Text

Issuance of Amendments Regarding Revision of the Technical Specifications to Relocate Specific Surveillance Frequency Requirements to a Licensee Controlled Program Using a Risk-Informed Justification
	ML110670536
Person / Time
Site:	Catawba
Issue date:	03/29/2011
From:	Jacqueline Thompson; Plant Licensing Branch II
To:	Morris J; Duke Energy Carolinas
	Thompson, Jon 415-1119
References
	TAC ME3722, TAC ME3723, TSTF-425
	Download: ML110670536 (168)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 March 29,2011 Mr. J. R Morris Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REVISION OF THE TECHNICAL SPECIFICATIONS TO RELOCATE SPECIFIC SURVEILLANCE FREQUENCIES TO A LICENSEE-CONTROLLED PROGRAM USING A RISK-INFORMED JUSTIFICATION (TSTF-425) (T AC NOS. ME3722 AND ME3723)

Dear Mr. Morris:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 263 to Renewed Facility Operating License NPF-35 and Amendment No. 259 to Renewed Facility Operating License NPF-52 for the Catawba Nuclear Station, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated March 31, 2010, as supplemented by letter dated November 30, 2010.

The amendments revise the Technical Specifications by relocating specific surveillance frequencies to a licensee-controlled document using a risk-informed justification.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

If you have any questions, please call me at 301-415-1119.

Sincerely, Jon Thompson, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413 and 50-414

Enclosures:

1. Amendment No. 263 to NPF-35

2. Amendment No. 259 to NPF-52

3. Safety Evaluation cc w/encls: Distribution via Listserv

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 263 Renewed License No. NPF-35

1.

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

A.

The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Renewed Facility Operating License No. NPF-35 filed by the Duke Energy Carolinas, LLC, acting for itself, and North Carolina Electric Membership Corporation (licensees), dated March 31,2010, as supplemented by letter dated November 30, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I;

8.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

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

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-35 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 263

,which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC, shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION G-JI~

Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-35 and the Technical Specifications Date of Issuance:

March 29, 2011

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 DUKE ENERGY CAROLINAS, LLC NORTH CAROLINA MUNICIPAL POWER AGENCY NO.1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 259 Renewed License No. NPF-52

1.

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

A.

The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Renewed Facility Operating License No. NPF-52 filed by the Duke Energy Carolinas, LLC, acting for itself, North Carolina Municipal Power Agency No.1 and Piedmont Municipal Power Agency (licensees), dated March 31, 2010, as supplemented by letter dated November 30, 2010, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (I) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and Oi) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D.

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

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachment to this license amendment, and Paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-52 is hereby amended to read as follows:

(2)

Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 259, which are attached hereto, are hereby incorporated into this renewed operating license. Duke Energy Carolinas, LLC, shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 90 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Gloria Kulesa, Chief Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to License No. NPF-52 and the Technical Specifications Date of Issuance:

March 29, 2011

ATTACHMENT TO LICENSE AMENDMENT NO. 263 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 259 RENEWED FACILITY OPERATING LICENSE NO. NPF-52 DOCKET NO. 50-414 Replace the following pages of the Renewed Facility Operating Licenses and the Appendix A Technical Specifications (TSs) with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages Insert Pages Licenses Licenses NPF-35, page 4 NPF-35, page 4 NPF-52, page 4 NPF-52, page 4 TSs TSs 1.1-6 1.1-6 3.1.1-1 3.1.1-1 3.1.2-2 3.1.2-2 3.1.4-3 3.1.4-3 3.1.4-4 3.1.4-4 3.1.5-2 3.1.5-2 3.1.6-3 3.1.6-3 3.1.8-2 3.1.8-2 3.2.1-3 3.2.1-3 3.2.1-4 3.2.1-4 3.2.1-5 3.2.1-5 3.2.2-3 3.2.2-3 3.2.2-4 3.2.2-4 3.2.3-1 3.2.3-1 3.2.4-4 3.2.4-4 3.3.1-9 3.3.1-9 3.3.1.10 3.3.1.10

