3F1013-01, License Amendment Request 316, Revision 0, Revise and Remove License Conditions and Revision to Improved Technical Specifications to Establish Permanently Defueled Technical Specifications

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License Amendment Request 316, Revision 0, Revise and Remove License Conditions and Revision to Improved Technical Specifications to Establish Permanently Defueled Technical Specifications
ML13316C083
Person / Time
Site: Crystal River Duke Energy icon.png
Issue date: 10/29/2013
From: Elnitsky J
Duke Energy Carolinas
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
3F1013-01 LAR 316, Rev 0, TAC MF3089
Download: ML13316C083 (191)
v • d • e


Text

{{#Wiki_filter:DUKE Crystal River Nuclear Plant ENERGY

  • 15760 W. Power Line Street Crystal River, FL 34428 Docket 50-302 Operating License No. DPR-72 10 CFR 50.90 October 29, 2013 3F1013-01 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001

Subject:

Crystal River Unit 3 - License Amendment Request #316, Revision 0, Revise and Remove License Conditions and Revision to Improved Technical Specifications to Establish Permanently Defueled Technical Specifications

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 September 26, 2013, "Crystal River Unit 3 -

License Amendment Request #315, Revision 0, Permanently Defueled Emergency Plan and Emergency Action Level Scheme, and Request for Exemption to Certain Radiological Emergency Response Plan Requirements Defined by 10 CFR 50"

3. CR-3 to NRC letter dated September 4, 2013, "Crystal River Unit 3 - License Amendment Request #313, Revision 1, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions and Response to Requests for Additional Information" (ADAMS Accession No. ML13255A056)

Dear Sir:

Pursuant to 10 CFR 50.90, Duke Energy Florida, Inc. (DEF), formerly known as Florida Power Corporation, hereby provides this License Amendment Request (LAR) to revise the Crystal River Unit 3 (CR-3) Facility Operating License (FOL) to remove and revise certain License Conditions. This LAR also proposes to extensively revise the CR-3 Improved Technical Specifications (ITS) in order to create the CR-3 Permanently Defueled Technical Specifications (PDTS). In Reference 1, the NRC acknowledged CR-3's certification of permanent cessation of power operation and permanent removal of fuel from the reactor vessel. Accordingly, pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel. Removing and revising CR-3 FOL License Conditions generally fall into three categories: 1) License Conditions are currently inconsistent with the non-operating status of the plant, 2) License Conditions were one time conditions that have been satisfied, or 3) Part 50 and Part 73 regulations have been established which accomplish the intent of or eliminate the applicability of the License Conditions.

U. S Nuclear Regulatory Commission 3F1013-01 Page 2 of 2 The ITS changes are being made since, in the permanently defueled condition, many of the existing portions of the ITS are no longer applicable for CR-3. The basis for determining continued applicability of ITS sections is evaluation of the CR-3 structures, systems, and components using the criteria in 10 CFR 50.36(c). Reference 2 submitted proposed changes to the Emergency Planning standards for CR-3. An attachment in that submittal summarizes the accident analysis performed to support those changes and are referenced in this LAR in support of conclusions justifying the changes requested herein. The changed pages provided in this LAR are based on the presumption that the changes to the Administrative Controls section of the ITS proposed in LAR #313, Revision 1 (Reference 3), will have been approved prior to approval of this LAR. Therefore, the amendment numbers on the majority of the PDTS revision bar pages included herein have been left blank. DEF requests NRC approval of this LAR by October 31, 2014, with a 30 day implementation period. There are no new regulatory commitments made within this submittal. The CR-3 Plant Nuclear Safety Committee has reviewed this request and recommended it for approval. If you have any questions regarding this submittal, please contact Mr. Dan Westcott, Licensing Supervisor, at (352) 563-4796. I declare under penalty of perjury that the foregoing is true and correct. Executed on October 29, 2013. oject Management and Construction JE/scp Attachments: A. Description of Proposed License Amendment Request, Background, Justification for the Request, and Regulatory Analysis B. Proposed Facility Operating License Page Changes, Strikeout and Shadowed Text Format C. Proposed Facility Operating License Page Changes, Revision Bar Format D. Proposed Technical Specification Page Changes, Strikeout and Shadowed Text Format E. Proposed Technical Specification Page Changes, Revision Bar Format F. PDTS Bases Pages for Information, Strikeout and Shadowed Text Format xc: NRR Project Manager Regional Administrator, Region I

DUKE ENERGY FLORIDA, INC. CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #316, REVISION 0 ATTACHMENT A DESCRIPTION OF PROPOSED LICENSE AMENDMENT REQUEST, BACKGROUND, JUSTIFICATION FOR THE REQUEST, AND REGULATORY ANALYSIS

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 1 of 44 DESCRIPTION OF PROPOSED LICENSE AMENDMENT REQUEST, BACKGROUND, JUSTIFICATION FOR THE REQUEST, AND REGULATORY ANALYSIS 1.0 Description of Proposed License Amendment Request Pursuant to 10 CFR 50.90, Duke Energy Florida, Inc. (DEF), formerly known as Florida Power Corporation, proposes to amend the Crystal River Unit 3 (CR-3) Facility Operating License (FOL) DPR-72 and Appendix A, the Improved Technical Specifications (ITS). This License Amendment Request (LAR) proposes to remove or revise certain License Conditions that are no longer applicable to CR-3 in the permanently defueled condition. This LAR also proposes to extensively revise the CR-3 ITS in order to create the CR-3 Permanently Defueled Technical Specifications (PDTS). 2.0

Background

CR-3 has been shutdown since September 26, 2009, when the plant entered the Cycle 16 refueling outage. In the process of creating a construction opening for replacement of steam generators during that outage, a delamination of the concrete shell of the containment was discovered. The construction opening and adjacent concrete shell of the containment was repaired during 2010 and 2011. During tensioning of the containment prestressing tendons following the concrete repair, delaminations occurred in two other sections of the containment shell. In consideration of performing a second repair of the containment shell, all fuel was removed from the reactor vessel and placed in storage in the Spent Fuel Pools (SFPs) as of May 28, 2011. On February 5, 2013, Duke Energy announced that CR-3 would be retired. By letter dated February 20, 2013, CR-3 informed the NRC of the permanent cessation of operation and that CR-3 had removed all fuel from the reactor vessel (Reference 6.1). By letter dated March 13, 2013, the NRC acknowledged CR-3's certification of permanent cessation of power operation and permanent removal of fuel from the reactor vessel (Reference 6.2). Accordingly, pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel. Chapter 14 of the CR-3 Final Safety Analysis Report (FSAR) described the complete scope of the event and accident analyses performed for CR-3 as an operating power plant. Based on the limitations applied to CR-3 under 10 CFR 50.82(a)(2), the majority of those accidents are no longer credible and were removed from the FSAR under 10 CFR 50.59. The only accidents that remain credible in the permanently defueled condition, with fuel stored in the SFPs, are a Fuel Handling Accident and a Radioactive Waste Handling Accident. The Radioactive Waste Decay Tank Rupture Accident described in the FSAR is no longer credible for CR-3 as all waste gas has been released and the tank relief valves have been removed. In LAR #315, Revision 0, (Reference 6.3), CR-3 proposed a revised Radioactive Waste Handling Accident and requested its approval for inclusion in the CR-3 FSAR. 3.0 Justification For The Request 3.1 License Conditions Removing and revising CR-3 FOL License Conditions generally fall into three categories: 1) License Conditions are currently inconsistent with the non-operating status of the plant, 2) License Conditions were onetime conditions that have been satisfied, or 3) Part 50 and Part 73

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 2 of 44 regulations have been established which accomplish the intent of or eliminate the applicability of the License Conditions. License Condition 2.B.(1) 2.B.(1) Duke Energy Florida, Inc., pursuant to Section 104b of the Act and 10 CFR Part 50, "Licetn ofiization Facilities," to possess,-use and operate the facility ~ This License Condition is proposed for revision to be consistent with the restriction of 10 CFR 50.82(a)(2) that CR-3 is no longer authorized to operate or place fuel in the reactor vessel. The revised License Conditions will read as follows: 2.B.(1) Duke Energy Florida, Inc., pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess and operate the facility as required for fuel storage; License Condition 2.B.(3) 2.B.(3) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Part 70, to eeeiVe, possess and d-ke at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and am.uAG requir:ed for reactor operation, as described in the Final Safety Analysis Report, as supplemented and amended; This License Condition is proposed for revision to be consistent with the restriction of 10 CFR 50.82(a)(2) that CR-3 is no longer authorized to operate or place fuel in the reactor vessel. The revised License Conditions will read as follows: 2.B.(3) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear material configured as reactor fuel, in accordance with the limitations for storage, as described in the Final Safety Analysis Report, as supplemented and amended; License Condition 2.B.(4) 2.B.(4) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30, 40, and 70 to FeeeiVe, possessaPnd use at any time any byproduct, source, and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation inamunts as required; This License Condition is proposed for revision to be consistent with the restriction of 10 CFR 50.82(a)(2) that CR-3 is no longer authorized to operate or place fuel in the reactor vessel. The revised License Conditions will read as follows: 2.B.(4) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30, 40, and 70 to possess at any time any byproduct, source, and special nuclear material as sealed neutron sources used previously for reactor startup, as fission detectors, and sealed

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 3 of 44 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; License Condition 2.C.(1) 2.C.(1) Maximum Power Level Duke Energy FGlorida, Inc. is authorized to operate the faciliV; at a steady state reacGt core poWer i8Vel not in 8Xcess of 2609 Megawatts (100 percent of rated core poWer This License Condition is proposed for elimination to be consistent with the restriction of 10 CFR 50.82(a)(2) that CR-3 is no longer authorized to operate or place fuel in the reactor vessel. The License Condition will read as follows: 2.C.(1) Deleted per Amendment No. License Condition 2.C.(2) 2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 24r, are hereb "Tenic DueEergy Florida, Inc. shall tefacility in accordance with the Technical Specifications. The Suve*!lanca Requiromonts contained OR the Appendix A T-echnical Specificationa and listed below are Rot Fequ"red to be perfoFrmed immediately upon imnplementation oe .A.mendment 119. The supfeillanne Requirem~ents shall be successfully demonstrated prior to the time and condition specified below for each.6 a) SR 3.3.8.2.b shall be successfully demonstrated prior to entering MODE 4 on the first plant Stant up following Refuel Outage 0-. b) SR 3.3.11.2, Functio 2, shall be succGessfully demon~strated no later than 31 days followingOA the imAPl8emotati9n date of the ITS. G) SR 3.3417.14, Functions 1, 2, 6, 10, 14, & 1:7 shall be successfully demonstrated Ano later than 31 days following the implementation date of the ITS. d) SR 3.3.17-.2, Fmuinction 10 shall be succe~ssfully demonstrated prior to entering MODE 3 on the firsit plant stak up following Refuel Outage 9-. e) SR 3.6.1.2 shall be successfully demonstrated prior to entering MODE 2 OR the first plant sta~t up fellewing Refue! Outage 9-. SSR 3.7.12.2 shal! be successfully demonsr~ated prior to entering MODE 2 on the first plant stant up following Refuel Outage G-. g) SR 3...1 hall be successfully demonstrated prior to entering MOIDE 2 on the first plant stant up following Refuel Outage 9. h) SR 3.8-2-23 shall bhe suGcossfully demonstrated prior to entering MODE 4 on the AFrst P'ant start up fellowing Refuel Outage 9-.

U. S. Nuclear Regulatory 3F1 013-01 Attachment A Page 4 of 44

6) SR 3.8.4.5 shall be successfully demoen~trated prior to entering MODE 4 on the first Diant start UD foliowing I.*euI I uuI ae to I)SR 3.8.7.1 shall be successfully demonstrated noe later than 7 days following the aimnplementation date of the ITS.

k) SIR 3.8.8.1 shall be successfully demonstrated no Iater than 7 daYs following the Smplementation date -of the ITS. The first paragraph of this License Condition is proposed for revision to be consistent with the restriction of 10 CFR 50.82(a)(2) that CR-3 is no longer authorized to operate or place fuel in the reactor vessel. The second paragraph of License Condition 2.C.(2) is proposed for elimination along with the onetime surveillance requirements that were satisfied as stated. This License Condition will read as follows: 2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. , are hereby replaced with the Permanently Defueled Technical Specifications (PDTS). Duke Energy Florida, Inc. shall maintain the facility in accordance with the Permanently Defueled Technical Specifications. License Condition 2.C.(3) 2.C.(3) Duke Energy Flori~da, Inc. shall not operate the reactor in operational Mondes I and :2 W .ith lossA than thrcee reactor coolantpupinoeaonntlsfyaayesores than three pump operation h-ave been.R subm-~rifted-by the licensees and approval has b9een granted by teCmmission by 1aRGmenmet to this; license This License Condition is proposed for elimination to be consistent with the restriction of 10 CFR 50.82(a)(2) that CR-3 is no longer authorized to operate or place fuel in the reactor vessel. The License Condition will read as follows: 2.C.(3) Deleted per Amendment No. License Condition 2.C.(5) 2.C.(5) p II 4 le A winin six mFn.ns of tfle date 0f issuance o. tni' ic..en., i'-lerid I'1.owe. G.;rporta,-. shall comple-te mondifications to the level indication of the ber~ated wa~ter stor~age tank, and installation of dual setpoint pilot operated relief valve on the pressurizer. This onetime License Condition is proposed for elimination since it has been satisfied. The revised License Condition will read as follows: 2.C.(5) Deleted per Amendment No. License Condition 2.C.(7) 2.C.(7) Prir to startup following the first regularly scheduled refueling Outage, FrIieda Power Cirperation*** shall difi to the satisfaction*of the Commission, the reactor coolant ss*tem flow indic-ation to Meet the single failure crfitnero with regard to pressure sensing lines to the flnow differential oaressure traRF;nsmittrs.

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 5 of 44 This onetime License Condition is proposed for elimination since it has been satisfied. The revised License Condition will read as follows: 2.C.(7) Deleted per Amendment No. License Condition 2.C.(8) 2.C.(8) Within thre. months of io.uanc. f this lic..ense, Florida PoWer Crperatin*** shall sub-mit to the Commisio a propsed sur.'eillance program for monGitrig the coentainment for the purposo;A of determining any future delamnination of the dome This onetime License Condition is proposed for elimination since it has been satisfied. The revised License Condition will read as follows: 2.C.(8) Deleted per Amendment No. License Condition 2.C.(9) 2.C.(9) Fire Protection Duke EneFrgy Florida, Inc. shall iad Maintain iffeGt all provisi*nsn of the approved fire protection prgama escribed in the Final Safety Analysis Repeot for the facility an asapoed in the Safet Evaluation Repeft dated July 2-7, 1979, January 22, 1981, Janua.y 6,I1983, July 18,,1985 and Macgh 16, 1988, subject to the following proVisions: The licensee mna" make cshanges to the approved fire protection programn without prior approvwal of the Com.mission only if those changes Would-not ;adverselyafc the ability to achieve and maintain safe shutdown in the event of a fire. This License Condition is proposed for elimination consistent with the restriction of 10 CFR 50.82(a)(2) that CR-3 is no longer authorized to operate or place fuel in the reactor vessel. This condition for making changes to the Fire Protection Program is no longer required to assure fire safety by maintaining the ability to achieve and maintain safe shutdown in the event of a fire. License Condition 2.C.(9), which is based on maintaining an operational Fire Protection Program, in accordance with 10 CFR 50.48, with the ability to achieve and maintain safe shutdown of the reactor in the event of a fire, is no longer applicable for CR-3. However, many of the elements that are applicable for the operating plant Fire Protection Program continue to be applicable during plant decommissioning. During the decommissioning process, a Fire Protection Program is required by 10 CFR 50.48(f) to address the potential for fires that could result in a radiological hazard. However, the regulation is applicable regardless of whether a requirement for a Fire Protection Program is included in the facility license. Therefore, a license condition requiring such a program for a permanently shutdown and defueled plant is not required. The revised License Condition will read as follows: 2.C.(9) Deleted per Amendment No.

U. S. Nuclear Regulatory 3F1013-01 License Condition 2.C.(10) Attachment A Page 6 of 44 2.C.(10) The design of the reacrtor GO olant pump rUDports need not include conSideation o the effects of postulated ruptures of the primnary reactor coolant loop piping and May be revised in accordance moith Florid Power Corperation'r,*** amon.dmont request ot Api 24, 19865 This License Condition is proposed for elimination consistent with the restriction of 10 CFR 50.82(a)(2) that CR-3 is no longer authorized to operate or place fuel in the reactor vessel. This License Condition has been implemented and the reactor coolant system is no longer subject to pressurization. The revised License Condition will read as follows: 2.C.(10) Deleted per Amendment No. License Condition 2.C.(11) 2.C.(1 1) A system of thermocouples added to the decay heat (DH) drop and Auxiliary/ Presurizer Spray (APS) lines, capable of detecting flow initiation, shall be operable for Modes 4 through 1. Channel checks, of the the~rmocouples, shall be peorf~Med on a monthly basis to demonstrate operability. if either the DH or APRS system thermocouples become finoperable, operability shall beA resotored-within; 30 days or the NRC shall be infomed, in a Special Report within the following fourteen (14) days, ot the nopereabilit' and the plans to restore operability. This License Condition is proposed for elimination consistent with the restriction of 10 CFR 50.82(a)(2) that CR-3 is no longer authorized to operate or place fuel in the reactor vessel. The reactor coolant system is drained and the function of these thermocouples is unnecessary. The revised License Condition will read as follows: 2.C.(11) Deleted per Amendment No. License Condition 2.C.(14) 2.C.(14) Wfigat;^n St*ateg eRp-G AGI..... -18 AGIA The licens~ee shall develop ai explesions ;and that inclu"-deA the-ne amanain GItrateaies ror aaaressina ]a~ge flFes and following keyArow (1.) Fire fighting responses strategy with the following elements:.

a.

Pre defined coordinated fire Frespnse strategy and guidancGe

b.

Assessment of mutual aid fire fighting assets c.f Designated staging areas for equipment and mnateril d.f Command and control

e.

TrFaining of response personnel (2.) Operations6 to mnitiqate fuel damnaqer considering the fellewing: v

a.

Protec,*ti and use e, personnel assets

b.

Communications

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 7 of 44 G. Mignimizing fire p*read

d.

Proedures for elmime entinegy efire response strategy

e.

m idetificatien of roadily availablo pro staged equipmenT

f.

Trtaining on integrated fire resporSe strategy 9M Spent fuel pool mitigation measures (3.) Action2 to minimize r0lease to inRlude consideratien of:

a.

W~ater spray scrubbing 190 Do-se to-onsAite Frespoders This section is proposed for elimination in its entirety. CR-3 has permanently ceased operation; therefore, the mitigation strategy license condition is no longer required. The NRC issued this license condition on August 23, 2007, to incorporate the requirements for the Interim Compensatory Measures (1CM) 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), section 50.54(hh) does no nt apply to a defueled reactor that has submitted the certification for permanent removal of fuel under 10 CFR 50.82(a). In the Federal Register notice for the Power Reactor Security Requirements Final Rule (74FR13926) the NRC states that, "Section 50.54(hh) requirements do not apply to decommissioning facilities for which the certifications required under 50. 82(a)(1) or 52. 1 10(a)(1) have been submitted." It also states, "The Commission notes that the 50.54 (hh) [requirements] do not apply to any current decommissioning facilities that have already satisfied the 50.82(a) requirements." On November 28, 2011, the NRC issued a letter to rescinded Item B.5.b of the ICM Order EA-02-26. Therefore, neither the ICM Order nor 10 CFR 50.54(hh) require continuation of B.5.b mitigation strategies for CR-3. The revised License Condition will read as follows: 2.C.(14) Deleted per Amendment No. License Condition 2.C.(15) 2.C.(15) Upon imlmnainof Amendment No. 230 adopting T-STF 118, Revision 3, the determ~ination Of control comnplex habitability envelope (CC HE) unfilteredaiinekg as required by Surveillance Requirement (SR) 3.7.12A1, in accredance With ITS; 6.6.2.21.3(i) and the assessment of CCHE habitability as requfired by ITS 5,65.2.21 3(ii), shall be considered mFet. Folloing implementatin: a) The frt performan*.e-of SR 347.12.4, in.accordancR.e with Specific*aRti 5.6.2.21.3(m), shall be within the specified Frequency of 6 years, plus the 18 month allo*wane of SR 3.0.2, as measured from May 18, 2007, the date of the moGst recent successful inleakage test. b) The first per9formance Of the periodic asseessment -of CCHE habitability, ITS 5.6.2.21.3(ui), shall be within 3 years, plus the 9 month allowance Of SR 3.0.2-,

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 8 of 44 as measured from May 18, 2007, the date of the mot rec*,ent SUcessful, s neakage teet-G) The Control Complex Habitability Envelope Integrity Program will be used to yorify' the integrity of the Control Complex boundar,'. Conditions that aro detife hto bedverse; shall be trede and used as pa~t of the :21 month4 as~sessment of the CCHE boundar,'. This aass-e-ssment will be pe~fermed withint 60 days of implementationi of Amendment. Reference 6.3 proposed changes to the Radiological Emergency Response Plan for CR-3. That correspondence includes a description of the revised accident analysis for the Fuel Handling Accident. The calculation accounts for radioactive material inventory in the most recently irradiated fuel elements in the pools after four years of decay. The calculation determined that the dose to occupants of the control room following this accident would be less than one millirem TEDE for an extended occupancy period. The calculation did not credit Control Complex ventilation isolation or the Control Complex Habitability Envelope Integrity Program to limit inleakage. Therefore, this license condition is proposed for elimination. This submittal also proposes to remove ITS 3.7.12, Control Room Emergency Ventilation System (CREVS) and ITS Program 5.6.2.21, Control Complex Habitability Envelope Integrity Program. The revised License Condition will read as follows: 2.C.(15) Deleted per Amendment No. License Condition 2.D 2.D Physical and-Gybel 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 plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Physical Security Plan, Revision 5," and "Safeguards Contingency Plan, Revision 4," submitted by letter dated May 16, 2006, and "Guard Training and Qualification Plan, Revision 0," submitted by letter dated September 30, 2004, as supplemented by letters dated October 20, 2004, and September 29, 2005. The licensee sha. ! fully iFmplemet and. mraintain Pi :effecr-"t all provisioFns of the Commissinr app-roed cyber s*urity plan (CSP), incluing changes made pur,,,sant t the authority of 10 C-F=R 50.90 and 10 CFR 50.54(p). The licensee's CSP was approved by License Amoendment No. 238, as supplemented by a change approved by -ic-ense-

Amendment hNo 212.

The second paragraph of this license condition, which addresses the Cyber Security Plan, is proposed for deletion in its entirety. CR-3 has permanently ceased operation and is prohibited from returning fuel to the reactor vessel or operating a reactor core. This license condition was added to comply with 10 CFR 73.54, which states: "By November 23, 2009 each licensee currently licensed to operate a nuclear power plant under part 50 of this chapter shall submit, as specified in §50.4 and §50.90 of this chapter, a cyber security plan that satisfies the

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 9 of 44 requirements of this section for Commission review and approval." (Emphasis added) Therefore, the Cyber Security Plan license condition is no longer required. The first paragraph of License Condition 2.D remains unchanged. The License Condition will read as follows: 2.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 plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Physical Security Plan, Revision 5," and "Safeguards Contingency Plan, Revision 4," submitted by letter dated May 16, 2006, and "Guard Training and Qualification Plan, Revision 0," submitted by letter dated September 30, 2004, as supplemented by letters dated October 20, 2004, and September 29, 2005.

U. S. Nuclear Regulatory 3F1013-01 License Condition 2.E Attachment A Page 10 of 44 2.E This l6 I lene i

  • s6ubject to the folloWnRg aRtitrust E=nergy Florida, Inc. (DEF):

VVeIditions aRd applies onlI to Duke (1) DEF will intercoAnect with and coordinate rescvwes of eer*gency bulk power with any entity Or cntitie proposing to engage in electric bulk poWer supply oI by means of the salee s in its serVice area* e R teI:rms that wl rvd ~nd eXeha~qe 4e=-QE -ost \\,..v luding a reaso-n-Rable return) On G9 to the benefits of reserve RIIUPARI LullA RL1 jiUPIAO +H8 9tH8FLIU pJ; LIGI ;j~u;IS (a) Interconnections will not be limited to low voltages when higher voltages are avalable from DEEF installed fiacilities i the area where itronn-.-.ectio is Aa desired, when the proposed arrangement is fo-und to be technically and economfically feasible-. (b) Emergecseie agreements Will not be limited to A fixed-a~mount, but emergency serVice providdudrsc gemnswl be furnished to the fullest extent available and desired where sucsh supply does not impair serVice to the supplier's customers (6) neeefd-*a pipe of the type of it and which would ion" would be one in reserve sharing arrangemnent available to any IIf II I II OrAOvip"llo t uii access Wnet1 0? rese erv ,hich the following conditions would obtain: (1) IDEF and each partficipant(s) shall provide to the other emergencyI power if and when available from its own generation, Or thrOUgh its tF.R-ransmission from: the generation of others to the extent it can do so without disruotfinGI service to its own customers. (2) The par-ticipant(s,) to the; reserve shaFrig arrangemnent shall, jofintly with -DEEF esbtablish from time to time the minmu reere to benstalld and/orF purchaseAd-asr necessary to mnaintain in total an adequate reliability of power supply oR the inecnetdsystem of DEE and participant(s). The Freseve responsibility thus determined shall be calculated as a percentage of peak load. No par-ticipant(s) to the i nterconnection shall be required to maintain moere than such percentage as, a percentage of its peak load; provided that if the reserve requirements of DEE are increased over and above the aRmou-,n t lDEE woul be required to maintain without suc rh interconnection then the other participant(s,) shall be required to carry or provide for as it reserve Fresposibility the fUll amou-nt in kilowAftA of operating reserve requirements eamceed-the ins-talle-d reserve Fequuemnenk

  • The use of the termn "service area" in no way indicate anainment Or allocation oe whoeslemarket areas. It is nteded only as a genera inictonofa aeawt~h~in the S-tate o-f lorid-a where DEE provides some class of electric service
    • In order to clarity the commitments,, certainl explanatory notes, have -been-ad-ded wherep RieG86sa~y.

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 11 of 44 (d) IntelSVRconV cn aAVI VOOrinatir l I I iI 1 ag~eerm~t ellay-~ ythis GGend tie-of they de Gion portaining to inter 6YStemA coordination. not emooaGy resiitF!Gt'e proVIG irlIU 41* lr p VI VV~~u iV UUVUIVpUU Iil LIii IIV iV U i III I L1U1 t LIIII ~ .IV II V*LIVI;y L11I. ccIdUtIFY '*GI~ as deeoe thsaeafe +.+A ov il"aIFYII condiion If it is non rest ricti-ve. (2) DEF will purchase froem or sell "buliiik power" to any other entity Or.. tities iOn the af-osaid re, enging in 9o prop sing to engago, n the gen*erAtono f e*ecFtr icpo.*,eR bulk, at.its COSt nc a reasnable return) when .uch transaction. would se..e ton educe the overall cots; of new bul*k power 6upply feoF tr ielf Or the other particsipant or panticipant6 to the transaction. This refers specifically to the oppGorunity to cnoornate On the planing t now generation, tranismission and associated facilities'. (a) it is nt*i to, be that thiS condition req;uires DEF to purchase or soilhbulk power fi finds such purchase or sale unfeasi ble or its costs in connection with such purc-hase or sale would exc-eed its beniefits therefrom. (b) if ,DEE engages*, !R coordinated*P. dievelopment of its bulk power supply system with that of any other bulk power supply system, by selling unit powenr at the cost of its, new power supply, or enae in joint ventures with the samne result, DEF= shall net re-fuse proportionial participation OR a GOmparable basis from the same unit to any other entity init enoie area (see Commitmenlt 1, supra) engagn ino proosngto engage in bulk power supply to the e~dent it is tehncalyfasibly to poid such unlit powAer from the unlit orF unis in question. (3) DEE will facilitate the eXchange of bulk power by transmission over its system between Or among two Or more entities with w~hich iisntroeced n terms which will fully compensate it for the useof itsyemo the extent that s:ubject arrangemfenits reasonablv can be accommodated from a funcr-tional and technical rtandeoint C,,nlpnan a nn (a) This onrdition applies to entities With whic.h

DEm*

nay be interconnectednthe futu-re as, well as, those to which it is noewinecnctd (b) DEE is obligated under this condition to transmit bulk Power for other entit,, the terms. sbtated aboave, and to inldnisplanning and construc~tion prog sLuffcient trFansRmiGsion capacity as required therefore, provided that such entities give DE*E

  • suffiAcit
  • aRdv;an**

n e as mnay be required to acommc the arranigemnent ftrom a func~tional and technical standpoinit and that the entities will be obligated to comApensa-;te-DEEF fully for the use Of it system 6V 0l R ethe date etheF (4) DEE will sell power i bulk to any entity in the aforesaid area now engaging in or proposfinig to enigage inthe retail distribution of electric power. (5) It s recognized that the foregoing conditions are to be implemented in a manner consitent-AR wAit !Re provisions Of Mne Federal roGwer An-tm and all rates, Gnarges or practices in connection therewith are to besbec

ote approval of regulator,' agece having jurisdic-tion over themr.

This License Condition is being proposed for deletion in its entirety. This License Condition was imposed to address antitrust concerns associated with the operation of CR-3 as part of the

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 12 of 44 former Florida Power Corporation system. Since the restrictions of 10 CFR 50.82(a)(2) no longer authorize operation of CR-3 to generate power, then the antitrust concerns are no longer valid for CR-3. The revised License Condition will read as follows: 2.E Deleted per Amendment No. 3.2 Technical Specifications 10 CFR 50.36, 'Technical specifications,' provides specific guidance for development of technical specifications for decommissioning plants and provides the regulatory basis for the PDTS proposed in this LAR. 10 CFR 50.36(c)(6), 'Decommissioning,' states: "This paragraph applies only to nuclear power reactor facilities that have submitted the certifications required by § 50.82(a)(1) and to non-power reactor facilities which are not authorized to operate. 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." This evaluation and justification for the proposed revisions to the CR-3 ITS to create the CR-3 PDTS will first establish which Limiting Conditions for Operation (LCOs) are required to be maintained. Based on PDTS LCOs that will be maintained, proposed changes to Sections 1.0, 'Use and Application,' 3.0, 'Limiting Condition for Operation (LCO) Applicability,' and 'Surveillance Requirement (SR) Applicability,' will be justified. Sections 2.0, 'Safety Limits (SLs),' 4.0, 'Design Features,' and 5.0, 'Administrative Controls,' will be treated separately identifying sub-sections that are proposed for deletion. 3.2.1 Limiting Condition for Operation: By letter dated August 25, 1989, Florida Power Corporation (FPC) proposed to amend Appendix A of Operating License No. DPR-72 to revise, in its entirety, the Crystal River Unit 3 Technical Specifications. The proposed amendment was based on guidance provided in the, "NRC Interim Policy Statement on Technical Specification Improvements for Nuclear Power Reactors," published on February 6, 1987 (52 FR 3788). During its review, the NRC staff relied on the NRC's Interim Policy Statement and later on NUREG-1430, "Standard Technical Specifications - Babcock and Wilcox Plants," which was issued in September 1992. The CR-3 Improved Technical Specifications were issued by letter dated December 20, 1993 (Reference 6.6). The Interim Policy Statement contained three specific criteria for inclusion of a LCO in the technical specifications. It also contained a provision that licensees should retain LCOs for a specified list of systems that operating experience and probabilistic risk assessment (PSA) had generally been shown to be important to public health and safety. In the final policy statement and subsequent revision to 10 CFR 50.36, that concept became Criterion 4. The current CR-3 ITS contains discussion in the Bases Applicable Safety Analyses, relative to operating experience and PSA results, as being determining factors for inclusion of particular LCOs. In discussion of LCOs that currently reference operating experience or PSA insights, in this LAR justification, reference to Criterion 4 is used instead, consistent with the latest revision of NUREG-1430, "Standard Technical Specifications - Babcock and Wilcox Plants." CR-3 anticipates that this will facilitate review of the proposed changes.

