE

TS Section 1.0, USE AND APPLICATION All PDTS in Section 1.0 are being deleted in their entirety, as identified in the table below.

Justification for deletion of these PDTS are as further described below and shown in Enclosure 2.

PDTS Being Deleted PDTS Being Revised/Maintained 1.0 USE AND APPLICATION 1.1 Definitions 1.2 Logical Connectors 1.3 Completion Times 1.4 Frequency There are no corresponding PDTS Bases sections associated with this PDTS section.

T

1.3 TECHNICAL ANALYSIS

TS Section 1.1, Definitions PDTS 1.1, Definitions, provides defined terms that are applicable throughout the PDTS and PDTS Bases. After transfer of spent fuel from the spent fuel pools to the ISFSI is complete, there will no longer be any applicable Limiting Conditions for Operation (LCOs) or Surveillance Requirements (SRs) in the CR-3 PDTS. As such, the definitions described below will no longer be needed. Therefore, deleting these definitions from the PDTS during implementation after the spent fuel is transferred from the SFP to the ISFSI is acceptable. The following definitions are being proposed for deletion because they will no longer have relevance to the plant PDTS after all fuel is removed from the spent fuel pools and stored in the onsite ISFSI.

Definitions Being Deleted Term Definition Being Deleted (summarized)

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

Amendment No. 247 removed all other definitions.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 11 of 38 PDTS Section 1.2, Logical Connectors PDTS 1.2, Logical Connectors, contains an explanation of the logical connectors used to discriminate between, and yet connect, discrete Conditions, Required Actions, Completion Times, Surveillances, and Frequencies throughout the PDTS.

PDTS Section 1.2 is being proposed for deletion in its entirety because all PDTS sections that contain Conditions, Required Actions, Completion Times, Surveillances, and Frequencies are also being deleted. As such, logical connectors will no longer appear in the PDTS and the section that describes them may be deleted.

PDTS Section 1.3, Completion Times PDTS 1.3, Completion Times, establishes the Completion Time convention throughout the PDTS and provides guidance for its use. The Completion Time is the amount of time allowed for completing a Required Action.

PDTS Section 1.3 is being proposed for deletion in its entirety because all PDTS sections that contain Required Actions and Completion Times are also being deleted. As such, Completion Times will no longer appear in the PDTS and the section that describes them may be deleted.

PDTS Section 1.4, Frequency PDTS 1.4, Frequency, defines the proper use and application of Frequency requirements throughout the PDTS. Each Surveillance Requirement has a specified Frequency, in which the Surveillance must be met in order to meet the associated LCO.

PDTS Section 1.4 is being proposed for deletion in its entirety because all PDTS sections that contain Surveillances and Frequencies are also being deleted. As such, Frequency will no longer appear in the PDTS and the section that describes it may be deleted.

T1.4

SUMMARY

Current Permanently Defueled Condition with Fuel in the Spent Fuel Pools PDTS Section 1.0, Use and Application, does not contain applicability requirements. As such, all parts of this section can be conservatively defined as being applicable at all times.

All Irradiated Fuel Stored in an Independent Spent Fuel Storage Installation Since PDTS Section 1.0 does not apply in the condition where all spent fuel is removed from the spent fuel pools and stored in the onsite ISFSI, the individual PDTS contained therein are not needed. As such, they may be deleted in their entirety with no impact on continued safe storage and maintenance of spent fuel in the ISFSI.

Conclusion Deleting all PDTS in Section 1.0 is acceptable.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 12 of 38 PDTS SECTION 3.0, LIMITING CONDITION FOR OPERATION APPLICABILITY SURVEILLANCE REQUIREMENT APPLICABILITY T

3.1 DESCRIPTION

The existing PDTS Section 3.0, Limiting Condition for Operation (LCO) Applicability, and Surveillance Requirement (SR) Applicability, contains general requirements applicable to all Specifications.

Because storage of spent fuel in an ISFSI does not rely on plant systems or activities addressed by PDTS, all PDTS sections that contain LCOs and SRs are being deleted. As such, the requirements of this section will no longer apply and are being proposed for deletion in their entirety.

T

3.2 PROPOSED CHANGE

PDTS Section 3.0, LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY; SURVEILLANCE REQUIREMENT (SR) APPLICABILITY All PDTS in Section 3.0 are being deleted in their entirety, as identified in the table below.

Proposed deletion of these PDTS (including the LCOs being deleted) are as further described below and shown in Enclosure 2.

PDTS Being Deleted PDTS Being Revised/Maintained 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY The corresponding PDTS Bases sections are also being deleted to reflect this change.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 13 of 38 T

3.3 TECHNICAL ANALYSIS

PDTS 3.0, LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY PDTS 3.0, Limiting Condition for Operation (LCO) Applicability, consists of LCO 3.0.1 and LCO 3.0.2. These LCOs establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

After the transfer of spent fuel from the spent fuel pools to the ISFSI, there will no longer be any applicable Limiting Conditions for Operation (LCOs) or Surveillance Requirements (SRs). As such, the LCOs described below will no longer be needed. Therefore, deleting these LCOs from the PDTS during implementation after the spent fuel is transferred from the SFP to the ISFSI is acceptable. The following LCOs are being proposed for deletion because they no longer have relevance to the plant PDTS after all spent fuel is removed from the spent fuel pools and stored in the onsite ISFSI.

LCOs Being Deleted LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the Modes or other specified conditions of the Applicability statement of each Specification).

LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met.

LCOs 3.0.1 and 3.0.2, are being proposed for deletion in their entirety. These two LCOs are no longer needed since all remaining PDTS to which they apply are being proposed for deletion.

Therefore, the proposed deletion of LCOs 3.0.1 and 3.0.2, is acceptable.

PDTS 3.0, SURVEILLANCE REQUIREMENT (SR) APPLICABILITY PDTS 3.0, Surveillance Requirement (SR) Applicability, consists of SR 3.0.1 through SR 3.0.4.

SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

After the transfer of spent fuel from the spent fuel pools to the ISFSI, there will no longer be any applicable Limiting Conditions for Operation (LCOs) or Surveillance Requirements (SRs). As such, the SRs described below will no longer be needed. Therefore, deleting these SRs from the PDTS during implementation after the spent fuel is transferred from the SFP to the ISFSI is acceptable. The following SRs are being proposed for deletion because they no longer have relevance to the plant PDTS after all spent fuel is removed from the spent fuel and stored in the ISFSI.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 14 of 38 SRs Being Deleted SR 3.0.1 establishes the requirement that SRs must be met during the Modes or other specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify the OPERABILITY of systems and components, and that variables are within specified limits.

SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per..." interval.

SR 3.0.3 provides the flexibility to defer declaring affected equipment inoperable or an affected variable outside the specified limits when a surveillance requirement has not been completed within the specified Frequency.

SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a Mode or other specified condition in the Applicability.

SRs 3.0.1, 3.0.2, 3.0.3, and 3.0.4 are being proposed for deletion in their entirety.

These four SRs are no longer needed since all remaining PDTS to which they apply are being proposed for deletion. Therefore, the proposed deletion of SRs 3.0.1, 3.0.2, 3.0.3, and 3.0.4 is acceptable.

T3.4

SUMMARY

Current Permanently Defueled Condition with Fuel in the Spent Fuel Pools PDTS Section 3.0, Limiting Condition for Operation (LCO) Applicability, and Surveillance Requirement (SR) Applicability, does not contain applicability requirements. As such, all parts of this section can be conservatively defined as being applicable at all times.

All Irradiated Fuel Stored in an Independent Spent Fuel Storage Installation PDTS Section 3.0 does not apply in the condition where all spent fuel has been removed from the spent fuel pools and is stored in an ISFSI. Since the PDTS to which PDTS section 3.0 applies are being proposed for deletion, the PDTS Section 3.0 requirements are no longer needed. As such, they may be deleted in their entirety with no impact on continued safe storage and maintenance of spent fuel in the ISFSI.

Conclusion Deleting all PDTS in Section 3.0 is acceptable.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 15 of 38 PDTS SECTION 3.7, PLANT SYSTEMS T3.

7.1 DESCRIPTION

The existing PDTS in Section 3.7, Plant Systems (PDTS 3.7.13, 3.7.14, and 3.7.15), contain Limiting Conditions for Operation (LCOs) that provide for appropriate functional capability of plant equipment required for safe maintenance and storage of fuel assemblies in the spent fuel pools.

After the transfer of spent fuel from the spent fuel pools to the ISFSI, there will no longer be any spent fuel assemblies in the spent fuel pools. As such, PDTS 3.7.13, 3.7.14, and 3.7.15 (along with their respective LCOs and Surveillance Requirements) will no longer be applicable.

Therefore, deleting these PDTS during implementation after all spent fuel has been transferred from the SFP to the ISFSI is acceptable. All PDTS in Section 3.7 are being proposed for deletion because they have no relevance to, and no longer apply to, the storage of spent fuel assemblies in an ISFSI.

T

37.2 PROPOSED CHANGE

PDTS Section 3.7, Plant Systems All PDTS in Section 3.7 are being deleted in their entirety, as identified in the table below (Note:

TS Sections 3.7.1, 3.7.2, 3.7.3, 3.7.4, 3.7.5, 3.7.6, 3.7.7, 3.7.8, 3.7.9, 3.7.10, 3.7.11, 3.7.12, 3.7.16, 3.7.17, 3.7.18, and 3.7.19 were previously deleted by Amendment No. 247.)

PDTS Being Deleted PDTS Being Revised/Maintained 3.7 PLANT SYSTEMS 3.7.13 Spent Fuel Pool Water Level 3.7.14 Spent Fuel Pool Boron Concentration 3.7.15 Spent Fuel Assembly Storage The corresponding PDTS Bases sections are also being deleted to reflect this change.

T3.

7.3 TECHNICAL ANALYSIS

Section 3.7 PDTS That Are Not Applicable When All Fuel Stored in ISFSI PDTS 3.7.13, 3.7.14, and 3.7.15 do not currently apply when no fuel assemblies are in the spent fuel pools.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 16 of 38 After all spent fuel has been removed from the spent fuel pools, there is no need to maintain spent fuel water level, verification of boron concentration, or spent fuel pool configuration. Safe load paths over the spent fuel pool are no longer necessary.

PDTS 3.7.13, Fuel Storage Pool Water Level, specifies requirements to ensure that the minimum water level in the spent fuel pools (the A pool, and B pool) meets the iodine decontamination factor assumptions used in the fuel handling accident (FHA) analysis of record.

The specified water level shields and minimizes the general area dose when the storage racks are filled to their maximum capacity. The water also provides shielding during the movement of spent fuel. PDTS 3.7.13 is applicable during movement of irradiated fuel assemblies in the spent fuel pools.

TS 3.7.14, Spent Fuel Pool Boron Concentration, specify requirements to ensure that the spent fuel pool boron concentration is 1925 ppm. The specified concentration of dissolved boron in the spent fuel pools preserves the assumptions used in the analyses of the potential criticality accident scenarios. This concentration of dissolved boron is the minimum required for fuel assembly storage and movement within the spent fuel pools. PDTS LCO 3.7.14 applies whenever fuel assemblies are stored in the spent fuel pools, until a complete spent fuel pool verification has been performed following the last movement of fuel assemblies in the spent fuel pools.

TS 3.7.15, Spent Fuel Assembly Storage, specifies restrictions on the placement of fuel assemblies within the spent fuel pools, in accordance with Figures 3.7.15-1 and 3.7.15-2 to ensure the reactivity (keff ) of the spent fuel pools will always remain < 0.95, assuming the pools to be flooded with unborated water. Specific categories of spent fuel are determined from initial enrichment versus fuel burn-up. These categories can be stored in the pools as determined by Figures 3.7.15-1 and 3.7.15-2. PDTS LCO 3.7.15 applies whenever any fuel assembly is stored in the spent fuel pools.

T3.7.4

SUMMARY

Current Permanently Defueled Condition with Fuel in the Spent Fuel Pools PDTS 3.7.13, 3.7.14, and 3.7.15 are related to assuring the appropriate functional capability of plant equipment required for safe storage and maintenance of spent fuel stored in the spent fuel pools. PDTS 3.7.13, 3.7.14, and 3.7.15 are applicable when fuel assemblies are stored or moved in the spent fuel pools.

All Irradiated Fuel Stored in an Independent Spent Fuel Storage Installation PDTS 3.7.13, 3.7.14, and 3.7.15 do not apply when all spent fuel assemblies are removed from the spent fuel pools and stored in an ISFSI. Therefore, these three PDTS will no longer be needed following the transfer of all fuel assemblies from the spent fuel pools to the ISFSI. As such, these three PDTS may be deleted in their entirety with no impact on continued safe storage and maintenance of irradiated fuel in an ISFSI at CR-3.

LCOs that provide for appropriate functional capability of facility equipment required for safe maintenance and storage of fuel assemblies in the ISFSI are specified in the applicable ISFSI TS.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 17 of 38 Conclusion Deleting PDTS 3.7.13, 3.7.14, and 3.7.15 in Section 3.7 is acceptable.

PDTS SECTION 4.0, DESIGN FEATURES T

4.1 DESCRIPTION

The existing PDTS Section 4.0, Design Features, contains descriptions and requirements for those features of the facility such as materials of construction and geometric arrangements which, if altered or modified, would have a significant effect on safety of the spent fuel pools and are not covered in the previous sections of the PDTS.