- 2 Remove Pages Insert Pages 3.3.1-11 3.3.1-11 3.3.1-12 3.3.1-12 3.3.1-13 3.3.1-13 3.3.1-14 3.3.1-14 3.3.1-15 3.3.1-15 3.3.1-16 3.3.1-16 3.3.1-17 3.3.1-17 3.3.1-18 3.3.1-18 3.3.1-19 3.3.1-19 3.3.1-20 3.3.1-20 3.3.1-21 3.3.1-21 3.3.1-22 3.3.2-10 3.3.2-10 3.3.2-11 3.3.2-11 3.3.2-12 3.3.2-12 3.3.2-13 3.3.2-13 3.3.2-14 3.3.2-14 3.3.2-15 3.3.2-15 3.3.2-16 3.3.2-16 3.3.2-17 3.3.3-3 3.3.3-3 3.3.4-2 3.3.4-2 3.3.5-2 3.3.5-2 3.3.6-2 3.3.6-2 3.3.9-3 3.3.9-3 3.3.9-4 3.4.1-3 3.4.1-3 3.4.3-2 3.4.3-2 3.4.4-1 3.4.4-1 3.4.5-3 3.4.5-3 3.4.6-2 3.4.6-2 3.4.6-3 3.4.7-2 3.4.7-2 3.4.7-3 3.4.8-2 3.4.8-2 3.4.9-2 3.4.9-2 3.4.11-3 3.4.11-3

- 3 Remove Pages Insert Pages 3.4.11-4 3.4.11-4 3.4.12-5 3.4.12-5 3.4.12-6 3.4.12-6 3.4.12-7 3.4.12-7 3.4.12-8 3.4.13-2 3.4.13-2 3.4.14-3 3.4.14-3 3.4.14-4 3.4.14-4 3.4.15-4 3.4.15-4 3.4.16-2 3.4.16-2 3.4.16-3 3.4.16-3 3.4.17-1 3.4.17-1 3.5.1-2 3.5.1-2 3.5.1-3 3.5.2-2 3.5.2-2 3.5.2-3 3.5.2-3 3.5.4-2 3.5.4-2 3.5.5-2 3.5.5-2 3.6.2-5 3.6.2-5 3.6.3-5 3.6.3-5 3.6.3-6 3.6.3-6 3.6.4-1 3.6.4-1 3.6.5-2 3.6.5-2 3.6.6-1 3.6.6-1 3.6.6-2 3.6.6-2 3.6.8-2 3.6.8-2 3.6.9-2 3.6.9-2 3.6.10-2 3.6.10-2 3.6.11-1 3.6.11-1 3.6.11-2 3.6.11-2 3.6.12-1 3.6.12-1 3.6.12-2 3.6.12-2 3.6.12-3 3.6.12-3 3.6.13-2 3.6.13-2 3.6.13-3 3.6.13-3 3.6.14-2 3.6.14-2 3.6.14-3 3.6.14-3

-4 Remove Pages Insert Pages 3.6.15-2 3.6.15-2 3.6.16-1 3.6.16-1 3.6.16-2 3.6.16-2 3.7.4-2 3.7.4-2 3.7.5-3 3.7.5-3 3.7.5-4 3.7.5-4 3.7.6-2 3.7.6-2 3.7.7-2 3.7.7-2 3.7.8-3 3.7.8-3 3.7.9-1 3.7.9-1 3.7.9-2 3.7.10-3 3.7.10-3 3.7.11-2 3.7.11-2 3.7.12-2 3.7.12-2 3.7.13-2 3.7.13-2 3.7.14-1 3.7.14-1 3.7.15-1 3.7.15-1 3.7.17-1 3.7.17-1 3.8.1-5 3.8.1-5 3.8.1-6 3.8.1-6 3.8.1-7 3.8.1-7 3.8.1-8 3.8.1-8 3.8.1-9 3.8.1-9 3.8.1-10 3.8.1-10 3.8.1-11 3.8.1-11 3.8.1-12 3.8.1-12 3.8.1-13 3.8.1-13 3.8.1-14 3.8.1-14 3.8.1-15 3.8.1-15 3.8.3-2 3.8.3-2 3.8.3-3 3.8.3-3 3.8.4-2 3.8.4-2 3.8.4-3 3.8.4-3 3.8.4-4 3.8.4-4 3.8.6-4 3.8.6-4 3.8.7-2 3.8.7-2 3.8.8-2 3.8.8-2