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 13 of 44 10 CFR 50.36(c)(2) states in part: "Limiting conditions for operation are the lowest functional capability or performance levels of equipment required for safe operation of the facility. A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria: Criterion 1. Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary." The reactor coolant pressure boundary performs no safety function with all fuel stored in the SFPs. No LCOs included in the CR-3 ITS to satisfy Criterion 1 are required in the CR-3 PDTS. "Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The only fission product barrier still performing a safety function is the cladding of the fuel stored in the SFPs. No LCOs associated with cladding of fuel in the reactor core, the reactor coolant pressure boundary, or the containment fission product barriers that were included in CR-3 ITS due to Criterion 2 are required in the CR-3 PDTS. Only those LCOs associated with fuel cladding in the SFPs are required. "Criterion 3. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier." The Fuel Handling Accident and the Radioactive Waste Handling Accident are the only design basis accidents which can occur in a permanently defueled plant. There are no structures, systems, or components which actuate to mitigate either of those accidents. No LCOs for equipment that was designed to mitigate other design basis accidents are required in the CR-3 PDTS. "Criterion 4. A structure, system, or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety." There are no LCOs in the CR-3 ITS that were included to satisfy Criterion 4 that are applicable to the protection of fuel stored in the SFPs. No Criterion 4 LCOs are required in the CR-3 PDTS. Each LCO in the CR-3 ITS was evaluated for applicability to a permanently defueled plant, and its significance to public health and safety. If an LCO is not applicable for the protection of nuclear fuel stored in the SFPs, it was not included in the proposed CR-3 PDTS. The following table identifies each LCO and the basis for removal or retention in the CR-3 PDTS:

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 14 of 44 CR-3 Improved Proposed Change and Basis Technical Specification LCOs 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 This specification defines the minimum shutdown margin in the Shutdown Margin reactor core for Modes 3, 4, and 5. Shutdown Margin was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.1.1 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.1.2 This specification defines the required accuracy for measured vs. Reactivity Balance predicted core reactivity balance. Reactivity balance was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.1.2 is not included in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.1.3 This specification defines the limits for reactor moderator Moderator Temperature temperature coefficient for the reactor core. MTC was included in Coefficient (MTC) ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.1.3 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.1.4 This specification defines the limits for reactor control rod CONTROL ROD Group alignment within each control rod group. CONTROL ROD Group Alignment Limits Alignment Limits were included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii) for fuel in the reactor core. LCO 3.1.4 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled and control rods are stored fully inserted in fuel assemblies in the fuel pools. This LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.1.5 This specification identifies that safety rods must be fully withdrawn Safety Rod Insertion in Modes 1 and 2. Safety Rod Insertion Limits were included in Limits ITS to satisfy Criterion 2 and Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.1.5 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled and control rods are stored fully inserted in fuel assemblies in the fuel pools. This LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.1.6 This specification defines the limits for axial power shaping control AXIAL POWER SHAPING rod alignment. APSR alignment limits were included in ITS to ROD (APSR) Alignment satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.1.6 is not proposed for inclusion in the PDTS since CR-3 is permanently

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 15 of 44 Limits shutdown and defueled and APSRs are stored fully inserted in fuel assemblies in the fuel pools. This LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.1.7 This specification defines the operability requirements for control Position Indicator rod absolute position indicator and relative position indicator Channels channels for control rods and APSRs. Control rod and APSR position indicator channels were included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii) for fuel in the reactor core. LCO 3.1.7 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled and there are no control rods or APSRs in use. This LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.1.8 This specification identifies conditions under which other LCOs PHYSICS TESTS may be suspended during physics testing. Physics Tests were Exception - MODE 1 included in ITS to satisfy Criterion 2 and Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.1.8 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and no Physics Tests can be performed. 3.1.9 This specification identifies conditions under which other LCOs PHYSICS TESTS may be suspended during physics testing. Physics Tests were Exception - MODE 2 included in ITS to satisfy Criterion 2 and Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.1.9 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and no Physics Tests can be performed. 3.2 POWER DISTRIBUTION LIMITS 3.2.1 This specification defines the insertion, sequence and overlap Regulating Rod Insertion limits for regulating control rods. The regulating rod insertion limits Limits were included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.2.1 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.2.2 This specification identifies that the APSRs will be positioned AXIAL POWER SHAPING according to the limits in the Core Operating Limits Report (COLR). ROD (APSR) Insertion The APSR Insertion Limits were included in ITS to satisfy Criterion Limits 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.2.2 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.2.3 This specification identifies that Axial Power Imbalance in the core AXIAL POWER shall be maintained within the acceptable operating limits specified IMBALANCE Operating in the COLR. The Axial Power Imbalance Operating Limits were

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 16 of 44 Limits included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.2.3 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.2.4 QUADRANT POWER TILT (QPT) This specification identifies that QPT shall be maintained less than or equal to the steady state limits specified in the COLR. The QPT was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.2.4 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.2.5 This specification identifies that FQ(Z) and F N shall be within the Power Peaking Limits AH limits specified in the COLR. The Power Peaking Limits were included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.2.5 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.3 INSTRUMENTATION 3.3.1 This specification identifies the requirements for the operability of Reactor Protection RPS channels for each RPS function as specified in the associated System (RPS) Table 3.3.1-1. The RPS functions were included in ITS to satisfy Instrumentation Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.3.1 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. Likewise Table 3.3.1-1 will not be included in the PDTS. 3.3.2 This specification identifies the conditions under which the RPS Reactor Protection manual trip function shall be operable. The ITS Bases do not System (RPS) Manual identify a 50.36 Criterion for the inclusion of this LCO. This Reactor Trip function provides a backup to the automatic trip function of LCO 3.3.1. LCO 3.3.2 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.3.3 This specification identifies the requirements for the operability of Reactor Protection the RTMs. The RTMs were included in ITS to satisfy Criterion 3 of System (RPS) - Reactor 10 CFR 50.36(c)(2)(ii). LCO 3.3.3 is not proposed for inclusion in Trip Module (RTM) the PDTS since CR-3 is permanently shutdown and defueled. 3.3.4 This specification identifies the conditions under which the CRD CONTROL ROD DRIVE Trip Devices shall be operable. The CRD Trip Devices were (CRD) Trip Devices included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.3.4 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.3.5 This specification identifies the requirements for operability of

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 17 of 44 Engineered Safeguards Actuation System (ESAS) Instrumentation ESAS Reactor Coolant System pressure and Reactor Building pressure instrumentation according to associated Table 3.3.5-1. The ESAS Instrumentation Channels were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.3.5 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. Likewise Table 3.3.5-1 will not be included in the PDTS. 3.3.6 This specification identifies the requirements for operability of Engineered Safeguards manual initiation channels of ESAS functions. The ESAS manual Actuation System (ESAS) initiation instrumentation functions were included in ITS to satisfy Manual Initiation Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.3.6 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.3.7 This specification identifies the requirements for ESAS automatic Engineered Safeguards actuation logic matrices to be operable. The ESAS automatic Actuation System (ESAS) actuation logics were included in ITS to satisfy Criterion 3 of 10 Automatic Actuation CFR 50.36(c)(2)(ii). LCO 3.3.7 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.3.8 This specification identifies the conditions under which the loss of Emergency Diesel voltage and degraded voltage function channels are required to be Generator (EDG) Loss of operable. The EDG LOPS Instrumentation was included in ITS to Power Start (LOPS) satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.3.8 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled and EDGs are no longer required to auto start in this condition. 3.3.9 This specification identifies the requirements for source range Source Range Neutron neutron flux channels to be operable. The ITS Bases do not Flux identify a 50.36 Criterion for the inclusion of this LCO. The Source Range Neutron Flux monitors provided a means to monitor core reactivity changes to trigger operator actions to respond to reactivity transients initiated during conditions when the RPS was not required to be operable. LCO 3.3.9 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.3.10 This specification identifies the requirements for intermediate Intermediate Range range neutron flux channels to be operable. The ITS Bases do not Neutron Flux identify a 50.36 Criterion for the inclusion of this LCO. Intermediate Range Neutron Flux channels were necessary to monitor core reactivity changes during reactor startup. LCO 3.3.10 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs.

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 18 of 44 3.3.11 This specification identifies the requirements for the EFIC System Emergency Feedwater instrumentation channels to be operable in accordance with Initiation and Control associated Table 3.3.11-1. The EFIC System Instrumentation was (EFIC) System included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). Instrumentation LCO 3.3.11 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. Likewise Table 3.3.11-1 is not proposed for inclusion in the PDTS. 3.3.12 This specification identifies the requirements for the manual Emergency Feedwater initiation switches to be operable for each EFIC function. The ITS Initiation and Control Bases do not identify a 50.36 Criterion for the inclusion of this (EFIC) Manual Initiation LCO. The EFIC Manual Initiation Functions were required by design as backups to the automatic initiation functions. LCO 3.3.12 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.3.13 This specification identifies the requirements for the automatic Emergency Feedwater actuation logic channels to be operable for each EFIC function. Initiation and Control The EFIC logic channels were included in ITS to satisfy Criterion 3 (EFIC) Automatic of 10 CFR 50.36(c)(2)(ii). LCO 3.3.13 is not proposed for inclusion Actuation Logic in the PDTS since CR-3 is permanently shutdown and defueled. 3.3.14 This specification identifies the requirements for the operability of Emergency Feedwater the EFIC vector valve logic channels. The EFIC vector valve logic Initiation and Control channels were included in ITS to satisfy Criterion 3 of 10 CFR (EFIC) - Emergency 50.36(c)(2)(ii). LCO 3.3.14 is not proposed for inclusion in the Feedwater (EFW) - Vector PDTS since CR-3 is permanently shutdown and defueled. Valve Logic 3.3.15 This specification identifies the requirements for one RB purge Reactor Building (RB) channel to be operable. The ITS Bases do not identify a 50.36 Purge isolation - High Criterion for the inclusion of this LCO. RB Purge Isolation - High Radiation Radiation was included in ITS to ensure safety analysis assumptions regarding RB isolation are bounded. LCO 3.3.15 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled and the inventory of radioactivity in the RB is significantly reduced compared to an operating plant. 3.3.16 Deleted by Amendment 199. Control Room Isolation - High Radiation 3.3.17 This specification identifies the PAM instrumentation that must be Post Accident Monitoring operable as shown in Table 3.3.17. PAM Instrumentation that (PAM) Instrumentation displays Type A variables were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). Non-Type A PAM instrumentation was included in ITS because it was considered important to reducing

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 19 of 44 public risk. LCO 3.3.17 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. Likewise Table 3.3.17-1 will not be included in the PDTS. 3.3.18 This specification identifies the Remote Shutdown System Remote Shutdown functions that must be operable as shown in Table 3.3.18-1. The System Remote Shutdown System was included in ITS to satisfy Criterion 4 of 10 CFR 50.36(c)(2)(ii). LCO 3.3.18 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. Likewise Table 3.3.18-1 will not be included in the PDTS. 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 This specification identifies that parameters for loop pressure, hot RCS Pressure, leg temperature, and RCS total flow shall be within limits for the Temperature, and Flow number of reactor coolant pumps in operation. RCS DNB limits Departure from Nucleate were included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). Boiling (DNB) Limits LCO 3.4.1 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.4.2 This specification identifies that each RCS loop average RCS Minimum temperature shall be a 5250F. RCS minimum temperature for Temperature for Criticality criticality was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.2 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.4.3 This specification identifies that RCS pressure, RCS temperature, RCS Pressure and and RCS heatup and cooldown rates be maintained within the Temperature (P/T) Limits limits specified in the Pressure and Temperature Limits Report. RCS P/T Limits were included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.3 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.4.4 This specification requires that two RCS loops shall be operable RCS Loops - MODE 3 and at least one in operation in Mode 3. RCS Loops - Mode 3 was included in ITS since it most closely satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.4 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.4.5 This specification requires that two loops consisting of any

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 20 of 44 RCS Loops - MODE 4 combination of RCS loops and decay heat removal loops shall be operable and at least one be in operation. RCS Loops - Mode 4 was included in ITS since it most closely satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.5 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.4.6 This specification requires one decay heat removal loop shall be RCS Loops - MODE 5, operable and in operation and either a) one additional decay heat Loops Filled loop operable, or b) one steam generator operable. RCS Loops - Mode 5 - Loops Filled was included in ITS to satisfy Criterion 4 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.6 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.4.7 This specification requires two decay heat removal loops shall be RCS Loops - MODE 5, operable and at least one loop in operation. RCS Loops - Mode 5 Loops Not Filled - Loops Not Filled was included in ITS to satisfy Criterion 4 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.7 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.4.8 Pressurizer This specification identifies when the pressurizer is required to be operable. It defines operability by a minimum required level and minimum heater capability. Pressurizer operability was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.8 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.4.9 This specification identifies when the two pressurizer safety valves Pressurizer Safety Valves are required to be operable. Pressurizer Safety Valves were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.9 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.4.10 This specification identifies when the pressurizer power operated Pressurizer Power relief valve is required to be operable. The PORV was included in Operated Relief Valve ITS to satisfy Criterion 4 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.10 is (PORV) not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.4.11 This specification identifies the conditions in which the LTOP Low Temperature system is required to be operable, and the system configuration Overpressure Protection required. The ITS Bases do not identify a 50.36 Criterion for the (LTOP) System inclusion of this LCO. The LTOP System was provided to protect the RCS from overpressurization transients during shutdown, in part, by providing a sufficient size RCS vent. In permanent

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 21 of 44 shutdown, the RCS is partially drained and adequately vented to prevent overpressurization. LCO 3.4.11 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.4.12 This specification identifies the limits on RCS operational leakage. RCS Operational Leakage The RCS Operational Leakage LCO was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.12 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.4.13 This specification identifies the maximum leakage for any PIV and RCS Pressure Isolation the requirement for the Automatic Closure and Interlock System Valve (PIV) Leakage (ACIS) to be operable. RCS PIV Leakage was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.13 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.4.14 This specification identifies the RCS leakage detection instruments RCS Leakage Detection that are required to be operable and in what conditions. The RCS Instrumentation Leakage Detection Instrumentation was included in ITS to satisfy Criterion 1 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.14 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.4.15 This specification requires that the specific activity of the reactor RCS Specific Activity coolant be within limits as provided on associated Figure 3.4.15-1. RCS Specific Activity was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.15 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. Likewise Figure 3.4.15-1 will not be included in the PDTS. 3.4.16 This specification requires that OTSG tube integrity be maintained Steam Generator (OTSG) and that all OTSG tubes satisfying the tube repair criteria be Tube Integrity plugged or repaired. OTSG Tube Integrity was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.4.16 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 This specification identifies the operating conditions under which Core Flood Tanks (CFTs) the CFTs are required to be operable. The CFTs were included in

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 22 of 44 ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.5.1 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.5.2 This specification identifies that two trains of ECCS shall be ECCS - Operating operable in operating Modes 1, 2, and 3. The ECCS Trains - Operating were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.5.2 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.5.3 This specification identifies that one train of ECCS shall be ECCS - Shutdown operable in Mode 4. The ITS Bases do not identify a 50.36 Criterion for the inclusion of this LCO. LCO 3.5.3 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.5.4 This specification identifies that the BWST shall be operable in Borated Water Storage operating Modes 1 through 4. The BWST was included in ITS to Tank (BWST) satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.5.4 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.6 CONTAINMENT SYSTEMS 3.6.1 This specification identifies that the containment shall be operable Containment in Modes 1 through 4. The containment was included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.6.1 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.6.2 This specification identifies the conditions under which the Containment Air Locks containment air locks are required to be operable. The containment air locks were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.6.2 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.6.3 This specification identifies that each containment isolation valve Containment Isolation must be operable in Modes 1 - 4. The containment isolation Valves valves were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.6.3 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.6.4 This specification identifies the limits on containment pressure Containment Pressure during operating Modes. Containment pressure was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.6.4 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 23 of 44 for the cladding of fuel stored in the SFPs. 3.6.5 This specification identifies the upper limit on containment air Containment Air temperature during operating Modes. The containment air Temperature temperature was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.6.5 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.6.6 Reactor Building This specification identifies the requirements for building spray Spray and Containment trains and containment cooling trains to be operable in Modes 1 Cooling Systems through 4. Reactor building spray and containment cooling trains were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.6.6 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.6.7 This specification requires the CPCS to be operable in Modes 1 Containment Emergency through 4. The CPCS was included in ITS to satisfy Criterion 3 of Sump pH Control System 10 CFR 50.36(c)(2)(ii). LCO 3.6.7 is not proposed for inclusion in (CPCS) the PDTS since CR-3 is permanently shutdown and defueled. 3.7 PLANT SYSTEMS 3.7.1 This specification identifies that the MSSVs shall be operable in Main Steam Safety Valves accordance with the associated Table 3.7.1-1. The MSSVs were (MSSVs) included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.1 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. Likewise Table 3.7.1-1 will not be included in the PDTS. 3.7.2 This specification identifies that each MSIV is required to be Main Steam Isolation operable in Modes 1 through 3. The MSIVs were included in ITS Valves (MSIVs) to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.2 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.3 This specification identifies the requirements for MFIV operability Main Feedwater Isolation and the capability for main feedwater isolation within the required Valves (MFIVs) time. The MFIVs were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.3 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.4 This specification identifies that each TBV is required to be Turbine Bypass Valves operable in Modes 1 through 3. The TBVs were included in ITS to (TBVs) satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.4 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled.

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 24 of 44 3.7.5 Emergency Feedwater (EFW) System This specification identifies that two EFW trains must be operable in Modes 1 through 3. The EFW System was included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.5 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.6 This specification identifies the minimum water volume to be Emergency Feedwater contained in the EFW tank. The EFW tank was included in ITS to (EFW) Tank satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.6 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.7 This specification identifies the requirements for operability of Nuclear Services Closed emergency SW pumps and SW heat exchangers. The Nuclear Cycle Cooling Water (SW) Services Closed Cycle Cooling Water System was included in ITS System to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.7 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.8 This specification identifies that two trains of DC must be operable Decay Heat Closed Cycle in Modes 1 through 4. The Decay Heat Closed Cycle Cooling Cooling Water (DC) Water System was included in ITS to satisfy Criterion 3 of 10 CFR System 50.36(c)(2)(ii). LCO 3.7.8 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.9 This specification identifies that two Nuclear Services Seawater Nuclear Services trains must be operable in Modes 1 through 4. The Nuclear Seawater System Services Seawater System was included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.9 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.10 This specification identifies that two Decay Heat Seawater trains Decay Heat Seawater must be operable in Modes 1 through 4. The Decay Heat System Seawater System was included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.10 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.11 This specification identifies that the Ultimate Heat Sink must be Ultimate Heat Sink (UHS) operable in Modes 1 through 4. The UHS temperature limit was included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.11 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.12 This specification identifies that the two CREVS trains must be Control Room Emergency operable in Modes 1 through 4. The CREVS was included in ITS Ventilation Systems to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii) since long term control room habitability is essential to mitigate accidents resulting

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 25 of 44 (CREVS) in atmospheric fission product release. LCO 3.7.12 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. See the accident analyses discussion at the end of this table. See Reference 6.3 for more details + 3.7.13 Fuel Storage Pool Water Level This specification is being retained since it meets the assumptions of the Fuel Handling Accident in the Final Safety Analysis Report for iodine removal efficiency. Pool level also provides shielding to reduce the general area radiation dose during both spent fuel handling and storage. Fuel Storage Pool Water Level continues to meet Criterion 2 of 10 CFR 50.36(c)(2)(ii) as this LCO preserves the current requirements for safe storage of irradiated fuel and provides additional protection to plant personnel. A note in the Required Action section is being removed. The note addresses the non-applicability of LCO 3.0.3 which is proposed for elimination from the PDTS with this LAR. 3.7.14 Spent Fuel Pool Boron Concentration This specification is being retained since it ensures ke, of < 0.95 in an unlikely event such as 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. The reduction in ke,, caused by the boron more than offsets the reactivity addition caused by credible mis-loading events. SFP Boron Concentration continues to meet Criterion 2 of 10 CFR 50.36(c)(2)(ii) as this LCO will prevent inadvertent criticality in the pools and preserves the current requirements for safe storage of irradiated fuel. A note in the Required Action section is being removed. The note addresses the non-applicability of LCO 3.0.3 which is proposed for elimination from the PDTS with this LAR. +/- 3.7.15 Spent Fuel Assembly Storage This specification is being retained. 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 and continues to meet Criterion 2 of 10 CFR 50.36(c)(2)(ii). The spent fuel assembly enrichment requirements in this LCO are required to ensure inadvertent criticality does not occur in the spent fuel pool. This LCO preserves the current requirements for safe storage of irradiated fuel. A note in the Required Action section is being removed. The note addresses the non-applicability of LCO 3.0.3 which proposed for elimination from the PDTS with this LAR. 3.7.16 Secondary Specific This specification identifies a limit on specific radioactivity of the secondary coolant as measured by dose equivalent 1-131. The ITS Bases do not identify a 50.36 Criterion for the inclusion of this

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 26 of 44 Activity LCO. However, the purpose is to limit offsite doses due to secondary side releases. Since CR-3 is permanently defueled, there are no longer primary or secondary inventories of any significant volume and no consequential energy to drive a release. LCO 3.7.16 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.17 Steam Generator Level This specification identifies that the water level in each steam generator must be less than or equal to the level shown in associated Figure 3.7.17-1. The ITS Bases do not identify a 50.36 Criterion for the inclusion of this LCO. However, the purpose is to limit secondary side steam generator inventory for a steam line break accident. Since CR-3 is permanently defueled, there are no longer primary or secondary inventories of any significant volume and no consequential energy to cause a steam line break. LCO 3.7.17 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. Likewise Figure 3.7.17-1 will not be included in the PDTS. 3.7.18 This specification identifies that the two control complex cooling Control Complex Cooling trains must be operable in Modes 1 through 4. The Control System Complex Cooling System was included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.18 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.7.19 This specification requires that stored diesel fuel oil, lube oil and Diesel Driven EFW (DD-starting air subsystems must be within limits stated in the EFW) Pump Fuel Oil, specification conditions. The DD-EFW Pump Fuel Oil, Lube Oil Lube Oil and Starting Air and Starting Air specifications were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.7.19 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 This specification identifies the requirements for maintaining AC Sources - Operating qualified offsite transmission circuits, the onsite Class 1 E electrical power distribution system and diesel generators to be operable in Modes 1 through 4. AC Sources - Operating were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.1 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.8.2 This specification identifies the requirements for maintaining a AC Sources - Shutdown qualified offsite transmission circuit, portions of the onsite Class 1 E electrical power distribution system required by specification 3.8.10, and one emergency diesel generator to be operable in

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 27 of 44 Modes 5 and 6. AC Sources - Shutdown were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.2 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.8.3 This specification requires that stored diesel fuel oil, lube oil and Diesel Fuel Oil, Lube Oil starting air subsystems must be within limits stated in the and Starting Air specification conditions. The Diesel Fuel Oil, Lube Oil and Starting Air specifications were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.3 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.8.4 This specification requires that two trains of DC electrical power be DC Sources - Operating operable in Modes 1 through 4. DC Sources - Operating were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.4 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.8.5 This specification identifies that DC electrical subsystems must be DC Sources - Shutdown operable to support the DC electrical distribution subsystems required by specification 3.8.10. DC Sources - Shutdown were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.5 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.8.6 This specification requires that the battery cell parameters for the Battery Cell Parameters two battery trains must be within the limits of the associated Table 3.8.6-1. Battery Cell Parameters were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.6 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. Likewise Table 3.8.6-1 will not be included in the PDTS. 3.8.7 This specification requires that two A train and two B train inverters Inverters - Operating be operable in Modes 1 through 4. Inverters - Operating were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.7 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.8.8 This specification requires that inverters needed to support the Inverters - Shutdown onsite ClasslE AC Vital bus electrical power distributions subsystem required by specification 3.8.10 be operable. Inverters - Shutdown were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.8 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.8.9 This specification requires that Train A and Train B AC, DC, and Distribution Systems - AC Vital bus electrical power distribution subsystems be operable Operating in Modes 1 through 4. Distribution Systems - Operating were

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 28 of 44 included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.9 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.8.10 This specification requires that the necessary portion of AC, DC, Distribution Systems - and AC Vital bus electrical power distribution subsystems be Shutdown operable to support equipment required to be operable in Modes 5 and 6. Distribution Systems - Shutdown were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.8.10 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.9 REFUELING OPERATIONS 3.9.1 This specification identifies that boron concentrations in the RCS Boron Concentrations and refueling canal be maintained within the limit specified in the COLR. RCS Boron Concentration was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). LCO 3.9.1 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. 3.9.2 This specification requires that two Source Range Neutron Flux Nuclear Instrumentation Monitors must be operable in Mode 6. Source Range Nuclear Flux Monitors were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.9.2 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.9.3 This specification establishes the controls required for containment Containment Penetrations penetrations during movement of recently irradiated fuel within the containment. Containment Penetration controls were included in ITS to satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). LCO 3.9.3 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled. 3.9.4 This specification requires that one DHR loop shall be in operation Decay Heat Removal when the refueling canal level is > 156 ft plant datum. In the (DHR) and Coolant refueling mode, DHR was included in ITS to satisfy Criterion 4 of Circulation - High Water 10 CFR 50.36(c)(2)(ii). LCO 3.9.4 is not proposed for inclusion in Level the PDTS since CR-3 is permanently shutdown and defueled. 3.9.5 This specification requires that two DHR loops be operable and at Decay Heat Removal least one DHR loop shall be in operation when the refueling canal (DHR) and Coolant level is < 156 ft plant datum. In the refueling mode, DHR was Circulation - Low Water included in ITS to satisfy Criterion 4 of 10 CFR 50.36(c)(2)(ii). Level LCO 3.9.5 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled.

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 29 of 44 3.9.6 This specification requires the refueling canal water level be Refueling Canal Water maintained at > 156 ft plant datum during movement of irradiated Level fuel assemblies within containment. Refueling Canal Water Level was included in ITS to satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii) during refueling activities. LCO 3.9.6 is not proposed for inclusion in the PDTS since CR-3 is permanently shutdown and defueled, and this LCO does not provide protection for the cladding of fuel stored in the SFPs. Supporting Accident Analyses In Reference 6.3, DEF provided accident analysis summaries for postulated accidents as well as a summary of atmospheric dispersion factor derivation for those accidents. Reference 6.3 is also proposing a Radioactive Waste Handling Accident to be considered a new design basis accident for CR-3. Fuel Handling Accident The Fuel Handling Accident calculation determined the projected dose from the rupture of all 208 fuel pins in a fuel assembly due to falling onto the spent fuel racks. The Control Complex Ventilation system is assumed to remain operating in its normal configuration of outside air intake and discharge. This is as opposed to swapping to the designed accident response of automatic or manual isolation of intake and discharge and placing the HEPA and charcoal filters in service. No equipment is required to mitigate the effects of this event. The results of the calculation predicts a Control Room dose of 1.3E-04 rem TEDE compared to the 10 CFR 50.67 limit of 5 rem TEDE. Radioactive Waste Handling Accident The proposed Radioactive Waste Handling Accident is the rupture of a high integrity cask (HIC) containing used demineralizer resin. A conservative source term was developed based on the resin shipment from the last 5 1/2 years with the highest curie content excluding Cobalt 60 and added to it was the highest Cobalt 60 content from the same group of waste shipments. Since no CR-3 equipment is credited with providing any mitigating actions, no additional LCOs or Programs are required to be retained for mitigating this accident. 3.2.2 Use and Application The Use and Application section is being proposed for revision to retain and/or revise only those Definitions, Logical Connectors, Completion Times, and Frequency descriptions that are used in the three LCOs that are being retained in the CR-3 PDTS.

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 30 of 44 1.0 USE AND APPLICATION 1.1 Definitions Term Definition ACTIONS No Change ALLOWABLE THERMAL This definition is not proposed for inclusion in the PDTS since the POWER term is not used in any PDTS specification. CR-3 is not allowed to operate at any thermal power AXIAL POWER This definition is not proposed for inclusion in the PDTS since the IMBALANCE term is not used in any PDTS specification. This term only has meaning for an operating reactor core. AXIAL POWER SHAPING This definition is not proposed for inclusion in the PDTS since the RODS (APSRS) term is not used in any PDTS specification. APSRS remain stored in the fuel pools, but are not relied upon to perform any design function. CHANNEL CALIBRATION This definition is not proposed for inclusion in the PDTS since the term is not used in any PDTS specification. There is no instrumentation credited in the analyses of the accidents that remain credible. CHANNEL CHECK This definition is not proposed for inclusion in the PDTS since the term is not used in any PDTS specification. There is no instrumentation credited in the analyses of the accidents that remain credible. CHANNEL FUNCTIONAL This definition is not proposed for inclusion in the PDTS since the TEST term is not used in any PDTS specification. There is no instrumentation credited in the analyses of the accidents that remain credible. CONTROL RODS This definition is not proposed for inclusion in the PDTS since the term is not used in any PDTS specification. CONTROL RODS remain stored in the fuel pools, but are not relied upon to perform any design function. CORE ALTERATION This definition is not proposed for inclusion in the PDTS since the term is not used in any PDTS specification. This term has no meaning when there is no reactor core. CORE OPERATING This definition is not proposed for inclusion in the PDTS since the LIMITS REPORT (COLR) term is no longer used and the program that required the COLR has been eliminated from the PDTS. DOSE EQUIVALENT This definition is not proposed for inclusion in the PDTS since the 1-131 term is not used in any PDTS specification. This term is used to

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 31 of 44 express dose from a mixture of iodine isotopes created in an operating core and contained in plant primary or secondary coolant. The value of DOSE EQUIVALENT 1-131 was used previously for dose analysis of accidents involving the primary and secondary coolant releases. Those accident conditions no longer apply to the permanently shutdown and defueled plant. E-AVERAGE This definition is not proposed for inclusion in the PDTS since the DISINTEGRATION term is not used in any PDTS specification. This value was ENERGY calculated to determine radioactivity in the reactor coolant system for the purpose of estimating dose to the public from a steam generator tube rupture. This accident is not longer possible. EFFECTIVE FULL This definition is not proposed for inclusion in the PDTS since the POWER DAY (EFPD) term is not used in any PDTS specification. EMERGENCY This definition is not proposed for inclusion in the PDTS since the FEEDWATER INITIATION term is not used in any PDTS specification. The EFIC system has AND CONTROL (EFIC) no function in the permanently shutdown plant. RESPONSE TIME ENGINEERED SAFETY This definition is not proposed for inclusion in the PDTS since the FEATURE (ESF) term is not used in any PDTS specification. The ESF equipment RESPONSE TIME has no function in the permanently shutdown plant. LEAKAGE This definition is not proposed for inclusion in the PDTS since the term is not used in any PDTS specification. This definition was used to specify different sources of leakage from the Reactor Coolant System that could limit operation. This limitation no longer applies to a permanently shutdown plant. MODE and This definition and table are not proposed for inclusion in the PDTS Tablel.1-1 MODES since operating modes are not used in any PDTS specification. Modes are defined operating or refueling conditions as contained in Table 1.1-1. These terms do not apply to a condition with all fuel in the SFPs. NUCLEAR HEAT FLUX This definition is not proposed for inclusion in the PDTS since the HOT CHANNEL FACTOR term is not used in any PDTS specification. This definition only (FQ(Z)) applies to conditions within an operating reactor core. NUCLEAR ENTHALPY This definition is not proposed for inclusion in the PDTS since the RISE HOT CHANNEL term is not used in any PDTS specification. This definition only FACTOR (F'H) applies to conditions within an operating reactor core. OPERABLE - This definition is not proposed for inclusion in the PDTS since the OPERABILITY term is not used in any PDTS specification. There are no systems or components required to be operable in the PDTS.