After the transfer of spent fuel from the spent fuel pools to the ISFSI, there will no longer be any fuel assemblies in the spent fuel pools. As such, certain design features will have no relevance to, and no longer apply to, the storage of fuel assemblies in an ISFSI. Therefore, deleting these PDTS during implementation after the spent fuel is transferred from the SFP to the ISFSI is acceptable.

Since this proposed amendment is premised on spent fuel no longer being stored in the spent fuel pools, a new design feature specification is being proposed stating that spent fuel shall not be stored in the spent fuel pools.

AREVA submitted an Application for Amendment 14, to the standardized NUHOMS Certificate of Compliance (CoC) No. 1004 for Spent Fuel Storage Casks, Revision 0, on April 16, 2015 (Reference 5). AREVA submitted a revision to the above referenced Application for Amendment 14 of this CoC on November 11, 2015 (Reference 6). This revision requests the deletion of the License Condition to require returning a dropped transfer cask or DSC to the Spent Fuel Pool for inspection. AREVA has requested approval of the CoC amendment by October 2016. Once the CoC is revised, there will no longer be a requirement to return spent fuel to the spent fuel pools.

T

4.2 PROPOSED CHANGE

PDTS Section 4.0, Design Features PDTS being retained and/or revised are 4.1 and 4.3 as further described below and shown in (the currently existing design features description in PDTS 4.3 is being deleted and replaced with a new design feature specification.)

Note: PDTS 4.2 was previously deleted under Amendment No. 247.

TS Being Deleted TS Being Revised/Maintained 4.0 DESIGN FEATURES 4.1 Site Location 4.3 Fuel Storage There are no corresponding PDTS Bases sections associated with this PDTS section.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 18 of 38 T

4.3 TECHNICAL ANALYSIS

PDTS 4.1, Site PDTS 4.1, Site Location, provides a description of the location of CR-3. This PDTS section is being retained.

PDTS 4.3, Fuel Storage PDTS Section 4.3, Fuel Storage, provides a description and requirements regarding prevention of criticality of spent fuel in the spent fuel pool storage racks and new fuel in the new fuel storage racks, prevention of spent fuel pool drainage, and spent fuel capacity limitations.

All currently existing requirements in PDTS Section 4.3 are being deleted in their entirety. All new fuel will be transferred off-site to another licensed reactor prior to offloading the spent fuel from the pools to the ISFSI.

After the transfer of spent fuel from the spent fuel pool to the ISFSI, there will no longer be any fuel assemblies in the pools. Therefore, the design features associated with fuel storage in the spent fuel pools and the new fuel storage racks is no longer applicable and may be deleted.

Since this proposed amendment is premised on spent fuel no longer being stored in the spent fuel pools, a new design feature specification is being proposed to be added to PDTS Section 4.3 stating that spent fuel shall not be stored in the spent fuel pools.

The new specification will read:

Spent fuel shall not be stored in the spent fuel pool.

Reactor spent fuel stored in dry canisters within the ISFSI is subject to the TS applicable to the associated storage canister system. The storage canister types used at CR-3 is the Transnuclear Model NUHOMS-32PTH1 under Certificate of Compliance (CoC) No. 1004, Amendment 14. The associated TS for this storage canister require (for assurance that the spent fuel will continue to meet the requirements for storage) that the reactor spent fuel in the canister be returned to the reactor spent fuel pools for inspection of the DSC and transfer cask (TC) in the event of a canister drop during handling. A discussion of this capability (returning the fuel to the spent fuel pool) appears in the Updated Final Safety Analysis Report (UFSAR) for the NUHOMS storage system.

The UFSAR for the NUHOMS storage system, Section 5.1.1.9 Removal of Fuel from the DSC (and several other sections), states the following.

If it becomes necessary to remove fuel from the DSC [dry shielded canister] prior to off-site shipment, there are two basic options available at the ISFSI or reactor site. The fuel assemblies could be removed and reloaded into a shipping cask using dry transfer techniques, or if the applicant so desires, the initial fuel loading sequence could be reversed and the plant's spent fuel pool utilized.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 19 of 38 To address the inconsistency with the requirements of NUHOMS CoC 1004 regarding return of the fuel to the spent fuel pool, a licensing action regarding these requirements is being separately requested by AREVA. The certificate holder of CoC No. 1004, AREVA, Inc.,

submitted a proposed change to the storage canister TS on November 11, 2015 (Reference 4).

The proposed change submitted by AREVA would eliminate the requirement to return the fuel to the spent fuel pool and therefore eliminate the need for the spent fuel pool to be available to meet the storage canister TS requirement. The revision to the CR-3 PDTS, Section 4.3 will align these two documents.

T4.4

SUMMARY

Current Permanently Defueled Condition with Fuel in the Spent Fuel Pools PDTS Section 4.0, Design Features, does not contain applicability requirements. As such, all parts of this section are conservatively assumed to be applicable at all times.

All Irradiated Fuel Stored in an Independent Spent Fuel Storage Installation PDTS 4.1 will remain germane with the spent fuel pool emptied of fuel assemblies. As such, this PDTS section is being retained.

The existing PDTS Section 4.3 describes design features associated with fuel storage in the spent fuel pools and storage of new fuel in the new fuel storage racks. After all spent fuel is removed from the spent fuel pools the existing information in PDTS Section 4.3 is no longer applicable and may be deleted. Adding a new design feature stating that spent fuel shall not be stored in the spent fuel pools documents the premise on which this proposed amendment is based (i.e., spent fuel no longer being stored in the spent fuel pool). This will also comport to the AREVA COC 1004, Amendment 14 requirement that fuel is no longer required to be returned to the spent fuel pools. Should the AREVA Amendment not be approved, this revision to the PDTS will not be implemented.

Conclusion Retaining PDTS 4.1, and adding the proposed design feature specification (regarding prohibition on spent fuel storage in the spent fuel pool) to PDTS 4.3, ensures appropriate requirements for the associated design features.

Deleting the currently existing (no longer applicable) design features in PDTS Section 4.3 is acceptable since no fuel will be stored in the spent fuel pools.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 20 of 38 PDTS SECTION 5.0, ADMINISTRATIVE CONTROLS T

5.1 DESCRIPTION

The existing PDTS Section 5.0, Administrative Controls, contains provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner.

After the transfer of spent fuel from the spent fuel pools to the ISFSI, there will no longer be any fuel assemblies in the spent fuel pools. As such, the associated administrative controls will have no relevance to and no longer apply to the storage of fuel assemblies in an ISFSI.

Therefore, deleting the associated PDTS during implementation after the spent fuel is transferred from the SFP to the ISFSI is acceptable.

T

5.2 PROPOSED CHANGE

PDTS Section 5.0, Administrative Controls All PDTS in Section 5.0, with the exception of TS 5.8, High Radiation Area, are being deleted in their entirety, as identified in the table below and as further described and as shown in. Pertinent information being deleted from the PDTS is being relocated to either the Quality Assurance Program Description (QAPD) or the CR-3 procedure that is the equivalent to a Technical Requirements Manual, (Compliance Procedure CP-0500, SPECIAL ACTIONS and REPORTING REQUIREMENTS). PDTS 5.8 is being retained unchanged.

TS Being Deleted TS Being Revised 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.2 Organization 5.3 Unit Staff Qualifications 5.6 Procedures, Programs, and Manuals 5.7 Reporting Requirements 5.8 High Radiation Area There are no corresponding PDTS Bases sections associated with this PDTS section.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 21 of 38 T

5.3 TECHNICAL ANALYSIS

NRC Administrative Letter 95-06 provides a discussion concerning the relocation of Technical Specification administrative controls to a quality assurance (QA) program. The NRC considers relocating these requirements to the quality assurance program acceptable because of the controls imposed by 10 CFR 50, Appendix B, the existence of an NRC approved quality assurance program, and the quality assurance program change control process in 10 CFR 50.54(a). The CR-3 quality assurance program is described in the Quality Assurance Program Description (QAPD) located in Section 1.7 of the CR-3 FSAR.

CP-0500 is part of the Final Safety Analysis Report (FSAR) and therefore subject to the requirements of 10 CFR 50.59. Maintaining relocated requirements in accordance with the change control process in 10 CFR 50.59 provides adequate control based on the ISFSI-only status of the facility. With the transfer of the spent fuel to the ISFSI, the administrative controls pertaining to the safe storage of spent fuel within the spent fuel pools are no longer needed or applicable.

PDTS 5.1, Responsibility PDTS 5.1, Responsibility, provides a description and requirements regarding certain key operational management responsibilities.

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

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

5.1.2 The Shift Supervisor shall be responsible for the shift command function The requirements of PDTS 5.1.1 are being removed from the PDTS and relocated to the QAPD.

The remaining requirements of Section 5.1.1 related to the responsibilities of the General Manager Decommissioning will be removed from the PDTS and relocated to the QAPD.

Relocating these responsibilities to the QAPD is consistent with the intent of NRC Administrative Letter 95-06. Therefore, the proposed removal of PDTS 5.1 is acceptable.

DEF is proposing to delete PDTS 5.1.2 in its entirety. The Shift Supervisor requirements in PDTS 5.1.2 are being eliminated. With removal of all of the spent fuel from the spent fuel pools, a need for the Shift Supervisor for spent fuel management no longer exists. The position of Shift Supervisor described in PDTS 5.1.2 is a holdover from the control room function of supervising multiple functions of an operating nuclear power plant. With the limited requirements for supervision of the passive fuel storage at the ISFSI or with respect to the decommissioning of the former power generation facility, the Shift Supervisor position is no Requirements Relocated to QAPD

U.S. Nuclear Regulatory Commission 3F0816-01 Page 22 of 38 longer required and the proposed deletion of PDTS 5.1.2 is acceptable. Therefore, Section 5.1.2, which provides a description of responsibility of the Shift Supervisor will be deleted from the PDTS.

PDTS 5.2, Organization PDTS 5.2, Organization, provides a description and requirements regarding onsite and offsite organizations and facility staffing. Descriptions include lines of authority and staff responsibilities. Requirements include the associated PDTS 5.2.2.a, which specifies a minimum shift crew composition staffing requirement for Certified Fuel Handlers and Non-Certified Operators. PDTS 5.2 also specifies requirements for fuel handling operations and supervision.

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

a.

Lines of authority, responsibility, and communication shall be established and defined from the highest management levels through intermediate levels to and including 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 General Manager Decommissioning 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 General Manager Decommissioning 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 Requirements relocated to QAPD Retained portion relocated to QAPD

U.S. Nuclear Regulatory Commission 3F0816-01 Page 23 of 38 The unit staff organization shall include the following:

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

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

c.

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

d.

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

e.

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

f.

The Shift Supervisor shall be a Certified Fuel Handler.

The requirements of PDTS 5.2.1 are being removed from the PDTS and relocated to the QAPD, with the exception of the certified fuel handler trainer requirements in PDTS 5.2.1.c, which is being eliminated.

Section 5.2.1, Onsite and Offsite Organizations, provides a general discussion of the site organization which assures safe facility operations and safety of the nuclear fuel. Section 5.2.1.c provides requirements for organizational freedom of the Certified Fuel Handler trainers, and the health physics and quality assurance personnel. DEF proposes to eliminate the portion of Section 5.2.1.c pertaining to Certified Fuel Handler trainers. Following the transfer of all spent fuel to the ISFSI, and the new PDTS 4.3 prohibition from storing fuel in the spent fuel pools, there will no longer be a need for Certified Fuel Handlers; therefore this proposed deletion is acceptable.

The remainder of section 5.2.1 will be removed from the PDTS and relocated to the QAPD to provide an equivalent description of the requirements for organizational freedom of the health physics and quality assurance personnel. Providing onsite and offsite organization descriptions in the QAPD is consistent with NRC Administrative Letter 95-06. Therefore, the proposed removal and relocation are acceptable.

DEF is proposing to delete PDTS 5.2.2 in its entirety. Section 5.2.2, Unit Staff, currently specifies the organizations and positions for activities affecting the safe storage of irradiated fuel in the spent fuel pools. The QAPD addresses the necessary organizational requirements for CR-3 after all spent fuel has been transferred to ISFSI. Therefore, the deletion of PDTS 5.2.2 after the fuel has been moved from the spent fuel pools to the ISFSI will have no impact and is acceptable.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 24 of 38 PDTS 5.3, Facility Staff Qualifications PDTS 5.3, Unit Staff Qualifications, provides a description and requirements regarding qualifications of the facility staff. It also specifies that an NRC approved training and retraining program for the Certified Fuel Handlers (CFH) shall be maintained.

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 General Manager Decommissioning.

The requirements of PDTS 5.3.1 are being removed from the PDTS and relocated to the QAPD.

DEF is proposing to delete PDTS 5.3.2 in its entirety. PDTS 5.3.2 specifies requirements for a Certified Fuel Handler training program. Following the transfer of all spent fuel to the ISFSI, and the new PDTS 4.3 prohibition from storing spent fuel in the spent fuel pools, there will no longer be a need for Certified Fuel Handlers, which obviates the need for the associated training program. Therefore, this proposed deletion is acceptable.

PDTS 5.6, Procedures, Programs, and Manuals PDTS 5.6, Procedures, Programs, and Manuals, provides a description and requirements regarding programs and manuals that are to be established, implemented, and maintained.

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 by 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.