- 5 Remove Pages 3.8.9-3 3.8.10-2 3.9.1-1 3.9.2-2 3.9.3-2 3.9.4-2 3.9.5-2 3.9.6-1 3.9.7-1 5.5-15 Insert Pages 3.8.9-3 3.8.10-2 3.9.1-1 3.9.2-2 3.9.3-2 3.9.4-2 3.9.5-2 3.9.6-1 3.9.7-1 5.5-15 5.5-16

4.

(2)

TechD~ii!L$l?ecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 263.vhich are attached herelo, are hereby incorporated into this renewed opera illig license. Duke Energy Carolinas, LLC shall operate the facility in accordance with the Technical Specifications.

(3) Updated Final Safety Analysis Report The Updated Final Safety Analysis Report supplement submitted pursuant to 10 CFR 54.21(d), as revised on December 16,2002. describes certain future activities to be completed before the period of extended operation. Duke shall complete these activities no later than December 6, 2024, and shall notify the NRC in writing when implementation of these activities is complete and can be verified by NRC inspection.

The Updated Final Safety Analysis Report supplement as revised on December 16, 2002. described above, shall be included in the next scheduled update to the Updated Final Safety Analysis Report required by 10 CFR 50.71 (e){4), following issuance of this renewed operating license. Until that update is complete, Duke may make changes to the programs described in such supplement without prior Commission approval, provided that Duke evaluates each such change pursuant to the criteria set forth in 10 CFR 50.59 and otherwise complies with the requirements in that section.

(4) Antitrust Conditions Duke Energy Carolinas, LLC shall comply with the antitrust conditions delineated in Appendix C to this renewed operating license.

(5) Fire Protection Program (Section 9.5.1, SER. SSER #2, SSER #3, SSER #4, SSER #5)"

Duke Energy Carolinas, LlC Shall implement and maintain in effect all prOvisions of the approved fire protection program as described in the Updated Final Safety Analysis Report, as amended. for the facility and as approved in the SER through Supplement 5, subject to the following provision:

The licensee may make changes to the approved fire protection program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in tI'Je event of a fire.

"The parenthetical notation following the title of this renewed operating license condition denotes the section of the Safety Evaluation Report and/or ;ts supplement wherein this renewed license condition is discussed.

Renewed License No. NPF-35 Amendment No. 263

-4 (2)

Tect;!likrpl Sps!clt!~t!onll,

,The T~tlnic¥I:S.,..lfir"~ns contilined I~ ~p",8ndlA A. as ~ \\t'tougtl "

NnlndmentNo. 259 which.rv:attached'~""o, arl ntnll)y 1n~led Inlo

. (

11I~ ntl'lClWfllCf openKlfl9........ ~. EI'ICkVY CIIfOIl~as, LlC lIh,.y operlle' \\he

, tldRty I" lIecordanc8 wttn 1h8 TIICtnIca/ $'pecil'icationl.

j

, " '(3)'

yPdlllj'CS EI!!II'§i;tx,A"~:R8P2rt.

, The Updated An.1 Safety Analyll. RepOrt lupplemenl lubmltllH:f ~""nl to 10 CFR &4I.21(d). as nrvfled pt1 Di:IcemOef 1** 2D02,.d"cftbes ClltlltJn future IICtlYllln '0 be c:anpivltd ~o"',I'" pertod QI IxttndedoPe;allart. Duke II'IaR complete thes**~lVitl" I"K1..... than F.br.... ry Z4. 20Z8, lind shalt notify lhe NRC., wrltln~ wtItn Implflmentallon 01 lhese actlvlUe, II. complete and c.. be