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 32 of 44 PHYSICS TESTS This definition is not proposed for inclusion in the PDTS since the term is not used in any PDTS specification. These tests are only applicable to a reactor core undergoing startup. PRESSURE AND This definition is not proposed for inclusion in the PDTS since the TEMPERATURE LIMITS term is no longer used and the program that required the PTLR REPORT (PTLR) has been eliminated from the PDTS. QUADRANT POWER This definition is not proposed for inclusion in the PDTS since the TILT (QPT) term is not used in any PDTS specification. This definition only applies to conditions within an operating reactor core. RATED THERMAL This definition is not proposed for inclusion in the PDTS since the POWER (RTP) term is not used in any PDTS specification. This definition only applies to an operating reactor core. REACTOR PROTECTION This definition is not proposed for inclusion in the PDTS since the SYSTEM (RPS) term is not used in any PDTS specification. The RPS system has RESPONSE TIME no function in the permanently shutdown plant SHUTDOWN MARGIN This definition is not proposed for inclusion in the PDTS since the (SDM) term is not used in any PDTS specification. This definition only applies to a loaded reactor core. STAGGERED TEST This definition is not proposed for inclusion in the PDTS since the BASIS term is not used in any PDTS specification. This definition applies to the performance of surveillance tests on systems with multiple subsystems or channels. There are no surveillance requirements in the PDTS for operating systems. THERMAL POWER This definition is not proposed for inclusion in the PDTS since the term is not used in any PDTS specification. This definition only applies to conditions within an operating reactor core. 1.2 LOGICAL CONNECTORS - Being retained as is. 1.3 COMPLETION TIMES BACKGROUND This is proposed for revision to remove reference to plant operation and replace it with reference to the management of irradiated fuel. DESCRIPTION This explanation is proposed for revision to remove discussion of conditions that will not exist in a permanently defueled plant. EXAMPLES This section is not proposed for inclusion. The examples are no longer necessary that describe the Completion Times that do not remain in the PDTS. The Actions that remain in the PDTS must all be completed IMMEDIATELY. IMMEDIATE Remains unchanged.

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 33 of 44 COMPLETION TIME 1.4 FREQUENCY DESCRIPTION This explanation is proposed for revision to remove discussion of surveillance performance situations that do not exist in the PDTS. EXAMPLES This section is proposed for revision to remove discussion of surveillance performance situations that do not exist in the PDTS, and to explicitly address those that do exist. 3.2.3 Safety Limits 10 CFR 50.36(c)(1), 'Safety limits, limiting safety system settings, and limiting control settings,' states the following in sub-paragraph (i)(A): "Safety limits for nuclear reactors are limits upon important process variables that are found to be necessary to reasonably protect the integrity of certain of the physical barriers that guard against the uncontrolled release of radioactivity. If any safety limit is exceeded, the reactor must be shut down." The Safety Limits apply only to the reactor core and reactor coolant system pressure. None of the Safety Limits apply to fuel stored in the SFPs. Since these limits only apply to the plant while in operation, they are proposed for removal from the PDTS. The proposed change also removes Figure 2.1.1-1 which defines Safety Limit 2.1.1.3. The actions for Safety Limit violations are also proposed for removal along with the Safety Limits. 3.2.4 Limiting Condition for Operation (LCO) Applicability and Surveillance Requirement (SR) Applicability LCO and SR Applicability conditions that only apply to an operating plant are proposed for elimination. LCO and SR Applicability conditions that can apply to a permanently defueled plant are being reworded to remove references to plant operating conditions. 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 Proposed for revision to be consistent with the permanently defueled condition by removing reference to operational modes. LCO 3.0.2 Proposed for revision to be consistent with the permanently defueled condition by removing reference to 3.0.x LCOs that are proposed for removal from the PDTS. LCO 3.0.3 This LCO requires placing the plant in a condition that is applicable to an operating plant. LCO 3.0.3 is not proposed for inclusion in the PDTS since it contains actions that cannot be met for a permanently defueled plant. LCO 3.0.4 This LCO prohibits entering an operational mode or specified

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 34 of 44 applicability condition unless the LCOs are met. LCO 3.0.4 is not proposed for inclusion in the PDTS since all actions in the PDTS have a completion time of "Immediately." This makes LCO 3.0.4 unnecessary. LCO 3.0.5 This LCO allows equipment removed from service or declared inoperable to be returned to service for the purpose of testing. LCO 3.0.5 is not proposed for inclusion in the PDTS since there are no LCOs for equipment to be operable or in operation in the PDTS. LCO 3.0.6 This LCO addresses the actions required for a supported system when the support system LCO is not met. LCO 3.0.6 is not proposed for inclusion in the PDTS since there are no LCOs for equipment to be operable or in operation in the PDTS. LCO 3.0.7 This LCO addresses Physics Tests Exception LCOs (3.1.8 and 3.1.9). LCO 3.0.7 is not proposed for inclusion in the PDTS since Physics Tests are not applicable to a permanently defueled plant. LCO 3.0.8 This LCO addresses the actions available when a required snubber is found to be unable to perform its function. LCO 3.0.8 is not proposed for inclusion in the PDTS since there are no LCOs for equipment to be operable or in operation in the PDTS. 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 This SR identifies the requirements for performing SRs and the conditions for failing to satisfy the SR. SR 3.0.1 is proposed for revision to remove references to operating modes and inoperable equipment since there are no LCOs for equipment to be operable or in operation in the PDTS. SR 3.0.2 This SR provides an allowance for extending the frequency for performance of a SR to 1.25 times the nominal frequency. SR 3.0.2 is proposed for revision to remove conditions for frequencies that do not exist in PDTS LCOs. SR 3.0.3 SR 3.0.3 will remain unchanged. SR 3.0.4 This SR establishes restrictions for changing operational modes or entering specified conditions when SRs are not satisfied. SR 3.0.4 is proposed for revision to remove references to operating modes. 3.2.4 Design Features Design Feature 4.1, 'Site,' is proposed for a minor revision to remove the word "emergency" in the statement that the site has an '...ample supply of emergency power...' In the permanently shutdown and defueled plant and proposed PDTS there are no requirements for emergency

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 35 of 44 power supplies. No equipment is credited to operate to mitigate design basis accidents; therefore, no emergency power is required. Furthermore, the Loss of Spent Fuel Pool Normal Cooling event described in Reference 6.3 demonstrates that more than 107 hours are available before the SFPs heat up to 212 0F, providing an abundance of time to restore cooling. Design Feature 4.2, 'Reactor Core,' is proposed for elimination since it describes the design of the CR-3 reactor core fuel assemblies and the control rods. These descriptions are no longer applicable since all fuel has been permanently removed from the reactor vessel. Design Feature 4.3, 'Fuel Storage,' is being retained without changes since it will continue to apply to CR-3 in the permanently defueled condition and with fuel stored in the SFPs. 3.2.5 Administrative Controls CR-3 LAR #313, Revision 1, dated September 4, 2013 (Reference 6.4) proposed changes to the Administrative Controls section of the ITS. When LAR #313, Revision 1 is approved, those changes will be retained in the PDTS. This LAR proposes additional changes to Administrative Controls sub-sections 5.2, 'Organization,' 5.6, 'Procedures, Programs and Manuals,' and 5.7, 'Special Reports.' Only the additional proposed changes are described below. 5.0 ADMINISTRATIVE CONTROLS 5.2.1 Onsite and Offsite Organizations Commas have been added to sub item (c) to differentiate between the three classifications of personnel who are assured of organizational freedom to perform their assigned functions. This change is considered editorial only and does not change the meaning of the item. 5.6.2.3 In two places, dose rate limitations on releases of Iodine-1 31 and Offsite Dose Calculation Iodine-1 33 are proposed for removal. Calculations have Manual (ODCM) determined that due to the time period since the reactor was last operated (> 4 years), spent fuel no longer has these two isotopes present. See the Fuel Handling Accident summary in Section 3.2.1. For additional details on the accident analysis see Reference 6.3. 5.6.2.12 The VFTP is proposed for elimination from the PDTS since neither Ventilation Filter Testing the Auxiliary Building nor Control Complex ventilation filters are Program (VFTP) credited to provide any decontamination factor for the remaining postulated accidents for CR-3 in the permanently defueled condition. See the accident summaries in Section 3.2.1. For additional details on the accident analyses see Reference 6.3. 5.6.2.14 The Diesel Fuel Oil Testing Program is proposed for elimination Diesel Fuel Oil Testing from the PDTS since neither the emergency diesels nor the diesel Program driven emergency feedwater pump perform any safety function in the permanently shutdown and defueled plant. The emergency

U. S. Nuclear Regulatory 3F1013-01 Attachment A Page 36 of 44 diesel and emergency feedwater pump LCOs are proposed for removal from the PDTS. 5.6.2.16 Safety Function Determination Program (SFDP) The SFDP is proposed for elimination since none of the three LCOs remaining in the PDTS rely on the operability of any active equipment or systems to satisfy the LCO. 5.6.2.21 Control Complex Habitability Envelope Integrity Program The Control Complex Habitability Envelope Integrity Program is proposed for elimination from the PDTS since isolation of the habitability envelope is not credited in the dose analyses for the remaining postulated accidents. See the accident summaries in Section 3.2.1. For additional details on the accident analyses see Reference 6.3. 5.7.2 Special Reports are proposed for elimination since the only Special Reports remaining special report is due to the inoperability of Post Accident Monitoring (PAM) Instrumentation, and the PAM LCO is being proposed for elimination from the PDTS. 4.0 Regulatory Analysis 4.1 No Significant Hazards Consideration Determination Pursuant to 10 CFR 50.90, 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 CR-3 Facility Operating License (FOL) and associated Improved Technical Specifications (ITS) to the CR-3 Permanently Defueled Technical Specifications (PDTS) to reflect the permanent cessation of reactor operation. On February 20, 2013, CR-3 submitted a certification of permanent cessation of power operations pursuant to 10 CFR 50.82(a)(1)(i) (ADAMS Accession No. ML13056A005). By letter dated March 13, 2013, the NRC acknowledged CR-3's certification of permanent cessation of power operation and permanent removal of fuel from the reactor vessel. Accordingly, pursuant to 10 CFR 50.82(a)(2), the 10 CFR Part 50 license for CR-3 no longer authorizes operation of the reactor or emplacement or retention of fuel in the reactor vessel (ADAMS Accession No. ML13058A380). In support of this condition, the CR-3 FOL and associated ITS are being proposed for revision to be consistent with this permanently shutdown and defueled condition. The existing CR-3 ITS contain Limiting Condition for Operation (LCOs) that provide for appropriate functional capability of equipment required for safe operation of the facility, including the plant being in a defueled condition. Because the CR-3 Part 50 license no longer authorizes emplacement or retention of fuel in the reactor vessel, the LCOs (and associated Surveillance Requirements (SRs)) that do not apply in a defueled condition are being proposed for deletion. The remaining portions of the ITS are being proposed for revision, into the PDTS, to provide an acceptable level of safety derived from the reduced scope of postulated design basis accidents associated with a defueled plant, as described in the CR-3 safety analyses.

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 37 of 44 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?

Response: No CR-3 has permanently ceased operation. The proposed amendment would modify the CR-3 FOL and ITS by proposing to delete certain License Conditions (LCs) and ITS that are no longer applicable to a permanently defueled facility, while modifying the remaining portions to correspond to the permanently shutdown condition. Changes proposed to LCs will make them consistent with the non-operating status of CR-3. Other proposed LCs changes will eliminate LCs that were designed for one time implementation and have been satisfied, or are no longer required due to changes to Part 50 or Part 73 regulations that accomplish the same result or eliminate the requirement for the LC. The proposed changes to the ITS are consistent with the criteria set forth in 10 CFR 50.36 for the contents of ITS. Chapter 14 of the CR-3 Final Safety Analysis Report (FSAR) described the design basis accident (DBA) and transient scenarios applicable to CR-3 during power operations. With the reactor in a permanently defueled condition, the spent fuel pool and its cooling systems are dedicated only to spent fuel storage. In this condition, the spectrum of credible accidents is much smaller than for an operational plant. As a result of the certifications submitted by CR-3 in accordance with 10 CFR 50.82(a)(1), and the consequent removal of authorization to operate the reactor or to place or retain fuel in the reactor vessel in accordance with 10 CFR 50.82(a)(2), the majority of the accident scenarios originally postulated in the FSAR are no longer possible and have been removed from the FSAR under 10 CFR 50.59. 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 50.34(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 the termination of reactor operations at CR-3 and the permanent removal of the fuel from the reactor vessel (following 4 years of decay time after shutdown) and purging of the contents of the waste gas decay tanks, 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 (with the exception of the passive spent fuel pool structure).

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 38 of 44 The deletion of ITS definitions and rules of usage and application, that are currently not applicable in a defueled condition, has no impact on facility SSCs or the methods of operation of such SSCs. The deletion of design features and safety limits not applicable to the permanently shutdown and defueled status of CR-3 has no impact on the remaining DBA (the Fuel Handling Accident in the Auxiliary Building) or the proposed Radioactive Waste Handling Accident. The removal of LCOs or SRs that are related only to the operation of the nuclear reactor or accidents do not affect mitigation of the applicable DBAs previously evaluated since these DBAs are no longer applicable in the defueled mode. The safety functions involving core reactivity control, reactor heat removal, reactor coolant system inventory control, and containment integrity are no longer applicable at CR-3 as a permanently defueled plant. The analyzed accidents involving damage to the reactor coolant system, main steam lines, reactor core, and the subsequent release of radioactive material are no longer possible at CR-3. Since CR-3 has permanently ceased operation, the generation of fission products has ceased and the remaining source term will decay. The radioactive decay of the irradiated fuel since shutdown of the reactor have reduced the consequences of the Fuel Handling Accident (FHA) to levels well below those previously analyzed. The relevant parameter (water level) associated with the fuel pool provides an initial condition for the FHA analysis and is included in the PDTS. The spent fuel pool water level, spent fuel pool boron concentration, and spent fuel pool storage LCOs are retained to preserve the current requirements for safe storage of irradiated fuel. Fuel pool cooling and makeup related equipment and support equipment (e.g., electrical power systems) are not required to be continuously available since there is sufficient time to effect repairs, establish alternate sources of makeup flow, or establish alternate sources of cooling in the event of a loss of cooling and makeup flow to the spent fuel pool. The deletion and modification of provisions of the Administrative Controls do not directly affect the design of SSCs necessary for the safe storage of irradiated fuel or the methods used for handling and storage of such fuel in the fuel pool. Deletion of Programs are administrative in nature and do not affect any accidents applicable to the safe management of irradiated fuel or the permanently shutdown and defueled condition of the reactor. The proposed LC revisions reflect the CR-3 functions that are still authorized in the permanently defueled condition, and remove authorizations that suggest the reactor can be placed in operation. LCs that are being removed due to their one time applicability being previously satisfied have no bearing on future functions at CR-3. Other LCs are being removed that are not required by regulation for a permanently defueled and decommissioning plant. These changes cannot increase the probability or consequences of any accident that remains credible. The probability of occurrence of previously evaluated accidents is not increased, since extended operation in a defueled condition is the only operation currently allowed, and is therefore bounded by the existing analyses. Additionally, the occurrence of postulated accidents associated with reactor operation is no longer credible in a permanently defueled reactor. This significantly reduces the scope of applicable accidents. Therefore, the proposed amendment does not involve a significant increase in the probability or consequences of an accident previously evaluated.

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 39 of 44

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

Response: No The proposed changes have no impact on facility SSCs affecting the safe storage of irradiated fuel, or on the methods of operation of such SSCs, or on the handling and storage of irradiated fuel itself. The removal of ITS that are related only to the operation of the nuclear reactor or only to the prevention, diagnosis, or mitigation of reactor-related transients or accidents cannot result in different or more adverse failure modes or accidents than previously evaluated because the reactor is permanently shutdown and defueled, and CR-3 is no longer authorized to operate the reactor. The proposed deletion of requirements of the CR-3 ITS do not affect safe storage of nuclear fuel. The proposed PDTS continue to require proper control and monitoring of safety significant parameters. The proposed restriction on the fuel pool level is fulfilled by normal operating conditions and preserves initial conditions assumed in the analyses of the postulated DBA. The spent fuel pool water level, spent fuel pool boron concentration, and spent fuel pool storage LCOs are retained to preserve the current requirements for safe storage of irradiated fuel. The proposed amendment does not result in any new mechanisms that could initiate damage to the remaining relevant safety barriers for defueled plants (i.e., fuel cladding and spent fuel cooling). Since extended operation in a defueled condition is the only operation currently allowed, and therefore bounded by the existing analyses, such a condition does not create the possibility of a new or different kind of accident. 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?

Response: No Because 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, as specified in 10 CFR 50.82(a)(2), the occurrence of postulated accidents associated with reactor operation are no longer credible. The only remaining credible accident is a FHA. The proposed amendment does not adversely affect the inputs or assumptions of any of the design basis analyses that impact a FHA. The proposed changes are limited to those portions of the LCs and ITS that are not related to the safe storage of irradiated fuel. The requirements for SSCs that have been deleted from the CR-3 ITS are not credited in the existing accident analysis for the remaining applicable postulated accident; and as such, do not contribute to the margin of safety associated with the accident analysis. Postulated DBAs involving the reactor are no longer possible because the reactor is permanently shutdown and defueled and CR-3 is no longer authorized to operate the reactor. Therefore, the proposed amendment does not involve a significant reduction in a margin of safety because the current design limits continue to be met for the accident of concern.

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 40 of 44 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. 4.2 Environmental Impact Evaluation 10 CFR 51.22(c)(9) provides criteria for and identification of licensing and regulatory actions eligible for categorical exclusion from performing an environmental assessment. A proposed amendment to an operating license for a facility requires no environmental assessment if the amendment changes a requirement with respect to use of a facility component within the restricted area provided that (i) the amendment involves no significant hazards consideration, (ii) there is no significant change in the types or significant increase in the amounts of any effluents that may be released offsite, and (iii) there is no significant increase in individual or cumulative occupational radiation exposure. DEF has reviewed this LAR and has determined that it meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22, no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of the proposed license amendment. The following is the basis for this determination: (i) The proposed license amendment does not involve a significant hazards consideration, as described in the significant hazards evaluation. (ii) As discussed in the Justification for the Request and the No Significant Hazards Consideration, this change does not result in a significant change or significant increase in the amounts of any effluents that may be released offsite. These changes do not authorize any additional releases due to ongoing decommissioning activities, and will not result in any additional releases due to analyzed accidents. (iii) The proposed LAR does not result in a significant increase to the individual or cumulative occupational radiation exposure because this is not a physical change to plant equipment and does not require operator or other actions that could increase occupational radiation exposure. Therefore, the proposed LAR does not result in a significant increase to the individual or cumulative occupational radiation exposure. 4.3 Applicable Regulatory Requirements/Criteria CR-3 has been designed and constructed taking into consideration the proposed 10 CFR 50.34 Appendix A, "General Design Criteria for Nuclear Power Plant Construction Permits," as published in the Federal Register (32FR10213) on July 11, 1967 which are applicable to this unit. The CR-3 FSAR identifies each Criterion and provides a short discussion of how the requirements remain satisfied. The applicable Criteria for the permanently defueled condition are discussed below: Criterion 1 - Quality Standards (Category A) - Those systems and components of reactor facilities which are essential to the prevention of accidents which could affect the public health and safety or to the mitigation of their consequences shall be identified and then designed, fabricated, and erected to quality standards that reflect the importance of the safety function to be performed. Where generally recognized codes or standards on design, materials, fabrication, and inspection are used, they shall be identified. Where adherence to such codes or standards does not suffice to assure a quality product in keeping with the safety function, they shall be supplemented or modified as necessary. Quality assurance programs, test

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 41 of 44 procedures, and inspection acceptance levels to be used shall be identified. A showing of sufficiency and applicability of codes, standards, quality assurance programs, test procedures and inspection acceptance levels is required. Criterion 2 - Performance Standards (Category A) - Those systems and components of reactor building facilities which are essential to the prevention of accidents which could affect the public health and safety or to the mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. The design bases so established shall reflect: (a) appropriate consideration of the most severe of these natural phenomena that have been recorded for the site and the surrounding area and (b) an appropriate margin for withstanding forces greater than those recorded to reflect uncertainties about the historical data and their suitability as a basis for design. Criterion 3 - Fire Protection (Category A) - The reactor facility shall be designed (1) to minimize the probability of events such as fires and explosions, and (2) to minimize the potential effects of such events to safety. Noncombustible and fire resistant materials shall be used whenever practical throughout the facility, particularly in areas containing critical portions of the facility such as containment, control room, and components of engineered safety features. Criterion 5 - Records Requirements (Category A) - Records of the design, fabrication, and construction of essential components of the plant shall be maintained by the reactor operator or under its control throughout the life of the reactor. Criterion 11 - Control Room (Category B) - The facility shall be provided with a control room from which actions to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit access, even under accident conditions, to equipment in the control room or other areas as necessary to shut down and maintain safe control of the facility without radiation exposures of personnel in excess of 10 CFR 20 limits. It shall be possible to shut the reactor down and maintain it in a safe condition if access to the control room is lost due to fire or other cause. Criterion 12 - Instrumentation and Controls Systems (Category B) - Instrumentation and controls shall be provided as required to monitor and maintain variables within prescribed operating ranges. Criterion 17 - Monitoring Radioactvity Release (Category B) - Means shall be provided for monitoring the containment atmosphere, the facility effluent discharge paths, and the facility environs for radioactivity that could be released from normal operations, from anticipated transients, and from accident conditions. Criterion 18 - Monitoring Fuel and Waste Storage (Category B) - Monitoring and alarm instrumentation shall be provided for fuel and waste storage, and handling areas for conditions that might contribute to loss of continuity in decay heat removal and to radiation exposures. Criterion 66 - Prevention of Fuel Storage Criticality (Category B) - Criticality in new and spent fuel storage shall be prevented by physical systems or processes. Such means as geometrically safe configurations shall be emphasized over procedural controls.

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 42 of 44 Criterion 67 - Fuel and Waste Storage Decay Heat (Category B) - Reliable decay heat removal systems shall be designed to prevent damage to the fuel in storage facilities that could result in radioactivity release to plant operating areas or the public environs. Criterion 68 - Fuel and Waste Storage Shielding (Category B) - Shielding for radiation protection shall be provided in the design of spent fuel and waste storage facilities as required to meet the requirements of 10 CFR 20. Criterion 69 - Protection Against Radioactivity Release From Spent Fuel and Waste Storage (Category B) - Containment of fuel and waste storage shall be provided if accidents could lead to release of undue amounts of radioactivity to the public environs. Criterion 70 - Control of Releases of Radioactivity to the Environment (Category B) - The facility design shall include those means necessary to maintain control over the plant radioactive effluents, whether gaseous, liquid, or solid. Appropriate hold up capacity shall be provided for retention of gaseous, liquid, or solid effluent, particularly where unfavorable environmental conditions can be expected to require operational limitations upon the release of radioactive effluents to the environment. In all cases, the design for radioactivity control shall be justified (a) on the basis of 10 CFR 20 requirements for normal operations and for any transient situation that might reasonably be anticipated to occur and (b) on the bases of 10 CFR 100 dosage level guidelines for potential reactor accidents of exceedingly low probability of occurrence except that reduction of the recommended dosage levels may be required where high population densities or very large cities can be affected by the radioactive effluents. The above criteria were developed to protect the public from operational and accident concerns. CR-3 has been shutdown since September 26, 2009, and as a result, any consequences from any design basis accidents that remain credible or postulated beyond design basis events are significantly below the EPA Protective Action Guidelines as described in LAR #315 (Reference 6.3). The intent of this amendment request is to create the Permanently Defueled Technical Specifications (PDTS) from the CR-3 (Operations Phase) Improved Technical Specifications which will not have any impact on any design criteria. The requirements in the PDTS will continue to assure that the fuel storage and handling activities maintain the fuel safely. Effluent monitoring will continue to be controlled by the Offsite Dose Calculation Manual. 10 CFR 50.82(a)(1) requires that when a licensee has determined to permanently cease operations the licensee shall, within 30 days, submit a written certification to the NRC, consistent with the requirements of 50.4(b)(8), and once fuel has been permanently removed from the reactor vessel, the licensee shall submit a written certification to the NRC that meets the requirements of 50.4(b)(9). CR-3 submitted the required certifications by letter dated February 20, 2013. The NRC acknowledged receipt of the required certifications by letter dated March 13, 2013. 10 CFR 50.36 establishes the requirements for Technical Specifications. 50.36(c)(6), Decommissioning, identifies that the 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. This LAR applies the principles identified in 50.36(c)(6), Decommissioning, for a facility which has submitted certification required by 50.82(a)(1) and proposes changes to

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 43 of 44 the Technical Specifications appropriate for the CR-3 permanently defueled condition. This LAR also includes changes to definitions that will apply to the decommissioning plant. 10 CFR 50.2 provides the definition of safety-related structures, systems, and components (SSCs) and 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 50.34(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 the termination of reactor operations at CR-3 and the permanent removal of the fuel from the reactor vessel (following 4 years of decay time after shutdown) and purging of the contents of the waste gas decay tanks, 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 (with the exception of the passive spent fuel pool structure). 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." Certain License Condition changes proposed in this LAR reinforce the actions identified in 50.51(b)(1) to continue to maintain the facility for the storage, control, and maintenance of the spent fuel in a safe condition, without the authority to operate the plant. Other proposed License Condition changes align with the restrictions applicable to the facility in the permanently defueled condition to conduct activities in accordance with NRC regulations. 5.0 Precedent This proposed amendment is consistent with the Zion Nuclear Power Station, Unit Nos. 1 and 2, license amendment issued on December 30, 1999 to modify the Zion Nuclear Power Station Facility Operating Licenses (DPR-39 and DPR-48) and Technical Specifications to reflect the permanently shutdown status of the plant (Reference 6.5).

U. S. Nuclear Regulatory Attachment A 3F1013-01 Page 44 of 44 6.0 References 6.1 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) 6.2 NRC to CR-3 letter, "Crystal River Unit 3 Nuclear Generating Plant Certification of Permanent Cessation of Operation and Permanent Removal of Fuel From the Reactor," dated March 13, 2013. (ADAMS Accession No. ML13058A380) 6.3 CR-3 to NRC letter, "Crystal River Unit 3 - License Amendment Request #315, Revision 0, Permanently Defueled Emergency Plan and Emergency Action Level Scheme, and Request for Exemption to Certain Radiological Emergency

Response

Plan Requirements Defined by 10 CFR 50," dated September 26, 2013. 6.4 CR-3 to NRC letter, "Crystal River Unit 3 - License Amendment Request #313, Revision 1, Revision to Improved Technical Specifications Administrative Controls for Permanently Defueled Conditions and Response to Requests for Additional Information," dated September 4, 2013. (ADAMS Accession No. ML13255A056) 6.5 Letter from U.S. Nuclear Regulatory Commission to Zion Nuclear Power Station, Unit Nos. 1 and 2, "Safety Evaluation for License Amendments 180 and 167 to License Nos. DPR-39 and DPR-48 Respectively," dated December 30, 1999. (ADAMS Accession Nos. ML003672704 and ML003672696) 6.6 NRC to CR-3 letter, "Crystal River Unit 3 - Issuance of Amendment RE: Improved Technical Specifications (TAC NO. M74563)," dated December 20, 1993. (ADAMS Accession No. ML020710149)

DUKE ENERGY FLORIDA, INC. CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #316, REVISION 0 ATTACHMENT B PROPOSED FACILITY OPERATING LICENSE PAGE CHANGES, STRIKEOUT AND SHADOWED TEXT FORMAT B Subject to the conditions'itd requirements incorporfted herein,, the Conmision hereby licenses: (1) Duke Energy Florida, Inc., pursuant to Section 104b of the Act and 10 CFR Part 50,.1 Licensing of Production and .Utilization Facilities,t to possess,-use and operate the facilitYt"_ as required for fuel storage (2) T1he licensees 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, Inc., pursuant to the Act and 10 CFR Part 70, to "eeeive, possess and u at any time special nuclear materialXs reactor fuel, in accordance with the limitations for stodre c-and azu.t= vquired for rca:tcr aperatien as describZ'n the Final Safety Analysis Report, as supplemented and amenu nfi-guorned rdfr /-lasfission detectors. andl (4) Duke Energy Florida, Inc., pursuant to the t and 10 CFR Parts 30, 40 and 70 to reeeive, possess at any time any oyproduct, source and special nr material as sealed neutron sources for reactor startup, ealed sources for reactor instrun tation an-radiation m.-itring equipment in amounts as required; --used previously (5) Duke Energy Florida, Inl., pursuant the Act and 10 CFR Parts 30, 40 and 70, to receive, poss ss and use in amounts as required any oyproduct, source or pecial nuclear material without restriction to chemical or p sical form, for saaple analysis or instrument calibration o, associated with radio-and to possess and use at any time any byproduct, source, and special nuclear material as sealed sources for radiation monitoring equipment calibration (6) Duke Energy Florida, Inc., pursuant to the Act and 10 CfM Parts 30 and 70, to possess, out not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. 2.B.(7) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30 and 70, to receive and'possess, but not r-separate, that by-product and special nuclear materials associated with four (4) fuel assemblies (B$ll Ideatifir-Rm, 16' cation Nunbers lA-0], 04, 05 and 36 which were previously irradiated in the Oconee Nuclear Station, Unit No. 1) acquired by Florida Power Corporatioifrom Duke Power Company for use as reactor fuel in th$ facility.

c.