Requirements Relocated to QAPD Requirements Relocated to QAPD

U.S. Nuclear Regulatory Commission 3F0816-01 Page 25 of 38 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 doe=se calculation methodologies, the radioactive effluent controls program, and radiological environmental monitoring activities. The OCDM shall contain:

1. The methodology and parameters used in the calculation of offsite doses resulting from

3. Shall be submitted to the NRC the change was implemented.

5.6.2.4 Not Used 5.6.2.5 Not Used 5.6.2.6 Not Used 5.6.2.7 Not Used 5.6.2.8 Not Used 5.6.2.9 Not Used 5.6.2.10 Not Used 5.6.2.11 Not Used 5.6.2.12 Not Used 5.6.2.13 Not Used 5.6.2.14 Not Used 5.6.2.15 Not Used 5.6.2.16 Not Used 5.6.2.3 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 methodology and parameters used in the calculation of offsite doses resulting from.

3. Shall be submitted to the NRC the change was implemented.

Relocated to CP 0500

U.S. Nuclear Regulatory Commission 3F0816-01 Page 26 of 38 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 under 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 t the updated FSAR or Bases that require NRC approval pursuant to 10 CFR 50.59.

The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with th FSAR Proposed changes that meet the criteria of Specification 5.6.2.17.a or Specification 5.6.2.17.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71.

5.6.2.18 Not Used 5.6.2.19 Not Used 5.6.2.20 Not Used 5.6.2.21 Not Used Note PDTS 5.6.2.4 through 5.6.2.16 and PDTS 5.6.2.18 through 5.6.2.21 were previously deleted in Amendments 149, 191, 244 and 247.

DEF is proposing to relocate PDTS 5.6.1, Procedures, and 5.6.2.3, Offsite Dose Calculation Manual, to the CR-3 procedure that is the equivalent to a Technical Requirements Manual, (Compliance Procedure CP-0500, SPECIAL ACTIONS AND REPORTING REQUIREMENTS),

or the QAPD, and delete PDTS 5.6.17 Technical Specification (TS) Bases Control Program.

All other Programs and Manuals have been previously deleted.

PDTS 5.6.2.3 Offsite Dose Calculation Manual (ODCM), currently specifies how to document, review, and approve changes to the ODCM. DEF proposes to remove this requirement from the Technical Specifications and relocate it to CP-0500. This requirement will be maintained in accordance with 10 CFR 50.59. Since the intent of this section is to ensure that the ODCM continues to meet the requirements of 40 CFR 190, 10 CFR 20, 10 CFR 50.36(a), and 10 CFR 50, Appendix I, and since this requirement will be maintained in CP-0500, the relocated requirement will continue to be subject to regulatory controls. This change is consistent with similar relocations approved by NRC of former TS requirements into CP-0500 (including CR-3 License Amendment 149, Improved Technical Specifications, approved by NRC on December 20,1993)

U.S. Nuclear Regulatory Commission 3F0816-01 Page 27 of 38 PDTS 5.6.2.17, specifies the process for changes to the PDTS Bases. Currently, the PDTS Bases are all related to storage of spent fuel in the spent fuel pools, specifically the requirements in PDTS 3.7, which are being deleted as described above. Following transfer of all spent fuel to the ISFSI, the spent fuel pools will no longer be used for spent fuel storage.

Since all the PDTS Bases will be deleted, there will no longer be a need for a PDTS Bases Control Program. Therefore, the proposed deletion of PDTS 5.6.2.17 is acceptable.

PDTS 5.7, Reporting Requirements PDTS 5.7, Reporting Requirements, provides a description and requirements regarding reports that are to be submitted in accordance with 10 CFR 50.4.

DEF is proposing to remove PDTS 5.7, Reporting Requirements, from the Technical Specifications and relocate the requirements to CP-0500 in their entirety. The two requirements are PDTS 5.7.1.1.b, Annual Radiological Environmental Operating Report, and TS 5.7.1.1.c, Radioactive Effluent Release Report. CP-0500 is part of the FSAR and therefore subject to the requirements of 10 CFR 50.59. Maintaining these relocated requirements in accordance with 10 CFR 50.59 provides adequate control based on the ISFSI-only status of the facility.

Therefore, the proposed removal and relocation of the requirements are acceptable.

PDTS 5.8, High Radiation Area PDTS 5.8, High Radiation Area, provides a description and requirements regarding controls applied to high radiation areas in place of the controls required by paragraph 20.1601(a) and (b) of 10 CFR 20 (as provided in paragraph 20.1601(c) of 10 CFR 20). PDTS 5.8 will remain germane with all spent fuel stored in the ISFSI. As such, it is being retained as-is with no changes being proposed.

T5.4

SUMMARY

Current Permanently Defueled Condition with Fuel in the Spent Fuel Pools PDTS Section 5.0, Administrative Controls, does not contain applicability requirements. As such, all parts of this section are conservatively as assumed to be applicable at all times.

All Irradiated Fuel Stored in an Independent Spent Fuel Storage Installation PDTS Section 5.0 describes administrative controls associated with fuel storage in the spent fuel pools. After the transfer of spent fuel from the spent fuel pools to the ISFSI, there will no longer be any fuel assemblies in the spent fuel pools. However, there may continue to be high radiation areas in the facility. Therefore, with the exception of PDTS 5.8, High Radiation Area, this PDTS section is no longer required and may be deleted. Pertinent requirements will be relocated to either the Quality Assurance Program Description (QAPD) or the CR-3 procedure that is the equivalent to a Technical Requirements Manual (CP-0500) and controlled in accordance with 10 CFR 50.54(a) and 10 CFR 50.59, respectively. Appropriate administrative controls for spent fuel storage within an ISFSI are specified in the applicable ISFSI storage system TS.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 28 of 38 PDTS 5.8 will remain applicable following the complete transfer of spent fuel from the spent fuel pool to the ISFSI because high radiation areas may continue to exist in the facility. As such, it is being retained unchanged.

Conclusion Deleting PDTS 5.1, 5.2, 5.3, and 5.6 and relocating the pertinent requirements of those sections discussed above to either the QAPD or CP-0500 is acceptable.

Retaining PDTS 5.8, unchanged, continues to ensure appropriate requirements for high radiation areas.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 29 of 38

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration Pursuant to 10 CFR 50.90, Duke Energy Florida, LLC, previously known as Duke Energy Florida, Inc. (DEF), requests an amendment to Facility Operating License Number DPR-72 for Crystal River Unit 3 (CR-3). The proposed amendment would revise the Operating License and associated Permanently Defueled Technical Specifications (PDTS) to reflect removal of all CR-3 spent nuclear fuel from the spent fuel pools (SFP) and its transfer to dry cask storage within an Independent Spent Fuel Storage Installation (ISFSI).

By letter dated February 20, 2013, DEF submitted a certification of permanent removal of fuel from the reactor vessel pursuant to 10 CFR 50.82(a)(1)(ii). Therefore, the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel. An update to the Post-Shutdown Decommissioning Activities Report (PSDAR) for CR-3 dated November 16, 2015, documented that DEF expects to have all spent fuel transferred to the ISFSI by the end of February 2018. Transfer of fuel out of the SFP supports decommissioning of CR-3, which involves the eventual dismantlement of the SFP. In support of this condition, the CR-3 license and associated PDTS are being proposed for revision, in accordance with 10 CFR 50.36(c)(6), to comport to facility possession with all spent nuclear fuel in dry storage within an ISFSI at CR-3 using casks certified for use under a general 10 CFR 72 license.

The existing CR-3 PDTS contain Limiting Conditions for Operation (LCOs) that provide for appropriate functional capability of equipment required for safe storage and management of irradiated fuel with fuel stored in a spent fuel pool (SFP). As such, the existing PDTS provide a level of control in excess of that needed for safe storage and management of irradiated fuel with fuel stored in an ISFSI. The majority of the existing PDTS are only applicable when irradiated fuel assemblies are within the SFP. Once all spent fuel assemblies have been transferred to the ISFSI, all remaining LCOs (and associated Surveillance Requirements (SRs)) will no longer be applicable and are being proposed for deletion (along with deletion and/or relocation of certain design requirements and administrative requirements). The changes being proposed reflect the removal of all spent fuel from the SFP. The proposed changes will result in PDTS that will be applicable to CR-3 after the last spent fuel assembly has been removed from the SFP and placed within the ISFSI.

DEF has evaluated the proposed amendment to determine if a significant hazards consideration is involved by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment, as discussed below:

1.

Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

The proposed amendment would modify the CR-3 facility operating license and PDTS by deleting the portions of the license and PDTS that are no longer applicable to a facility with no spent nuclear fuel stored in the spent fuel pools, while modifying the remaining portions to correspond to all nuclear fuel stored within an ISFSI. This amendment will be implemented within 60 days following DEFs submittal of written notification to the NRC that all spent fuel assemblies have been transferred out of the spent fuel pools and placed in dry storage within the ISFSI.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 30 of 38 The definition of safety-related structures, systems, and components (SSCs) in 10 CFR 50.2 states that safety-related SSCs are those relied on to remain functional during and following design basis events to assure:

1. The integrity of the reactor coolant boundary;

2. The capability to shutdown the reactor and maintain it in a safe shutdown condition; or

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR 50.43(a)(1) or 100.11.

The first two criteria (integrity of the reactor coolant pressure boundary and safe shutdown of the reactor) are not applicable to a plant in a permanently defueled condition. The third criterion is related to preventing or mitigating the consequences of accidents that could result in potential offsite exposures exceeding limits. However, after all nuclear spent fuel assemblies have been transferred to dry cask storage within an ISFSI, none of the SSCs at CR-3 are required to be relied on for accident mitigation. Therefore, none of the SSCs at CR-3 meet the definition of a safety-related SSC stated in 10 CFR 50.2. The proposed deletion of requirements in the PDTS does not affect systems credited in any accident analysis at CR-3.

Section 14 of the CR-3 Final Safety Analysis Report (FSAR) described the design basis accidents (DBAs) related to the spent fuel pools. These postulated accidents are predicated on spent fuel being stored in the spent fuel pools. With the removal of the spent fuel from the spent fuel pools, there are no remaining spent fuel assemblies to be monitored and there are no credible accidents that require the actions of a Certified Fuel Handler, Shift Manager, or a Non-certified Operator to prevent occurrence or mitigate the consequences of an accident.

The proposed changes do not have an adverse impact on the remaining decommissioning activities or any of their postulated consequences.

The proposed changes related to the relocation of certain administrative requirements do not affect operating procedures or administrative controls that have the function of preventing or mitigating any accidents applicable to the safe management of irradiated fuel or decommissioning of the facility.

Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 31 of 38

2.

Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

The proposed changes eliminate the operational requirements and certain design requirements associated with the storage of the spent fuel in the spent fuel pools, and relocate certain administrative controls to the Quality Assurance Program Description or other licensee controlled document.

After the removal of the spent fuel from the spent fuel pools and transfer to the ISFSI, there are no spent fuel assemblies that remain in the spent fuel pools. Coupled with a prohibition against storage of fuel in the spent fuel pools, the potential for fuel related accidents is removed. The proposed changes do not introduce any new failure modes.

Therefore, the proposed amendment does not create the possibility of a new or different kind of accident from any previously evaluated.

3.

Does the proposed amendment involve a significant reduction in a margin of safety?

The removal of all spent nuclear fuel from the spent fuel pools into storage in casks within an ISFSI, coupled with a prohibition against future storage of fuel within the spent fuel pools, removes the potential for fuel related accidents.

The design basis and accident assumptions within the CR-3 FSAR and the PDTS relating to safe management and safety of spent fuel in the spent fuel pools are no longer applicable. The proposed changes do not affect remaining plant operations, systems, or components supporting decommissioning activities.

The requirements for systems, structures, and components (SSCs) that have been removed from the CR-3 PDTS are not credited in the existing accident analysis for any applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis.

Therefore, the proposed amendment does not involve a significant reduction in a margin of safety.

Based on the above, DEF concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 32 of 38 5.2 Applicable Regulatory Requirements/Criteria General Design Criteria The US Atomic Energy Commission (AEC) issued their Safety Evaluation (SE) of the Crystal River Unit 3 (CR-3) on July 5, 1974 with supplements dated January 13, 1975, December 3, 1976, December 30, 1976, and January 28, 1977. The SE, Section 3.1, Conformance with AEC General Design Criteria, described the conclusions the AEC reached associated with the General Design Criteria in effect at the time. The AEC stated:

The facility was designed and constructed to meet the AEC's GDC, as originally proposed in July 1967. The Commission published the revised GDC in 1971 just before the FSAR was filed. We conducted our technical review against the present version of the GDC and we conclude that the plant design acceptably conforms to the current criteria.

These General Design Criteria (GDC) that remain applicable are discussed in detail in the CR-3 FSAR, Section 1.4. Much detail was removed as systems, structures and components were abandoned.

Design Basis Accidents (DBAs)

Chapter 14 of the CR-3 FSAR described the safety analysis aspects of the plant that were evaluated to demonstrate that the plant could be decommissioned safely and that radiological consequences from postulated accidents do not exceed the guidelines of 10 CFR 50.67. The analyzed accidents were based on the Atomic Energy Guide for the Organization and Contents of Safety Analysis Reports, June 1966.

FSAR Chapter 14 currently contains descriptions of design basis safety analysis accidents that pertain to the permanently defueled condition with fuel stored in the spent fuel pools. With the station permanently shut down, with spent fuel stored within the spent fuel storage pools, only a single Standby Safeguards Analysis retains a design basis accident within its scope.