'wrtrkfd by NRC. inspection, The Updaled,FII"\\liI si!"ty AnalytlsR8Poh liupplem,'n, III rMect 'on December' 8, z002. dftcrlOed abovv. lhall be Included In 'hi nexl sCheduilKr update to the Updated Final Safely AnaIySI. ReP9r1 reqund by 10 CFR 5Q;71(vX4). fol~lnglssU8nce'of lhl. renewed operliUt'lg tlgenae;UnClllhIl,

update II complele, D\\ItIe""y ~k' CharivVa 10 the progrilml dllaibed, ;tlluch supplement wHhoIA prior CommjAIon ipproVIII, provlCted'tat DtAIt,eVlluat"s eich such change' P\\nuent 10 'he cnt"riIt..t fort'lln 10 CFR 50.59 rind otherwise compile, wllh,be reqU""'"'"tI in thaI section, (4)

AntbfYl' Com211ions Cuke J;I"MJl"gY CaroINs, LlC If)IIIt compty with the antitrusl cond\\llon$ deineatud in Appendix C '0 'his rene\\f!'Vd ~peretlng,bnse.,

(~)

ElreP121ectlon pfogrDtl"l (~ectlon9.5.1. SER, SSER2. S$ER #3, SSER #4, SS,ER #6)'

Cuke Energy Camhs, llC IhaII1mPlement 1Ifld mal",1II1n In elfeet.1I promionl of 'he approved flre pt:ol<<1lon program as d~Cftbed In Ole VPdIaled FNI.Sefely AnalysIS F(epO~**\\I'lII1'lfIndIJd; for the fatlllly' and '.. approved" the SER tnroU9h Supph,,"ent 5, sub'ect to lhefoilowtnIJ provision:

, The tlcenl" 'may make criarigel 10 the IIPproyed Hr. p.:oltc'Clon pf09ram withOut pnor appro.... ' of th" CQmmlnkm' only If thOM c"'nget would nQt adVeniety aneet the IIblltv 10 achl""'" IIrtd mainlain.efe shutdtM'n In the

..... n' of a flrv.

"The parenthetical notaten following the title of 'his rVllOWltd oJ)Ol1Iling license oondltlOn denotes the sectloh rA the Safety EYBlua,lon Report pndfor its supplement' wharetrt this rontwOd ftcelin condItion II dllcl.ISlOd.

Renowed llcenoe No, NPF*5~

Amendment No, 259

Definitions 1.1 1.1 Definitions (continued)

THERMAL POWER TRIP ACTUATING DEVICE OPERATIONAL TEST (TADOT)

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

A TADOT shall consist of operating the trip actuating device and verifying the OPERABILITY of required alarm, interlock, and trip functions. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the required accuracy.

Catawba Units 1 and 2 1.1-6 Amendment Nos. 263, 259

SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be within the limit specified in the COLR.

APPLICABILITY:

MODE 2 with kelf < 1.0, MODES 3, 4, and 5.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

SDM not within limit.

A.1 Initiate boration to restore SDM to within limit.

15 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM is within the limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.1.1-1 Amendment Nos. 263, 259

Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1.2.1

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

The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.

Verify measured core reactivity is within +/- 1 % Aklk of predicted values.

FREQUENCY Once prior to entering MODE 1 after each refueling In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.1.2-2 Amendment Nos. 263, 259

Rod Group Alignment Limits 3.1.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COIVIPLETION TIME C.

Required Action and associated Completion Time of Condition B not met.

C.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months D.

More than one rod not within alignment limit.

D.1.1 Verify SDM is within the limit specified in the COLR.

OR D.1.2 Initiate boration to restore required SDM to within limit.

AND D.2 Be in MODE 3.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 1 hour 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.1.4.1 Verify individual rod positions within alignment limit.

FREQUENCY In accordance with the Surveillance Frequency Control Program Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter when the rod position deviation monitor is inoperable

( continued)

Catawba Units 1 and 2 3.1.4-3 Amendment Nos. 263, 259

Rod Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.4.2 Verify rod freedom of movement (trippability) by moving each rod not fully inserted in the core> 10 steps in either direction.

In accordance with the Surveillance Frequency Control Program SR 3.1.4.3 Verify rod drop time of each rod, from the fully withdrawn position, is ~ 2.2 seconds from the beginning of decay of stationary gripper coil voltage to dashpot entry, with:

b.