This license shall be deemecl to contain and is subject to th -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

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

Facility Operating License No. DPR-72 Amendment No. 243 of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: 2.C.(1) Maximum P..... ,Le Dulke Energy Florida, Inc. is autho.;Ied to operate the facility at a steady State reec.or core poWeF level not *in emeec of 2600 Megawats (100 perent of rated care peweF I

2. C. (2) T

,_..replaced with the Permanently Defuieled echnical Specifications Technical Specifications (PDTS) The nical Specifications contained in App ix A, as revised through Amendment No.14-3 lare hereby i.epe.rted in t. i,.n'"Duke Energy Florida, Inc. shall epeFs8aee facilit in accordance with thee.echnical Soecifications. maintainPermanentl Defueled The SuL Jeiane I.. quiremente eontind in the A.ppen.ix A Technical Speeifi.ti.n. and listed bcloW WrS not required to be perfermed immediately upon imlmnainof Amendment 110. The Sur.vcillancc RequireSmnt Shall be sucesecefully demanstratad 9XWF ftoe trotme and senlkitnr speeOWCO DCIOw WO eaeWf. I a) b) d) e) SIR 3..8.2.b shall be suc.essfully de.on.trated pior. to entering MODE 4 on the first plant start up following Refuel Outage 0. SIR 3.3.711.2, Function 2, shall be sucessfully demonstrated no later thOn 31 days fo..llo-ing the implementation date of the ITS. SR 3.3.47.1, Functions 1, 2, 6,10, 14, & 17 shall be uccafutly dmon-tr0tcd n. later than 31 days following the idlmottandte of the ITS. SIR.3.3.4 7.2, Function 10 shall be successfully demonstraed prior to entering MODE 3 on, the fiFst plent star" up f...i..ng Refuel Outage 9, SIR 3.66.11.2 shall be suooesfully demointrated pior to enter-iRg MOGM 2 on the firs plant staV t up fellovW IIng Refuel Outage 0. SIR3.7.12.2 shall be successfully demonstrated PRiOr to aenting MODM 2 on the first plant start up following Refuel Outage 0. SR 3.8.1.109 shall be aucacosfully demonsrated prorF to enteArn MODE 2 on the firM plant start up folloWing Refuel Outage 0. SIR3.8.3.3 shall be successfully demonstrated prior to entering MODE 4 on the first Wlant start uo-folleipna Refuel Q Atau A. h) Facility Operating License No. DPR-72 Amendment No. 243

- 4a -

1) 2

.6.4.5 AWEl be suceeasfull, demonstrated Mr~ to enermr. NGOE 4 on the fis l ftt tt-up 0oli e~u uae2 SA 6R3.8.7.d--she~l be-seemoosafull, wdwistvated no; later theIai I ~ ~ ~ ~ 7 ILIThU1U I ~.~.t1C.1 C.Ii i V

  • V VV 1,A 0

-g be Semee L.,,.21 ssfully J..,rnatretwd gem vatte thaus 2.C.(3) n.,L,- Cn-rro-171^AAý In^ -Ehoh _A___ Jl_ Nodes 1 ond ! with 4wow wagwav foo. _E` 9

0.

vwv i Until -then !*r!!-? 10-64en We bee.. w weengees wwww wrro wwwo 4d&-be" A L-a-_ I _4v 1-11 -0 2.C.(4) DELETED per Amendment No. 20 dated 7-3-79. 2.C.(5) Withdit six m.ositts oF the dote of issuanee of thfta llccnc. flRed Power ftporatoie ~hall complete r,3ilficaton: to the levc1 hidicatilmi Af ttIq boart A weter sterage tank. and installation of I

    • w

. S

r. *r we-r

,olo-on Aprit ieq,

ieuts, wis na"

--r+/-critas Powex umpcrat2:osv, was eftartweet to Hatezyy Pl-ida, 144c. Facility Operating License No. DPR-72 Amendment No. 243 I

5-2.C.(6) Deleted per Amendment No. 21, 7-3-79 2C(7) 2 GOW~ h OFg~GI ~euedmw"e4g... FlndeP ed ify tthsMr....1 -anL0 U QMIw*

  • AIL L

Me .. Lk.W IS,. system flow ~ndicetiGA to meot the single failure crtitein wit rcgcd to pressure sensing unes tn the flew+/- differential Droccurcw-tranambtAm~ v -4 2.C.(8) Within Weec maenths of "cusaw of this liconse, RFloida Powvr Corpe~eticn shall SUbFit to the CommcI a. ro 5ed GUr08i anc program fo ro icri 11g t Wentain*mnt for the purPoco of det erMinig any future dlamMI* w at of the dome I 2.C.(9) F~ eo~ -Duke En=-ergy Florida, 'PG. ctol implementA and-m-aintaSin inA eff"s all PFGicicnc of the approyed Amr proeotin progrom as decoribed in the Final Sfety Anaa~ Reportfo the f a eility and as ap provod in th Safot-yEva Elu-at*- --As Rpori, daed July 27., 1:7 9, Januar22 191 mnay 6, 19e3, July 18,14086, and March 16, 1988, subjootlto hce fellewiAngpr.lco The lieensee may make changes to the approved Amr proecetion pmc~am wthou prr ppmova of the Commission cnly Wf thwe Ghanges would Pat odvemeoly fc i.~ -LH I..-...1 J.A -9 -9 6....6A

_ &L. -

i ~ -9 IA.. I UVWS I4~1 9 2.C.(10) no ac~an OT inc ro~o~r v~ciru LIWTIO ~ iucorr~ c~c~z ncx ~uuu ~nwcvr~ion or iria v i I I I Meetsi Ill ~ liatd F IUSl I I1 Me V iY MMl "i Wei 1 W pipingil ow " 'a I 1986. (Added per Amft. M5, 23486)- 2.C.(11) A system o4 "Fcoouples added to the de"a heat (DH) d"o and AWdIan; PrOSciizr SpraY (GPS) mwe, capable of deledik@ flow Wien, &l be opcroblo Modes 4 through 1. ChIArII checVe of the thoI r oIples shell be podormod on a met hi basis-t demeitstIa eeperIIitV etiwt If M F ie 4 eH AN Is the T.wuploc become lneparable, oporabilit shal be retored withi 30 ots OF fth N4RG shagl be informed, in a Speeial Report YWith the fo~lnoi'g fourteen (14) days, of the inoccrabeWt~ and the almse to rectoe AaA-en-obll-NW. ifA-md1. 9146-9. 1 2705 2.C.(12) Deleted per Amendment No. 237 Saw- 'be naze "Florida Power eorporabion" was changed to W1....... I Facility Operating License No. DPR-72 Amendment No. 243 I

-5C-2.C.(14) wagagen guaway ueense weneive i m neensee Onci oc~eIGB nad tRWFFtmn SrrOaMOM Ier aaarcFMO IrOFS maft G I expDienefl and #Wa ineldo the fzI:.ne key emeas diF "--" *P.......... (1w) I-,*G_ L f II J I-Ira ianuna rooDorz~o suwoa~ WRfl me rouo~.ing oornonrs: PFO dofintd cMRdinAto-d-Weo FScrM stmna. and auidenca Or".......... IGFJ tF....... A ~~r~w~nt ~~.f mad.,rI n .i,,.r in~nin,'rt

7777

"S

  • Iti-~*
  • ,l~tt w

e~ u~agnotcaat'Fig aOF80 fOr GqUlprnen Ond1 MatFno (2-.) Oporations to mnibtiet Wu damager oodrn UofFwr~ a& PRotootln and use of porconnel Sassta

b.

Gw nobn G-. MIinfimizig OFir pr a-ldwnetifitn of readily mv~oabe pile ftagod eqpment + TSaoft fon oo mint 1 ato-mFa~wene m (3-) AAfimm #A rAhs, km oWgumA ineAuda 0,00AWAVANOMM a& 97 Water., spray. os..bi.. bi Dooo lo eonit responders 4dVl=e We I*I*INiN li*1511mel I VII

  • 2.C.(15)

~.Jonimleentation Of AFROndmontI Wag~ udop"i~ TSTF 448, Rwosin 3, .... determin.. ion of cont*ro, O,.p-habitability OnvOlp.O (CCHE) unfilered airinlolugo s rquiod y ur~mllnooe Roquiromont (SR) 3.7.12.4. i bcorana wit IT 6..242.3(i) and " aeew l 09911mon WfCII habitability 6s required by ITS 6.6.2.21.3(i1), shati be considered met; Foflowin e, s 0) b) e) The Iist porfonaneo of SIR 3.7.112A. in aordone i Spowficgati 5.6.2.21.3(i), 9ha W within the spioofied Frequenoe of 6 yews, plus thoe 18 mnth a"llwee, of SIR 8.0.2, as m.eau'ed krW-May 18, 2007, the date of the moot rocont suoeeeiful IRIloekage toct. The first p-- -Fomanc-of the p--itd;e agesamnt of, CCHE habitailityl RTS 6.6.2.21.3(ii), shall be

-hn3 er.plus the 9 mont ellwoeno S

.. 2* a measured from May 1A, 2007, the date of the moot! Feeent sjuocoosful inleekage test. The nl Cm AAebailitx po Integriiy Prgram wiAb used 1o Veriy the l IntcgreIt of th Control 4 Gmplve boundary Conditions that are identifiiod to be adverse 9hall be trorniod and uo D II AAI i*I i II I D *i X

  • ABSSoSotA! will bo pO~eformed vWthIn 60 days of implementaten*o Amendme.

Facility Operaiing License No. DPR-72 Amendment No. 230 I

-5d-2.D Physical and-Gvbe 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 2781.7 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Physical Security Plan, Revision 5," and 'Safeguards Contingency Plan, Revision 4," submitted by letter dated May 16, 2006, and "Guard Training and Qualification Plan, Revision 0," submitted by letter dated September 30, 2004, as supplemented by letters dated October 20, 2004, and September 29, 2005. t I mJ m Im I ml licznsee sflhll tully imp*lem*nt and maintain in AffAet all Pro'iIAsean of trl GCmmi**^*n ofppred eybeF o*eurAty plan (CSP), ineluding

  • ohnges mode pursuant t, the authoriy of 10 CFR 50.90 and 10 CFR 60.64(p). The lisensee's CSP wac approyed by Liccnee Amendment No. 238, as supplemented by a ehange approyed by I

Facility Operating License No. DPR-72 Amendment No. 242 I

DO NOT REMOVE is. %is~i iicf3n" is swojeet te the fallewing aitit4!ust conditin wvd applies enly to Duk:e Enefrgy RFicrde, Ine. (D!F): (1) DEF WAI inbternneet with t eerihzrcimnte reserveis seams of tite salke and esmar.e of emergerny bulk pei~ with my~ entity or ezntities in. its ser.'iee area' engagifig in er prepes"n to engog in elactrie bulk pftw StWW1y, an terms thmat wRI peavize for DEF= east '(i I " * - -a reaseonabl return) in eemnectior theerotib arml al~ow t I other artiei mt(q) ft4l: eegeeve eeerdirtebien-. s to the benfit Of i I ei (a) inbettenrncctione ;4 no be Jlaited to Iew vebae whenhig* veltayeg& ee avial fro wp i ML ta.. fae!Uities in thec area wnere intcroorameetion ie derireco, AMen the prepo.ed arronement is foind to be teeinioa~ly w.-- -- -4 -i

  • mr-,---

IL-r,-- v¶r--r-- I ~..flrJ .e~ V ~ V LW - S W IWE fa N ~ t, w~tegerney serviee provLdcd uier sueh~ agreemenczts w~il ce Emnrised to tew Ealesth entent availdale ari degired w-here sueh ismpl. I.. not imieii serviee to theo sewieir' eugtonmars. ~mrA~-r ~f th~ t~.-~- ~t r~r~a mt-.ftr ma ~rr~~m~nt ailolo o o~ pticeipcrt an~d wthzem woitl provide 0fal aocc te;- &i benef its of reserve czordimboior T weu~ld icc or ne.ttk e ftok-'awin Comtion v"Ou. obetai-DEF anal each Vortipat(s) she!l p..rov tde 1

  • flnl the othermt-e ow RZ ilbl who FINP- -

A~ %~WA ~..& ~ ~ h.&&'4~IA A ~.& ~M& takago.JA fte the geneato a 'fl "Ie eftent it ean deo go gieo~tzarpin oviee ....e. o ef t.hae erm .s.rvi.e area-in no wa

  • i no te. an ass i..n or a~leeatin of w~allaMA-1-f m-NNIOLU iap&,R q
  • 4 it is intmnach Cei1-Ay as I

C .L* WE ~. W LA* 4~ J W~

5.

& A& .U W**%. L.P U.i mm...... of .s........ -- ia croer to cinrir; umi~ coaraiuc~nrs - c:rtain ~ r5 otor~ r.ct~ rz~ 4 been esoea.. rerzzSor. - i A " Facility Operating License No. DPR-72 Amendment No. 243 i

-DO NOT REMOVE h t

  • J

......,.int. w i h Q. estb.. h A..511w 0%.S2ab 6w ho U I imtst~lid anr.,/r purecsed as ncecagcry to M~ifftain if! Wbtf& W1 aieqftbe rc~i~bility Of penter st~ky an the intereenneebetd wa~jtem of I b! 4ity thus 01 ___mnc Waal! be cealclatei as a percentag - f cc : ad. Noc partietpant(s) to the interewmti..n skiwal: be required to maintain were thmi such per~tcg Mai'--es a Weecntag of t ý1ý I ýý I .1. ý _ý Jic ý 1 ý

  • W~ Y'
  • ~

... au~. a.. a 1a. u-Jint DEF -- ul-be r..ui.rd to -aintain .ith. shirt~erwimerebticn then the other partteipent(s) shall: be required be eazcy or provide fb -its reserve reappmuibility the full anmunt tin kikcwatt .,f suh ineteasec Under no cirelwe-tr-r-es W.ill-N.A. ia sj krn at l*cz*rt*& reserve re'ju, iremetis ed the i--st* 1l c, re.. rve-r..... ire" nt... I (d) 3Enterc~wecticx eIre eootdhna im wo ~'ON~8~ W&~4. v2ae w .. A.. C. ,ý, -, - _--l'Off. -- Ord io.*.

  • .l.va V

beeS air. a.. e*a -~ aS -~ ac~~~~asr 1 ~+ noncrix~ ri rarcie PA r=r .. ill r-~~ fr-rn ~r +/-~1!1 "1~z:iIIr ~ t~ ~im~ I '-I l.r.I ni mew taferew-m iv ti .n r,* i 'f i n aI v J... 1 .. L.. ql q* l I i k l

4.

CS'.., be'fl ~fl'W.. zke~aZ aa a.aS. .a gacO aag... o~r partieipants to the trarnsLacticr Th~is refers speeifi-eall, to the eppertnitcj to eawdkote in the plaien"gof new, aeneraticr.. tranwaission and asseeiated fa:l~itice. (a) it is kok t1 be cwi 1ecle.. that this ,Inditton requirL-DEF to purehase or e36l btAk po.wer if it fiits a Pz.&hai of s.le i---2.e esibc or its eocts in eormcctti .,ili ..c ~uchaeor sa:Ie weald ekceeei its benef it I Facility Operating License No. DPR-72 Amendment No. 243

DO NO!f REMOVE t W

b.

ý ^6 d&&Ift btlk power supply wyetezm with that of myj o"b balk pom stu;i 5,5tem, byj 5eJling unit po at: the cost of its new power stq~iy, or enrgqea in joint vntereg with the saw result, DEF eh-all not refusne p~roprtional por t~ieption mn a eemparoib1e basis -Mn h seme unit to any othr.~ entity

is e~i ta~

(gee eaClwuit'et I si~ra) e~i i onzr prpzcsing W~ engage in btilkI pmftr supply Wc the extent i stene~ feasibly to proovize _u _A_ p~ fr-M t-he unit or units in questian-. O3) QEF AIl fae~l~itete the emehange zef bulk power by I rtranmiulas n oear Itos arm berwmen or awmng b" or~ owe~ zrtittes with,,t4mil It i nenec a-tee"s whieh W.il yul enspate it. ffzr-t+/- ~ -1%-~f its a-jtern to the externt --it Subjeet arrangezinte re~sncby ~n b aenundated from a firfteiondaln (a) This eandition apiplies to entities with i1h6 DEF may be intereenneeted it... e .9 wsell as thoge to w~himi it is now, inbftcrc.....teei. (W ULI-ts eb&*qawti wnor this emdktien to trensmit bulk newler for other aniic n the termsa tat& I acvand to include in its jplanning and eenatruetion prcgams uffiienttr~~ iaccr stapaity a requiiz therezfer, ro imc hctauc gta nttc ive' APP 00tff~ien aMR4c. c MZ=n;tie& be requirca t wecmmadcate the exremngemenb from a fmebinctcnla tzbeehilal 9bwvdpo!6tb &An t-h-t --- othar ent-ibies w411 be obligated to eampzenati DEF f-lly -ce the toe z I -1 -1 1 t,=u L L-I

  • ,,l..

l,

  • iL,,b rl'~a lm.
  • =,lr, bD q--

y c86,l tv in ,~~l,,,bl,, qqb i m d M.Wft.ft am,%- - iý- 21.0 r4boLivi.A .14~ &M

  • EJ C,

I-I f(ftI

  • 1r A~ wo1lif

%-Iri a-A% PM.i orzegoing emditions are to bee

nsistent ~itn tna prcviswna r-~

I TI~~U rrra~n -r praciccain cnnection tiwrewi it ae to~ be stojeeb-~ ies havine jueladiction ever tem. of reeit4atova Anevae --d -- d..... Facility Operating License No. DPR-72 Amendment No. 243

DUKE ENERGY FLORIDA, INC. CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #316, REVISION 0 ATTACHMENT C PROPOSED FACILITY OPERATING LICENSE PAGE CHANGES, REVISION BAR FORMAT B Subject to the conditions "and requirements incorpordted herein, the Commision hereby licenses: (1) Duke Energy Florida, Inc., . pursuant to Section 104b of the Act and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," to possess and operate the facility as required for fuel storage; (2) The licensees to possess the facility at tne designated location in Citrus County, Florida, in accordance with the procedures and limitations set forth in this license; (3) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Part 70, to possess at any time special nuclear material configured as reactor fuel, in accordance with the limitations for storage as described in the Final Safety Analysis Report, as supplemented and amended; (4) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70 to possess at any time any byproduct, source and special nuclear material as sealed neutron sources used previously for reactor startup, as fission detectors, and sealed sources for reactor instrumentation and to possess and use at any time any byproduct, source, and special nuclear material as sealed sources for radiation monitoring equipment calibration in amounts as required; (5) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30, 40 and 70, to receive, possess and use in amounts as required any oyproauct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radio-active apparatus or components; (6) Duke Energy Florida, Inc., pursuant to the Act and 10 CFR Parts 30 and 70, to possess, out not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility. 2.E. (7) Duke Energy Florida, Inc., pursuantto the Act and10 -R Ad Parts 30 and 70, to receive and'possess, but not 'e*- separate, that by-product and special nuclear materials associated with four (4) fuel assemblies (B&W Identifi-94mrji 15' cation Numbers 1A-Ol, 04, 05 and 36 which were previously irradiated in the Oconee Nuclear Statidn, Unit No. 1) acquired by Florida Power Corporationyfrom Duke Power Company for use as reactor fuel in thi facility.

c. This license shall be deemea to contain and is subject to the

-conditions specified in the following Comaission regulations in 10 CFR Cnapter 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

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

Facility Operating License No. DPR-72 Amendment No. of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: 2.C.(1) Deleted per Amendment No. 2.C.(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. are hereby replaced with the Permanently Defueled Technical Specifications (PDTS). Duke Energy Florida, Inc. shall maintain the facility in accordance with the Permanently Defueled Technical Specifications. Facility Operating License No. DPR-72 Amendment No. I I

- 4a - 2.C.(3) Deleted per Amendment No. 2.C.(4) DELETED per Amendment No. 20 dated 7-3-79. 2.C.(5) Deleted per Amendment No. Facility Operating License No. DPR-72 Amendment No. I I

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

- 5c-2.C.(14) Deleted per Amendment No. 2.C.(15) Deleted per Amendment No. Facility Operating License No. DPR-72 Amendment No.

- 5d-2.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 2781.7 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled: "Physical Security Plan, Revision 5," and "Safeguards Contingency Plan, Revision 4," submitted by letter dated May 16, 2006, and "Guard Training and Qualification Plan, Revision 0," submitted by letter dated September 30, 2004, as supplemented by letters dated October 20, 2004, and September 29, 2005. Facility Operating License No. DPR-72 Amendment No.

DO NOT REMOVE E. Deleted per Amendment No. I Facility Operating License No. DPR-72 Amendment No. I

-DO NOT REMOVE Deleted per Amendment No. I Facility Operating License No. DPR-72 Amendment No. i

DO NO!- REiW--E Deleted per Amendment No. Facility Operating License No. DPR-72 Amendment No. I

DUKE ENERGY FLORIDA, INC. CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #316, REVISION 0 ATTACHMENT D PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES, STRIKEOUT AND SHADOWED TEXT FORMAT

TABLE OF CONTENTS 1.0 1.1 1.2 1.3 1.4 2-.0-3.0 3.0 USE AND APPLICATION.................................... 1.1-1 Definitions......................................... 1.1-1 Logical Connectors.................................. 1.2-1 Completion Times.................................... 1.3-1 Frequency........................................... 1.4-1 -SAFETY LTMrTS-(S.s) .S-k -V I-i o l-at i-o n -....................................... LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY... 3.0-1 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY............ 3.0-4 I 3-;-1-. 3.2.3 3.2.4 3.2-.5-- REACTIVI-TY--CONTROL-SYST-EMS-........................ 1 SHUTDOWN-MARGIN-(DM).--..-...............

3. 1-1 Reac-t+vty-Bal-ance..................................

3.1. 2 Modera-tor Temperatre-Coe~Ffi*-ieit -(MTC-)-._.-.-.-.-...-.-;.-_3. -4 CONTROL-ROD-Group Alignment limits.................... 3.-6 -iafety Rod Ensertion Limits......................... 3-110 r,,L -Agnment Limits

3. 1 1.

Po stior-In -ator CFanewl-s................ 1-14 -PH-lYSI-TEST-S-Ex eep - MODE---.................... 3.--1-7 .PH-YS KI-S--T-ST &-Excep-- MODE W - 7 BUT ON LIMITS 1 -Re ng-Rod Insert-iu Li .it-, A XIA-H-POWER--SHAPING--ROBD-(APSR)-T--s-nse--i o-Lts_-;..- 4 AI-A-L POWER I)'ALA E Operating-Limits............. QUADRAN POOWER TILT (QPT)........................... 3.2 7 -Power eaking-Factors.............................. 3.3 - -INSTRUMENTATION................ ,Rea~etecr -- PYro t--t-i-on Systemv (RPS) I nstrumentation....._ .J3-1 T Re-ae r-Proteeion- -ys-tef-e -(RPS) - Reac-tor-T-rip IUI*,-I,,I fnTl*A"*, 0 3.3.9 - 3.3.10- -CONT-ROL--ROO , -i (CRD) -T-.ip Devi -Engineered-S-afeguards Actuation-System (E-SA S-Instrume-n-tat-ion ,Engi*-nee -red--Safegtmrds ActUati-oLn---SYZLCe _"_I.(ESAS)-- a 1 Ir t at.on-. .-E*g-nei r ,ered-Safeguards- -Aetua-i-orir--Sy (ESAS)-- Automatic-t^ L-. enerator ([DG) loss of Po -...Start-- LOPSY............................... 1 ...... 3.3 1-0 ...... 3.3 2-0 3.3 24 Soul - intermediate Raf wje--Nettti Fltw (continued) Crystal River Unit 3 Amendment No. IS2

TABLE OF CONTENTS 3-3 3.-3.11............. 3.3-.13 3.;.3_.14

3. 315.

3-r3-16~ 3.3-.47 3-.4.7 3-.4. 3-3-. 4. 4-3-.4*.6--- 3.4.7 3.4. 3.-4.1 3-A.-4-5 3 -;5.-............ 3.5.

3. 5.4-

-INSTRUMENTATION -- ont-inued) "1 A I-tfT1eTtJeIcy I eeuwa. Le .1.tat on .1 1I 1tU -{EFI.C) System instrumentation ....3.3- '\\k......-EFl - -Manual Initiation............... EmergencyIFeedwater--lnit-i-o and-Cnt.-,-1 P% le -etuatilon Logic

  • ,4*..pgtmý

--i-T.i i- *inn:*.,r.- -*-ýn.d.-,.--. 3-3-30 ...... 3.3-32 EFIC--Emergeny- -feedwate--(EFW) - Vector Val ve Loglic.................. Re-acetor--Bu-l di-ng- (RB) ---P ge-- olat -H4-gh aII tI o -I --- II -*L I -Cb ý-o-R om Iso at on H II it'o Pos*tAIccUdent --- fitorI-ng-PM-* ns-Lrumenta0i-i,D-.. -Remot-e-Shutdown System.............. -3;3--34 3.3-3.- 8 3.3 364 --REACTOR-~OOL-AN-T -E -- (ES-...................... RCS Pressure, -Temperature-, --and ano- -Departure -f4om-N"cle-ate C-IoI+/-I-nqI-I(N -Lint*s-&_........ RCS-H-i-mTemperature-f-or CriticZal-ity........3.4 -3 R e-S-r..s.su.re and Temperature (P/T),mt--.-. . 3.4 RCS-oops -MODE-3

3. 4-6

-RCS- -o-MODE-4 3.4-6 RC-S--Loo-ps - MODE-4,-Lop-F13e-4--a -RC-S-Lto-ops -MODE-_5-,-+Loo'ps--o-F-i 3-d-...... .-.4-10 R*S-£**'*

  • +

.3-* - RC-5-1o-o" - 11 J _ 11 F -1 3 -Press r ze 4--5 -1Pr u-er -2S etY-Va-Ives....................... 3.4-17 r UlI ruvvql [lJr.: 5 L U I-II-C IV r lv ,t-L Lo-Temperatu-re--Ove-rpres-s-ure-Pr-et-LTOP ysten..................... 4.-421-1 LRS-Opeationa--EAKAGE 3.4-2 2 -RCS--Pressiu Isolation VIve (,,,r) A 3.4-24 MlL-1 LeaKage-T-Detetion nstrt-atrtn-t-a-tO.-...- RCS-Spec+-i-fic-Ac-t vi-ty-4-j.- -3.5-1-3 S.-a... Generator '71 a . 'f 3cf -EMERGENCY CORE--OOL SYS-TEMS (ECS).- Core-Flood-Tanks---((FTs-)-........... EC-CS - Opera-t4fg................... ECCS - Shutdown.................. Bo-,.r-at-ed--Water --- agte-Tank -(BWST-)-.- .. 3. 3-; 5 3.;6 3.6.3-3-.-6-- ..Cofttai nment............. I La II IIII~I IL r~ II

  • U IL I

III I I_ I I UI-jý IXZ Containment -Isolation Valves... Containinment-- P su -............ 36-1 3-- 6-3.~6 8 3.6 16 ~.AJI I L~ II 1111t1 IL flU I CIII 1 J~ i a LL41 C... I ý111Fý I ý U I C....... (conti nued) Crystal River Unit 3 i i Amendment No. M

TABLE OF CONTENTS 37-,6- -CONTAINMENT - MS---K(,ont t,,,uj -Reactomr-Bt ldng-Spray and Containment Cooling--Systems. £ontai-nmment-ergency- -ump, C.ntrol S y s S)..........................

.b -t 3.7 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 3.7.13 3.7.14 3.7.15 3-.7.-16-.... .-7 3.7.19--- PLANT SYSTEMS............................... -a-n-Ste-Stea-Safety Val ves (MSSV)-.s -an-St~e"-M Isolation Vaal---esLCT &I _N~ M - a asolation Valves (MFIVs). Turbine Byp-as-s--Va-lves--(TBVs-)............ Emergency Leedwater _y)tLCIII........ UVV*I.

  • I L[

J I(All IF%. ucearSeVices ClsdCycle Coolin a _... ...... water-(5W)- System........... ec.. ay-H-t--CI-os-ed- -Cyc-1e-Coii-Cyg-(.-Syste................... LI H F A r R 3.7-1 v...-3.-7-6 3.7-6 ...- F7-9 ... 3 1- '1 -7 _'I Q rcay--HSe-at-w--awater-Sys-tem................. 3--7-21 -Syate-Hfle at-S)---....s).......................... 3-7-23 C R-oom-Emer~ey 4Vent*-1Tatton Fuel Storage Pool Water Level................... 3.7-2-7, Spent Fuel Pool Boron Concentration............. 3.7-2-8, Spent Fuel Assembly Storage..................... 3.7-30-G 3 -3n4r p c-f ^ -*"* -eodr-Sped-f i-Activiy...---- J Steam Generator Level.......................... 35 G.......... 3-3-7 _.-D iDesel -D--FW)- -Pump--F-Ie-l- -- i-1 -Lube +/-1and Starting Air 3.7 39 I*LI U JL LI k l,,,,,,,,..,,...,, 2. 3-8 3-.8.1 3.-8. 3 3-.8.-4 3.8.9 3-.&.* 0 E-LECTRIICAL--POWER- --SYSTEMS........................... 1 -AC Souw ces-0peO ratI1ngI -AC,,ot r § uo -1 -0 ... Diese--Fuel--i-,---Obei-Orl, -and S'.... ing Air 3 ' 14 DC-Sources- -Operat ing........................... Mn-, -)*n.... jou c- -Shutdown................................

u f_

t.te.r.-y.-..-. -Trwe-nrte-r-s - Ope-ra-ai-n-g-~-__ inverters-Shutdown................. fist-ribu-ti-o0n Sys-t&-eS-0perating-... __Distributi-o-n-Systems-Shutdown....... U-4-'L 3.823 3.8-33 3.9-1 9 3-~9. 1 I"11--I--111--I TkIp t'* rl r r"l i -r T A k I r" I~.... JLL... IS -- U-- JII P. I +/-II i...

Bo-ron 'Concentrati on..--.--.-.-....