Standby Safeguards Analysis Event Standby Safeguards Analysis Events are situations requiring operation of standby or engineered safeguards because of postulated equipment failures which would create demands beyond the capacity of the core and coolant boundary protective systems.

Safety Analyses are analyses performed to satisfy regulatory requirements. The safety analyses are integral to the plants design and licensing basis. The safety analyses demonstrate the capability to prevent or mitigate the consequences of accidents. Systems, structures, and components (SSC) that perform design basis functions are credited in the safety analyses for the purpose of mitigating the transient or accident.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 33 of 38 Chapter 14 of the CR-3 FSAR described the design basis accident (DBA) scenarios that were applicable with spent nuclear fuel stored in the spent fuel pools. However, with all spent fuel stored in dry casks within the ISFSI, the fuel handling accident scenario postulated in FSAR Chapter 14 is no longer applicable. The postulated accidents applicable with spent nuclear fuel stored in the spent fuel pool are listed in the following table (including the FSAR section they are described in), with a statement whether or not they are applicable in the condition of all spent nuclear fuel stored within an ISFSI.

FSAR Chapter 14 Postulated Accidents or Transients Applicability with All Fuel in ISFSI Standby Safety Features Analysis Fuel Handling Accident (FHA) (14.2.2.3)

NO FSAR Section 14.2.2.3 describes the fuel handling accident. With all spent fuel stored in dry casks within the ISFSI, the fuel handling accident scenario postulated in FSAR Chapter 14 is no longer possible and therefore no longer part of the licensing basis.

10 CFR 50.2, Definitions, Safety-Related Structures, Systems and Components 10 CFR 50.2 defines safety-related structures, systems, and components (SSCs) as those structures, systems and components that are relied upon to remain functional during and following design basis events to assure:

1. The integrity of the reactor coolant pressure boundary

2. The capability to shut down the reactor and maintain it in a safe shutdown condition; or

3. The capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in § 50.34(a)(1) or § 100.11 of this chapter, as applicable.

10 CFR 50.36, Technical Specifications In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of Technical Specifications (TS). In doing so, the Commission placed emphasis on those matters related to the prevention of accidents and mitigation of accident consequences; the Commission noted that applicants were expected to incorporate into their TS "those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity." (Statement of Consideration, "Technical Specification for Facility Licenses; Safety Analysis Reports," 33 FR 18610 (December 17, 1968).) Pursuant to 10 CFR 50.36, TS are required to include items in the following five categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCOs); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plant's TS.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 34 of 38 In September 1992, the Commission issued NUREG-1430 Standardized Technical Specifications - Babcock and Wilcox Plants, which was developed using the guidance and criteria contained in the Commission's Interim Policy Statement. Standard Technical Specifications (STS) were established as a model for developing improved TS for Babcock and Wilcox plants in general. STS reflect the results of a detailed review of the application of the interim policy statement criteria to generic system functions, which was published in a "Split Report" issued to the Nuclear Steam System Supplier (NSSS) Owners Groups in May 1988.

STS also reflect the results of extensive discussions concerning various drafts of STS, so that the application of the TS criteria and the Writer's Guide would consistently reflect detailed system configurations and operating characteristics for all NSSS designs. As such, the generic Bases presented in NUREG-1430 provide an abundance of information regarding the extent to which the STS present requirements that are necessary to protect public health and safety.

On July 22, 1993, the Commission issued its Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors, indicating that satisfying the guidance in the policy statement also satisfies Section 182a of the Atomic Energy Act of 1954, as amended (the Act),

and 10 CFR 50.36 (58 FR 39132). The Final Policy Statement described the safety benefits of the improved STS, and encouraged licensees to use the improved STS as the basis for plant-specific TS amendments, and for complete conversions to improved STS. Further, the Final Policy Statement gave guidance for evaluating the required scope of the TS and defined the guidance criteria to be used in determining which of the LCOs and associated surveillances should remain in the TS.

The final Commission Policy Statement established four criteria to define the scope of equipment and parameters to be included in the improved Standard Technical Specifications.

These criteria were developed for licenses authorizing operation (i.e., operating reactors) and focused on instrumentation to detect degradation of the reactor coolant system pressure boundary, process variables and equipment, design features, or operating restrictions that affect the integrity of fission product barriers during design bases accidents or transients. A fourth criterion refers to the use of operating experience and probabilistic risk assessment to identify and include in the Technical Specifications structures, systems, and components (SSCs) shown to be significant to public health and safety. These criteria, which were subsequently codified in changes to Section 36 of Part 50 of Title 10 of the Code of Federal Regulations (10 CFR 50.36)

(60 FR 36953), also pertain to the Technical Specification requirements for safe storage of spent fuel. A general discussion of these considerations is provided below.

Criterion 1 of 10 CFR 50.36(c)(2)(ii)(A) states that Technical Specification limiting conditions for operation must be established for "installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." Since CR-3 is no longer licensed to operate, this criterion is not applicable.

Criterion 2 of 10 CFR 50.36(c)(2)(ii)(B) states that Technical Specification limiting conditions for operation must be established for a "process variable, design feature, or operating restriction that is an initial condition of a design basis accident [DBA] or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The purpose of this criterion is to capture those process variables that have initial values assumed in the design basis accident and transient analyses, and which are monitored and controlled during power operation. Since CR-3 is no longer licensed to operate, and DBAs are no longer credible, this criterion is not applicable.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 35 of 38 Criterion 3 of 10 CFR 50.36(c)(2)(ii)(C) states that Technical Specification limiting conditions for operation must be established for structures, systems, or components (SSCs) that are part of the primary success path and which function or actuate to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. The intent of this criterion is to capture into Technical Specifications only those SSCs that are part of the primary success path of a safety sequence analysis. Also captured by this criterion are those support and actuation systems that are necessary for items in the primary success path to successfully function. The primary success path of a safety sequence analysis consists of the combination and sequences of equipment needed to operate (including consideration of the single failure criterion), so that the plant response to design basis accidents and transients limits the consequences of these events to within the appropriate acceptance criteria. Since fuel will have been removed from the spent fuel pool at the CR-3 facility prior to implementation of this amendment, this criterion is not applicable.

Criterion 4 of 10 CFR 50.36(c)(2)(ii)(D) states that Technical Specification limiting conditions for operation must be established for SSCs that operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The intent of this criterion is that risk insights and operating experience be factored into the establishment of Technical Specification limiting conditions for operation. Since fuel will have been removed from the spent fuel pool at the CR-3 facility prior to implementation of this amendment, this criterion is not applicable.

Addressing administrative controls, 10 CFR 50.36(c)(5) states that they "...are the provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner." The particular administrative controls to be included in the TS, therefore, are the provisions that the Commission deems essential for the safe operation of the facility that are not already covered by other regulations. Accordingly, the NRC staff determined that administrative control requirements that are not specifically required under Section 50.36(c)(5), and which are not otherwise necessary to obviate the possibility of an abnormal situation or an event giving rise to an immediate threat to the public health and safety, may be relocated to more appropriate documents (e.g., Quality Assurance Program, Technical Requirements Manual, Security Plan, or Emergency Plan), which are subject to regulatory controls. Similarly, while the required content of TS administrative controls is specified in 10 CFR 50.36(c)(5), particular details may be relocated to licensee-controlled documents, where other regulations provide adequate regulatory control.

10 CFR 50.36(c)(6), Decommissioning, applies only to nuclear power reactor facilities that have submitted the certifications required by § 50.82(a)(1). For such facilities, Technical Specifications involving safety limits, limiting safety system settings, and limiting control system settings; limiting conditions for operation; surveillance requirements; design features; and administrative controls will be developed on a case-by-case basis.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 36 of 38 Quality Assurance Program Description (QAPD) and CP-0500, SPECIAL ACTIONS AND REPORTING REQUIREMENTS The Quality Assurance Program Description (QAPD) and CP-0500 (CP-0500 is the CR-3 procedure that is equivalent to a Technical Requirements Manual) are appropriate candidates for relocations of administrative controls due to the controls imposed by such regulations as 10 CFR 50.59, Appendix B to 10 CFR Part 50, the existing NRC-approved QA plans and commitments to industry QA standards, and the established QA program change control process of 10 CFR 50.54(a). CP-0500 is part of the Updated Safety Analysis Report (USAR) and therefore subject to the requirements of 10 CFR 50.59.

Administrative Letter (AL) 95-06 NRC Administrative Letter (AL) 95-06, "Relocation of Technical Specification Administrative Controls Related to Quality Assurance," (http://www.nrc.gov/reading-rm/doc-collections/gen-comm/admin-letters/1995/al95006.html) provides guidance to licensees requesting amendments that relocate administrative controls to NRC-approved QA program descriptions, where subsequent changes are controlled pursuant to 10 CFR 50.54(a). AL 95-06 provides specific guidance in the areas of: (1) independent safety engineering group, (2) reviews and audits, (3) procedure review process, and (4) records and record retention.

Some relocations are specifically discussed in AL 95-06, while others are similar in nature.

Relocations not specifically discussed in AL 95-06 were assessed with respect to the appropriateness of the relocation. Editorial changes are allowed without basis by 10 CFR 50.54(a)(3).

On December 20, 1993, the NRC issued Amendment No. 149 to Facility Operating License No.

DPR-72 for CR-3. The amendment consisted of changes to the Technical Specifications and the license conditions for CR-3. As stated in the NRC safety evaluation accompanying Amendment 149, the amendment converted the previous custom TS (CTS) to the improved TS (ITS) and relocated certain requirements to other licensee-controlled documents. The CR-3 ITS were based on:

NUREG-1430, "Standard Technical Specifications (STS) Babcock and Wilcox Plants,"

Revision 0; 10 CFR 50.36, "Technical Specifications."

This proposed amendment deletes the portions of the previous CR-3 PDTS that are no longer applicable to a permanently defueled facility with all irradiated fuel in dry storage within an Independent Spent Fuel Storage Installation (ISFSI), while modifying the remaining portions to correspond to the SAFSTOR decommissioning condition, consistent with STS.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 37 of 38 10 CFR 50.51, Continuation of License 10 CFR 50.51(b) states Each license for a facility that has permanently ceased operations, continues in effect beyond the expiration date to authorize ownership and possession of the production or utilization facility, until the Commission notifies the licensee in writing that the license is terminated. During such period of continued effectiveness the licensee shall--

(1)

Take actions necessary to decommission and decontaminate the facility and continue to maintain the facility, including, where applicable, the storage, control and maintenance of the spent fuel, in a safe condition, and (2)

Conduct activities in accordance with all other restrictions applicable to the facility in accordance with the NRC regulations and the provisions of the specific 10 CFR part 50 license for the facility.

Words summarizing the intent of 10 CFR 50.51 is being added to the DEF Facility Operating License (Condition 2.G) to document the reason why the license remains effective after December 3, 2016.

10 CFR 50.82, Termination of License 10 CFR 50.82(a)(2) states Upon docketing of the certifications for permanent cessation of operations and permanent removal of fuel from the reactor vessel, or when a final legally effective order to permanently cease operations has come into effect, the 10 CFR part 50 license no longer authorizes operation of the reactor or emplacement or retention of fuel into the reactor vessel.

5.3 Precedent This proposed amendment is consistent with the license, and accompanying TS, issued to Zion Nuclear Power Station on January 14, 2015 (Reference 3), which was issued to reflect the unloaded spent fuel pool status of the plant.

5.4 Conclusion Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

6.0 ENVIRONMENTAL CONSIDERATION

DEF has evaluated this license amendment against the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. DEF has determined that this license amendment meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9) and 10 CFR 51.22(c)(10)(ii). This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50, that changes a requirement with respect to installation or use of a facility component located within the restricted area, and changes to recordkeeping, reporting, or administrative procedures or requirements.

U.S. Nuclear Regulatory Commission 3F0816-01 Page 38 of 38 However, (i) the proposed amendment involves no significant hazards consideration, (ii) there is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite, and (iii) there is no significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusions set forth in 10 CFR 51.22(c)(9) and 10 CFR 51.22(c)(10)(ii).

Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

7.0 REFERENCES

1. CR-3 to NRC letter, Crystal River Unit 3 - Certification of Permanent Cessation of Power Operations and that Fuel has Been Permanently Removed From the Reactor, dated February 20, 2013. (ADAMS Accession No. ML13056A005

2. CR-3 to NRC letter, Crystal River Unit 3 - Notification of Schedule Change for the Post-Shutdown Decommissioning Activities Report, dated November 16, 2015. (ADAMS Accession No. ML15322A117)

3. License Amendments 188 and 175 and associated NRC safety evaluation issued to Zion Nuclear Power Station Units 1 and 2, respectively, dated January 14, 2015 [ADAMS Accession No. ML14295A716].

4. Letter from AREVA TN to NRC Document Control Desk, Application for Revision to the Standardized NUHOMS System - CoC 1004, Amendment 14, Revision 1, Docket 72-1004, dated November 11, 2015. (ADAMS Accession No. ML15331A350)

5. AREVA to NRC Application for Amendment 14 to Standardized NUHOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks, Revision 0, dated April 16, 2015 (ADAMS Accession No. ML15114A056)

6. AREVA to NRC Application for Amendment 14 to Standardized NUHOMS Certificate of Compliance No. 1004 for Spent Fuel Storage Casks, Revision 1, Response to First Request for Additional Information, dated November 11, 2015 (ADAMS Accession Nos.