All reactor coolant pumps operating.

Prior to reactor criticality after each removal of the reactor head Catawba Units 1 and 2 3.1.4-4 Amendment Nos. 263, 259

Shutdown Bank Insertion Limits 3.1.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the limits specified in In accordance with the COLR.

the Surveillance Frequency Control Program Catawba Units 1 and 2 3.1.5-2 Amendment Nos. 263, 259

Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.6.2 Verify each control bank insertion is within the limits specified in the COLR In accordance with the Surveillance Frequency Control Program Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter when the rod insertion limit monitor is inoperable SR 3.1.6.3 Verify sequence and overlap limits specified in the COLR are met for control banks not fully withdrawn from the core.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.1.6-3 Amendment Nos. 263, 259

PHYSICS TESTS Exceptions 3.1.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and associated Completion Time of Condition C not D.1 Be in MODE 3.

15 minutes met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform a CHANNEL OPERATIONAL TEST on power range and intermediate range channels per SR 3.3.1.7, SR 3.3.1.8, and Table 3.3.1-1.

Prior to initiation of PHYSICS TESTS SR 3.1.8.2 Verify the RCS lowest loop average temperature is

> 541°F.

In accordance with the Surveillance Frequency Control Program SR 3.1.8.3 Verify THERMAL POWER is::. 5% RTP.

In accordance with the Surveillance Frequency Control Program SR 3.1.8.4 Verify SDM is within the limit specified in the COLR.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.1.8-2 Amendment Nos. 263, 259

Fa(X,Y,Z) 3.2.1 SURVEILLANCE REQUIREMENTS

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

During power escalation at the beginning of each cycle, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE SR 3.2.1.1 Verify F~(X,y,Z) is within steady state limit.

FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by

> 10% RTP, the THERMAL POWER at which F~(X,y,Z) was last verified In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.2.1-3 Amendment Nos. 263, 259

FQ(X,Y,Z) 3.2.1 SURVEILLANCE SR 3.2.1.2

~------------------------------N()lrE--------------~-------~~-----------

1.

Extrapolate F"6(X,Y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement. If F"6(X,Y,Z) is within limits and the 31 EFPD extrapolation indicates:

and E"6(X,Y,Z)EXTRAPOLATED > EMo(X,Y,Z)

F~(X,Y,Z)OPEXTRAPOLATED F~(X,Y,Z)OP then:

a.

Increase FMQ(X, Y,Z) by the appropriate factor specified in the C()LR and reverify F"6(X,Y,Z).:: F~(X,y,Z)oP; or

b.

Repeat SR 3.2.1.2 prior to the time at which F~(X,Y,Z).:: F6(X,Y,Z)oP is extrapolated to not be met.

2.

Extrapolation of FMQ(X,Y,Z) is not required for the initial flux map taken after reaching equilibrium conditions.

Verify F~(X.Y,Z).:: F~(X.Y,Z)oP.

FREQUENCY

()nce within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by :::.

10% RlrP. the lrHERMAL P()WER at which F~(X.Y,Z) was last verified In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.2.1-4 Amendment Nos. 263, 259

FQ(X,Y,Z) 3.2.1 SURVEILLANCE SR 3.2.1.3

NOTES----------------------------

1.

Extrapolate F~{X,y,Z) using at least two measurements to 31 EFPD beyond the most recent measurement. If F~(X,Y,Z) is within limits and the 31 EFPD extrapolation indicates:

F~(X,Y,Z)EXTRAPOLATED.::: F~(X,Y,Z) RPSEXTRAPOLATED, and EMo(X,Y,Z)EXTRAPOLATED > EMo(X,Y,Z)

FL (X Y Z)RPS L

RPS o

EXTRAPOLATED F o(X, Y,Z) then:

a.

Increase F~{X,y,Z) by the appropriate factor specified in the COLR and reverify F~(X,y,Z).::: F~{X,y,Z)RPS; or

b.

Repeat SR 3.2.1.3 prior to the time at which F~(X,y,Z)::: F~(X,y,Z)RPS is extrapolated to not be met.