,Nucl-ear-Instrumentation............ I IContainment Penetrations......... (continued) Crystal River Unit 3 iii Amendment No. IR

TABLE OF CONTENTS 3-.9--- REFUEtINGC-OPERATIONS& -.-. (continued)- 3--9;4--..... -eay-tHeat -Removal -(O-fHIand-Coot--ant o - -Fag.u -oU,_,--,,a,, ter Level 3.9 6 3-.9-5..... ..- De-cay- +eat -Removal--E,,R----and--an- +r-u-aion - Low-lWater-L eve1-e 3. 39.6 euelng Canal Water Leve'. m 3_1-4.0 DESIGN FEATURES........................................ 4.0-1 5.0 ADMINISTRATIVE CONTROLS................................ 5.0-1 Crystal River Unit 3 iv Amendment No. 2

TABLE OF CONTENTS B- -2. O--SAFETY -LIMITS--(SL) B--2_0)-1 B--... Retor-ore-s B-2-- B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY. B 3.0-1 B 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY.......... B 3.0-16 B-3. B--3-. L-1--- B 3.1.6 B--3.-1.6 a-3.-1.-7--- B-3.1.9 -REACTIVITY IONTROL SYSTEMS -u B I SH Ml Il'l) UTD....*I.... B ( D ) .......-. Re-a.ttA-,.-* y Ba-mce... -B-- .- -fti-6 lilUUcI aLUF lcuipicra LUva ýUI I -~l.~~~........UJ+/- ý_ i-, rn l fI n er% A-I-,,,,, -roup Al ignment Limits............ B 3.1 17 Safety*Rod-herti on Limi t................. -27 AXIAL---POWER--SI AP-ING-ROD--(AFSR) li gnment Li mi tsB 3.1 Position indicator Channels................... B 3.1 35 -- PHYSICS-EST. Exceptions Systems-MODE 1 - PHYSICS -TESTS Ex-eptions - MODE-2......B3.-48 B-3. B 2.1-B 3.2.2 B-3.2.3-a 3.2.4-- B-3-.2-.-5 B-3. 3 a-3.-3.- B-3-.3. B -- ----- B 3.-3-.6 - B-3.3.7 B 3.3.9 B-Th-3.11- -POWER DI-STRIBTI)N--EMI&......................... AXIAL -POWER S 'APING*-R -(APSRY -i 4n--rn4sB 3'.2--1-- ,AXI-AL IPOWER-- I.UB1 ALANC CE O0 peraUti 41 Ing L i m i -t. ,B-.-.-2-4-7 l-r-T'3 QUADRANT POWER TILT (QPT-........ C-E.. P4owe-r-Peak r -at r 1-- rSTRUMENTATION B 3.331 -~ ~ ~ te -R to, -Syste-{R ReactorI.- -Prte4 -S~ten, (RPS)-Mn-l ~eo B 3.3 ,*,,,............................. B

3. -

Reae tor-Pro, tecti stem ,RPS )--Reaeter T p Module- (RTM) B--3.-3-34 CO -rive 3R3ip

  • .3-*-3--

w LnIfneercl Uaregua as mctuation Sys em (ESAS)- Instrumentaton............... Engi-neered-Safeguards Actuat*i o Ist", I*1 "1 M--*'* i E ]* I I ""~lL'"C.I LI C.L J.I .I. L .1 1 ung lneered Safeguards Actuationl _7!stem '--(ESAS)--A-tmat ic Act ation-Log-ic.-

  • B -3.3-61 Li1l-I yjiZ- [%y U/ I ý 1

I '.JCI M: I CL Lu. I LLJSJJ - -t art -(+/-tOPS)-;............. -Soure--Range-e'... I3 ~ -C V ~ l l= 11C LlI UIl FI U r JVMC I m R3-.i-rr aR+/-3T -+/-nerie~nate~an ,wreutrom Ru -- e rgency leeawate ~ Initiation a**aOn-t"ro-(EFI-nstumentation.................. (continued) Crystal River Unit 3 V Amendment No. 1-82

TABLE OF CONTENTS f-3-.- 3-.-13 B-3.3.15 B-* -3.-38-1-7 8.-4. B-3-.34-_-- B-3.4. B 3.4.4 B-3.4.5 8-3-.4.* B-3-.-4.-7--- - B 3. 4. 8-B-3-.-4.1--- B--3

  • 4' 13 -

B 3.4.14 B 3.4-.1 B-3.-5 B 3-..-5 B-3.5.4 B- -3.-5-3...- B-3.6.1Z B-3.6.5 B-3. 1.5-........ I-NSTMNAMN ,cotinued) Emergeny-Fee-Ini ti ati on and -Co-T ..... C) Manual initiation................. B 3.3 100 (EFC) Autom-at-i--Actuati-on-tog+e...... 3-15 Emergency Feedwater Initiatio and -eot-rol )-Emergency Feedwater (EFW).-Vetor - -Valve Logi_ c 3.3-110 Reato. -r ,*iding (RB)--Purge Isolation-Radiation................ B -3 1 -- Con---4 Reom-Is-ol-at+o-n- -i-gh-Rdiatn----. 3.3 119 ,,,,. a,, ,On,..,., B 3.>---91 Post..........otonAI'toring (PAM) Instrumentation--B-3-.3-1-24 .emot.e-Shutdown-S-ys-te 3;3-145 REACTOR--(OOLAN-T-S-YSTEM---RS-)-............. B r 3.4 1 -- -RCS-Pressur-, -T-Temperatutre,*,,an-'4F10w-Dpa, r... u-re

f-rom-Nucleate Boil~n-g-'(DUNB)--tI-mis...

,,~~ ~ L,, I,, fit Ig L-.......... B 3v

  • __-R-CS--Mini-mum -Temperature-- rCr i t i -L CL I "LY

--RCS--P-ressu-reand Temperature (P/T) 1-t... B"- R..3B 3.4-17 .... RCS Loops-MODE 4........... B-3-.2-*2

  • w*

5, -3 4 - 7

cS--Lop-s-OE5 Loops Filled

RC-S--Loops - MODE Loops-Not-F4ed.. -

3. 4 ___. -re-s-sur-zer Pressuri*zer _*a.y Valves B 3.4 43 r I I. 1-1Z* rU V PCI I~ lO CLL U WC1 --- Low Temperatwe-Overpressure L-TOP ,System -RCS-Opett iona- -LEAKAGE..... RCS ,Pressure isolation V-a----

I 1_1 A-V_ %.I %j1%Vj. U 3.4 47 i--
onLlL,

.jil n n A rý Li J I B..

3. 4-53

{PIV)--Lea-ageQ -3 A. B3 2 '~ ~~~~~~~ J.'~If f'tn J?-. f i A D~.. 1J ~ L L I.J

  • I
    1 LM AIII At L2 I Ij vi

-7 1 CL yj" -RCS-Speti-fl-eý i vi ty................ ..-3 -7 JtdT~U~T~ aLU ILi .- M11 JUM 111=3 I. I LY _V T. 7 -lT7 EMERGE-NC_, CORE,OLI-S-Y&TEN---ECC)- -SI.... Core Flood Tanls--(FTs-)................ CC-S -O e ra t r a-ng...................... -. -CCS - Shutdown atoed Water Storage Tank- -(BWST-.... n r I B 3.5-B--3-.-5-9 B -3.5 21,a w-.- -q 0 3 *,c B 3.6-6 B3.6 15 -- *INMEN-S¥STEM;.......... Containment Containment-Ar-L-ocs-k.-.- 'LiIILa II1191f111L X-iLiv IL11VcVs -Contai nnen t P.res-s-ur-.-.... t u B 3. 632 r, -F m e -tI LII~~ eL I i ( ea C D.- _LJ~ U I-l -1 I U_-11 g .JpI a nd pu ol I. ih a flilft -Syste-so.l-.........S...e..... (conti nued) Crystal River Unit 3 vi Amendment No. M

TABLE OF CONTENTS B 3.6- -3+.6.7-.. B 3.7 B --3.-7. 1 - B-3--7. 3--- 1-3--*.--.---- 8-3.- 7.5-B- -3. 710 B-3-7.7-B 3.7-13 B 3.7.14 B 3.7-15 B--3 . B-3.7.v17 B-3-. 7.- &- Bj 3.8.-I-B-3.8---2 B -3.-8.5... B-3.8-.6 fONAThM TEMS.... eont~nued* - -Cnta-i-nment mergency p p1 Control (CPCS), U..-- V PLANT SYSTEMS...................................... B 3.7-1 Ma-i-Steam-safetyoValves (MSSVs-) 3....

Mai - -Ste-am-- The l-ati-oft V-a lves-9-M¶SI V-s9-......

3.

7 Ma4-n Feedwater-e-Isol-ati-onf-Vave---fL-B 3.7-13 .- u-.-i-eBVa........ Valves (T.Vs) B Turbine Bypass s................7 1: -- Th¶ereny edwter-EF~ -Sys-tem-. ~ .-- B 37--2-3 .Emergeney Feedwater---SI*- Te2)I' -.- B 3. 732u .Nu&1e-ar- -Services Cl-osed--Cyc-Te--Cool-ing Wate-r System (SW)- B--3-7-36 _____~~~~-ý -7UA-A~ J ~ --Nttc-e-ar Services eatrSstem-... ... ee-ay +e-Hat- -Se-awate-r -Sys-tem..... aSye--Hetem -H -Con-tro-- Room Emergency Ventilation Sy REVS)................. Fuel Storage Pool Water Level...... Spent Fuel Pool Boron Concentration Spent Fuel Assembly Storage........ -_1. 7-46 3-7 -5 .., 3 / 3-7--6-0 B 3.7-661 B 3.7-6-94 B 3.7-7-27 .. B 3,--7--7-7 3-3-8.7 ... r3;.--71* CL lU Ial .y JJI_ L II %,L VI Ly. -- Steam Generator Level............... -Con~trol--Compl1ex-Cooling System... tt--and -tarti ng AM r........ ................----13 ---- -ELECTRICAL-POWER-SYSTEMS........................... .-B--. AC-Sou.res-0pe rating........................ B---3 8-1 SAC-Sources Shutdown 13.-8--4 +)--Sources-Operating n+ -B-3--30 __- { ,--y t-re-s-htdw--......................... -3;-8--49 B-a-tte-ry-Cel Par ameters...... - inve.rter .S

8.

B-3. -65 bIti on S.......s-Operating............. is ib ion Systems Shu d,,.................. B 3.8 77 B-3-.-9-- B -3.9.2-B--3. 9. 3-B-3-.9.5 B-3.9.6- -REFUEL-I-NG -OPERATIO)NS -~--............... Boron Cr-- ntr t +o n _ +- +.++. -. - Nu-,lear -Instrtment-t io----.. Cont-a fnmernt Pe-net-rati-ons.......... ii -~ n '~It i rLI n ] a.,: A, a t o o B. B3. 48 ul; "y 11;1+1. I _IIIkU V O. I I. I I1j ",I I%,A kU I "Il I I,. --... reul-at-on-ig4¶Wate-r -evel C-i -on-Low-Water-Level Refuel +/-g -- na r--Water---eve-1................. Crystal River Unit 3 vii Amendment No. 18-2

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions

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

The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. Te rm ACTIONS Defi ni ti on ACTIONS shall be that part of a Specification that prescribes Required Actions to be taken under designated Conditions within specified Completion Times. AL-LOWAB-LE-A POWER-- A.OWA -LE--TH ,,,A.L O ri^L,\\ s,al b-- t--e-- maxi-mum reactor core heat tra,,fer rate to the reactor coolant permitted by consideration of the number ,nd un, fguration of reactor coolant pumps (RCPs) in operation. AXIAL-POWER IMBALANCE l -- ~ ~ ~ n -1 1-Ve-Tnet-power in -T-ne top -ha- -f --of-the -coreexpress-ed-.-a--. percentage-of n ATF-M TIIn.PIo l*ll N F T RAT_1D TIERMAL POWER RTP) m.,inus the power in the bo-ttom-half--of--the core expessed--as- -a-lpe -rentLCJe of--T-P. A__ cf 1 i 1L I ý -I L i- " I RODS--APSRs C+HANNEL -CALIBRATIONW u eto-ctrl--ta+/--u--powed+is-tibut4on---of the i-reacto-r,-c-ore-. -- The J A P 21C VSs-r --i--i-ne manuial-y---by--the--operator and--are -o-r t -ippabl e- --A- -CHANNEL--CALI-BRATION-shal-,-ethe -audjus-tment, -as e-ee a,--o the-I hannel outputt-sc,,-that-it responds---wi-th-hn-the--nee--ay range an acuac d-s a,,10 accuracy to-known values of the parame-te. he afnel montrs-e. CThe-C 'ANNEL CALIBRATION -shall encompass th--et-re,,,,, channl, i ncl udi-ng -the-reqtired---senso-r, (continued) a-t-afaY1-lLl 1P Unmet ons, -andG srna-1 melude-the-(ANLFUE1NLTEST.-Cal ibrat~o of--s-t-rument channels with resistance temperature detector.RTD)--or -,thermouple sensors may consist of an inpiace qualitative assessment of sensor behavior and normal calibrat40 -of- -L ereM-aining adjustable devices in the e Whenever e"s'"l--'e-'""-1is replaced, the next required inpl-1-e a ssessment consists of comparinmg (continued) Crystal River Unit 3 1.1-1 Amendment No. 149

Definitions 1.1 1.1 Definitions CHANNEL -CALIBRATION- -tbe -th sen s-i-rte-ict tnhe-- t I y -(,onti,,nued) installed sensing element.- The CHANNEL CAL IBZA-T11014 may be-pef-fov-rrned by -me-ans- -- f --ay serges--of squent ,-oe r a 9pp ing-- -or-a toal chane -step-s-- so -,,t-t-te-* e h-n-ea cal ibrated-. I CHTCIANNEL CALIBRATION shall also include testing o-f safety related Reactor Protection System (RPS)-, Engineered Safeguards Actuation System (ESAS),--and Emergency Feedwater initiation and Conto (EIC bypass functions for each channel affected by-thfe bypass pe-rat-10n. V I ,..LIALMI. fI j I f, -

  • IOLEU Iq~

.L2 L I-Il Iy L*II tl~wJ) C H E ANEL--C+-- -A CIINNEI -CHECl--sh--be-*- "..... U qu tati e s -b! observat-ion, o'Iof I beh a -a-vi u"r d uri ng.--p.e.r ati on --Thi-s de*termiat-ion shal-i-cl-.de where possible, comparison ,fthe chanel indication and-s-tattts--to-o-r i-ndi-e-at-an-s -w-status derivted from-4independent instrument cha.nels measuring-the same parameter.- CH-IANNEL FUNCTIONAL-TEST -A-A--HAN*NEL FNCTIONA E-ST--s-ha-be-_. a-.-- Analog channels the injecton-of--- -'mulated or actu-al--+g--- ,,,-a, n -tho -*he-e l- -z --el--ose -to thte--senor-as practi cabe--to -- ve OPERABILITY, including requi red alarms, interlocks, I UM-switt-tch 11%--te-+/-;; -t;n-of-simulated or actual signal-into the channel as close -to the sensor as prac-ti-eab-le -t-veri--f-y BUT,~ i n cud ing required al-arM-and-ti-ip f unIL *I.Ins Te-h-eSA5CH',.- E',FUNC-TIONAL-T EST -sh-a,,,-l -a!Iso include testing of EHAS safety -rel.atedbypass f-une,OnS--for each c hanne.l.affected-bybypa-ss operatlon. (continued) Crystal River Unit 3 1.1-2 Amendment No. 149

Definitions 1.1 1.1 Definitions ont-nuc-1IU) CONTROL -RODS-CORE ALTERATION --..... CORE -OPERATING LIMITS REPORT-(-OLR9- -- -- _-

CONTRO ROD--s t-+/- I-b-a-1-]--frl-l-1length safety and f-e-a -1.. --rods-tah--- ar-u sed to shut--down-the r-e-acto-r ad Joto power level during maneuvern operations--

.... -CORE---L-T--~AT--~I sha-I be the movement of any

uel,

-oure-sor other, eactivity control components-, with*n-the reactor vessel w-th the.vesse-l-head removed-and- -f-tw-- -th+/-- .- Suspens+on-of COREAL-TERATIONS-shaIl not preclu-ede COMp,et ion of movemen-t -of-a-componentn to -a--sa fe - pos -t 1-on-- .... The-C-R--i-s--the-unit -sp. " docment that

  • F'=L

,,~ ~ uu u,=r-r provides*, cycle, spec f c paramete r--* t- -for--the current reload cycle. Thse cyce, specifec -y sthlt I-be-dete, fned--for--e ad---]e-n LL I UIAII`- lVi I 1_ Ia ý I I 1%NuL I~i m 0 I a .m. I,1 Ilu o-prai-on-w -th-,n-these li~mi ts +s-addre-s-sed-n DOSE -EQUPIAENT 1-31 ofL I--L mici J-roe-uri.es-/g-ram) ~rodutee-the. -- mtyo-d--de isotopic mixture of i-131,- U= LIjdL _l'JI=IL~i LIL Null11 -th alon--ol i 3 1 2 1 11 3 3, 3 4 -, The thyro-i d dose ~J .r~. con lversion Tfatol U INJU be--the MIlisted in International Commit-tee--on Radi-ation-rte i on p-)-- p, -pIeme t--t-Parlt -I- +'I9--, 1-.4-- II.I T1TOl -i Af"rA1mTrpOM Fj., E - - -A V E R A G E D-I-NTEGRATI-ON--ENERGY-Equivalent in iarge-c Organs or lissues per +/--n'c-ie of Unit ActivItyI." -I sall ube the average (weighted in proportion -to- -th concentration 0of-ea"aonu-e-+/-de "i-N-reacto-r -eoo ant-a-the i e-of--samplII)-of-Ile sum-ofl the average beta and gammaeneigies-per disintegration (in M1eV) for isotopes, other than, i-odines--- with half lives > 1:5 minutes, making -up at--teas95%*of-the-tot-a- -not-iod-ne-activity i-n the-coolant-. EFFECTIVE--FULL POWER-DAY--(EFPD)--- - l l . I -I PD shall be the ratio ot--thec-number ot hours ofproduction of a given THERMAILOWER--to T~iE-RMAL-LOWE 1-to---he-RP"'-One--E, PD- -i-s-equ-i-va-I-en-t Vl l"' my con tinued (conti nued) Crystal River Unit 3 1.1-3 Amendment No. 149

Definitions 1.1 1.1 Definitions EFFECTIVE FUtLL -POWER -- DAlyEE-- P r c...r t i r-nue-d) -~ll*-1rr rea-t-o--e-re-at-- RT o for o ful Iday.", 1ne-EFPD-1 s 2__V_9 MWl*.t times 24 hours or 626-16. MWh-.) rL - rere nrrmn&trr ,rnar LI I L. LJIL.I~SJLJ~~... I I.. LL'VVI'. I L.I~ I I I1 LI J.L.. I\\L~rJIIjL I +/-IIL 3111L1 I UL% LI IL L I II1* JLI1 II.tI I 1JI) /RISI.J lA..-Jl I IMJL-(EFic)--RESPONSE --TI-ME--.... I IIL* I v I I I III YI II=I LI IC IIIUII LVI CU q J. pr li a LnII -ex-eeeds--i-ts---EFKC-act u at i on -setpu-i-nt--at--the---hanreI sensor until the emergency feedwater equipment-i-s capable of performing its -s-c-a..ety--fuInLtion (i e., va-ves-t-ravel to-their required positions-_ pump U I,, I I .I

  • pI 111 I !

II LIll I I I =1.4U I I ý_ %A vaL I U .3_ s-ar't-ing a---nd sequence loading del-ays-,-where ENGPNEERED-SAFETY-- rrArltnr zrrrN nrrnt-lfr applicable. The response time may be measured --- meanfso - of--- y---seri-e- -seuti al-, overtapp-i-ng, -or tot-aT s-teps- -so---tha*-the--e re eespo* s-e tine-is measttred;. The--ESF-RESPONSE -T-IME--sah-a---be-that--time-- from when the monitored parameter-exee ds its ES-actu-at-on - setpo-rnt--at- -the-* ...senso until--the ES--equipment is capable of perfoming -Its-5a*-,=Ly ftnet

  • o,---e-i,-the, vaves-travel -to-.t.e.-r I

I" I WINL - .LJI J I\\L.JI U ý- TIME. -%I. I, P P a ~- M-11i[ equLired-*L

values, fLC.*I.

t Hime Snli iicl u i~~i die-generator startig,-and-seqiuenee---load-i-ng de-ays i-,he-r\\ T-ap Ie----T he response time may I ve measureO by means -o any series or sequential-,- U V ýI I ýIFvI I Ii. II ui ý L UL t i %JL ~ ý3 CILI* II IIL II E -LEAK-AGE shall be. LEAKAGE-

a. Identified L[AKAGE l

= I -I .At~t, sucn as tnat from pump seals or va-l-ve-packi,,n, ,thati*,aptured and eu, ,du eted-to--eollection s ,sy-tems---or--- --sump or collecting ta-"o.- I r A I/ A Pr L LXIU I I LU LIC %-VI I LCL I I IIIICI I L CtLIIIUJai IC I C from sources thatI are both specifically located andy qu-anAl;Fied--and--Okown-not-to inte ewio te-rat on-of-l(ka ned) Aaltarti~n Q ta" S~nr ni-

  • A k-nwroalI-ro

ýV V Ic Ps bOun -aryt~~tt-(conti nued) Crystal River Unit 3 1.1-4 Amendment No. 2-H

Definitions 1.1 1.1 Definitions LEAKAGE- -.-- Reaector Cot lSy ys t^em -*%"RC I LI I Ou a u c~ami Vcllc i LaLI ~ U ~a I systeir Lprjmajzýsecondarv rFgKAGF. W. -IUI=-in-%-i-fi4ed-LEAKAGE -A-l -LEAKAGE-that -i Snot -id-ent-, i',ed- -EAKAGE. c.- -P-ressure Boundary L EAIAGE A Ar-ri I LL 'II'JM.-- T* AL"*T L Ui Illi{].l V LU -UI ag V LIUI*.IJ-) throgh--a-non isolable fault in -a-nIRCS-C-oMPOnent boy, pi* pewalriA vs--,-l MOBE- -A -OE--s-hall correspond to any one inclu,,.,ve ,,,,n-at-on-o f---o rereacti""....."- "o"vit-y o,- level, average reactor coolant temperature, and reac-toirhe-vesse----ead closure bolt tensioning NUCL-EAR-fEAT.-Ft-UX-HOT-CHANNEL-FACTOR (FQ (Z))-- FQl--j -lalI- -oa -1 4-near-power -ensi-ty in the core divided-by the core-averagee lU I I I U I IiICaI JVVV I U.IIT IM Ly1, 0.z oUIIII I1V IIUIIIIII1.I f-full---ro-di-mens-os-- NIfILEAR-ENTHALPY-RISE -- HOT-CHANNEL-FACTOR-(FH)- OPERABE-E-OPERABILI-TY.. nt-eglrl--ofT-4near power- -along th fuel rod on ,whicii,,t-rmm departu-re r--nu-elte-boilgrat i-c oc-c ur-s to t-t--verage-'-Fue-l rod-power-; A--SYStem

Subsyse, t ra oponet, or-ev-te sha-1 1 1bePERABLE-when -i-t-+/-s--ap

- of-perfo'-m--g -ts specified safety"and when all ne-essi--a-ttendan- , -eontro-s-, norma-l or emergency-e-Iect-ri -po i---*-oing-and seal water, lubrication"-- "-.-aux y ,Mtn,,- W "4 W-.4 -I.-

    • 1 I F/ll l

i L l( I-1 "I

l iI g i i

i

i.

l a L ll subsystem, train, componIetlL, or dev ice LU 1I Iorm -specified safety function(s) are also capable of-efrin theirdelte sppor-t-uneti-on~s-)-. l PHYSJS-- -TESTS-. NIIN GM4-M-1 r-PRTfl-4-04!r-C WArrmn I n m 1 1-m 1..

t...

I.. r-l --

  • the-reactor core and related instrumentation.-

(continued) Crystal River Unit 3 1.1-5 Amendment No. 2-23

Definitions 1.1 1.1 Definitions PlYSICS T[Si-S- -c-ontlntted)- These tests are.

a.

i n Chapt-er-13--'* 11T+t-i1 Tes-ts-and O-pe-rat +/--n-"-o-f--thie--FSAR--

b. Authorized under the provisions o 10CR50. 59; or q
c. Otherwise approved by the Nucllear Regulatory Commission.

PRESSURE AND TEMPERATURE LIMITS REPORT-*-PT-LR)---- --. -ri nr-r -L II1EV r I LI* 13 Lil: UII I L 3tJVI. I I I U*I..U LIdIsIure L LI-Ia L p---ro-i-des----the- -reactor -.-- e-p-res-s-ure-anld Eempera-urlimits, imcludimg heatup and cooldown rates, for the current reactor vessel fluence period. Th"Iese Vp~rC-Ssu-re andI temperature limit I-ll V-'E lI I LIE-Shall be determined for each fluence period in L I II I VV I LII [J)VL%* I I I GL I lull J. .4L..L r I aIiIL operation within these operating limits is audrdrese+nLC .43 "S rss ure -and Temperature-4mit-s. -- QPT shall be defined by the following equation-and is exp1essed as alpercentaIe. QUADRANT POWER -TU-T ( Q)P T ).................. I, N t -rv'='-- Ally,, lAJI V 'JU0LII aIiL I tu L- -1i ^.......... I nV~l y ru~lowi I dlo Ia UdUdllLan I RATED-THERMAL-P-WER-(-RTP) fl r A r~fnn nnn~~IrTfnkl n"r r'* -- I---1 1 L L IN I I 311ai I VC CL LULCL I I VCL LUI LUI l I IVQaL LI a1131 )C rate to the rL... nnr nrcnn&acr *rrnar ... L...1 1 L... ... L...... SYSTEM- (RPS) RESPONSE-TIME--- - --- I I Iý; I\\I J I\\L.J I I% JL. I LI'IL 3110.1I I IJCV LII I .L LIIIIV I IIL II VGc.I -from when thee moni*to.red pGarameter-eeds I-t-R-P-S -tr-i -sY 3 ]int at tLIh ch.iiiiV e se,-sor UiiL il "1 J Ve I V'.. LI I power I 1 J*/JVVVI 13 i IILVeI I UpLe d L thlle L ILI-- rod Fll*'l*,9*-*-p, sred by means of any series of sequential overlapping, or total steps so that the entire response-t +me-As--meastfwed-II II I I I It I*IM lWlMl* I I IM I II HI-I I -rec-t-+v-tr-by---whi ch the reactor-is-st-i-a--or (continued) Crystal River Unit 3 1.1-6 Amendment No. 2-2-8 Crystal River Unit 3 1.1-6 Amendment No. 2-H6

Definitions 1.1 1.1 Definitions SIUTDOWN-MARG-IN- (SDM-.. -woul--be-sube iiea4---from-t--present--eond-ition (--eonlt-i-u )} - -a -ur-n-- a -C-W ONTROL RODS (sfety anldregttlatýl-& are f-u+/-Iy-i inse rted except for *k-.,,-s-ng"-e-C ROD of h'igh'e-st -reaetivity worth, which is as-st -to be fully withdrawn-;--a-d -b---In MODES 1 and 2, the fuel and moderat-r tempe ratures--are-changed to-the-post-trip-RCS* average temperature. D .U.t With any CONTROL RODS not c-apabl-e-of being fullTy +/-n~rte---hereaetivity -wocrth of-these CONTROL -ba ,counted for in the determination-of STACGEREI) TEST-BASIS. -A--STAGGERED--T-EST-zST--sha-l-I-eonsos-t-- ofthe tes-ti-n-g--of- -one -of-the--s-ytems-,--subsystems-, channels, or othe-r-des-ignated components during the -i__te-rva+/- specified by-the Surveill-ane. Frequeney; so-- -that--a-T--s"y-em*s -- ulbsystems-,- channes,,or other designated components are tested-uri-ng n Surveilla e Frequency interval-s-, where--n-f l +s-te~ t-otal u-mmer-VF 3Y3-te11-, subsystemsy-ehIarnnels-,-or-other-des i gnvated mponents--i--the associ-ated-fnt-t.-on-THERMAL--POWER - -----------THERMAL --POWER--shI---be the--,,atota1 reactor core heat t-r-ansfer -rte ,tothie reactor coolant. Crystal River Unit 3 1.1-7 Amendment No. 14-9

Definitions 1.1 Tabl-e--.1-.-i-(page-- of----) MODES-(-) MODES TITLE REACTIVITY %,-R*T-*-, AVERAGE C-ONDI-TION THERMAL REACTOR-COOLANT (K.> POWER I-EMErATURE I Power OtperJain NA 2 Startup O --5 NA 3 IlHot--Standluy <-0-.99 NA e-280 4 Hot <-0.-99 NA 2-80->-T >-200 od Shutdown() -NA 6 Ref Lt-+/-neg) NA NA NA (a-) (b) (4>-- Wi-th-fuel -i-n- -the--reactor vessel. Excl ud4-n dec-ay-heat. --A-tl- --re--ator---ves-Ihead closure bolts fully tensioned. -nereactor-vessel head ,osure bo lessthan-ul-ytensioed-Crystal River Unit 3 1.1-8 Amendment No. 14,9

Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use. BACKGROUND Limiting Conditions for Operation (LCOs) specify minimum requirements for ensuring safe oeaino hanlin ijd sr~a bfnuiankkA".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 uit-fis in a MODE-orr-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 unitf ljiis not within the Specification Applicability. If-sta m ar "d'sco"v......d that require entry intomor t-an one-Condition-alita-t t hihn-a sigl-Speei-fc-ation (mu~t~pte--Condit-ions-)-, - the-Re +rred-A-t~eons*Fer -eac-h onU I,,,, must be performed with1-the asseoiated Completi Ti-me.-- When-i n,mu,*,t*,pl e Condition*s* , separate--Compi-etien Times -are t-ratked -f-or-each Condrtion startInIg Ili I I Will th ti" me of--r Cthe suation-that-req-ired entry into th*e Cond it-i-on-- O--ee a*-e'diti-has been entered, subsequent trains, subsystems, components, or variables expressedi-n--the Condi ti-onedi-scovered-to-be-Ii,.ope-rable or not within limits-, wi4-I-t-,result in s a try into the Condition, unless I*pei.ally-stated.- -The-Requ*re~d--tA,,ms--ofhe--Condition conti nue to-appl-y -to--ea*h add-, ti-,,al fai u re, wi th. Co oTis based on initial en, into the Condition. (conti nued) Crystal River Unit 3 1.3-1 Amendment No. 14-9

Completion Times 1.3 1.3 Completion Times D[SCRIPTION -1 I effI-Ue)- "tfm~~ t r a n S 'hqt e M Fr~Mr~npn"Pw-I -vaVr-aoe, exprAsseo in Ic.1 r_ On, I3 - -- v-r- ) inope-a bl e, ot Wi,,thn III ILa, L,,11 - e-tIoIe I -m-a be e-te-ded-;... To-appt-metion Time-extension two criteria must fmrst be mt. The subsequent .n.. per-ab.