ML15114A050, ML15114A051)

7. CR-3 to NRC letter, Crystal River Unit 3 - License Amendment Request #321, Revision 0, ISFSI Only Physical Security Plan, dated May 24, 2016. (ADAMS Accession No. ML16152A045)

DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / 72 - 1035 LICENSE NUMBER DPR - 72 LICENSE AMENDMENT REQUEST #323, REVISION 0, PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS FOR THE INDEPENDENT SPENT FUEL STORAGE INSTALLATION TO REFLECT PERMANENT REMOVAL OF SPENT FUEL FROM THE SPENT FUEL POOLS ENCLOSURE 2 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS - REDLINE AND STRIKEOUT VERSION

Facility Operating License No. DPR-72 Amendment No.

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 DUKE ENERGY FLORIDA, LLC DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

License No. DPR-72

1.

The Nuclear Regulatory Commission (the Commission) having found that:

A.

The application filed by Florida Power Corporation*** (the licensee), as supplemented by letter dated December 9, 1976, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.

Construction of the Crystal River Unit 3 Nuclear Generating Plant (facility) has been substantially completed in conformity with Provisional Construction Permit No.

CPPR-51 and the application, as amended, the provisions of the Act and the rules and regulations of the Commission; C.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; D.

There is reasonable assurance: (i) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; E.

The licensee is financially qualified and Duke Energy Florida, LLC is technically qualified and financially qualified to engage in the activities authorized by this operating license in accordance with the rules and regulations of the Commission; F.

The licensee has satisfied the applicable provisions of 10 CFR Part 140, Financial Protection Requirements and Indemnity Agreements, of the Commission's regulations;

- - On April 29, 2013, the name "Florida Power Corporation" was changed to "Duke Energy Florida, Inc." On August 1, 2015, Duke Energy Florida, Inc. converted to a limited liability company and the name was changed to "Duke Energy Florida, LLC."

Facility Operating License No. DPR-72 Amendment No.

G.

The issuance of this operating license will not be inimical to the common defense and security or to the health and safety of the public; H.

After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of Facility Operating License No. DPR-72 subject to the conditions for protection of the environment set forth herein is in accordance with 10 CFR Part 51, (formerly Appendix D to 10 CFR Part 50), of the Commission's regulations and all applicable requirements have been satisfied; I.

The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30, 40 and 70, including 10 CFR Sections 30.33, 40.32 and 70.23 and 70.31.

2. Facility Operating License No. DPR-72, issued to the licensee, is hereby amended in its entirety to read as follows:

A.

This amended license applies to the Crystal River Unit 3 Nuclear Generating Plant, a pressurized water nuclear reactor and associated equipment (the facility), owned by the licensee and operated by Duke Energy Florida, LLC. The facility is located on the Gulf of Mexico, about seven and one-half miles northwest of the town of Crystal River, Citrus County, Florida, and is described in the Final Safety Analysis Report as supplemented and amended and the Environmental Report as supplemented and amended.

B.

Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

(1)

Duke Energy Florida, LLC, pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess and use the facility; (2)

The licensee to possess the facility at the designated location in Citrus County, Florida, in accordance with the procedures and limitations set forth in this license; (3)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear material configured as reactor fuel, in accordance with the limitations for storage as described in the Final Safety Analysis Report, as supplemented and amended; (4)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to possess at any time any byproduct, source and special nuclear material as sealed neutron sources used previously for reactor startup, as fission detectors, and sealed sources for reactor instrumentation and to possess and use at any time any byproduct, source, and special nuclear material as sealed sources for radiation monitoring equipment calibration in amounts as required; Revised page submitted 2-24-77

Facility Operating License No. DPR-72 Amendment No.

(5)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; (6)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (7)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to receive and possess, but not separate, that byproduct and special nuclear materials associated with four (4) fuel assemblies (B&W Identification Numbers 1A-01, 04, 05 and 36 which were previously irradiated in the Oconee Nuclear Station, Unit No. 1) acquired by Florida Power Corporation*** from Duke Power Company for use as reactor fuel in the facility.

C.

This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Deleted per Amendment No. 247 (2)

Technical Specifications The Technical Specifications contained in Appendix A are hereby replaced with the Permanently Defueled Technical Specifications (PDTS). Duke Energy Florida, LLC shall maintain the facility in accordance with the Permanently Defueled Technical Specifications, as revised through Amendment No. 249.

(3)

Deleted per Amendment No. 247 (4)

Deleted per Amendment No. 20 dated 7-3-79 (5)

Deleted per Amendment No. 247 (6)

Deleted per Amendment No. 21, 7-3-79 (7)

Deleted per Amendment No. 247 (8)

Deleted per Amendment No. 247 (9)

Deleted per Amendment No. 247 (10) Deleted per Amendment No. 247

- - On April 29, 2013, the name "Florida Power Corporation" was changed to "Duke Energy Florida, Inc." On August 1, 2015, Duke Energy Florida, Inc. converted to a limited liability company and the name was changed to "Duke Energy Florida, LLC."

Added per Amdt. 15, 7-24-78

Facility Operating License No. DPR-72 Amendment No.

(11) Deleted per Amendment No. 247 (12) Deleted per Amendment No. 237 (13) Deleted per Amendment No. 229 (14) Deleted per Amendment No. Mitigation Strategy License Condition The licensee shall develop and maintain strategies for addressing large fires and explosions that include the following key areas:

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

a.

Pre-defined coordinated fire response strategy and guidance

b.

Assessment of mutual aid fire fighting assets

c.

Designated staging areas for equipment and materials

d.

Command and control

e.

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

a.

Protection and use of personnel assets

b.

Communications

c.

Minimizing fire spread

d.

Procedures for implementing integrated fire response strategy

e.

Identification of readily-available pre-staged equipment

f.

Training on integrated fire response strategy

g.

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

a.

Water spray scrubbing

b.

Dose to onsite responders (15) Deleted per Amendment No. 247 D.

Physical Security The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The combined plan, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "ISFSI Physical Security Plan, ISFSI Safeguards Contingency Plan and ISFSI Guard Training and Qualification Plan, Revision 0, submitted by letter dated May 24, 2016.

Facility Operating License No. DPR-72 Amendment No.

E.

Deleted per Amendment No. 247 F.

In accordance with the requirement imposed by the October 8, 1976, order of the United States Court Appeals for the District of Columbia Circuit in Natural Resources Defense Council v. Nuclear Regulatory Commission, No. 74-1385 and 74-1586, that the Nuclear Regulatory Commission "shall make any licenses granted between July 21, 1976 and such time when the mandate is issued subject to the outcome of the proceedings herein," the license issued herein shall be subject to the outcome of such proceedings.

G.

This amended license is effective as of the date of issuance. Facility Operating License No. DPR-72, as amended, shall expire at midnight, December 3, 2016.

Duke Energy Florida, LLC submitted the 10 CFR 50.82(a)(1) notification to the Nuclear Regulatory Commission on February 20, 2013. Per 10 CFR 50.51(b), the Facility Operating License No DPR-72 continues in effect until the Commission notifies the licensee that the License has been terminated.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by Roger S. Boyd, Director Division of Project Management Office of Nuclear Reactor Regulation Attachments:

Appendices A & B - Technical Specifications Date of Issuance: Jan 28 1977 Amdt. #97 March 31, 1987

CRYSTAL RIVER UNIT 3 PDTS LOEPage 1 of 1 12/16/15 PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS List of Effective Pages (Through Amendment 249 and ITS Bases Revision 89)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189, 190, 226, 235, 238, 239, 242, 243, 245 and 246 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page Amendment Page Amendment i

247 ii 247 1.1-1 247 1.2-1 149 1.2-2 149 1.2-3 149 1.3-1 247 1.4-1 247 1.4-2 247 3.0-1 247 3.0-2 247 3.7-1 247 3.7-2 247 3.7-3 247 3.7-4 247 3.7-5 247 3.7-6 247 3.7-7 247 4.0-1 247 4.0-2 247 4.0-3 247 5.0-1 249 5.0-2 249 5.0-3 244 5.0-4 249 5.0-5 149 5.0-6 149 5.0-7 244 5.0-8 149 5.0-9 249 5.0-10 244 5.0-11 247 5.0-12 247 5.0-13 247 5.0-14 247 5.0-15 247 Appendix B - Part II 1-1 190 2-1 190 3-1 190 3-2 190 4-1 190 4-2 190

Crystal River Unit 3 i

Amendment No. 247 TABLE OF CONTENTS 1.0 USE AND APPLICATION.................................... 1.1-1 1.1 Definitions......................................... 1.1-1 1.2 Logical Connectors.................................. 1.2-1 1.3 Completion Times.................................... 1.3-1 1.4 Frequency........................................... 1.4-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY............ 3.0-2 3.7 PLANT SYSTEMS....................................... 3.7-1 3.7.13 Fuel Storage Pool Water Level................... 3.7-1 3.7.14 Spent Fuel Pool Boron Concentration............. 3.7-2 3.7.15 Spent Fuel Assembly Storage..................... 3.7-4 4.0 DESIGN FEATURES........................................ 4.0-1 5.0 ADMINISTRATIVE CONTROLS................................ 5.0-1

Crystal River Unit 3 ii Amendment No. 247 TABLE OF CONTENTS B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY... B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY............ B 3.0-2 B 3.7 PLANT SYSTEMS....................................... B 3.7-1 B 3.7-13 Fuel Storage Pool Water Level....................... B 3.7-1 B 3.7.14 Spent Fuel Pool Boron Concentration................. B 3.7-4 B 3.7-15 Spent Fuel Assembly Storage......................... B 3.7-7

Definitions 1.1 Crystal River Unit 3 1.1-1 Amendment No. 247 1.0 USE AND APPLICATION 1.1 Definitions

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

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

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

Logical Connectors 1.2 (continued)

Crystal River Unit 3 1.2-1 Amendment No. 149 1.0 USE AND APPLICATION 1.2 Logical Connectors PURPOSE The purpose of this section is to explain the meaning of logical connectors.

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

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

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

Logical Connectors 1.2 (continued)

Crystal River Unit 3 1.2-2 Amendment No. 149 1.2 Logical Connectors (continued)

EXAMPLES The following examples illustrate the use of logical connectors.

EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

LCO not met.

A.1 Verify...

AND A.2 Restore...

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

Logical Connectors 1.2 Crystal River Unit 3 1.2-3 Amendment No. 149 1.2 Logical Connectors EXAMPLES EXAMPLE 1.2-2 (continued)

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. LCO not met.

A.1 Trip...

OR A.2.1 Verify...

AND A.2.2.1 Reduce...

OR A.2.2.2 Perform...

OR A.3 Align...

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

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

Completion Times 1.3 Crystal River Unit 3 1.3-1 Amendment No. 247 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use.

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

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

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

Frequency 1.4 (continued)

Crystal River Unit 3 1.4-1 Amendment No. 247 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

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

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

Frequency 1.4 Crystal River Unit 3 1.4-2 Amendment No. 247 1.4 Frequency EXAMPLES The following example illustrates the type of frequency statement that appears in the Permanently Defueled Technical Specifications (PDTS).

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform (activity).

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

LCO Applicability 3.0 Crystal River Unit 3 3.0-1 Amendment No. 247 3.0 LIMITING CONDITION FOR OPERATION (LC0) APPLICABILITY LCO 3.0.1 LCOs shall be met during specified conditions in the Applicability, except as provided in LCO 3.0.2.

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

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

SR Applicability 3.0 Crystal River Unit 3 3.0-2 Amendment No. 247 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the specified conditions in the Applicability for individual Specifications, unless otherwise stated in the SR. Failure to meet a Surveillance, whether such failure is experienced during the performance of the Surveillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Frequency shall be failure to meet the LCO except as provided in SR 3.0.3.

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

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

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

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

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

Fuel Storage Pool Water Level 3.7.13 Crystal River Unit 3 3.7-1 Amendment No. 247 3.7 PLANT SYSTEMS 3.7.13 Fuel Storage Pool Water Level LCO 3.7.13 The fuel storage pool water level shall be 156 ft Plant Datum.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fuel storage pool water level not within limit.

A.1 Suspend movement of irradiated fuel assemblies in fuel storage pool.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify the fuel storage pool water level is 156 ft Plant Datum.

7 days

Spent Fuel Pool Boron Concentration 3.7.14 Crystal River Unit 3 3.7-2 Amendment No. 247 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Boron Concentration LCO. 3.7.14 The spent fuel pool boron concentration shall be 1925 ppm.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool boron concentration not within limit.

A.1 Suspend movement of fuel assemblies in the spent fuel pool.

AND A.2.1 Initiate action to restore spent fuel pool boron concentration to within limit.

OR A.2.2 Verify by administrative means a Storage Pool A and Storage Pool B spent fuel pool verification has been performed since the last movement of fuel assemblies in the spent fuel pool.

Immediately Immediately Immediately

Spent Fuel Pool Boron Concentration 3.7.14 Crystal River Unit 3 3.7-3 Amendment No. 247 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pool boron concentration is 1925 ppm.

7 days

Spent Fuel Assembly Storage 3.7.15 Crystal River Unit 3 3.7-4 Amendment No. 247 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Assembly Storage LCO 3.7.15 The combination of initial enrichment and burnup of each spent fuel assembly stored in Storage Pool A and Storage Pool B, shall be within the acceptable region of Figure 3.7.15-1 or Figure 3.7.15-2.