2.

Extrapolation of F~{X,y,Z) is not required for the initial flux map taken after reaching equilibrium conditions.

Verify F~{X,y,Z).::: F~(X,y,Z)RPS.

FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by.:::

10% RTP, the THERMAL POWER at which F~(X,y,Z) was last verified In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.2.1-5 Amendment Nos. 263, 259

F6H(X,Y) 3.2.2 SURVEILLANCE REQUIREMENTS

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

During power escalation at the beginning of each cycle, TH ERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE SR 3.2.2.1 Verify FM6H (X,Y) is within steady state limit.

FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by.:::.

10% RTP, the THERMAL POWER at which F~H (X, Y) was last verified In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.2.2-3 Amendment Nos. 263, 259

SURVEILLANCE REQUIREMENTS continued)

SURVEILLANCE S R 3.2. 2.2

NOTES--------------------------------

1.

Extrapolate F~(X,Y) using at least two measurements to 31 EFPD beyond the most recent measurement. If F~H(X,y) is within limits and the 31 EFPD extrapolation indicates:

M l

SURV F L!.H(X,Y)EXTRAPOLATED'::' F aH(X,Y)

EXTRAPOLATED and EML!.H(X,Y)EXTRAPOLATED

> EMaH(X,Y)

Fl (X y)SURV l (X y)SURV L!.H, EXTRAPOLATED F aH then:

a.

Increase F~(X,y) by the appropriate factor specified in the COLR and reverify F~H(X,Y).::: F~H(X,y)SURV; or

b.

Repeat SR 3.2.2.2 prior to the time at which F~H(X,y).::: F~H(X,y)SURV is extrapolated to not be met.

2.

Extrapolation of F~H(X,y) is not required for the initial flux map taken after reaching equilibrium conditions.

FaH(X,Y) 3.2.2 FREQUENCY Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after achieving equilibrium conditions after exceeding, by.::.

10% RTP, the THERMAL POWER at which F~H(X,Y) was last verified In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.2.2-4 Amendment Nos. 263, 259

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits specified in the COLR.

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

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILITY:

MODE 1 with THERMAL POWER ~ 50% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

AFD not within limits.

A.1 Reduce THERMAL POWER to < 50% RTP.

30 minutes SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.2.3.1 Verify AFD within limits for each OPERABLE excore channel.

FREQUENCY In accordance with the Surveillance Frequency Control Program Once within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and every 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months thereafter with the AFD monitor alarm inoperable Catawba Units 1 and 2 3.2.3-1 Amendment Nos. 263, 259

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4. 1

NOTES-------------------------------

1.

With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER

<75% RTP, the remaining three power range channels can be used for calculating QPTR.

2.

SR 3.2.4.2 may be performed in lieu of this Surveillance.

3.

This SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after exceeding 50% RTP.

Verify QPTR is within limit by calculation.

In accorda nce with the Surveillance Frequency Control Program Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter with the QPTR alarm inoperable S R 3.2.4.2

N 0 TE S-----------------------------

Only required to be performed if input from one or more Power Range Neutron Flux channels are inoperable with THERMAL POWER:: 75% RTP.

Verify QPTR is within limit using the movable incore detectors.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.2.4-4 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.2

NOTES------------------------------

1.

Adjust NIS channel if absolute difference is > 2%.

2.

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is:: 15% RTP.

Compare results of calorimetric heat balance calculation to Nuclear Instrumentation System (NIS) channel output.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.3

NOTES------------------------------

1.

Adjust NIS channel if absolute difference is:: 3%.

2.

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is:: 15% RTP.

Compare results of the incore detector measurements to NIS AFD.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.3.1-9 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.4

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

This Surveillance must be performed on the reactor trip bypass breaker prior to placing the bypass breaker in service.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.5 Perform ACTUATION LOGIC TEST.

In accordance with the Surveillance Frequency Control Program S R 3.3. 1.6

N OT E ----------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is.::: 75% RTP.

Calibrate excore channels to agree with incore detector measurements.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.7

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

Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into MODE 3.