  • t-a-.---.. Mttst--exi-s-t - oncutire-nt.- wit the f,,

f rst -iroperab41ty;, and ...... M......s-t--emin-perab-r--r-not-withinlimilwts--af -tethe f-i-rst inoperability is resolved. T-4--- trvt-2 0%0". t,ýA I m 1I=f -r ~ a-n-2 P--, Ar L .>JtL ~JI~JI tII

  • I

~ A .I

    • J*

1 ~LI ~L 9 A-L~ fl WL l 'll L "j.I.Iýl C LilZEZ i ,1CIIL I lII ./JI

IAL, I I I L%

,.f110. I I IC tlmfit-ed-to--the- -more--res~+/-ete--f-e~e-i*. 'he stated Completion lime, as measured trom the i-M-tial-entry into the Condi o p-l].us.a..r.. ad---on 24 -ho-rd;- -or -The stated Completion Time as measured from discovery of-the--sub-sequent .*,,perab 1 i*, ty. -b-.ý ..T he- -ab*ove--Coipl et-ion-Tdme-ext-e s+0-d o-no-t-apply-to-thes-e Specif-i-cations that have exceptionsthat al-owcmpetely se-p~a-at-e--re--er, ,,ttry- -i-n-t-o--the-Condi-o..n--*-i"-r-ae-o -t ra+n, subsystem,-, omponent-o var-i-ab1e-p -i+/-n th Cond i t "- and- -separate-traCk--ng Vof-C-omppetion-Tes-based o ts re-entry-These exceptions are stated in idividual-Spec-ifi cati-ins-. Th-e-bove Compl etion Time extension-does--not apply-to--a tfonfptet or---l-Ime WIlt a m 1,Tmp r ýr;1p YAr P M1 I-- 1 Q it it I-- A " I k ^_ I I... l " I-r MI IC .:.4..--.-.J n-.- T '-,-I II -- I -,I -I I f-rom--a-peiu copletion of the Required Action versusth tine-oof "^-i'-- on -o-ime -odif ifd* by the ph rase "-from-discovery I (conti nued) Crystal River Unit 3 1.3-2 Amendment No. 2-29

Completion Times 1.3 1.3 Completion Times (continued) EXAMPLES -The-f-o't-lowi-ng -^-,,,,l-e- ,1ustrate the use o-f- -l-tion TI-Imes with dif fe types of Condi-ti.s ad changi ng Codi-tio-s. EXAMPL_-E-1-.3--1 CONDI-TION rE-DAMTON COMPLETION-TIME B--- RequI red LUJ.Be Ui I OLD J. I 6-houIIrs etm*t and __associ-ated AND _-Comp-l-et-i-Or Timeot B.2 Be in MOD5-36-hours -Condition B has two Required Actions. Each Required Action has---ts-Ownf -s aa--omplet-Ir-on Time. -ac1,Completon-T-iime is refeteced -to-t-h me t-Cond-ti-t.m--B-s-entered-.- eed Acti-ons of ConditionB-hours ODE5rs-- allowed for eah. MO-. -3 anda 4 hou rs)---i-s-allowed-Ior--reahing-' Codtiom -B-wa--entered. IfMODE*3 a C.- nrL r -are--t-o--be-4MODE i-n -i--total-of" ousi Otal--Of 36 hur (not "*).5--o the. '-ti-me--that: %JUL. r I eached -LII e-or i s reae1-1e--4-nt---3---hou~r.s-,- LI I% L IM III a I i'J V U I.I I~ .IIIII 1.L .1 I c iI-InL J II It I I be c-atu-e-~ 3-6-hour-s-.- a r, ".r I ri" nn 2,FI-- r If-f L2... nn*lr-ir -2.. while in,.,- 3,--the-t-ime--allowed SI I ý I I I I -j I-IWUL .J I ' Li I I%- I A L JV I IJU Ia3. (continued) Crystal River Unit 3 1.3-3 Amendment No. 14,9

Completion Times 1.3 1.3 Completion Times (continued) EXAMRL-ES--- -f-eolrti-nted) EXAMPLE 1.--3-2 -ACT-IONS__ CONDI-T-ION REQUIRED-ACTION COMPLETION T-IME A-7--Ehe-ptmp A. 1 Restore pump to -ay inpeabe. OPERABLE status. B.--Require B.1 Be in MODE 3. 6--h-o- /-r -11 -,a k d --aSet f~-iated AND -opeti-on -Ti.-e-not B.2 Be in MODE 5. 3J6JhouIrs -When a --pu-mpis delared inoperable,. ConditionA is entered. i-f-thepupino restore to OPERABLE status within -days, C-ond-t-i-oii--B--i-s- -a]- ente-red --and-the -Com-p]-et-ion Ti-me tco-lcks-for Required Actions B.1 and 3.2 staI rt. fthe inope-rab-Thf-pump is restored to OPERABLE status -after Cofnd4tion-B-+/-s-n d-- ndi-ti-on--A-and--B are exited-,-and ther-efore-- The R~equl-re-d M-AetIIZOs-o-C`Dnd-t i-on -B- --May--be terminated-- When--a -seond- -pump--is-declared-1-noperab-le-whrI-e-te-first pump-A-i-still inoperable, Condition A is not re entered*for t-hl-se-cond pump---L-CO-- 3-.-O-.3--3is entered, since the ACTIONS d,-o not in-c-lude----Condtion -for more than one--Anoperabul p-np. The--Completion Time clock for Condition A -does-s-top aft-er-tC 3--.-O;-,sentered,,but continues to be..,ac1"d-f-1rom the--t*me-,Condition A was initially entered. Wh-il-e- -+/-n-1&O 3-.03, if one of the inoperable pumps is restored to OPERABLE status and the Completion Time -f-o ~=,,uLt h a rmplton-T-i J - J CdiU iomI I p A ~ ha not J exired ~LGO~ 3.0. may be exte andU~II opera-tio-continued in accordance wit h Co o I i -W--m-eO--LTJW.J, it one ot the inoperable pumps is res-t-o-red-to-OP VRB BL' status and the-Comp et-on-T-iie--or c on--A -has expi-red---LCO-3.-0.3-may-be-and (continued) Crystal River Unit 3 1.3-4 Amendment No. 149

Completion Times 1.3 1.3 Completion Times EX)MnFLES -EXAMir E-4r,-2 (continued) ,perat-ion -cont+nued--iaccordance with C Co,,plt,, H-,,, ,,,,forCondition B is tracked-from the time--he %AJII1.1I L ILII I %'.j111 JIV ICLI LII I II111C ~ I ~l -n----restoringore of the pumps to OPE.RABE--a Condi-iomA-CompletiOn, Time is not reset, but continues from the-time ~~,,he fi rst pump was decl ared i-nopera T-h-Compl'etion TiMe may be extended if the-pump restored-o OPERABL-E-sta -was the, first inoperable pumnp. MLA-f 2hour extension to the stated 7 days is allowed,- provided thi-s does not result in the second pump being inoperable for >--7--dS-. (continued) Crystal River Unit 3 1.3-5 Amendment No. 149

Completion Times 1.3 1.3 Completion Times EX~AMPLES %,eontiinueud) EXA¶PL&4A-3 -ACTIONS___ CONITIN nREQUIRED AACTION COMPLETION TTIME A. -One A. Restore 7-dy Function X Function X train -t-r-a~nto OPERABLE ope.rabl*. status. B-- One B--;i Restore 72-hours -.. F*n-t-in-ti Y ,uncton Y train -t-ra--in to-OPERAtLE ... nope.rab..e-- status. CCOe Resto 72-hou-rs -Fn-t.-o..-X , Function X--train -t rai.n

oOPERABLE inope-rable.

tat s.-- .AND OR One

  • C.2 Restore 72

.hot... -F*ftit-on-Y Function Y train ..t.ra. to OPERABLE ---mnope-abl-e.- status. (continued) Crystal River Unit 3 1.3-6 Amendment No. 229

Completion Times 1.3 1.3 Completion Times EXAMPES- /EXAMPnE-I.-1-3 (r .-..t--1u eU- -Wh-en--one -tnctti+ X X-trai-n and one Fun,,*tiLon -Y-tran--a-re in~ope-rable, Condaition A and Condition B are concurrently applicabl-e.--The Completion Times for Condition A and Condit ton B are traeked separately for ea*h train 3tart*i-g from-the time eaeh-train declwasda-red, nopera15 ,e--a-the r _ A - A -_ I I IIU I L I I I VVL3 -I I Lý I

  • _U 3r%

Z IJCI C L. -- IIIJ CI L I I I I I IIIC I 3 e.s-ta.-ished for Condition C and tracked from the time setnd-tra`n -was deal --^ - -"derra-Ihe

  • } I

-tl l-. -I M.. I_ M.*It II -U I* I VII I MJL .* I , I M .*.l~l T r= n AJ A F*_ An r F A 11 1 U r% '"P 1--&-w -*the--s-pe-e-f-ed 4-f l n1-I P TI.m-(r-A1i1--I-.a. wVIIiI I L LI tI II U ill l FIII ILItl II.. U II n L I.I A I LLl w L l I.- 1 I LW -me-fo-Required ti-A-A- --- Inot exp omeration may continue ordane with Condition A,*-- remaining Completion Time in Condition A is measured from the time the affected train was de-l'a-red inoperable E.---e--- mn4tital--efntry---nto-CondIt~ow A)-.- 1-t-.-s--possalternate be tween -'onl6t io,, A, aU,-and-l ~l-Tever LI I L IJ I 0. L I tI I oU I U I UIIL I IM I I ItIh IO H IIrowe er witot-ever restorin systems, to meet= the LCO. ,,oever,, U ,,J II Iy aJ VVUU I U IIL ,,3I 3La I IL ,I LI.I LI IC UO. , a LUi C-ompI-et-ion--Tities-...

The,

,o-re-,- there contrl--t-i mit th, maximum ti.me--fallowed-for any combinat-ion.--o Condi+ti-on-- -tha-t-esstin.a-single contiguous occurrenee of-If a-ilI'ng-- -0 Mee-t-the L-Q-These admi ni strativecontrols shalI ensu, tha**--te C-ompnl.,et-ion T-imes-- r--ho Conditions-are--not-i nappropri ately extended. (conti nued) Crystal River Unit 3 1.3-7 Amendment No. 22,9

Completion Times 1.3 1.3 Completion Times EXAMPLES EXAMtvPft-rr3-4 -ACTIONS IMr I 1IIInr-I ArTr*L~l CODMITIO fXEQUIRED-AUC TION COMPLETION TIME11 A I-@-One -r-Mor e A.1 Restore valve(s) 4 hou.rs -valves to OPERABLE B.- Req red I-1----Be-in-- E 3. .Acti-m-and -assoiated AND -e-t .on ..... met - n tB.2 -B -n-O E -- r o r -A-s-inl-e-Completion Time is WJ-'A f=^ SJI MWiky ^r=A~t I t-tEle sa e time. mne Completzion lime associated oiLthCondioti-on A is based on the initial entry int ConditonnJ----A A -I-nat tracked on a per valve basis. Dedl-aring subsequent lalves--inoperab'b e -9wh'*" 1eond-i"-ion" k i-s I-,,, z u u , n-A-* i--11 --in--ef-fect, -does -not--t-r --the--t-rackig -of--septra-e Comp-etý o-Tmes.- -Once-one--hf-ee "vales... has been restored to OPERABLE stats, the-Condition A Completion Time-is-not-rese-tbt---ontirttes fr*mte he -first-v& e--ws-dec-lred-irnoperable.- The Completion -Time imay--be--extended-1'f--the valve restored-to OPERABLE -stat us- -was--the1-f-irs-t-1 nope-r'abI e valve. The Condition A Completion Time may be extended-forup--to A 4---vo.... provied^ this d .es o r valve- -be~ng---i-nprb-e-f->--- 4-h-s-. -I I " I I II I JIL L"CI ýU1TICI VII I i1m1 U1 T X1ISIL IUU Ilt l LIIn rALXItI3 U1 j exp,-es-,while one or more valves are still -Inoperable, Condition B is entered. (continued) (conti nued) Crystal River Unit 3 1.3-8 Amendment No. 49

Completion Times 1.3 1.3 Completion Times EXAMPLE -(continuted) EXAMPLE --. 3-5 "ACTIONS NOTE Separte ,ondit-ion entry is allowed for each inopera-le v-alye-. C TONTION TON COMPLETIONT-TME A. One or more A.1 Restore valve to hours valTves OPERABEstat-s-. ....... ifloperabl-e7. B-. Requ i-red B-3e-.n--MOBE-3-- 6--ho... __as-so-i-a-ted AND Completion

Time--ot

-. 2 -Be-n--MODE-4.- 12 hours The Note--above the-ATINS tab e i s-a-method-of,-modi-fy n how the CoI leti ime-s raked-.f-this -method--of mod-ify+/-ing Io-the Completion Time is-ttraeked--was--app*ieab-e on-ly-to-a specifi c Condition-,--the-Note---wou-ld-appear i-that Conditi-on-,-rather--thaft-at--the-o-o-tf h -T* --The-Note-o "-Vt r% be--entered-- seprate-lyfor e-achnoperable-va.ve-,--and Completion Times tracked -on a ---per-v-a-le -basis.- hen-a valve is declared inoperable, Cond-i-t-ionAi-nt-redand, its Completion Time-starts-I Subsequent valves are declared inoperable* Condition AM-S entered for ea*h valve and separate Eompletion Times start a-nd -a-re- --k---- each valve. L.I LIlC %AljIIIJICLI UII I lille a33% I CL LCLtU V ILII a VI IIVI III Condi-tion-A BsxptresreCn t o is for-thata-vae-.- II ,,Comp!etion, Ti mes ,associ ated wi th subsequent valvesi-n (continued) Crystal River Unit 3 1.3-9 Amendment No. 149

Completion Times 1.3 1.3 Completion Times EXAMPLES-- EXAPLE.3-5'CntiLIeIUC ..Conitin A-expire Cond-i-ti on -- is--etered-separartely-for each valve and separate Completion Times st-art and-are

  • W-Pl--d-
f.

r -Qp ,-1., TZ 7..1,- ýL-- .J A. IVC. aI Va VVC LIL t-LQU3Z C"LI lflL Condion--B--+s-,restored to OPERABLE status, Condition B is ext.d-^ for-o +/-1-hat a-Wve-Jii IL* Li ~II* iuLt I1"1 L~ii 3

  • AQIIIJi I

', i iU*. .iVTTUi Li p i

  • ..i l~

L IUi entry and tra.ing of.epa.e Completion Ti-me "-Com-e-*-- Time-extensions do not apply. S-EXAPL-E 1.-3-M6 -ACTIONS CONDI TION REQ E AC-TION COMPLETI-N- -TIME A.-One-channl A.!

Perform, Once-per

..... pe-ra-*e -. SR 3- .x-.x-. 8--hours OR A 'I n-0.. " +* " r% R1ed-uee THIERMAL &-u-- POWER to s---< 50% RTP. B. Reque B. -Be -In MODEr 3. 6-hot-s -etme-nt ___--mpet-k I*..... Z +__ J LIILI y IIIL J II IL IJII l / JI I I1 3 C LI I I JU LVVCCII I\\ CHU I I U A etion-A.'-T-oreA Required Action A ^ h a " Comeion-T h--r-e-2 extens 5lo,,-per SR--3.-;D--2~- tO--each performance- - t-fman-e-- A--'- A*,*,=-atrt~-*, e f -f--R , red Aon.--A 1*-Is-:-ftl-ow*-a -the-Requi red Action (continued) Crystal River Unit 3

1. 3-10 Amendment No. 1-49

Completion Times 1.3 1.3 Completion Times EX-AMFPES-EV AM Frl I 1. 3-6 "conti nued) +s- -. not- -me~t- -w~-Vi 8s-1 -1t 1 e -01emp"Iet i oen-T-me--(4m 1 -ud i-ng-L the-2-5 extension allowed by SR 3.0.2), Condt.ion-B is1.3 Ie* nt, ered. If Requred Action A.2 is followed and the Completion Tim 8-hours is not met, Condition B is e.tered-. If-afte-r--entry into Co.ditio-B, Required Actio-A.1 or A.2 f f-AJ A ___I I13 IIIC L y %.WAI II I L I V.I I i I3 a C I LCLI O.II ~JUUPCI CLL I Ui I IIiaLy LIICII corit-nu -in-Cond t-Ion -A-. -EXAMPLEL. 3-7 AC-TIONS CONDI TIOIN RE[QUIRED ACT.IONI COMPLETION TIME A-One A.! Verify affeeteI d I1 hour .. su-b-systm -u tbsYSt-em ...... n ,rab1-e. --.isolated. AND Onee-pe-r 8-houtrs t-he~reaftter AND A-.-.Restore subsyste, 72--hou-s to OEAL R-.--equired B. 1 Be in MODE 3. 6 hou.rs --- Ac-tion --nd ... -a-s-s-i ated AND -Tm oB-e M-3 -hou-rt (continued) Crystal River Unit 3 1.3-11 Amendment No. 149

Completion Times 1.3 1.3 Completion Times EXAMPL-E-5

EAMPLE (continued)

CRequired ActionmA.!ehas-two Completion Times. The-1 "hour omp-le-tortTi-me--beg-,ns --at e-t-h-t-time-thel-ond-it'--i--enered eac*, ,Once per 8 hours thereafter" interval benuo completion of Required Action A.1. I-f-afte,,Condition A is entered, Re,*uiredAction A.! is not me-t-+-withi-either the initial 1 hour or any subsequent 8-hoKu-tnterval from-the previous performance- .inel.i..*g-the 2-5%-extension allowed by SR 3.0.2),3Condito-0 B is entered. The Completion Time clock for Cond -A-doesnot stop aft,*e r-Cndti,, T nL Cot~ueSae ( re Codit-ion A was initially entered. If Required Action" Is met after Condition B is entered... Cond-it.i-.*n I

  • I*

and operation-may -ontinue-in.a-c-cordanceWIth i-totion -A-provided the Completion Time for Required Action A.2 has--ot expired-; IMMEDIATE When "Immediately" is used as a Completion Time, the COMPLETION TIME Required Action should be pursued without delay and in a controlled manner. Crystal River Unit 3

1. 3-12 Amendment No. 149

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements. DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR. The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, "Surveillance Requirement (SR) Applicability." The "Specified Frequency" consists of the requirements of the Frequency column of each SR, as well as certain Notes in the Surve-tanc-*ee-lumntthat modify perfo.rmane requireme...t. - - -I I .J I L II"L I IIa VVI1I I C a U. I VC I I II I.%IU IJ IJ U I I r-U %,I I I L, Frequency could expire), but wherei-,t is not possible or.not desired that it be performed until ,1sometime after the* associated Specification is within its Applicability, represent-pot-enti-a SR 3.0.4 conflict-*_-.0_Tavoid these co-nfli+/-cts, the SR (ii e., the Surveillance or the -frequency"- .-..stated such, -t-at--L-i--I s only ,"required" when 4t--e--be--and ,With an SR satisfied, SR-+/-30.4 imposes restriction. (continued) Crystal River Unit 3 1.4-1 Amendment No. 149

Frequency 1.4 1.4 Frequency (continued) EXAMPLES -Freque*neayN-tnres -Freqencie are specified. n these ex1 Tmples 2 -- I I I k.y JI L-I I. ,J1JL.,. I, I I ,IL I II tI I U/I,. .JI VIUi¥ j 1,, I J ,I-2.--afd-3-, EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Perform C ^"N',N,.,. L C,,LE,.

  • ~tx

-ya 12 hours Example 1.4-1 contains the type of SR most-oftem-encountered in the T^-einiea Spei-fi*at*onýETS). The Frequency specifies an interval (12 hours) 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 hours, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for operational flexibility.--The-mat*rement-o ,fthis interval continues at all-times, even when the SR is,-ot--, ti-,,r-to be-me-t--p J ,,,,,,such,aswhen the equipment is inperable, a var4-ab-l-,-i4-outs-+/-ei-specified limits, or the unit i4s outside theOpp-abl-i-t-y-' f-a' e---pecifi-ea on).- s-pe-i-f-i-e-by-WR..2 is exceeded while-t tvi--i i L Z IIIa MOF)-r--other-,s-ipei-edcodItonA i I-tle Appi-ilabi-i+t- -of te--Speci fi cat.ion, ---aCI-the%-Ie rformana -of-t-he--S* -T-uiveil-anee is*-snotherwise modified (refer to Example 1.4-33) -then SR-3-w.-.-*--be.omet appliab. i I I m * -r tne interval as specithed by -3.0.- -2is exceeded whi le tCe--un-It -is not in a MODE or other specified ConditioIn-i-the-App-i -aeli--t--of- -the-Specifi cati on for whi"ch -performanCe of - the--- "s requi-r'.-ed, the Surveillance must be performed within--the-Frequency-requirements of SR 3.0.2 prior to entry -ito--te--MOOE-o-ote specifi ed .onditon.. -F-a+lure-to-do so woful d-resu -4n vi o w*aa-,-on of-SIR-3 0.4 (continued) Crystal River Unit 3 1.4-2 Amendment No. 1-49

Frequency 1.4 1.4 Frequency EXAM-PLES-o-S Ce~oti ftu-ed) EXAMPLE_1-_4-2 SURVE *_LAINCE--REQUIREMENTS-SURVEILtANE-E PREQUENEY VeY --within limits. Once within +/--*-hous-f-te-r >-25%--RT-P AND 24-houtrs tereafter ,,aph-s--wo -Feut,,-- +/-.- - t-is a one.-,,me performance Frequency, and tHe second is -of-te-type--shown in Example 1.4 1. The logic-al--eon-nee-to-r "AND" indicates tat-bothy requ rementFseqnus-t-c-I-et... E-ac-h-ti-me rea-c-tor--power --i-s --inceasedf-ro -a power-eve' %- 1R 2-5% 2 RT*P, t-he--Srleu-s-t , be-peiformed-withih 12---hows=-- The-use -- f-' C e"--iudicates a single performance will satisfy the specified Frequency 'assum g m--..e o-oe-, Preqen-e-e-N--r connf~ected by _1A0D") -T.hi-s -type of Frequency ".-1! -C 1 n ~ 2 '11 I-n 1 -Thereafter" indiates future e-s tab-i-syhed-per SIR 3.0. -- but condit-ion-- is first met (Jie-, ex-te-nsion aloe Vy J--U-.-L-- performances must be M!**:

  • =

the "once" performance in-th-s LA(cuII I +/-l I ý1C.% LL.JI VIW'V'I A. %LI CIC a LU . J10 IN I r I LI I;. measurement of -%-both--+trVaI-s-tops_---1 1--.intervals

start, upon reactor power reaching 25% RTPT (continued)

Crystal River Unit 3 1.4-3 Amendment No. 14,9

Frequency 1.4 1.4 Frequency EXAMPLES-.----.. V A ll rl 3 I- -SURVEI-LANCE -REQUIREMENTS SURVEItLANCE FREQUENCY No-t required to be performed until 12-]hours---al er -. %J/O I'I V PerfIlorm 1channel adjustment. 7--days 2EJ.n~2~atn4s.saL .4-- r I I I IIIL l l lJ ___1 VVI ICLII V.I IILJL LIIV_ U11 IL Up% 1 J;l CXL IJII I. _5k1 pp P%-*%I -R in-PT-R Wiwi-M As-the--Note-modifi-ie-s--thecr**equired-l e,, -taheeof it is construed to be part of the "specified Frequency." Should the I day interval be exceeded while sJp~t (AL I~.jII Ia ~ flI I LII I.J I~JLI~ (Al IIJVTJ ..L.~. 11JU1.J (Al L~I puncl~ I ca%-ic

f

/O Ixi r LVJ PCI I [III Lim~ .ul VC]I I I ti% I Iic Surv~l-anee- -- i-s-7 stillI considered tor be within the !"speeiffied Preqenc-< -- Thiefrefore-,--T-i-the -Su-rve-i-}laffe--was-into .- L ~-J.. i 7 ~ A ro/ C 2 n ~ ~ L l u ll* y-T LV LI l ll LII I I U a y %ýpJ IU ' ) J/O l Jl J J. U,L ) II IL I 1 VI but-operation-wasthe2*R % Ra, u -to&d--n-t--onht--i-t.-,--a fa-i-1-ttre--of--th-e -S-R-or-- fai re--to---me-et--tte--LC-O-.--- -K soj -no I, RZIT Ni:f i 1.I L r_0 ir Lr.%rI F=R. RdMIPI MIIIN F4l 1

  • I I

P I -1 l Inn-3w-_ n.-t-MmIt n rlnrl nrD^Irn-cPot F. No. I I - U U*IE*

  • exterd 12 hou, s vv I th pow-1, ý 25/6 IN, F-

.on.-. reaches 25% RTP, 12 hours would be allow -fo* r ,omp et*,n, Surveillance.-- 9 if the Surveillance was not performed within thi-s-12 hour

interval, there would then be

,a failure to prform a--he--seefed F-requemey,--and--t+te-p5-of-SR 3.0.3 would apply. Crystal River Unit 3 1.4-4 Amendment No. 149

SLs 2.0 2--.I .S.FETY .LI-I.T-S-SL- -.. I nteSt 1..-* R.eactor Core-St-2.11 .1 1,,and 2-;--the maximum t-oc-a---ful--pcen-te-rlin temperatu-re-s-h-a-ll-be -s*-508O-&--(6T.5-E-3~)-XA(fu-rnup MWD/-MT-U)-°F.- Operation within this .l--imi e-ts---ured-by compliance with the AXI---POWER-E protective li-rn-its-pserved by the* Reactor Protection System setpoifntsr -i-n--LCG 3-.3;-1. "Reactor Protection System -(RPS) othrumen-tat4n-,- -as specified in the LR.

2. 1.1.2-- - n--MODES-I-and 2, the departure from nucleate boiling ratio--(DNBR)shall be maintained greater than t lh i of-+/-.-3--Nil-the ---BAW.2 correlation=61.18 for the BW co'ra-'I 1.132 f -or'

.tht-"B .T.l c.^ Operationr wi thiln ths 1511mit i1S ensu red by compliancee lhSL4. and with the AXIAL POWER aMBALAN protecti-ve -Ji-mits-y --preserved-by,--th~e-RP-S-set poi nts n L-Ca---33.1,-sse 4-e i-i-the--C-OL-R-MO 2H 1 3f 'and- -*-iRea - Coo.. a.*- -Syste-ft--fRS)--c-ore outle temp.e. atuire-and pres-su eshall be mai-ntied-above and to the left of the SL shown in Figure-,-.1_*,,, p----ten--MODES1----2---3--4--tand 5, the RCS pressure shael-----setpoi n-tained <- 275-& psi-g-.. L The-fol-wi g co-ac.t3-s.---be -ompleedh- ..2.1.In.MODE !..or 2, if SL.2..1..1..1, SL 2.1.1.2 or SL.2.1.12 is vi olat-ed, - be -n -- mOD4E 3-ihn--1-hur- -ou2,,,if SL 42A.2 is violated, restorecd-be -in MODE 3 with-in 1 -h-ou.r. (conti nued) Crystal River Unit 3 2.0-1 Amendment No. 21--

SLs 2.0 2-.2 SL Viol-at-ions -------- (crntinued) ........... 2*.-~~I 1DS3 n ,ifS 2.12 i sviol ated-*---estore--eomp-1+ancee Crystal River Unit 3 2.0-2 Amendment No. 201

SLs 2.0 600 610 620 Reactor Outlet Tempereture. 0F Figure 2.1.1 1 (page 1 of 1 ) Reactor Coolant System BNB Safety Crystal River Unit 3 2.0-3 Amendment No. 49

LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during theMODS-S-ei-other-specified conditions in the Applicability,-exept-ýas provided-in LCO-3-.-0.. C 3.0.--7 and LCQ--3.-8. LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met 7-,except as provided-i--nLO-3.0 5 and 3.0.6. If-t-he-CO--i-s-met or is no longer applicable prior to expiration of theperd-ompletion-_^ -m t n-is ot requirhed,-tnnless otherwise stated.. When- -a is not ,-exeept- -as pUrovI-ded- -n-the assoi-ate -ACTIONS -and-an associated C_,IO,- i-S-not-m et-- or prov-idedv--the--unit sh-a3--e-- p e--in-a- MOIE-or ^-'-er spe-ifi-ed-ondi-ti-on---ihi-c-L-the*-S- ",ion i"s not applicable. ,Ation sha-ll be in i tiated wi t,,,in-1--hou-rc-o p1l~ the--tmt- -as-pple "~-~---i-ni" 154-a:. -as-- a-b 1 e, t n,. b.--. MOOE-1 wi.thin 13 hours, and e_ NUDE 5"with-37-- hours. -Except-ions to this Specification are stated in the individual Spec-i fi cat-i-ons-. Where--**r-reetive measures are completed ,that "permi-t operati-on-i n accodancew*th .-t,,he LC' or A-.-IONS--,eomplet+on o-f-e-ati-ons-required by L-O 3.0. 3 s*not required. O-3.-0.3 is only applicabl~e-nMODE5--, 3, an-L-CO--3 0-.-4------ W---hen-anLCO is not met, entry into a MODE or ot spefi-ed ndition in the Applicability shall only be made*. a.- -- When the associ-atedAC IONS, to be entered permit cont ntted-eration ""he MODE-or other-spee-ffi-ed condition in the Ap,.,i,.bihity-for--n lmed period--of-ti-mei--or (continued) Crystal River Unit 3 3.0-1 Amendment No. 2-24

LCO Applicability 3.0 3.0 LCO APPLICABILITY .(c--.Ontinued--- -b - At-tar^ rn t-r

rnaL, 1"=cc r

pt 1rl C f esults. determination of the acceotabilitl' F -4 enterina the MODE or other snecified condj*jAn jn the Anolicability. nd establishment of:jý manaaement getions. if aggrooriate. excentions this Snecification are stated in the individual I i . I i I I i I ,e-, -- -When--an all1owance is stated in the indiVidual I IIvalue. oarameter. o, other Soerzifirzation I -l-ri~See1ttc~on-s1F-f~t-DreentchIanaes in f MO LS or A*L Mp" nap* Af --a 5Wutdown~ ~f the-un+/-t~ S-5 ,Equpment removed fr.m service or de*,ared-i-nroerable-to -omplMy-*it hAUiCTeM t-- be returned -to-se-rvi-e-. ider adm+n-strative control, solely to perform testing requireed-to demonst.rate -its OEABILIT ,-tV equipment-,-or-variables to be with+n limits. This is-a1,, except+on-to LCO-3.00.2 for the system returned o S.vI.ee under administrative control testing;. -Whe.-s.upported--sys-temLCO is not met solely due to suppo-rt-system -CO not being met, -the-Condito and Requ i red.- n associated with this supported system are r-equired to be entered. Only the support .sys.tem Spe-ife-it+on-ACTI-ONS-are-requ-red to be entered. Th-i-s-+ an-e-xf-ept~on-to-LC&+/-O-3 the supportd t system_. -- thi-s ever.t.-*additional evaluations and limi,,tations may be requ redi-* -aweordaie w th, Specification 5, 6.2.16-v"S-afety Function Dete rmi nation Program z -I-f-a---ossof--s-fety function is determined to exist by this program, the appropriate Conditions and Required Ac tions- -the Speci fication -in -whic-- hthe'loss ,,1,-ty-u. t-* onexi-t-are required to be entered-. (continued) Crystal River Unit 3 3.0-2 Amendment No. 2.15

LCO Applicability 3.0 3.0 LCO APPLICABILITY LcO 3.a.-6....... When--a--stpport--system's Required Action directs a supported -(-ontM-i.ue..... stem-o-tbe--delared i"operable or directs entry into Cod~~os-an-' Requi-:red Actions for a supported sys-tem, -the -a4-ieab-e-Conditions ad- -e-Actionsshall be ent.e.red app- +abTo-rdnew C th id O i -o3san0T re2-CO 3.0.7..............- PYSICS--TE-STS EXcetion LCOS (SCC-3.--8--and --a---ow -spe-eif:4 d Teech~nicaal Specifications (TS) requirements to--be--sus-pende to permit -perfo-rmance of.sp.a-.-tes-s-and operati-ons-.--- Unl.eSS othle r.i -sp-cified, all other TS requirements --rema*-n-unchanged---Comp-ian -wi FItP -PS TEjTS Exeptiov s i-s optional-. -W*-hrena -YSS TEST -xept --LCO L is desired to be met but is not met-,-the ACT-IONS-of te-PHNICS TESTS EA-fxCept-C-ot sLhall bn m et,. When-- a PHYSICS TEST--Except4i--t(-CO-i-s--inot --desired to -be -net,- entry--itto-- MODE or other specified condition in the Appi-1-cab-i-SMII t---a--onl- -eraein accordance wt te appl i-c-abl-e Spec-if-i-cati-ons. Whnon r oe-e-**-raed snubto-efone LE s A Iný."a--met-IUUU It lII 111V O- "III--e -o w-t-e s ci ie L&.. Iod 1th1 e r uis., snuAb111ber 6 sAV-Crystal River Unit 3 3.0-3 Amendment No. 224

SR Applicability 3.0 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY SR 3.0.1 SRs shall be met during the MOE) WI-or--other-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. Surveillances do not have to be performed-o n-e.-u-i -oi---va-riab -s-1utsaIde s-pec*f-ie+/-- -t-imi-ts-- 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. Fm'r Frqteti-es-speei flevd as "once, -th-v- extensondoes -not--ap-y- -...........I-f----a-----R eqt-re-d --Ae-t on -S-uer-u es--pe-- S-uive-i-l ante or-its--CompletionTime requires e-- odi-e-performan---on-a e--peP-r-- .-.- -_bas-s,--,-above-Frequency extension app-l-i-e-sto--e-ach performance after the initial performan-ee-. -Exceptions- -to-LIhI--Speci fi cati on are--stated --iit-the I-nd-i-vi-du-a-l-- Spee-fi-I-ea-ei-ons. 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 hours or up to the limit of the specified Frequency, whichever is areater. This delay period is permitted to allow performance of the Surveillance. A risk evaluation shall be performed for any Surveillance delayed areater than 24 hours 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. (continued) Crystal River Unit 3 3.0-4 Amendment No. 203

SR Applicability 3.0 3.0 SR APPLICABILITY SR 3.0.3 (continued) 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 MODE or other 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. When an LCO-is not--met-ue--to--Surveilances not having-betm-et-*-ý rv--to %JI %,21-I1* .*1.. L*%. I I*

  • 1%,

I, % I III I..11 r'uI I A .l* I III-shel -orly be made ,,, a i.gu, a,,g --. - --- , LC-Q,.. v.-t This orovision shall not orevent entrv-into MODES or other specnii,,c*djjgos in the,M21,,tab,,,ty h-- v Illy¥ 1 I UII -l -LII:L-- Jll I I Il lU I Crystal River Unit 3 3.0-5 Amendment No. 215

Fuel Storage Pool Water Level 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Fuel Storage Pool Water Level LCO 3.7.13 The fuel storage pool water level shall be Ž 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 A.1 NOTE water level not within LEO 3.0.3 is not limit. a--i-abl Suspend movement of Immediately irradiated fuel assemblies in fuel storage pool. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify the fuel storage pool water level is 7 days 156 ft Plant Datum. Crystal River Unit 3 3.7-2-71 Amendment No. -149

Spent Fuel Pool Boron Concentration 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Boron Concentration LCO. 3.7.14 The spent fuel pool boron concentration shall be Ž1925 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 NOTE concentration not LCO 3.0.3 is not applicable. within limit. A.1 Suspend movement of Immediately fuel assemblies in the spent fuel pool. AND A.2.1 Initiate action to Immediately restore spent fuel pool boron concentration to within limit. OR A.2.2 Verify by Immediately 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. Crystal River Unit 3 3.7--52-8 Amendment No. 193

Spent Fuel Pool Boron Concentration 3.7.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pool boron 7 days concentration is Ž 1925 ppm. Crystal River Unit 3 3.7--2-93 Amendment No. 149 1

Spent Fuel Assembly Storage 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Assembly Storage LCO 3.7.15 APPLICABILITY: 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. 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 A.1 NOTE [ LCO not met. LCO 3.0.3 is not app! i ab! -e-Initiate action to Immediately move the noncomplying fuel assembly to an acceptable configuration. Crystal River Unit 3 3.7-304 Amendment No. -19,3

Spent Fuel Assembly Storage 3.7.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify by administrative means the initial Prior to enrichment and burnup of the fuel assembly storing the is in accordance with Figure 3.7.15-1 or fuel assembly Figure 3.7.15-2. in Storage Pool A or Storage Pool B. Crystal River Unit 3 3.7-3-15 Amendment No. +93 1

Spent Fuel Assembly Storage 3.7.15 45 40 35 30 725 CL20 15 10 5 0 CCat goryo e-__ _ (between cur (ees) ~Cat.egory F (bene~lth curve) 2 2.5 3 3.5 4 4.5 5 Initial Enrichment, Weight Percent U235

1.