APPLICABILITY: Whenever any fuel assembly is stored in Storage Pool A or Storage Pool B of the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Requirements of the LCO not met.

A.1 Initiate action to move the noncomplying fuel assembly to an acceptable configuration.

Immediately

Spent Fuel Assembly Storage 3.7.15 Crystal River Unit 3 3.7-5 Amendment No. 247 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the initial enrichment and burnup of the fuel assembly is in accordance with Figure 3.7.15-1 or Figure 3.7.15-2.

Prior to storing the fuel assembly in Storage Pool A or Storage Pool B.

Spent Fuel Assembly Storage 3.7.15 Crystal River Unit 3 3.7-6 Amendment No. 247

1.

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

2.

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

3.

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

Figure 3.7.15-1 Burnup versus Enrichment Curve for Spent Fuel Storage Pool A 0

5 10 15 20 25 30 35 40 45 2

2.5 3

3.5 4

4.5 5

Initial Enrichment, Weight Percent U235 Burnup, MWD / Kg U Category F (beneath curve)

Category B (above curve)

Category A (between curves)

Spent Fuel Assembly Storage 3.7.15 Crystal River Unit 3 3.7-7 Amendment No. 247

1.

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

2.

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

3.

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

4.

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

Figure 3.7.15-2 Burnup versus Enrichment Curve for Spent Fuel Storage Pool B 0

5 10 15 20 25 30 35 40 45 2

2.5 3

3.5 4

4.5 5

Initial Enrichment, Weight Percent U235 Burnup, MWD / Kg U Category BE (below curves)

Category B (above curves)

Category BP (between curves)

Design Features 4.0 (continued)

Crystal River Unit 3 4.0-1 Amendment No. 247 4.0 DESIGN FEATURES 4.1 Site The 4,738 acre site is characterized by a 4,400 foot minimum exclusion radius centered on the Reactor Building; isolation from nearby population centers; sound foundation for structures; an abundant supply of cooling water; an ample supply of power; and favorable conditions of hydrology, geology, seismology, and meteorology.

4.2 Not Used 4.3 Fuel Storage Spent fuel shall not be stored in the spent fuel pool.

Design Features 4.0 (continued)

Crystal River Unit 3 4.0-2 Amendment No. 247 4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

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

b.

keff 0.95 if fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.6 of the FSAR;

c.

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

d.

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

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

a.

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

b.

keff 0.95 is fully flooded with unborated water, which includes an allowance for uncertainties as described in Section 9.6 of the FSAR;

c.

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

d.

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

Design Features 4.0 Crystal River Unit 3 4.0-3 Amendment No. 247 DESIGN FEATURES 4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 138 feet 4 inches.

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

Responsibility 5.1 Crystal River Unit 3 5.0-1 Amendment No. 249 5.0 ADMINISTRATIVE CONTROLS 5.1 Responsibility 5.1.1 The General Manager Decommissioning shall be responsible for overall facility functions and shall delegate in writing the succession to this responsibility during his absence.

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

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

Organization 5.2 (continued)

Crystal River Unit 3 5.0-2 Amendment No. 249 5.0 ADMINISTRATIVE CONTROLS 5.2 Organization 5.2.1 Onsite and Offsite Organizations Onsite and offsite organizations shall be established for facility staff and corporate management, respectively. The onsite and offsite organizations shall include the positions responsible for activities affecting the safe handling and storage of nuclear fuel.

a.

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

b.

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

c.

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

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

a.

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

b.

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

Organization 5.2 Crystal River Unit 3 5.0-3 Amendment No. 244 5.2 Organization 5.2.2 Unit Staff (continued)

c.

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

d.

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

e.

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

f.

The Shift Supervisor shall be a Certified Fuel Handler.

Unit Staff Qualifications 5.3 Crystal River Unit 3 5.0-4 Amendment No. 249 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 General Manager Decommissioning.

Not Used 5.4 Crystal River Unit 3 5.0-5 Amendment No. 149 5.0 ADMINISTRATIVE CONTROLS 5.4 Not Used

Not Used 5.5 Crystal River Unit 3 5.0-6 Amendment No. 149 5.0 ADMINISTRATIVE CONTROLS 5.5 Not Used

Procedures, Programs, and Manuals 5.6 (continued)

Crystal River Unit 3 5.0-7 Amendment No. 244 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:

Procedures, Programs and Manuals 5.6 (continued)

Crystal River Unit 3 5.0-8 Amendment No. 149 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

Procedures, Programs and Manuals 5.6 Crystal River Unit 3 5.0-9 Amendment No. 249 5.6 Procedures, Programs and Manuals 5.6.2.3 ODCM (continued)

2.

For tritium and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrems/yr to any organ;

h.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I;

i.

Limitations on the annual and quarterly doses to a member of the public from tritium and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50, Appendix I; and

j.

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

Licensee Initiated Changes to the ODCM:

1.

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

a.

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

b.

A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR Part 190, 10 CFR 50.36a, and Appendix I to 10 CFR Part 50 and not adversely impact the accuracy or reliability of effluent dose, or setpoint calculations.

2.

Shall become effective after review and acceptance by the on-site review function and the approval of the General Manager Decommissioning; and

Procedures, Programs and Manuals 5.6 (continued)

Crystal River Unit 3 5.0-10 Amendment No. 244 5.6 Procedures, Programs and Manuals 5.6.2.3 ODCM (continued)

3.

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

5.6.2.4 Not Used 5.6.2.5 Not Used 5.6.2.6 Not Used 5.6.2.7 Not Used 5.6.2.8 Not Used 5.6.2.9 Not Used 5.6.2.10 Not Used 5.6.2.11 Not Used

Procedures, Programs and Manuals 5.6 Crystal River Unit 3 5.0-11 Amendment No. 247 5.6 Procedures, Programs and Manuals 5.6.2.12 Not Used 5.6.2.13 Not Used 5.6.2.14 Not Used 5.6.2.15 Not Used 5.6.2.16 Not Used 5.6.2.17 Technical Specifications (TS) Bases Control Program Changes to the Bases of the TS shall be made under appropriate administrative controls and reviewed according to the review process specified in the Quality Assurance Plan.

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

a.

A change in the TS incorporated in the license; or

b.

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

The Bases Control Program shall contain 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 5.6.2.21 Not Used

Reporting Requirements 5.7 (continued)

Crystal River Unit 3 5.0-12 Amendment No. 247 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.

Reporting Requirements 5.7 Crystal River Unit 3 5.0-13 Amendment No. 247 5.7 Reporting Requirements 5.7.1.2 Not Used 5.7.2 Not Used

High Radiation Area 5.8 (continued)

Crystal River Unit 3 5.0-14 Amendment No. 247 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 from the radiation source or from any surface penetrated by the radiation but less than 500 rads/hr at 1 meter from the radiation source or from any surface penetrated by the radiation shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor or health physics supervision. Doors shall remain locked except during periods of access by personnel.

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.

High Radiation Area 5.8 Crystal River Unit 3 5.0-15 Amendment No. 247 5.8 High Radiation Area (continued) 5.8.3 For individual high radiation areas with radiation levels of

> 1000 mrem/hr at 30 cm from the radiation source or from any surface penetrated by the radiation but less than 500 rads/hr at 1 meter from the radiation source or from any surface penetrated by the radiation, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that are not 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 ITS Bases LOEPage 1 of 1 10/1/15 IMPROVED TECHNICAL SPECIFICATION BASES List of Effective Pages (Through Amendment 247 and ITS Bases Revision 89)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189, 190, 226, 235, 238, 239, 242, 243, 245 and 246 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page Revision B 3.0-1 89 B 3.0-2 89 B 3.0-3 89 B 3.0-4 89 B 3.7-1 89 B 3.7-2 89 B 3.7-3 89 B 3.7-4 89 B 3.7-5 89 B 3.7-6 89 B 3.7-7 89 B 3.7-8 89 B 3.7-9 89 B 3.7-10 89 B 3.7-11 89 B 3.7-12 89

LCO Applicability B 3.0 Crystal River Unit 3 B 3.0-1 Revision No. 89 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCO 3.0.1 and 3.0.2 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

LCO 3.0.1 LCO 3.0.1 establishes the Applicability statement within each individual Specification as the requirement for when the LCO is required to be met (i.e., when the unit is in the specified conditions of the Applicability statement of each Specification).

LCO 3.0.2 LCO 3.0.2 establishes that upon discovery of a failure to meet an LCO, the associated ACTIONS shall be met. The Completion Time of each Required Action for an ACTIONS Condition is applicable from the point in time that an ACTIONS Condition is entered. The Required Actions establish those remedial measures that must be taken within specified Completion Times when the requirements of an LCO are not met. This Specification establishes that completion of the Required Actions within the specified Completion Times constitutes compliance with a Specification.

The Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore variables to within specified limits. Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS.

Completing the Required Actions is not required when an LCO is met or is no longer applicable within the associated Completion Time, unless otherwise stated in the individual Specifications.

SR Applicability B 3.0 (continued)

Crystal River Unit 3 B 3.0-2 Revision No. 89 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY BASES SR 3.0.1 through SR 3.0.4 establish the general requirements applicable to all Specifications and apply at all times, unless otherwise stated.

SR 3.0.1 SR 3.0.1 establishes the requirement that SRs must be met during specified conditions in the Applicability for which the requirements of the LCO apply, unless otherwise specified in the individual SRs. This Specification is to ensure that Surveillances are performed to verify that variables are within specified limits. Failure to meet a Surveillance within the specified Frequency, in accordance with SR 3.0.2, constitutes a failure to meet an LCO.

Variables are assumed to be within limits when the associated SRs have been met. Nothing in this Specification, however, is to be construed as implying that variables are within limits when the requirements of the Surveillance(s) are known not to be met between required Surveillance performances.

Surveillances do not have to be performed when the specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified.

SR 3.0.2 SR 3.0.2 establishes the requirements for meeting the specified Frequency for Surveillances and any Required Action with a Completion Time that requires the periodic performance of the Required Action on a "once per..."

interval.

SR 3.0.2 permits a 25% extension of the interval specified in the Frequency. This extension facilitates Surveillance scheduling and considers facility conditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities).

SR Applicability B 3.0 (continued)

Crystal River Unit 3 B 3.0-3 Revision No. 89 BASES SR 3.0.2 The 25% extension does not significantly degrade the (continued) reliability that results from performing the Surveillance at its specified Frequency. This is based on the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs.

SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring an affected variable outside the specified limits when a Surveillance has not been completed within the specified Frequency. A delay period of up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SR 3.0.2, and not at the time that the specified Frequency was not met.

This delay period provides an adequate time limit to complete Surveillances that have been missed. This delay period permits the completion of a Surveillance before complying with Required Actions or other remedial measures that might preclude completion of the Surveillance. The basis for this delay period includes consideration of facility conditions, adequate planning, availability of personnel, the time required to perform the Surveillance, the safety significance of the delay in completing the required Surveillance, and the recognition that the most probable result of any particular Surveillance being performed is the verification of conformance with the SRs.

Failure to comply with specified Frequencies for SRs is expected to be an infrequent occurrence. Use of the delay period established by SR 3.0.3 is a flexibility which is not intended to be used as an operational convenience to extend Surveillance intervals. While up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or the limit of the specified Frequency is provided to perform the missed Surveillance, it is expected that the missed Surveillance will be performed at the first reasonable opportunity.

SR Applicability B 3.0 Crystal River Unit 3 B 3.0-4 Revision No. 89 BASES SR 3.0.3 If a Surveillance is not completed within the allowed delay (continued) period, then the variable is considered outside the specified limits and the Completion Times of the Required Actions for the applicable Specification Conditions begin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the variable is outside the specified limits and the Completion Times of the Required Actions for the applicable Specification Conditions begin immediately upon the failure of the Surveillance.

Completion of the Surveillance within the delay period allowed by this Specification, or within the Completion Time of the ACTIONS, restores compliance with SR 3.0.1.

SR 3.0.4 SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a specified condition in the Applicability.

This Specification ensures that variable limits are met before entry into specified conditions in the Applicability for which these ensure safe storage of nuclear fuel. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring the specified condition in the Applicability.

The provisions of SR 3.0.4 shall not prevent entry into specified conditions in the Applicability that are required to comply with ACTIONS.

The precise requirements for performance of SRs are specified such that exceptions to SR 3.0.4 are not necessary. The specific time frames and conditions necessary for meeting the SRs are specified in the Frequency, in the Surveillance, or both. This allows performance of Surveillances when the prerequisite condition(s) specified in a Surveillance procedure require entry into the specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance.

Fuel Storage Pool Water Level B 3.7.13 (continued)

Crystal River Unit 3 B 3.7-1 Revision No. 89 B 3.7 PLANT SYSTEMS B 3.7.13 Fuel Storage Pool Water BASES BACKGROUND The water contained in the spent fuel pool provides a medium for removal of decay heat from the stored fuel elements, normally via the spent fuel cooling system. The spent fuel pool water also provides shielding to reduce the general area radiation dose during both spent fuel handling and storage.

Although maintaining adequate spent fuel pool water level is essential to both decay heat removal and shielding effectiveness, the Technical Specification minimum water level limit is based upon maintaining the pool's iodine retention effectiveness consistent with that assumed in the evaluation of a fuel handling accident (FHA). The fuel handling accident described in FSAR Section 14.2.2.3 (Ref. 2), assumes that a minimum of 23 feet of water is maintained above the stored fuel. This assumption allows the use of the pool iodine decontamination factor used in the associated offsite dose calculation.