Perform COT.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.3.1-10 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.8

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

This Surveillance shall include verification that interlocks P-6 (for the Intermediate Range channels) and P-10 (for the Power Range channels) are in their required state for existing unit conditions.

Perform COT.

NOTE-----

Only required when not performed within the Frequency specified in the Surveillance Frequency Control Program or the previous 184 days Prior to reactor startup Four hours after reducing power below P-10 for power and intermediate range instrumentation Four hours after reducing power below P-6 for source range instrumentation AND In accordance with the Surveillance Frequency Control Program

( continued)

Catawba Units 1 and 2 3.3.1-11 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 SURVEILLANCE SR 3.3.1.9

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

Verification of setpoint is not required.

Perform TADOT.

FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.3.1.1 0 ---------------------------------NOTE----------------------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

SR 3.3.1.11 -------------------------------NOTE----------------------------

1.

Neutron detectors are excluded from CHANNEL CALIBRATION.

2.

Power Range Neutron Flux high voltage detector saturation curve verification is not required to be performed prior to entry into MODE 1 or 2.

3.

Intermediate Range Neutron Flux detector plateau voltage verification is not required to be performed prior to entry into MODE 1 or 2.*

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program (continued)

This Note applies to the Westinghouse-supplied compensated ion chamber neutron detectors. The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors which do not require detector plateau voltage verification. Therefore, this Note does not apply to the fission chamber neutron detectors.

Catawba Units 1 and 2 3.3.1-12 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.12 Perform CHANNEL CALIBRATION.

I n accordance with the Surveillance Frequency Control Program SR 3.3.1.13 Perform COT.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.14 ------------------------NOTE-----------------------------

Verification of setpoint is not required.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.1.15 -----------------------------NOTE------------------------------

Verification of setpoint is not required.

Perform TADOT.

NOTE-----

Only required when not performed within previous 31 days Prior to reactor startup SR 3.3.1.16 ---------------------------------NOTE-------------------------------

Neutron detectors are excluded from response time testing.

Verify RTS RESPONSE TIME is within limits.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.3.1-13 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 SURVEILLANCE FREQUENCY SR 3.3.1.17 Verify RTS RESPONSE TIME for RTDs is within limits.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.3.1-14 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1. Manual Reactor Trip 1,2 2

B SR 3.3.1.14 NA NA 3(a), 4(a). 5(a) 2 C

SR 3.3.1.14 NA NA

2. Power Range Neutron Flux

a.

High 1.2 4

D SR3.3.1.1 SR3.3.1.2 SR3.3.1.7 SR3.3.1.11 SR 3.3.1.16

$110.9%

RTP 109% RTP

b.

Low 1 (b),2 4

E SR3.3.1.1 SR 3.3.1.8 SR 3.3.1.11 SR 3.3.1.16

$27.1% RTP 25% RTP

3.

Power Range Neutron Flux High Positive Rate 1.2 4

D SR3.3.1.7 SR 3.3.1.11

$6.3%RTP with time constant

2 sec 5%RTP with time constant

2 sec (continued)

(a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.

(b)

Below the P-10 (Power Range Neutron Flux) interlocks.

Catawba Units 1 and 2 3.3.1-15 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 2 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

4. Intermediate Range Neutron Flux 1 (b), 2(c) 2 F,G SR3.3.1.1 SR 3.3.1.8(1)(m)

SR 3.3.1.11{~(m)

~31% RTP*

~38%RTP 25% RTP 2(d) 2 H

SR 3.3.1.1 SR 3.3.1.S(I)(m)

SR 3.3.1.11 (Q(m) s 31% RTP*

~38% RTP 25%RTP

5. Source Range Neutron Flux 2(d) 2 I,J SR 3.3.1.1 SR 3.3.1.S(I)(m)

SR 3.3.1.11(1)(m) s 1.4 E5 cps'*

~ 1.44 E5 cps 1.0 E5 cps 3(a), 4(a), 5(a) 2 J,K SR 3.3.1.1 SR 3.3.1.7(I)(m)

SR 3.3.1.11 (I)(m) s 1.4 E5 cps"

~ 1.44 E5 cps 1.0 E5cps

6.