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

2.

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

3.

Category F: Fuel from this category must be stored in a one-out-of-two checkerboard configuration with fuel from Category B or empty water cells. Category F fuel stored in a checkerboard pattern with either Category B fuel or empty water cells must be separated from Category A fuel by a transition row of Category B fuel. Figure 3.7.15-1 Burnup versus Enrichment Curve for Spent Fuel Storage Pool A Crystal River Unit 3 3.7-=-H6 Amendment No. 2-2-7 1

Diesel Driven EFW Pump Fuel Oil, Lube Oil and Starting Air 3.7.19 45 40 35 30 -25 4o 15 10 5 0 Category B above curves "aeg0 CP 01Z 110, 00-ýý(below curves) 2 2.5 3 3.5 4 4.5 5 Initial Enrichment, Weight Percent U235

1.

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

2.

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

3.

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

4.

Fuel of any enrichment and burnup including fresh, unburned fuel may be stored in Pool B if surrounded by eight empty water cells. Category BE fuel assemblies must be separated by two adjacent empty cells in Pool B. Figure 3.7.15-2 Burnup versus Enrichment Curve for Spent Fuel Storage Pool Crystal River Unit 3 3.7--3-37 Amendment No. 272-7 1

Design Features 4.0 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 emergency-power; and favorable conditions of hydrology, geology, seismology, and meteorology. 4.2 Reactor Ee"fjJ A

3
L r-.~

61 a I ~J~I r~3~IIIIJI I~3 The reactor shall con.tain 1 fuel asse.bl-es. Each.. -,t ,n. *. f e .J,,* ,*o = *,,l~. shall consist of a matrix of Zircaley'-4 or 45-fuel reds with an, initial coemnosition of natural or slichthly ienihed rnu InT have b m l d with aonl.eable NRC staff...... L L I - n_ "--I ... ~ J-.-

4.

2.2 EONTROL-RODS The reactor core shall eecotain 60 satety am-d reaulatim EONCTROL nROD asemlies am 8J AX/IAL POWERI 516 ,APTIN ROD~I (APSRl as I-P I The material shall be silver imemum eacimium or imeeme! as fto-bv thy-ýNRE-(continued) Crystal River Unit 3 4.0-1 Amendment No. 210

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

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 oool:

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.

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

d.

A nominal 21.125 inch center to center distance between fuel assemblies placed in the storage racks. (continued) Crystal River Unit 3 4.0-2 Amendment No. 193

Design Features 4.0 4.0 DESIGN FEATURES 4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 138 feet 4 inches. 4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1474 fuel assemblies and six failed fuel containers. Crystal River Unit 3 4.0-3 Amendment No. 93

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

a.

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

b.

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

c.

The individuals who train the Certified Fuel Handlersi] carry out health physics[,j 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 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. (continued) Crystal River Unit 3 5.0-2 Amendment No. i4-4 Crystal River Unit 3 5.0-2 Amendment No. -244

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

2.

For Iodine 131, Iodine 133, tritium-and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to a dose rate of 1500 mrems/yr to any organ;

h.

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

i.

Limitations on the annual and quarterly doses to a member of the public from iodine 131, iodine 133, tritiumT 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 0DCM:

1.

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

a.

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

b.

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

2.

Shall become effective after review and acceptance by the on-site review function and the approval of the Plant Manager; and (continued) Crystal River Unit 3 5.0-9 Amendment No. 244

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.12 Ventdlation Filter Testi ng rrogram (v, I I I/wUts'Vedd A pr ra sill be established to implement te followi@

reuIred 4tsin f

Ei Cntrol Ro\\ mmTergeny V=entilation System.R[V.,SV)j WI J lI {.I I1 I t J, U I I JII l ai,.,* I rI I U IllJ I o;I,* I G

  • vI U

I ~t J a.~~~~~~~ Desrt for eac t n~ of N~tr the 1R[V thta 2-,l t Regulator Gudn.2 eso,17, and in aecordancel and 4789 cfm.poed199)

a.

Bemonsrarte for eaeh train of the fRlEV that test-'~-~ of thI-."T effleien~ ariult airL~ EUyPAi-fii.4Ji ysii ~~~~~96_.tested inacranewt Rgltr Guie1.52,ne Reisio 2 and14 ANS 145H 1975 at the system flowrate of between 37,800w and 47, 850 cfm.

b. Demonstrate for each train of the CR[VS that a-lbr tosy ofthe larboraor tdestnb ier i ofow ASw 3038 aRe ANproved 95) at a temperature of at a

o d relativ 37*800

e. Demonstrate for each train of thCR[VS ha thet arsr dabrop castbon adobs -' <h earb4" wateorbaug when otested in accordance iýth Rgultory Guide 1.52, Revision 2, 1978, and ANS N50 975at hesysem W oratue of b0Etwend r7el00ian Th povsinsof SR% 3.0. an oSd30.dae pntappicabl tof thes 't
t.

frequeBncIeshat t.h I &W ANSI-* N510' 19r75 at*r-the ....... flwrt of.. be'^ e 3780 a 47,850I /"fIIJJ. IJ. -J* J G tl UV

L*

V t Vt l /, The I pU "-r ,U f SR 3n.0. n R303aeapial oteVT test E;I*JI*L fr qu n ie E= I*I t;-I IIG l I t i; .l(*. 1 IIJI* E; tE;* 5.6.2.13 Not Used (continued) Crystal River Unit 3 5.0-11 Amendment No. 244

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.14 Diesel Fuel Oil Testing Program A diesel fuel oil testing program to implement required testitig-of both new fuel oil and stored fuel oil shall be established. Th program shall inelude sampling and testing rqieetand acceptance criteria, in accordance with applieable AS-M Standards. THe purpose of the program is to establish -the foalloWing

a. Aeeeptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has the following properties within limits of ASTM D 975 for Grade No. 2 B fuel oil.
1. Kinematic Viscosity,
2. Water and Sediment,
3. Flash Point,
4. Specific Gravity API;
b.

Other properties of ASTM D 975 for Grade No. 2 B fuel oil are within limits within 92 days follon sapling and addition of new fuel to storage tanks.

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

5.6.2.15 Not Used (continued) Crystal River Unit 3 5.0-12 Amendment No. -244

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.16 Safety Function Determination Program (SFDP) $,tN This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCQ 3.0.6, an evaluation shall be made to determine if loss of safety functio exists. Additionally, other appropriate limitations and remedia or compensatory actions may be identified to be taken as a result of the support system inoperability and corresponding exception t C~

  • I*IC II JII

.. 3U*JIJJI.

  • .LC I373*

IICI* L JE I I LI E:Il I aI I~ i (C9U I. I VJ f L i Urn ll O I

.II I lJ t.l

,ilII l .=y =*.* II IllJJ~lIIJII .*f (i~ ~i n 2.JlI* AI*'.,/

  • I, l

l.j /,,*I=2 %-11 t program implements the r e irements of LC) 3.0.6. The SFDP shall contain the following;

a.

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

b. P"rovisios for ensuring the plant is maintained in a safe Lendition if a loss of function condition exists;-
c. Proviion to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a resul-t of multiple support system inoperabilities; and
d.

Other appropriate limitations and remedial or compensatory A n .-f c*. ni"

f.

r-t 4-r n 4 -t n 2cI--1r r n4nr f I llll I' ~I AI. I LJI I III L I* I i fll I I I V;i IJ CI ,~~ I lL* 3MI I I*' IIt* II.U*, LUI,,PI I C*..1lL II, actions. i-f single failure, a safety function assumed in the accidt.. I analysis cannot be performed. For the purpose of this program, a losso safety function may whe a support system is inoperable andt. a. A required system redundant to the system(s) supported byY the inoperable support system is also inoperable); or

b.

A required system redundant to the system(s) in turn supported by the inoperable supported system is also ,11operable, o C. A required system redundant to the support system(s) for the supported systems (a) and (b) above is also inoperable. I (continued) Crystal River Unit 3 5.0-13 Amendment No. 2144

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.16 SFDP (continued)ý The -FDr identifies where a less of safety f "-t-om exists. f less of safety function is determined to exist by this program, the appropriate Conditians and Required Actions of the LEO 4m whih the loss of safety funtion exists are reqfu red to be entered-. 5.6.2.17 Technical Specifications (TS) Bases Control Program Changes to the Bases of the TS shall be made under appropriate administrative controls and reviewed according to the review process specified in the Quality Assurance Plan. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

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

A change to the updated FSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR. Proposed changes that meet the criteria of Specification 5.6.2.17.a or Specification 5.6.2.17.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71. 5.6.2.18 Not Used 5.6.2.19 Not Used 5.6.2.20 Not Used Crystal River Unit 3 5.0-14 Amendment No. -M4

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.21 Control Complex Habitability Envelope Integrity Program Nb7tiý`-Ie'd A Control Complex Habitability Envelope integrity Program shall be established and implemented to ensure that CCIIE habitability s m--intained such that, with an OrERABLE Control Room Emergency VenilaionSystem (CREVS), CCIIE occupants can control the reactor safely under normal conditions and maintain it in a saf condition followin a rdiological event, hazardous chemical release, or a challenge from smoke. The program shall ensure that adequate radiation protection, is provided to permit acces-s and occupancy of the CCIIE under design basis accident (DRA) conditions without personnel receiving radiation expsue in excess of 5 rem total effective dos eqiaent (TD)for the duration of the accident. The program shall incalude the following elements. I+--The definition of the CCIIE and the CCIIE boundary. 2-r-Requirements for maintaining the CCIIE boundary in its design ondition including configuration control and preventive. mai ntenance. 9-r-Requirements for (i) determining the unfiltered air in leakage past the CCIIE boundary into the CCIIE in accordance with th testing methods and at the Frequencies specified in Sections CE1 and C.2 of Regulatory Guide 1.197, "Demonstrating Conto Room Envelope integrity at Nuclear rower Reactors," Revision 0, May 2003, and (ii) assessing CCIIE habitability at the Frequencies specified in Sections CE1 and C.2 of Regulatorry Guide 1.197, Revision 0. 4-r-The Control Complex Habitability Envelope integrity rrogram will be used to verify the integrity of the Control Complex boundary. Conditions that are identified to be adverse shl be trended and used as part of the 24 month assessment of the CCIIE boundary.

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

These limits shall be stated in a manner to allow direct comparison to the unfiltered air in leakage measured by the teting described in paragraph 3. The unfiltered air in leakage limit for radiological challenges is the in leakage flow rate assumed in the licensing basis analyses of DDA consequences. Unfiltered air in leakage limits for hazardous chemicals andý smoke must ensure that exposur of CCIIE occupants to these hazards will be withi h supin in the licensing basis..

6. The poions of SR 3.0.2 are applicable to the Frequencie-s for assessing CCIIE habitability, determining CCIIE unfiltere in, leakage as required by paragraph 3.

Crystal River Unit 3 5.0-15 Amendment No. 2-44

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

a.

Not Used

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

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

c.

Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted prior to May 1 of each year, and in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program, and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV B.1. (continued) Crystal River Unit 3 5.0 1ý Amendment No. -244

Reporting Requirements 5.7 5.7 Reporting Requirements 5.7.1.2 Not Used 5.7.2 Special Re.orts NrottfU§e'd Special Reports shall be submitted in a*cordance with 10 CFR 50.4 within the time period specified for each report. The following Special Reports shall be submitted;

a.

When a Special Report is required by Condition D or F of LCO 3.3.17, "Post Accident Monitoring (PAM) instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanne alternate method of= monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status. Crystal River Unit 3 5.0-177k-3, Amendment No. 24

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), alternative methods are used to control access to high radiation areas. Each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation (measured at 30 cm) is > 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

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

b.

A radiation monitoring device that continuously integrates the radiation dose in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance. 5.8.2 In addition to the requirements of Specification 5.8.1, areas with radiation levels Ž 1000 mrem/hr at 30 cm shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor or health physics supervision. Doors shall remain locked except during periods of access by personnel. Direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area. (continued) Crystal River Unit 3 5.0o-18-1ý Amendment No. 244

High Radiation Area 5.8 5.8 High Radiation Area (continued) 5.8.3 For individual high radiation areas with radiation levels of > 1000 mrem/hr at 30 cm, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that are not be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device. Crystal River Unit 3 5.0-19iý5 Amendment No. 244

DUKE ENERGY FLORIDA, INC. CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #316, REVISION 0 ATTACHMENT E PROPOSED TECHNICAL SPECIFICATION PAGE CHANGES, REVISION BAR FORMAT

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 i Amendment No.

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

Definitions 1.1 1.0 USE AND APPLICATION 1.1 Definitions NOTE--------------------------------- The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications and Bases. 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. Crystal River Unit 3 1.1-1 Amendment No.

Completion Times 1.3 1.0 USE AND APPLICATION 1.3 Completion Times PURPOSE The purpose of this section is to establish the Completion Time convention and to provide guidance for its use. BACKGROUND 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 COMPLETION TIME When "Immediately" is used as a Completion Time, the Required Action should be pursued without delay and in a controlled manner. Crystal River Unit 3 1.3-1 Amendment No.

Frequency 1.4 1.0 USE AND APPLICATION 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements. DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated LCO. An understanding of the correct application of the specified Frequency is necessary for compliance with the SR. The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, "Surveillance Requirement (SR) Applicability." The "Specified Frequency" consists of the requirements of the Frequency column of each SR. (conti nued) Crystal River Unit 3 1.4-1 Amendment No.

Frequency 1.4 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 hours Example 1.4-1 contains the type of SR encountered in the PDTS. The Frequency specifies an interval (12 hours) 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 hours, an extension of the time interval to 1.25 times the stated Frequency is allowed by SR 3.0.2 for flexibility. Crystal River Unit 3 1.4-2 Amendment No.

LCO Applicability 3.0 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY LCO 3.0.1 LCOs shall be met during specified conditions in the Applicability. LCO 3.0.2 Upon discovery of a failure to meet an LCO, the Required Actions of the associated Conditions shall be met. Crystal River Unit 3 3.0-1 Amendment No.

SR Applicability 3.0 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 hours 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 hours and the risk impact shall be managed. If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. SR 3.0.4 Entry into a specified condition in the Applicability of an LCO shall only be made when the LCO's Surveillances have been met within their specified Frequency, except as provided by SR 3.0.3. Crystal River Unit 3 3.0-2 Amendment No.

Fuel Storage Pool Water Level 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Fuel Storage Pool Water Level LCO 3.7.13 The fuel storage pool water level shall be Ž 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 A.1 Suspend movement of Immediately water level not within irradiated fuel

limit, assemblies in fuel storage pool.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify the fuel storage pool water level is 7 days 156 ft Plant Datum. Crystal River Unit 3 3.7-1 Amendment No.

Spent Fuel Pool Boron Concentration 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Boron Concentration LCO. 3.7.14 The spent fuel pool boron concentration shall be Ž1925 ppm. APPLICABILITY: When fuel assemblies are stored in the spent fuel pool and a spent fuel pool verification has not been performed since the last movement of fuel assemblies in the spent fuel pool. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool boron A.1 Suspend movement of Immediately concentration not fuel assemblies in within limit, the spent fuel pool. AND A.2.1 Initiate action to Immediately restore spent fuel pool boron concentration to within limit. OR A.2.2 Verify by Immediately 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. Crystal River Unit 3 3.7-2 Amendment No.

Spent Fuel Pool Boron Concentration 3.7.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.14.1 Verify the spent fuel pool boron 7 days concentration is Ž 1925 ppm. Crystal River Unit 3 3.7-3 Amendment No.

Spent Fuel Assembly Storage 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Assembly Storage LCO 3.7.15 APPLICABILITY: 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. 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 A.1 Initiate action to Immediately LCO not met. move the noncomplying fuel assembly to an acceptable configuration. Crystal River Unit 3 3.7-4 Amendment No.

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

Spent Fuel Assembly Storage 3.7.15 45 40 35 30 ý25 I.. 15 10 5 0 Categury B above curve) Category A, between cur ies) Category F (beneith curve) 2 2.5 3 3.5 4 4.5 5 Initial Enrichment, Weight Percent U235

1.

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

2.

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

3.

Category F: Fuel from this category must be stored in a one-out-of-two checkerboard configuration with fuel from Category B or empty water cells. Category F fuel stored in a checkerboard pattern with either Category B fuel or empty water cells must be separated from Category A fuel by a transition row of Category B fuel. Figure 3.7.15-1 Burnup versus Enrichment Curve for Spent Fuel Storage Pool A Crystal River Unit 3 3.7-6 Amendment No.

Diesel Driven EFW Pump Fuel Oil, Lube Oil and Starting Air 3.7.19 45 40 35 30 -25 .0~Q 15 Category B (3bove curves (beteen(brelowurv -000, ýCae oyM 00ý.00 00ý(eo uv; 10 5 0 2 2.5 3 3.5 4 4.5 5 Initial Enrichment, Weight Percent U235

1.

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

2.

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

3.

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

4.

Fuel of any enrichment and burnup including fresh, unburned fuel may be stored in Pool B if surrounded by eight empty water cells. Category BE fuel assemblies must be separated by two adjacent empty cells in Pool B. Figure 3.7.15-2 Burnup versus Enrichment Curve for Spent Fuel Storage Pool Crystal River Unit 3 3.7-7 Amendment No.

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

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

a.

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

b.

kef ! 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.

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

d.

A nominal 21.125 inch center to center distance between fuel assemblies placed in the storage racks. (continued) Crystal River Unit 3 4.0-2 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES 4.3.2 Drainage The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 138 feet 4 inches. 4.3.3 Capacity The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1474 fuel assemblies and six failed fuel containers. Crystal River Unit 3 4.0-3 Amendment No.

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

a.

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

b.

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

c.

The individuals who train the Certified Fuel Handlers, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their ability to perform their assigned functions. 5.2.2 Unit Staff The unit staff organization shall include the following:

a.

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

b.

Shift crew composition may be less than the minimum requirement of 5.2.2.a for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. (continued) Crystal River Unit 3 5.0-2 Amendment No.

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

2.

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

h.

Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10 CFR 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 Plant Manager; and (continued) Crystal River Unit 3 5.0-9 Amendment No.

Procedures, Programs and Manuals 5.6 5.6 Procedures, Programs and Manuals 5.6.2.12 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 Crystal River Unit 3 5.0-11 Amendment No.

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

a.

Not Used

b.

Annual Radiological Environmental Operating Report The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM). The Annual Radiological Environmental Operating Report shall include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

c.

Radioactive Effluent Release Report The Radioactive Effluent Release Report covering the operation of the unit shall be submitted prior to May 1 of each year, and in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the ODCM and Process Control Program, and in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV B.1. (continued) Crystal River Unit 3 5.0-12 Amendment No.

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

High Radiation Area 5.8 5.0 ADMINISTRATIVE CONTROLS 5.8 High Radiation Area 5.8.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), alternative methods are used to control access to high radiation areas. Each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation (measured at 30 cm) is > 100 mrem/hr but < 1000 mrem/hr, shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

a.

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

b.

A radiation monitoring device that continuously integrates the radiation dose in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c.

An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance. 5.8.2 In addition to the requirements of Specification 5.8.1, areas with radiation levels > 1000 mrem/hr at 30 cm shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor or health physics supervision. Doors shall remain locked except during periods of access by personnel. Direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area. (continued) Crystal River Unit 3 5.0-14 Amendment No.

High Radiation Area 5.8 5.8 High Radiation Area (continued) 5.8.3 For individual high radiation areas with radiation levels of > 1000 mrem/hr at 30 cm, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for purposes of locking, or that are not be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device. Crystal River Unit 3 5.0-15 Amendment No.

DUKE ENERGY FLORIDA, INC. CRYSTAL RIVER UNIT 3 DOCKET NUMBER 50-302 / LICENSE NUMBER DPR-72 LICENSE AMENDMENT REQUEST #316, REVISION 0 ATTACHMENT F PDTS BASES PAGES FOR INFORMATION, STRIKEOUT AND SHADOWED TEXT FORMAT

LCO Applicability B 3.0 B 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY BASES LCO 3.0.1 through LEG 3.0.8 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 MODE or other 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-

a.

Ccompletion of the Required Actions within the specified Completion Times constitutes compliance with a Specification-;--and.

b. Completion of the Required Actions is mot required

..he. an LEE is met within the specified Completion Times, unless otherwise specified. There are two basie types of Requqred Actioms. The first type-of The Required Action specifies a time limit in which the LCO must be met. This time limit is the Completion Time to restore a, inoperable system or componant to OP[ABL[ status or to restore variables to within specified limits. if this type of Required Action is not completed within the specified Completion Timne, a shutdowm may be required to place the unit in a MODE or condition in whichth Spe-ifi.ati.n is not appliab**. (Whether stated as a Required Action or not, correction of the entered Condition is an action that may always be considered upon entering ACTIONS. The second type of Required Action specifies th-e (continued) Crystal River Unit 3 B 3.0-1 Revision No. 66 Crystal River Unit 3 B 3.0-1 Revision No. 66

LCO Applicability B 3.0 BASES LCO 3.0. 2 remedial measures that permit continued operation of the (*cntinued) unit that is not further restricted by the Completion Time,. In this case, eompliance with the Required Aetions provides an acceptable level of safety for continued operation. ..mpleting the Required Actions i not required when an LC is met or is no longer applicable within the associated Completion Time, unless otherwise stated in the individual Specii fications. e The nature of some Required Actio*s of some Conditio* s necessitates that, once the Condition is entered, the Required Actions must be completed even though the associated Conditions no longer exist. The individual Specification's ACTIONS specify the Required Actions whr this is the case. An example of this is in LCO 3.4.3, "RC. rressure and Temperature (r/T. ) Limits." The Completion Times of the Required Actions are also ipp i cab..=l el we aII syl or component is removed from ser Vic intenIionally. Reasons for intentionally relying on the ACTIONS include, but are not limited to, performanceof Surveillances, preventive maintenance, corrective maintenance, or Investigation of operational problems. Entering ACTIONS for these reasons must be done in. a manner that does not compromise safety. Intentional entry into ACTIONS should not be made for operational convenience. Alternatives that would not result in redundant equipment being mnoperable should be used instead. individual Specifications may specify a time limit for performing an SR when equipment is removed fro sevc or bypassed for testing. in this case, the Completion Times of the Required Actions are applicable when this time limit expires, if the equipment rean emoved from service or bypassd When a change in MODE or other specified condition is required to comply with Required Actions, the unit could enter a MODE or other specified condition in which anothe Specificeation becomes applicable. in this case, the Completion Times of the associated Required Actions would apply from the point in time that the new Specification becomes applicable and the ACTIONS Condition(s) is entered (conti nued) Crystal River Unit 3 B 3.0-2 Amend ment-RevisiomnNo. 149

LCO Applicability B 3.0 BASES LCO 3.0.3 LEO 3.0.3 establishes the actions that must be implemented when an LEE) is met met and:

a. An asseciated Required Action and Completiom Time is met met and ne other Co-dition applies, or h.

The odition of the unit is not specifically addressed by the associated ACTIONS. This means that ne cmbination of Conditions stated in the ACTIONS can be made that exactly correspends to the actual onditio of the unit. S.metimes, possible combimations of Conditions are such that entering LEO 3.0.3 is warranted, in su.h eases, the ACTIONS specifically state a Co.dition corresponding to such cmbinatios ad also that LEO 3.0.3 be entered This Specificatio-delineates the time limits for placing the unit in a safe MOD[ or other specified condition when operation cannot be maintained within the limits for safe operation as defined by the LEE) and its ACTIONS. it is no~t intended to be used as an operational covenee. e that permits routine voluntary removal of redundamt systems or ,wmponents from service in lieu of other alternatives that would not result in redundant systems or components being imepe~rable-.- Upon entering LEO 3.0.3, 1 hour is allowed to prepare for an orderly shutdown. These oreoaratio ns emit of a I".... di~. ol,,s. a,,d allow time to Germi te operator to ecoordimat total time of-7 h....- frmtr into. LEO 3..... if at I =- time al-% low,,,.. for an orderly olant s~hutdown, commencingaa Uleadi d,,e-- er(Lmea mavy be deJKlaLyed. umtl t,.. ha -tll tim. II* (continued) Crystal River Unit 3 B 3.0-3 Revision No. 50

LCO Applicability B 3.0 BASES LCO 3.0.3 The time limits specified to reach lower MODES of operatio (continued) permit the shutdown to proceed in a controlled and orderl-y manner that is well within the specified maximum ecooldown rate and within the capabilities of the unit, assuming ta on.ly the minimum required equipment is OPERABLE. This reduces thermal stresses on compoenets of the Reactor Coolant System and the potential fmor a plant upset that eould challenge safety systems under conditions to which this Specification applies. The use and interpretation of specificd times to complete the actions of LCO) 3.0.3 are Lansistent with the discussion of Section 1.3, Completion Ti-mets. A unit shutdown required in accordance with LCO) 3.0.3 may be terminamteed and LCO 3.0.3 exited if any of the following oeeurs. ai. The LCO is now met. b. A Condition exists for which the Required Actions hav now been performed. t -ACTIONS exist that do not have expired Completion Times. These Completion Times are applicable from the point in time that the Condition is initially entered and not from the time LCO 3.0.3 is exited. The time limits of Specification 3.0.3 allow 37 hours for the unit to be in MODE 5 when a shutdown is required durig MODE 1 operation. if the unit is in a lower MODEo operation when a shutdown is required, the time limit for reaching the next lower MODE applies. if a lower MODE is reached in less time than allowed, however, the total allowable time to reach MODE 5, or other applicable MODE, isr -me reduced. For example, if MODE 3 is reached in 2 hours-, -then the time allowed for reaching MODE 4 is the next 1:1 hours, because the total time for reaching MODE 4 isno reduced from the allowable limit of 1:3 hours. Therefore, if remedial measures are completed that would permit a return to MODE 1, a penalty is not incurred by having to reach a-lower MODE of operation in less than the total time allowedi in MODES 1, 2, 3, and 4, LCO) 3.0.3 provides actions for Conditions not covered in other Specifications. The requiremets of LCO) 3.0.3 do not apply in MODES 5 and 6 beauses the unit is already in the most restrictive Condition required by LCO 3.0.3. The rqients of= LCO 3.0.3 do not apply in other specified nerditions of the Applicability (unless in MODE 1, 2, 3, or 4) because the ACTIONS of individual Specifications su:fficiently define the remedial measures to be taken r., (contianued) Crystal River Unit 3 B 3.0-4 Revision No. 50

LCO Applicability B 3.0 BASES LCO 3.0.3 (conti ued) Exceptions to [CO 3.0.3 are provided in instances hr requiring a unit shutdown, in accordance with LO303 would not provide appropriate remedial measures for the associated condition of the unit. An example of this is i Specification 3.7.13, "Fuel Storage Pool Water Level."' Specification 3.7.13 has an Applicability of "Durinrg movement of irradiated fuel assemblies in fuel storage pool.' Therefore, this Specification can be applicable + any or all MODES. If the LCO and the Required Actions of Specifiation 3.7.13 are not met while in MODE 1, 2, 3, or 4, there is no safety benefit to be gained by placing the unit in a shutdown condition. The Required Action of Specification 3.7.13 of "Suspend movement of irradiated fuel assemblies in fuel storage pool" is the appropriate Required Action to complete in lieu of the actions of [CL.03 These exceptions are addressed in the individual Specifit I o ,*e f eltl LCO 3.0.4 [LO 3.0.4 establishes limItations P, 4-- n-- [CO 3.O.4.aalw RA^M -R-- -r 1% e r (cotined (conti nued) Crystal River Unit 3 B 3.0-5 Revision No. 5-5

LCO Applicability B 3.0 BASES -r I I (eemt1mued) f2la t

1.