APPLICABLE The minimum water level in the fuel storage pool meets the SAFETY ANALYSES assumptions of the FHA described in FSAR Section 14.2.2.3.

The resultant 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months dose to a person at the exclusion area boundary and the 30 day dose at the low population zone are much less than 10 CFR 50.67 (Ref. 4) limits.

Although the water level above a damaged assembly lying on top of the fuel storage racks may be less than 23 feet, an extrapolation of the iodine removal efficiency factors indicates that the iodine removal factor used in the dose calculations will still be conservative at water levels as low as 21 feet (Ref. 5). The 23 foot criteria above the fuel in the racks will ensure at least 21 feet above the damaged assembly.

Fuel Storage Pool Water Level B 3.7.13 (continued)

Crystal River Unit 3 B 3.7-2 Revision No. 89 BASES APPLICABLE Fuel storage pool water level satisfies Criterion 2 of the SAFETY ANALYSES NRC Policy Statement.

(continued)

LCO The specified water level of 23 feet over the top of the irradiated fuel assemblies seated in the storage racks (156 ft plant datum) preserves the assumptions of the FHA analysis (Ref. 2). As such, it is the minimum level allowed during movement of fuel within the fuel storage pool.

APPLICABILITY This LCO is only applicable during movement of irradiated fuel assemblies in the fuel storage pool. This is consistent with the safety analysis which assumes the FHA initiating event to be the drop of an irradiated fuel assembly. Control of heavy loads, i.e., damaging the fuel assembly as a result of dropping a heavy load onto it, is not addressed by the safety analysis or this Technical Specification. Plant procedures are relied upon to prevent the dropping of heavy loads onto spent fuel.

ACTIONS A.1 With the fuel storage pool level less than the minimum required level, the movement of fuel assemblies in the fuel storage pool is immediately suspended. This effectively precludes the occurrence of a fuel handling accident.

Fuel Storage Pool Water Level B 3.7.13 Crystal River Unit 3 B 3.7-3 Revision No. 89 BASES SURVEILLANCE SR 3.7.13.1 REQUIREMENTS The water level in the fuel storage pool must be checked periodically. Since there is no mechanism for inadvertently lowering the level during normal operations (changes in level are procedurally controlled) and there is a low level alarm should pool level drop to approximately 24 feet above the stored fuel assemblies, a 7 day Frequency is sufficient to provide assurance of adequate water level. The Frequency is based on engineering judgment and industry-accepted practice.

REFERENCES

1.

Deleted.

2.

FSAR, Section 14.2.2.3.

3.

Deleted.

4.

10 CFR 50.67.

5.

FPC Calculation N-00-0001.

Spent Fuel Pool Boron Concentration B 3.7.14 (continued)

Crystal River Unit 3 B 3.7-4 Revision No. 89 B 3.7 PLANT SYSTEMS B 3.7.14 Spent Fuel Pool Boron Concentration BASES BACKGROUND As described in the Bases for LCO 3.7.15, "Spent Fuel Assembly Storage," fuel assemblies are stored in the high density region of the spent fuel pool storage racks in accordance with criteria based on initial weight-percent enrichment and discharge burnup. Although the water in the spent fuel pool is normally borated to 2000 ppm, the criteria that limit the storage of a fuel assembly to specific rack locations (criticality analysis) are conservatively developed without taking credit for the boron in the pool water.

APPLICABLE The acceptance criteria for the fuel storage pool SAFETY ANALYSIS criticality analyses is that a keff of < 0.95 must be maintained for all postulated events. The storage racks are capable of maintaining this keff with unborated pool water at a temperature yielding the highest reactivity (assuming the storage restrictions of LCO 3.7.15 are met). Most abnormal storage locations will not result in an increase in the keff of the racks. However, it is possible to postulate events, such as the mis-loading of an assembly with a burnup and enrichment combination outside the acceptable area in Figure 3.7.15-1 and 3.7.15-2, or dropping an assembly between the pool wall and the fuel racks, which could lead to an increase in reactivity. For such events, credit is taken for the presence of boron in the pool water since the NRC does not require the assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accident (double contingency principle). The reduction in keff, caused by the boron more than offsets the reactivity addition caused by credible accidents.

The concentration of dissolved boron in the fuel storage pool satisfies Criterion 2 of the NRC Policy Statement.

Spent Fuel Pool Boron Concentration B 3.7.14 (continued)

Crystal River Unit 3 B 3.7-5 Revision No. 89 BASES LCO The required concentration of dissolved boron in the fuel storage pool of 1925 ppm preserves the assumption used in the analyses of the potential accident scenarios described above. This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the fuel storage pool.

APPLICABILITY This LCO is applicable whenever fuel assemblies are stored in the spent fuel pool, until a complete spent fuel pool verification has been performed following the last movement of fuel assemblies in the spent fuel pool. This LCO does not apply following the verification since the verification would confirm that there are no misloaded fuel assemblies.

With no further fuel assembly movement in progress, there is no potential for a misloaded fuel assembly or a dropped fuel assembly and the reactivity of the racks alone is adequate to preserve assumptions of the criticality analysis.

ACTIONS A.1, A.2.1, and A.2.2 When the concentration of boron in the fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident. This is most efficiently achieved by immediately suspending the movement of fuel assemblies within the pool. This Action does not preclude movement of a fuel assembly to a safe position.

Additionally, action must be initiated immediately to restore pool boron concentration to within the LCO limit or a pool verification performed. Either of these Actions will restore compliance with the LCO or demonstrate the need for the LCO does not currently exist.

SURVEILLANCE SR 3.7.14.1 REQUIREMENTS This SR verifies that the concentration of boron in the fuel storage pool is within the required limit. This is accomplished by sampling representative samples of the pool.

Spent Fuel Pool Boron Concentration B 3.7.14 Crystal River Unit 3 B 3.7-6 Revision No. 89 BASES SURVEILLANCE SR 3.7.14.1 (continued)

REQUIREMENTS Operating experience has shown significant differences between boron measured near the top of the pool and that measured elsewhere. As long as this SR is met, the analyzed events are fully bounded. The 7 day Frequency is acceptable because no major replenishment of pool water is expected to take place over this period of time.

REFERENCES

1.

Criticality Safety Evaluation of the Pool A Spent Fuel Storage Racks in Crystal River Unit 3 With Fuel of 5.0% Enrichment, S. E. Turner, Holtec Report HI-931111, December 1993.

2.

Criticality Safety Analysis of the Westinghouse Spent Fuel Storage Racks in Pool B of Crystal River Unit 3, S. E. Turner, Holtec Report HI-992128, May 1999.

3.

Criticality Safety Analysis of the Crystal River Unit 3 Pool A for Storage of 5% Enriched Mark B-11 Fuel in Checkerboard Arrangement with Water Holes, Holtec Report HI-992285, August 1999.

4.

Criticality Evaluation of CR3 Spent Fuel Pool Storage Racks with Mark B-12 Fuel, Holtec Report HI-2022907, September 2002.

5.

Progress Energy Engineering Change EC No. 52456, Documentation of Acceptability to Receive and Store Mk-B/HTP Fuel.

6.

Criticality Analysis of Additional Patterns for Crystal River 3 Pools A and B, Holtec Report HI-2063559, September 2006.

Spent Fuel Assembly Storage B 3.7.15 (continued)

Crystal River Unit 3 B 3.7-7 Revision No. 89 B 3.7 PLANT SYSTEMS B 3.7.15 Spent Fuel Assembly Storage BASES BACKGROUND This document describes the Bases for the Spent Fuel Assembly Storage which imposes storage requirements upon irradiated and unirradiated fuel assemblies stored in the fuel storage pools containing high density racks. The storage areas, which are part of the Spent Fuel System, governed by this Specification are:

a.

Fuel storage pool "A" and

b.

Fuel storage pool "B".

In general, the function of the storage racks is to support and protect new and spent fuel from the time it is placed in the storage area until it is shipped offsite.

Spent fuel is stored underwater in either fuel storage pool A or B. Only fuel pool A has the capability to store failed fuel in containers. Spent fuel pool A features high density poison storage racks with a 10 1/2 inch center-to-center distance capable of storing 542 assemblies. Fuel pool A is capable of storing fuel with enrichments up to 5.0 weight percent U-235 (Ref. 1, 6, 7, 8 and 9) without exceeding the criticality criteria of Reference 3 providing the fuel has sufficient burnup and required storage configuration.

Spent fuel pool B also contains high density racks having a 9.11 inch center-to-center distance capable of storing 932 assemblies. Fuel pool B is capable of storing fuel with enrichments up to 5.0 weight percent U-235 (Ref. 2, 7, 8 and

9) without exceeding the criticality criteria of Reference 3, providing the fuel has sufficient burnup and required storage configuration. New and low burnup fuel may be placed into pool B if surrounded by empty storage cells.

This is primarily for, but not restricted to, fuel inspection and reconstitution activities (Ref. 9).

Spent Fuel Assembly Storage B 3.7.15 (continued)

Crystal River Unit 3 B 3.7-8 Revision No. 89 BASES BACKGROUND Both of the spent fuel pools are constructed of reinforced (continued) concrete and lined with stainless steel plate. They are located in the fuel handling area of the auxiliary building.

New fuel storage requirements are addressed in Section 4.0, "Design Features".

APPLICABLE The function of the spent fuel storage racks is to support SAFETY ANALYSES and protect spent fuel assemblies from the time they are placed in the pool until they are shipped offsite. The spent fuel assembly storage LCO was derived from the need to establish limiting conditions on fuel storage to assure sufficient safety margin exists to prevent inadvertent criticality. The spent fuel assemblies are stored entirely underwater in a configuration that has been shown to result in a reactivity of less than or equal to 0.95 under worse case conditions (Ref. 1, 2, 6, 7, 8 and 9). The spent fuel assembly enrichment requirements in this LCO are required to ensure inadvertent criticality does not occur in the spent fuel pool.

Inadvertent criticality within the fuel storage area could result in offsite radiation doses exceeding 10 CFR 50.67 limits.

The spent fuel assembly storage satisfies Criterion 2 of the NRC Policy Statement.

LCO Limits on the new and irradiated fuel assembly storage in high density racks were established to ensure the assumptions of the criticality safety analysis of the spent fuel pools is maintained.

Limits on initial fuel enrichment and burnup for both new and for spent fuel stored in pool A have been established.

Two limits are defined:

1.

Initial fuel enrichment must be less than or equal to 5.0 weight percent U-235, and

Spent Fuel Assembly Storage B 3.7.15 (continued)

Crystal River Unit 3 B 3.7-9 Revision No. 89 BASES LCO

2.

For new, low irradiation, and spent fuel with initial (continued) enrichment less than or equal to 5.0 weight percent and greater than or equal to 3.5 weight percent, fuel burnup must be within the limits specified in Figure 3.7.15-1.

Figure 3.7.15-1 presents three areas of required fuel assembly burnup as a function of initial enrichment.

a. Category B: Fuel with enrichment-burnup combinations in the area above the upper curve can be stored with no restrictions except as noted below. That is, this fuel can be stored next to fuel with enrichment-burnups that fall into Categories A, B or F provided there are no restrictions on that fuel type preventing it.

Category B has the same burnup-enrichment requirements for Pools A and B.

b. Category A: Fuel with enrichment-burnup between the curves can be stored in any configuration with fuel above the lower curve. That is, this fuel may be stored next to fuel with enrichment-burnups that fall into Categories A or B.

c. Category F: Fuel with enrichment-burnup combinations below the lower curve must be stored in a one-out-of-two checkerboard configuration with fuel that has enrichment-burnup combinations above the upper curve (Category B) or with empty watercells that contain no fuel. Areas of Category F fuel stored in the checkerboard combination with Category B fuel or empty water cells must be separated from areas of Category A fuel by a transition row of Category B cells. The acceptability of storing this fuel in the checkerboard configuration is documented in References 6, 7, 8 and 9.

Fuel enrichment limits are based on avoiding inadvertent criticality in the spent fuel pool. The CR-3 spent fuel storage system was initially designed to a maximum enrichment of 3.5 weight percent. Enrichments of up to 5.0 weight percent are permissible for storage in spent fuel pool A as long as the fuel burnup is sufficient to limit the worst case reactivity in the storage pool to less than or equal to 0.95.

Fuel burnup reduces the reactivity of the fuel due to the accumulation of fission product poisons. Reference 1

Spent Fuel Assembly Storage B 3.7.15 (continued)

Crystal River Unit 3 B 3.7-10 Revision No. 89 BASES LCO documents that the required burnup varies linearly as a (continued) function of enrichment with 10500 megawatt days per metric ton uranium (Mwd/mtU) required for fuel with 5.0 weight percent enrichment and 0 burnup required for 3.5 weight percent enriched fuel.

Similar types of restrictions have been established for Pool B.

1.

Initial fuel enrichment must be 5.0 weight percent U-235, and

2.

For fuel with initial enrichment 5.0 weight percent and 2.0 weight percent, fuel burnup must be within the limits specified in Figure 3.7.15-2.

a. Category B: Fuel with burnup-enrichment combinations in the area above the upper curve can be stored with no restrictions except as noted below. That is, this fuel can be stored next to fuel with burnup-enrichments that fall into Categories B or BP. Category B has the same burnup-enrichment requirements for Pools A and B.

b. Category BP: Fuel with burnup-enrichment combinations in the area between the lower and upper curves must be stored in the peripheral cells of the pool. A peripheral cell is defined as the outermost of the first two storage cells closest to the spent fuel pool wall that has a fuel assembly located in it. If the storage cell closest to the spent fuel pool wall is kept empty of fuel, then the second storage cell from the spent fuel pool wall may be filled with lower burnup fuel meeting the requirements of Category BP fuel.

c. Category BE: Fuel of any burnup with an enrichment 5.0 weight percent, including fresh, unburned fuel, fuel from Category BP or fuel with burnup-enrichment combinations in the area below the lower curve can be placed in Pool B, but must be surrounded by eight empty water cells. Category BE fuel assemblies must be separated by two adjacent empty cells in Pool B.

Spent Fuel Assembly Storage B 3.7.15 (continued)

Crystal River Unit 3 B 3.7-11 Revision No. 89 BASES APPLICABILITY In general, limiting fuel enrichment of stored fuel prevents inadvertent criticality in the storage pools. Inadvertent criticality is dependent on whether fuel is stored in the pools and is completely independent of plant MODE.

Therefore, this LCO is applicable whenever any fuel assembly is stored in high density fuel storage locations.

ACTIONS A.1 When the configuration of fuel assemblies stored in the spent fuel pool is not in accordance with Figure 3.7.15-1 or Figure 3.7.15-2, immediate action must be taken to make the necessary fuel assembly movement(s) to bring the configuration into compliance. The Immediate Completion Time underscores the necessity of restoring spent fuel pool fuel loading to within the initial assumptions of the criticality analysis.

The ACTIONS do not specify a time limit for completing movement of the affected fuel assemblies to their correct location. This is not meant to allow an unnecessary delay in resolution, but is a reflection of the fact that the complexity of the corrective actions is unknown.

SURVEILLANCE SR 3.7.15.1 REQUIREMENTS Verification by administrative means that initial enrichment and burnup of fuel assemblies in accordance with Figure 3.7.15-1 and Figure 3.7.15-2 is required prior to storage of spent fuel in storage pool A or pool B (as applicable). This surveillance ensures that fuel enrichment limits, as specified in the criticality safety analyses (Ref. 1, 2, 6, 7 and 8), are not exceeded. The surveillance Frequency (prior to storage in high density region of the fuel storage pool) is appropriate since the initial fuel enrichment and burnup cannot change after removal from the core.

Spent Fuel Assembly Storage B 3.7.15 Crystal River Unit 3 B 3.7-12 Revision No. 89 BASES REFERENCES

1.

Criticality Safety Evaluation of the Pool A Spent Fuel Storage Racks in Crystal River Unit 3 with Fuel of 5.0%

Enrichment, S. E. Turner, Holtec Report HI 931111, December 1993.

2.

Criticality Safety Analysis of the Westinghouse Spent Fuel Storage Racks in Pool B of Crystal River Unit 3, S. E. Turner, Holtec Report HI-992128, May 1999.

3.

NUREG 0800, Standard Review Plan, Section 9.1.1 and 9.1.2, Rev. 2, July 1981.

4.

10 CFR 50.67.

5.

CR-3 FSAR, Section 9.6.

6.

Criticality Safety Analysis of the Crystal River Unit 3 Pool A for Storage of 5% Enriched Mark B-11 Fuel in Checkerboard Arrangement With Water Holes, S. E.

Turner, Holtec Report HI-992285, August 1999.

7.

Criticality Evaluation of CR3 Spent Fuel Pool Storage Racks with Mark B-12 Fuel, Holtec Report HI-2022907, September 2002.

8.

Progress Energy Engineering Change EC No. 52456, Documentation of Acceptability to Receive and Store Mk-B/HTP Fuel.

9.

Criticality Analysis of Additional Patterns for Crystal River 3 Pools A & B for Progress Energy, Holtec Report No. HI-2063579, September 2006

DUKE ENERGY FLORIDA, INC.

CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50 - 302 / 72 - 1035 LICENSE NUMBER DPR - 72 LICENSE AMENDMENT REQUEST #323, REVISION 0, PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS FOR THE INDEPENDENT SPENT FUEL STORAGE INSTALLATION TO REFLECT PERMANENT REMOVAL OF SPENT FUEL FROM THE SPENT FUEL POOLS ENCLOSURE 3 FACILITY OPERATING LICENSE AND TECHNICAL SPECIFICATIONS - REVISION BAR VERSION

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555 DUKE ENERGY FLORIDA, LLC DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.

License No. DPR-72

1.

The Nuclear Regulatory Commission (the Commission) having found that:

A.

The application filed by Florida Power Corporation*** (the licensee), as supplemented by letter dated December 9, 1976, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.

Construction of the Crystal River Unit 3 Nuclear Generating Plant (facility) has been substantially completed in conformity with Provisional Construction Permit No.

CPPR-51 and the application, as amended, the provisions of the Act and the rules and regulations of the Commission; C.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the rules and regulations of the Commission; D.

There is reasonable assurance: (i) that the activities authorized by this operating license can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the rules and regulations of the Commission; E.

Duke Energy Florida, LLC is technically qualified and financially qualified to engage in the activities authorized by this operating license in accordance with the rules and regulations of the Commission; F.

The licensee has satisfied the applicable provisions of 10 CFR Part 140, Financial Protection Requirements and Indemnity Agreements, of the Commission's regulations; G.

The issuance of this operating license will not be inimical to the common defense and security or to the health and safety of the public;

- - On April 29, 2013, the name "Florida Power Corporation" was changed to "Duke Energy Florida, Inc." On August 1, 2015, Duke Energy Florida, Inc. converted to a limited liability company and the name was changed to "Duke Energy Florida, LLC."

H.

After weighing the environmental, economic, technical, and other benefits of the facility against environmental and other costs and considering available alternatives, the issuance of Facility Operating License No. DPR-72 subject to the conditions for protection of the environment set forth herein is in accordance with 10 CFR Part 51, (formerly Appendix D to 10 CFR Part 50), of the Commission's regulations and all applicable requirements have been satisfied; I.

The receipt, possession, and use of source, byproduct and special nuclear material as authorized by this license will be in accordance with the Commission's regulations in 10 CFR Part 30, 40 and 70, including 10 CFR Sections 30.33, 40.32 and 70.23 and 70.31.

2. Facility Operating License No. DPR-72, issued to the licensee, is hereby amended in its entirety to read as follows:

A.

This amended license applies to the Crystal River Unit 3 Nuclear Generating Plant, a pressurized water nuclear reactor and associated equipment (the facility), owned by Duke Energy Florida, LLC. The facility is located on the Gulf of Mexico, about seven and one-half miles northwest of the town of Crystal River, Citrus County, Florida, and is described in the Final Safety Analysis Report as supplemented and amended and the Environmental Report as supplemented and amended.

B.

Subject to the conditions and requirements incorporated herein, the Commission hereby licenses:

(1)

Duke Energy Florida, LLC, pursuant to Section 104b of the Act and 10 CFR Part 50, Licensing of Production and Utilization Facilities, to possess and use the facility; (2)

The licensee to possess the facility at the designated location in Citrus County, Florida, in accordance with the procedures and limitations set forth in this license; (3)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear material configured as reactor fuel, in accordance with the limitations for storage as described in the Final Safety Analysis Report, as supplemented and amended; (4)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to possess at any time any byproduct, source and special nuclear material as sealed neutron sources used previously for reactor startup, as fission detectors, and sealed sources for reactor instrumentation and to possess and use at any time any byproduct, source, and special nuclear material as sealed sources for radiation monitoring equipment calibration in amounts as required; (5)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components;

(6)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility; (7)

Duke Energy Florida, LLC, pursuant to the Act and 10 CFR Parts 30 and 70, to receive and possess, but not separate, that byproduct and special nuclear materials associated with four (4) fuel assemblies (B&W Identification Numbers 1A-01, 04, 05 and 36 which were previously irradiated in the Oconee Nuclear Station, Unit No. 1) acquired by Florida Power Corporation*** from Duke Power Company for use as reactor fuel in the facility.

C.

This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1)

Deleted per Amendment No. 247 (2)

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

(3)

Deleted per Amendment No. 247 (4)

Deleted per Amendment No. 20 dated 7-3-79 (5)

Deleted per Amendment No. 247 (6)

Deleted per Amendment No. 21, 7-3-79 (7)

Deleted per Amendment No. 247 (8)

Deleted per Amendment No. 247 (9)

Deleted per Amendment No. 247 (10) Deleted per Amendment No. 247 (11) Deleted per Amendment No. 247 (12) Deleted per Amendment No. 237 (13) Deleted per Amendment No. 229 (14) Deleted per Amendment No.

(15) Deleted per Amendment No. 247

- - On April 29, 2013, the name "Florida Power Corporation" was changed to "Duke Energy Florida, Inc." On August 1, 2015, Duke Energy Florida, Inc. converted to a limited liability company and the name was changed to "Duke Energy Florida, LLC."

Added per Amdt. 15, 7-24-78

D.

Physical Security The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p).

The combined plan, which contains Safeguards Information protected under 10 CFR 73.21, is entitled: "ISFSI Physical Security Plan, ISFSI Safeguards Contingency Plan and ISFSI Guard Training and Qualification Plan, Revision 0, submitted by letter dated May 24, 2016.

E.

Deleted per Amendment No. 247 F.

In accordance with the requirement imposed by the October 8, 1976, order of the United States Court Appeals for the District of Columbia Circuit in Natural Resources Defense Council v. Nuclear Regulatory Commission, No. 74-1385 and 74-1586, that the Nuclear Regulatory Commission "shall make any licenses granted between July 21, 1976 and such time when the mandate is issued subject to the outcome of the proceedings herein," the license issued herein shall be subject to the outcome of such proceedings.

G.

This amended license is effective as of the date of issuance. Facility Operating License No. DPR-72, as amended, shall expire at midnight, December 3, 2016.

Duke Energy Florida, LLC submitted the 10 CFR 50.82(a)(1) notification to the Nuclear Regulatory Commission on February 20, 2013. Per 10 CFR 50.51(b), the Facility Operating License No DPR-72 continues in effect until the Commission notifies the licensee that the License has been terminated.

FOR THE NUCLEAR REGULATORY COMMISSION Original Signed by Roger S. Boyd, Director Division of Project Management Office of Nuclear Reactor Regulation Attachments:

Appendices A & B - Technical Specifications Date of Issuance: Jan 28 1977 Amdt. #97 March 31, 1987

CRYSTAL RIVER UNIT 3 PDTS LOEPage 1 of 1 XX/XX/XX PERMANENTLY DEFUELED TECHNICAL SPECIFICATIONS List of Effective Pages (Through Amendment 249 and ITS Bases Revision 89)

Amendment Nos. 159, 164, 166, 171, 173, 181, 189, 190, 226, 235, 238, 239, 242, 243, 245 and 246 amended the CR-3 Operating License, only, and did not effect changes to the ITS LCOs or Bases.

Page Amendment Page Amendment i

247 ii 247 4.0-1 247 5.0-1 247 5.0-2 247 Appendix B - Part II 1-1 190 2-1 190 3-1 190 3-2 190 4-1 190 4-2 190

Crystal River Unit 3 i

Amendment No. 247 TABLE OF CONTENTS 4.0 DESIGN FEATURES........................................ 4.0-1 5.0 ADMINISTRATIVE CONTROLS................................ 5.0-1

Crystal River Unit 3 ii Amendment No. 247

Design Features 4.0 Crystal River Unit 3 4.0-1 Amendment No.

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

4.2 Not Used 4.3 Fuel Storage Spent fuel shall not be stored in the spent fuel pool.

High Radiation Area 5.8 (continued)

Crystal River Unit 3 5.0-1 Amendment No.

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 from the radiation source or from any surface penetrated by the radiation but less than 500 rads/hr at 1 meter from the radiation source or from any surface penetrated by the radiation shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor or health physics supervision. Doors shall remain locked except during periods of access by personnel.

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.

High Radiation Area 5.8 Crystal River Unit 3 5.0-2 Amendment No.

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

> 1000 mrem/hr at 30 cm from the radiation source or from any surface penetrated by the radiation but less than 500 rads/hr at 1 meter from the radiation source or from any surface penetrated by the radiation, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that are not 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.

3F0816-01, License Amendment Request 323, Revision 0, Permanently Defueled Technical Specifications for the Independent Spent Fuel Storage Installation to Reflect Permanent Removal of Spent Fuel from the Spent Fuel Pools

Contents

Text

Subject:

References:

Dear Sir:

Enclosures:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 TECHNICAL ANALYSIS

SUMMARY

1.1 DESCRIPTION

1.2 PROPOSED CHANGE

1.3 TECHNICAL ANALYSIS

SUMMARY

3.1 DESCRIPTION

3.2 PROPOSED CHANGE

3.3 TECHNICAL ANALYSIS

SUMMARY

7.1 DESCRIPTION

37.2 PROPOSED CHANGE

7.3 TECHNICAL ANALYSIS

SUMMARY

4.1 DESCRIPTION

4.2 PROPOSED CHANGE

4.3 TECHNICAL ANALYSIS

SUMMARY

5.1 DESCRIPTION

5.2 PROPOSED CHANGE

5.3 TECHNICAL ANALYSIS

SUMMARY

5.0 REGULATORY ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

7.0 REFERENCES

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