182. "A-Aetivitea4tt7 Comsiderat+oi eemnletime r the LCO wouý GTIONS Contp-:h - L- - .. -1 A I - L- - w

  • ..w "1

tha wuldreu and-eom----ens unvalal. N-...C 93 0L.. mrovir ive ee sd rto of -fl a er (conti nued) Crystal River Unit 3 B 3.0-6 Revision No. 5

LCO Applicability B 3.0 BASES LC-*

3..4 The Te.hni.al Scif"-atioms all Econtinued eauioment unavailable in MODE 1If ZF fce I

n A L -I -i dee ~md

  • g

==5tft The orovisions of this Soeci ieation should not b status before-enter.. -seitd EJ orr-o- The nr iions of LECO 3.0.4 shall mot prevent-ehanges i t ieo (continued) (conti nued) Crystal River Unit 3 B 3.0-7 Revision No. 55

LCO Applicability B 3.0 BASES LCO

34.

C 3.

0.

salse tealwneo retrn e--uipment 3 Q 4 ton sente undtr admODEstraotiver cntrolsd wendit hs ee mued) Amp! jea fro sej c tocoy with thAC)mtmt CTO 3..1 Tn-He sole0. purponse unfi thie Cpedifation is teolprvede anti teLCeptiot metL o.0.. t h allw ,th npor withi tes Aodeabtrat the Techniea[ SWLT cfteation.

b. Theilme do mot have tfother veripment or h

sseae Theeatl tv0-V-- ariables are within limts ACmtIis) liit-d to the time1 aolThelyrefoessr utolzn perfor the. allwe a~s vlThiso Specificaton S does4 nF=ot provide timeto herfor anyoth preven tived ornoretv Anvexample ofweer deontatn must be~I~ oft the eqsuipe bein TY rtrned to serielais eoenn asonitainment iOPEALation valvitatl has benclsed tod coplswthR ie Actmionand with bhe rfeoened toLErfrGhes An example of deamonistratingte eomr[RelI ofe ot herb proeof theiripe condifetion i to perirh ogice to function and LE).di.2t the approprwersonedrn the performance of stodmsrae an ThR OnEAILT anohe chneothesm tri sysem C-That variables are with( limitn is ri or to thelaimgte ab solutlyated sr au mnto pe*rform th allwe r ~ .Ti pefl mde o reoed t l =ve ha, beei, y1se e ph wio th Reqire pA, exapl of dhSemonstrationg the OPRBLT of¢"*-* ofxcpio 3.iet 2the appopitersp,, urn the performance Off,* tod,,,,t t -a SR-em '......... e '-e in' the. sam sys-rturem. (*. II IE; U[ I*r'* J L.L I UI ilE; I* I* II*III II(contIn u ed Crystal River Unit 3 B 3.0-8 Revision No. 5-5

LCO Applicability B 3.0 BASES LCO 3.0.6 LCO 3.0.6 establishes an exception to LCO 3.0.2 for suppor-t systems that have a Specification specified in the Technical Specifications (TS). This exception is necessary because LCO 3.0.2 would require that the Conditions and Required Actions of the associated inoperable supported syste~m. Specification be entered solely due to the inoperability-of the support system. This exeption is justrified because th actions that are required to ensure the unit is ma.intained in a safe condition ar 'pcfied in the support system Spec"fication's Required Ations. These Required Actions may incluide eneig the supported system's Conditions and Required Actions or may spec~ify other Required Actions. W.hen a support system. isinperable and there is an LCO) specified for it in the TS, the supported system.s) are -requi ed to be declared inoperable if determined to be ... operable as a result of the support system inoperabiity. However, it is not necessary to enter into the supported systems' Conditions and Required Actions unless directed to. do so by the support system's Required Actions. The eunfusion and inconsistency of interpretation of requirements related to the entry into multipl~e Specification's Conditions and Required Actions are eliminated by providing all the actions that are necessar-y to ensure the unit is maintained in a safe condition in th-e support system's Required Actions. 'owever, there are instances where a support system's Required Action may either direct a supported system to be declared inoperable or direct entry into Conditions and Required Actions for the supported system. This may occur immediately or after some specified delay to perform som other Required Action. Regardless of whether i im"mediate or after some delay, when a support system's Required Action directs a supported system to be declared inoperable or directs entry in Conditions and Required Actions for a supported system, the applicable Condition and Required Actions shall be entered in accordance with and appropriate actions are taken. Upon failure to meet two or more LCOs at the same time, an evaluation shall be md (continued) Crystal River Unit 3 B 3.0-9 Revision No. 5-5

LCO Applicability B 3.0 BASES LCO 3.0.6 to determine if loss of safety function exists. (cormtinued) Additionally, other limitations, remedial actions, or-wempensatory actions may be identified as a result of th-e support system inoperability and corresponding exception to entering supported system Conditions and Required Actions The SrorP implements the requirements of LCO 3.0.6. Cross train checks to verify a loss of safety function fo~r those support systems that support multiple and redundant safety systems are required. The cross train check verifies that the supported systems of the remaminig OPE[ABLE support-systems are OP[RABL[, therebyesrn safety function is retained. if this evaluation dtrmines that a lss o -safety fuction exists, the appropriate Conditions and Required Actions of the Specification in which the loss of safety functiom exists are required to be entered.. W.hen a support system becomes inoperable, its associated LCQ ACT ION ah-re entered. Supported system LCQ ACTIONS are not required to be entered when the supported system becomes inoperable solely due to the support system bein inoperable. While the support systemisnoealth Completion Time for the suppore yse defines the operating window. Should another system become inoperable that supports the same supported system, then its LCQ AaTIQNS are also entered, however, the most recent inoperable suppor-t system LCO ACTIONS may not receive the full benefit of it-s Completion Time. This is because the most restrictive Completion Time is arssociated with the supported system, even though its LCO ACTIONS were not formally entered. Therefore, operation must be limited in accordance with the limiting Completion Time, regardless of entering the ACTIONS of-a-LEO.r The following examples are provided for clarification. (conti nued) Crystal River Unit 3 B 3.0-10 Revision No. 5S

LCO Applicability B 3.0 BASES I:EO 3 Eeomtimued) SUPPORT 5UPPORTED Above is a graphieal representation of the relationships for support and supported SSCs and related LCE~s for a sihgle@ train. SSC A-41 ,,d-A1.2 support SSC , '--hich in turn supports SSC A. SSc A2.1 and-A2.2 support E A2, ""wh'eh in turn supports A.For the purpose of the followig examples eaeh support SSC is required to be OPERABLE in order to declare its associated supported SSE OPERABL[. (conti nued) Crystal River Unit 3 B 3.0-11 Revision No. 5-5

LCO Applicability B 3.0 BASES LCO 3.0.6 -Eeoitiiued Px -f WheM-A is deelared inoperable, then the ACTIONS for tat SSE are entered *o,'. ,he AICTIONS for A -re not entered even though, that SSE is determined inoperable (no eas ,ading). I, the event that Ak beomes inoperable-'*** ý prior to exiting the Aetion Statement for does met get the full benefit of its own Completiom Time -@4)). Furthermore, A is still inoperable from the time I ha-t&-was initially delared inoperable @T*,--A-A-mtm-t be restored to OP[RABLE prior to exceeding the Completion Tim-s a (@T)--% Tim ,s e ia e with r%, ,.I L*.*;%, /1,1 (continued) Crystal River Unit 3 B 3.0-12 Revision No. 5S

LCO Applicability B 3.0 BASES LCO 3.0.6 LAI~I (eentinued) When-A is declared inoperable then the ACTIONS for that SSE are etered (@TW. Thie ACTIONS for A are not entered even though that SSE is determined inoperable (n .as.adi-g). I. the event that -Ai. beeemes inoperable "T-prior to exiting the ACTIONS ..t get the full be.efit of its . Co mpletion Time T-* Furthermort-, A is still inoperable from the time that-A was 4nitially delared inoperable,.T,.e). The ACTIONS for ae e..te ,d.,,, eVen thoug being inopable results in the SSC for A inoperable, because of no caseading. - must be restored to OPERADL[ prior to exceeding the Completion Time associated with A a)--. (continued) Crystal River Unit 3 B 3.0-13 Revision No. 5-5

LCO Applicability B 3.0 BASES LEO 3.0.6 (conti nued) -LE-Xatmple 3 Whem-is declared inoperable then the ATIONS for that SCS are ,tered (@T3). The ACTIONS for A are not entered even though that SSE is determnined inoperatble (no-cascading). i' the event that A,-* be.omets-inperabl T)- ri texiting the ATIONS for Ai (@T2 not et te full benefit of its owm Completion Time (@.4-- Furthermore, A is still inoperable from the time that A-wa; initially deelared inoperable (E@.,)--The AETIONS-fo~r are not entered even though that SSE is determined inoperable Eno eascadimg). A2,z must be restored to PERABLE prior to exceeding the Cmpletion Time associated (conti nued) Crystal River Unit 3 B 3.0-14 Revision No. 5

LCO Applicability B 3.0 BASES LCO 3.0.6 Econti Iued) -Er-Xampl e 4 Whe,- is delared,,operable them the ACTIONS for that SSE are etered The ACTIONS for A are not etered even though thatL S-SC is determined inoperable (no -casading). in the evet that becomes inoperable (,) prior to exiting the ACTIONS for Ai -0)--2,

the, A-get the full benefit of its own Completio, Time on.

Furthermore, A-still inoperable from the time that A was initially detlar ioperable ,E,@T- ,A *must

  • be restored to OPERABL[ prior to exceeding its Completion Time as soci4ated-(ýff3--.

(continued) Crystal River Unit 3 B 3.0-15 Revision No. 5-5

LCO Applicability B 3.0 BASES LCO 3.0.7 There are certain special tests and operations required to be erored at var.Ious timmes over the life of the141 unit. These special tests and Operations aeecsryto ,demonstrat~e select unmitt pe-rformance characeter istics,. PHYICSTETS xcptins C~ (Secfication 3.1.8 and 3.1.9) al Io. I se I iedl ITS Iequirements to be susp endle t o. peI IIt perforaeI so f t-fh w ese special a! tests and oeain which ohriecould not be prore if rqire to compl wihth euieets o)f ths6T.Ulesohews specifiedl,-I all oI Sl reqIuiments remain unchanged. This* VVI I 1 1' IrI l I I 7r!! r!nr I.~l jC Lio 4 LL49 &3seeiaLd with or reqird o e canedtopefr th pecial test or Compliance with riiYSICS TESTS Exetio LCO is optoal. A specal peration maly be pe~rformed either under the ,I LII Lthe IECI Iei I,E TLIE ST Ix, Ui II It +/-U II Io r3 desi red to perform the seiloraonunder the proison ofl the-IPHYSIC TEST Exception LCO, the iit!ýT it of th P I IYSE &uuL ý I L1 uLý III LEO sh. l be reqire tobe met, unless specified in the r'lYSICS TESTS Exception LCO). LCO 3.0E. 8 LCO 3.0.8 establishes conditions under which sysems UIIEI C JEII I 54 ,4~ Ly If I uk L Iu WII icI.l &5nL I a*l I*

  • Lc 31 IPULJJI 3*

l* p 1 VVIC~l I**1-UhLL i Ui ). mI ni %..lSL&L*Efi LU/1 I IbII LII 4 SdppinE* L~ll 3JLIin -isr notcnsidered to be inope*rablelsolelyedue1to one o bmeaued a imte lnthoftieisllowed for ar ti-"i'fl -- -*Il--ti-2 t -1r u* o .--r located outsid~ed of thrb o! te ITSudrlcne otrol LCeor asIls.ee Iia ed Urt I* Pl. FJ JII il. I; I 0-. I M .T isEII;IIL-'*P, .~ 1 3.0.8i appIe tod snbersthaoft hae sesmi function' only. It d t to-snbbersmhatalsobha desig funtion to itigte seam/ater hammer or othr raniet-lad. ThVt-mesnubbrequrent dorno

such, tside ae-app ae r c

l by the lic ll 1ensee Gl

  • s ec

,,I*__ E T When at sn se"ubber isa tbernrdica able o of v nofucinl fo tetn srmitenm/aner oamr is dicvee to not be functonal ite mutedeterdomine whether any sytes fr equirelb the afec ed m Crystale Rive Unitq 3F B 3.-1 Rn tmutb eevisioneNod wh*ethe am, reurete1fee-- is nt cnsiere tobe noprabe slel du (ontinued Crysta Rive UnitI.UU; 3 Bl..l 3. 0-16. Revision/ No. 66IEII IIII11* I;I

LCO Applicability B 3.0 BASES LCO 3.0.8 snubberis) for system OIERABLILITY, aI d whether the (continued) plntisi a MODE ospcf ed cndition in the Applj IICab Ility that r equire the supported systems) :to If an analysis determines that the supported system(s) do not requI re t snubberIEs to be Iunctional in order to support the OPERABILITY of the system(s), LCO 3.0.8 Is not needed. IF the LCO(S) associate Wit anyIGL supporte s lystems) are not urrenly applcable (.e.. aI n I, met 4 mI a MOD orU I-I. eaIVItio---_ l ell IIl II o , ItLh 'JI C I L J.U.U i1 IIUL-neede. Ifthe sported system(s) are inope~rable for -reasons other1 thman snubbers, LCO 3.0.8 cannot be used. CI 0.8 " s a 1 allowance, nt a uement. When-a -snubber is nonfuncti--o , any.su t e systcm(s) may be declared inoperable instead oF uig LCO 3.0.-8.7 Every time the prov ision of LC0 3.0.8 are used, CR 3 will r-nfrff that at les on train Ear S!tSS!Tnm

F Systr"rt 7

,.,,nLC u EI d, s-uu I en R-3u wil by ooooe in o!!ý ofo P IU L I W1 II

I
  • I E I*11

-I Vw5U IE I*II I 111 IU ILw postul ated desig' od te hnsimclas LCO 3.0.8 does not a o non-seismic snubbers. A -record of the desigýn function1%3 of W theinperable subber (i.e., seismmc vs. non seismic), imleenato of 1~!P! abl Tie 2

--kn--

E-t- -available on a recoverable b asi e NRC staff inspection. The applicable action for each snubber ILCO 3.O.8.a, LCO 3.0.8.b or egnri e aion required, will be listed in the Eq-uipmentDatabase A l-of...... 1-GLII .L l W II Il llý I rT ICTCýl If the allowed time expires and the snubber(s) are unable to perform their associated support funmtions)*, theaffected supprI Led system's LCO-%s)ý must.beeclared. not met and the Co ions and Required A o etered in accordance with LCO 3.0.2. LCE 3.0.8.a applies when one or more snubbers are not ceapable of providing their associated support funtion(s) to~~~ a sigl trai or UbsysteM Of a Mutperano subs-ys.tm supported system or to a i ngle,..in er !the upored syspotem ocurn while the snbbr o)ure not caab e ofprorigthihsscae suppor pucto n I US"Owl 5M"W UI ý ma*III on the 1* ow

b4i l*

of ai sm el UI em 5,y 5 urIII I I i tmh* am I1 r I NIU A,. i I I A I. I li.i the s.p-I-e sy t m o c r ii-l1a .... u,,,,.,. Es-- -) alr-e not (continued) Crystal River Unit 3 B 3.0-17 Revision No. 66

LCO Applicability B 3.0 BASES LCO 3.0.8 due to the availability of the redundant train of the (continued) supported-system. At least one Emergency Feedwater Train (including a mi.imum set of supportirg equipment required f-or its successful operation) not associated with the inoperable snubberCs) will be available when LCQ 3.0.8.a used at CR 3. LCO) 3.0.8.b applies when one or more snubbers are not capable of providing their associated support function(s) to more than one train or subsystem of a multiple train o~r subsystem supported system. LCO 3.0.8.b allows 12 hours t-o restore the snubber~s) before declaring the supported system inoperable. The 12 hour Completion Time is reasonable based on the low probability of a seismic event eancurrent with an event that would requir opration of the supported system occurin while the snbe5) are not capable of performing their associated support function. Fo, snubber; that impact both trains of FFW, 1I.P/rOR ecoling capability must be verified before utilizing 3.0.8.b. At least e Emergency Feedwater Traill c a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber~s) or an alternative means of core cooling, I'rI/rORV cooling, will be available when LCO 3.O.8.b isy used at CR 3. LCO 3.0.8 reursthat risk be assessed and managed. industry and.... NR guidance on the implementation of 10 CFR 50.6..a5')( (the Maintenace. Rule) does not address se.smc rI. owever, use of LCO 3.0.80 should be considered with respect to other planrt maintenance activities, and integrated into the existing Maintenance Rule process to the extent, possible so that maintenance on any umaffected train orsubsystem is properly controlled, and emergen issues ar e properly addressed. The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and Lumponents when one or more snubbers are not able to perform their associated support function. Crystal River Unit 3 B 3.0-18 Revision No. 66

SR Applicability B 3.0 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 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 ad ,omponents, and 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. Vari abl es are assumed. to, be,"- S .ystems-a d mponentsare assumed to be ,PERA..L-when the associated SRs have been met. -Nothing in this Specification, however, is to be construed as implying that-*.V l*L0.s a-raeg.wnithin limits w hen -t~e,~_&É gea uiimn-Ohw A i8l1 e known not. be, m be*tW-- performances systems or .ompome.ts are OPERABLE when:

a.

The systems or mponents are k..wn to be "oaperable, although still meeting the SRsy o~r

b.

The reqirements of the Surveillances) are known not to be.et betwee, required Surveillan, e performances. Surveillances do not have to be performed when the unit is in a MODE or other specified condition for which the requirements of the associated LCO are not applicable, unless otherwise specified. The SRs asso.iated with a "HYSI* S TEST Exc.ep ... LE are only applicable when the P'YSICS TEST Exeeption LEO is used as am allowable exIeption to the requirementls of a Speeifieation. .LI VI I

  • aI u

I -V IIH L IIaVU LUtvc V , C JIH 1 1 UII tII IJ_. aui.L -C equipment beeause the ACTIONS define the remedial measure that apply. SRs have to be et, in acordamce with SR 3.0.2i prior to returning equipment to OPERALE status. Upon completion of maintenance, appropriate pos-t maintenanee testing -s required to declare equipmemt OPERABLE. This imc'-des ineeting applicable SRs + accordance -with SIR 3.0.2. Post mainternanee testing may not be possible in the current MODE or other specife (conti nued) Crsa Rvr U it Jl~ - U -

  • 3=

B 3.-II

',.I.VI 1*

19 RevI

llfl4 i

sio n No... 66 Crystal River Unit 3 B 3.0-19 Revision No. 66

SR Applicability B 3.0 BASES SR 3.0.1 conditions i" the Appli-ability due to the nees t (*,,t,,ued) parameters not having been established. In the"e situations, the equipment may be considered OPERABLE provided testing has been satisfaetorily completed to the extent possible and the eq~uipmn iset otherwirse believe to be incapable of performi,, its function. This will allow operation to proeede to a MODE or other speeified eenditio where other necessary post maintemamee tests ean be eemplted-. 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 plant operating ciqtconditions that may not be suitable for conducting the Surveillance (e.g., transient conditions or other ongoing Surveillance or maintenance activities). The 25% extension does not significantly degrade the 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. The ex.eptions to SR 3.0.2 are these Surve4,lances for whivh the 25% extension of the i.terval specified in the Frequency does not apply. These exceptions are stated in the ,nd*,idual Specifi-ations. The requirement, of regulations take precedence over the TS. Therefore, when a test interval is specified in the regulations, the test in1terval cannot be extended by the TS, and the SR include a Note in the Frequeny stat..g, "SR 3.0.2 4s not app.,,able." An example of an ex^eptio, when the test interval is not specified in the regulations is the NOTE in the Containment Leakage Rate Testing Program, "SR 3.0.2 is not applieable." This exeeption is provided because the program alreadyY includes extension of test imterval. As stated in SR 3.0.2, the 25% extension also does not apply to the initial portion of a periodi* Completion Ti -that requires performance on a "oce per..." basis. The (continued) Crystal River Unit 3 B 3.0-20 Revision No. 66

SR Applicability B 3.0 BASES SR 3.0.2 25...... ex e -

'e eae ne af e th (e *iud erfora o^f*='2 the ea ramn e

Rfequrtediita ,-etior., whether it is a partiular Surveillance or some o0ther remedial action, is e nidered a Single action with a single Completion Ti me. One reason frnot -allowin t' cheki~ he tausof=redundýant or diverse components ory a---..... fu tion of the inoperable*e .pment in an tlterhative manner. The provi ions of SR 3.0.2 are not intended to be used repately erey a anopeatinalconeninceto extend S urvei11anc, e iterval or' perdic CoEmpleti on Ti me inlte rval s In tl~~ ~ ~ ~ ~ ~ ~ ~i. V* I-I..jQcra %-a H e bieyond thosetw speified. SR 3.0.3 SR 3.0.3 establishes the flexibility to defer declaring .affected eqU^i Ment*%21 I -11

  • 1 I

=1-1 i abe or 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 hours or up to the limit of the specified 1 requency, whichever is greater, applies from the point in time that it is discovered that the Surveillance has not been performed in accordance with SIR 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 compl etion of the Surveillance. The basis for this delay period includes consideration of utn-it facijlit conditions,' adequate planning, availability of peronniel, 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. When a Surveillance with a reenybsdnot on time intervals, but uponspecifie umilt conditions, operating I I.)IC I IU Vl~ I ~ JI I.. 1-1*- I1I. 1%J *.. a %-#1~ I,11-I % /1 I IJ I.. ~~1IJ..,,1I:, e Lmpions et. is discojvered tonthaebnprfrd when' specified, S.R 3.0.3 allows for the full delaypro sp~eified, the missed Survei'llance should be performed at thefirt easonable opportunity. (conti nued) Crystal River Unit 3 B 3.0-21 Revision No. 9

SR Applicability B 3.0 BASES SR 3.0.3 (continued) S_ 3.0.3 provides a time limit fr, allowances for the proa, veillace..s that become apibl as a ~nseqence f MOE changes imposed by Reui edAtions. 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 hours 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. The determination of the first reasonable -'L -I 1 LJ--*.7I4-' 1 pmI aIi I Lv UI I I *I UtII ~L I UfI I-T, IIal9c3 I I UT cUI I 1UI I L UVUI9I1 IFI p11 TT, down to perform ththe Surveillance) an m actos n aalysi I ssumtions, in addition to unit conditios rJ I*11111fJl l,-VtliI]L I Ihi I" nf *I JI rI--ui:I, Cl rl hr i r: III I E*I IIfl* -ILI t % I i -th-------.m .i tl im-r... l I\\E 1_IelL I. a Jm IUE LIS*1E 3 3 Id L~b ll3L I I IY

  • UIUImLCl t

I. LEIIJ It yumaI yfI11 Guide-1182, "Assesin and~ Mangn m,,RikBfr Maintenanc-e Activities a1t Nucear rwe Plants." Thi Regulatory Guide addresses consideration of temporary and ac ~ ~ ~ ~ ~ ~ mmtion

thehod, n

riskmagentcio upoad incldin plnt hutown Themised urvillnceshould be treatedajs aan e~meren codtion as discussed in Reuatory Guide 1.182. The risk1 eZvaluaion-WI C ma us quanitatve, .I u i I i IC i l ua,.. L I4

5.

V 31 1G Ed -d I -U "I Id Li-iBU I imprtaceof the component. Missed Surveillances for the results WofL I r*is evl t Ion determIIIIIneI* the risk icrease is significant, this aevaluation should be to dete -rn te safIest. cou r s. of aI o l missed S urveillances will be placed in the licensee's Corrective Action Prrogram.:Refer to-the Master Surveillance Plan for additional gudnefor missed surveillances. If a Surveillance is not completed within the allowed delay period then the equipm is considered oiinpedrabl e or the variable is cons der outside the specified limits and the Completion Times of the Required Actions for the applicable Specification Conditions be qin immediately upon expiration of the delay period. If a Surveillance is failed within the delay period, then the equipment is ino-perable, or 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. (continued) Crystal River Unit 3 B 3.0-22 Revision No. 66

SR Applicability B 3.0 BASES SR 3.0.4 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. This Specification ensures that system and componenrt OPERABILITY' requirements and variable limits are met before entry into MODES or other specified conditions in the Applicability for which these systems and components ensure safe operation of the unit storage of nuclear fuel. The provisions of this Specification should not be interpreted as endorsing the failure to exercise the good practice of restoring thesystems or components to OP[RABLE status before entering an asso.iated MODE or other specified condition in the Applicability. A provision is ineluded to allow entry into a MODE or other specified condition in the Applicability when an LCO is nort met due to Surveillance not being met in acordance with i C-/ 3.0.4.

However, in certain

,irumsta,,es, failing to meet an SR will mot result in SR 3.0.4 restricting a MODE change or other spe-ified conditio "n hange. When a syste",- subsystem, division, component, devi*,, or varlable is inoperable or outside its specifiedI limits, the associated SSREs) are not required to be performed, per SR 3.0.1, which states that surveillan.es do not have to be performed on inoperable equipment. When equipment is inoperable, 5R 3.0.4 does not apply to the asso.iated SRE-) since the requremnt for the SRES) to be performed is removed.

Therefor, failing to perform the Surveillancees) within the specified Frequency does not result in an SR 3.0.4 restriction to changing MODES or other specified conditin of the Applicability.

foweve., since the LCO is not met in this instance, LEO 3.0.4 will govern any restrictions that may (or may not) apply to MODE or other specifed condition changes. SR 3.0.4 does not restrict changing MODES or other specified conditions of the Applicability when -a Surveillance has not been performed within the specified Frequency, provided the requirement to declare the LCO not met has been delayed in accordance with SR 3.0.3. (continued) Crystal River Unit 3 B 3.0-23 Revision No. 66

SR Applicability B 3.0 BASES SR 3.0.4 (continued) The provisions of SR 3.0.4 shall not prevent entry into MODES or other specified conditions in the Applicability that are required to comply with ACTIONS. in addition, the pr~ 1 1~1 s of SR 3.0.4 shall net prevent ehanges in MODBESS or the,,speified n:Hdis in the Appliability that -result from any unit shutdown. In this context, a umit -shutdown is defined as a ehange in MODE or other specified -eeditiem in the Applicability associated with transitioning from MODE 1 to MODBE 2, MODE 2 to MODE 3, MOD 3 to MODE 4, and MODE 4 to MODE 5. 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 MODE or oter specified condition in the Applicability of the associated LCO prior to the performance or completion of a Surveillance. A Surveillance that could not be performed until after emtering the LCO's Applicability, would have its Frequenc-y specified such that it is not "due" until the specific conditios needed are met. Alternately, the Surveillance may be stated in the form of a Note, as not required (tobe met or performed) until a particular event, eondition, or time has been reached. Further discussion of the specific formats of SRs' annotation is found in Section 1.4, Frequeney,- Crystal River Unit 3 B 3.0-24 Revision No. 66

Fuel Storage Pool Water Level B 3.7.13 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. in the ev fuel pool cooling is lost when the pool is 1402F, full core is discharged under the conditions of Refee' e1 the pool volume provides approximately 8 hours before boiling would o,,ur (Ref. 1). 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 SAFETY ANALYSES The minimum water level in the fuel storage pool meets the assumptions of the FHA described in FSAR Section 14.2.2.3. The resultant 2 hour 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. (continued) Crystal River Unit 3 B 3. 7-6561_ Revision No. -H

Fuel Storage Pool Water Level B 3.7.13 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. Required Aetiom A.1 is modified by a Note indicatimg that LCO 3.0.3 does not apply. If moving irradiated fuel assemblies while MODS 1, 2, 3, a.d 4, the fuel movement is indepemdent of reaetor aperatioms. Therefore, placing the reater in a shutdown condition in the event of an inability to suspend movemnent of irradiated fuel assemblies does nothing to compensate for the Required Action not met. it is inappropriate to subjeet the plant to a shutdown transient in this condition. in MOBE[ 5 and 6, LCO 3.0.3 not applicable. (continued) Crystal River Unit 3 B 3.7-6-72 Revision No. 5-

Fuel Storage Pool Water Level B 3.7.13 BASES SURVEILLANCE REQUIREMENTS SR 3.7.13.1 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. Whe" ref.elimg perati s are ta'-ig place, the level in t,,e fu.l peal is at equilibrium with that in the refueling canal a*d in the reactor vessel. The level in the refuel..g .anal is verifqed daily by the perfermamee of SR REFERENCES

1.

^SAR, Section 9.3.1.

2.

FSAR, Section 14.2.2.3.

3.

Deleted.

4.

10 CFR 50.67.

5.

FPC Calculation N-00-0001. Crystal River Unit 3 B 3.7-68.3 Revision No. H

Spent Fuel Pool Boron Concentration B 3.7.14 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 Ž 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 SAFETY ANALYSIS The acceptance criteria for the fuel storage pool criticality analyses is that a kffof < 0.95 must be maintained for all postulated events. The storage racks are capable of maintaining this k with unborated pool water at a temperature yielding the hig est reactivity (assuming the storage restrictions of LCO 3.7.15 are met). Most abnormal storage locations will not result in an increase in the k eff 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 k , 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. LCO The required concentration of dissolved boron in the fuel storage pool of Ž 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. (continued) Crystal River Unit 3 B 3.7-694 Amendment Revision No. +93

Spent Fuel Pool Boron Concentration B 3.7.14 BASES 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. The Required Actions are modifled by a Note indicating that LEO 3.0.3 does met apply. if moving irradiated fuel assemblies while i' O.fB* 1, 2, 3, or 4, the fuel movement is independent of reacto peaion. Therefore, placing the reacter in a shud.w cnditon in the event of an inability to suspend movement of fuel assemblies does mothing to eompemsate for the Required action not met. it is therefore inapropiate to subjeet the plant to a shutdown transiemnt imthis condition. in MODES 5 and 6, LEE) 3.0.3 is not app! ieab! e-.- 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. (continued) Crystal River Unit 3 B 3.7 -705 Revision No. -23

Spent Fuel Pool Boron Concentration B 3.7.14 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.

Crystal River Unit 3 B 3. 7-7-16 Revision No. 67

Spent Fuel Assembly Storage B 3.7.15 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. b. Fuel storage pool "A" and Fuel storage pool "B". In general, and protect the storage the function of the storage racks is to support new and spent fuel from the time it is placed in 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). (continued) Crystal River Unit 3 B 3.7-H-2Z Revision No. 67

Spent Fuel Assembly Storage B 3.7.15 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 (continued) Crystal River Unit 3 B 3. 7--7-38 Revision No. 67

Spent Fuel Assembly Storage B 3.7.15 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 documents that the required burnup varies linearly as a 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 (continued)

Crystal River Unit 3 B 3.7-7-49_ Revision No. fr7

Spent Fuel Assembly Storage B 3.7.15 BASES LCO (continued)

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. 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 RequiredAtion A.! is rnodifiedba Note imdieatih N C .01.-3 does IIUt aply. S I me the design ba1Is accident af conen ini1 thi ~eciicdon~san inadvertent e.tcahy and sicee III , L I/ L. I 1 If1 I I I I I. i 1 I M ,LI thI re is. I Ire o II or Requied A-tions annot be met. (continued) Crystal River Unit 3 B 3. 7--7-510 Revision No. 67

Spent Fuel Assembly Storage B 3.7.15 BASES ACTIONS A.1 (continued) 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. 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. (continued) Crystal River Unit 3 B 3. 7-7611 Revision No. 67

Spent Fuel Assembly Storage B 3.7.15 BASES REFERENCES (continued)

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. (continued) Crystal River Unit 3 B 3. 7--7-712 Revision No. 6-7}}

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