Overtemperature t>.T 1,2 4

E SR 3.3.1.1 SR 3.3.1.3 SR 3.3.1.6 SR3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.17 Refer to Note 1 (Page 3.3.1-19)

Refer to Note 1 (Page 3.3.1-19)

The ~ 31% RTP Allowable Value applies to the Westinghouse-supplied compensated ion chamber Intermediate Range neutron detectors.

The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors.

The ~ 38% RTP Allowable Value applies to the replacement fission chamber Intermediate Range neutron detectors.

- The ~ 1.4 E5 cps Allowable Value applies to the Westinghouse-supplied boron triflouride (BF3) Source Range neutron detectors. The BF3 neutron detectors are being replaced with Themno Scientific-supplied fission chamber neutron detectors. The ~ 1.44 E5 cps Allowable Value applies to the replacement fission chamber Source Range neutron detectors.

(a) With Reactor Trip Breakers (RTBs) closed and Rod Control System capable of rod withdrawal.

(b) Below the P-10 (Power Range Neutron Flux) interlocks.

(c) Above the P-6 (Intermediate Range Neutron Flux) interlocks.

(d) Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(I)

If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(m) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the NOMINAL TRIP SETPOINT (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to detemnine the as-found and the as-left tolerances are specified in the UFSAR.

Catawba Units 1 and 2 3.3.1-16 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 3 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

7.

8.

Overpower 6 T Pressurizer Pressure 1,2 4

E SR 3.3.1.1 SR 3.3.1.3 SR 3.3.1.6 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16 SR3.3.1.17 Refer to Note 2 (Page 3.3.1-20)

Refer to Note 2 (Page 3.3.1-20)

a.

Low 1 (e) 4 L

SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

~ 1938(f) psig 1945(f) psig

b.

High 1,2 4

E SR 3.3.1.1 SR 3.3.1.7 SR3.3.1.10 SR3.3.1.16

~2399 psig 2385 psig

9.

Pressurizer Water Level-High 1(e) 3 L

SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10

~ 93.8%

92%

10. Reactor Coolant Flow-Low

a.

Single Loop 1 (g) 3 per loop M

SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR3.3.1.16

~ 89.7%

91%

b.

Two Loops 1(h) 3 per loop L

SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1,16

~ 89.7%

91%

(continued)

(e)

Above the P-7 (Low Power Reactor Trips Block) interlock.

(f)

Time constants utilized in the lead-lag controller for Pressurizer Pressure - Low are 2 seconds for lead and 1 second for lag.

(g)

Above the P-8 (Power Range Neutron Flux) interlock.

(h)

Above the P-7 (Low Power Reactor Trips Block) interlock and below the P-8 (Power Range Neutron Flux) interlock.

Catawba Units 1 and 2 3.3.1-17 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 4 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

11. Undervoltage RCPs

12. Underfrequency RCPs 1 (e) 1(e) 1 per bus 1 per bus L

L SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.16 SR 3.3.1.9 SR 3.3.1.10 SR 3.3.1.16

~ 5016 V

~ 55.9 Hz 5082 V 56.4 Hz

13. Steam Generator (SG) Water Level-Low Low 1,2 4 per SG E

SR 3.3.1.1 SR 3.3.1.7 SR 3.3.1.10 SR 3.3.1.16

~ 9% (Unit 1)

~ 35.1%

(Unit 2) of narrow range span 10.7%

(Unit 1) 36.8%

(Unit 2) of narrow range span

14. Turbine Trip

a.

Stop Valve EH Pressure Low

10) 4 N

SR 3.3.1.10 SR 3.3.1.15

~ 500 psig 550 psig

b.

Turbine Stop Valve Closure

10) 4 0

SR 3.3.1.10 SR 3.3.1.15

~ 1% open NA

15. Safety Injection (SI)

Input from Engineered Safety Feature Actuation System (ESFAS) 1,2 2 trains P

SR 3.3.1.5 SR 3.3.1.14 NA NA (e)

Above the P-7 (Low Power Reactor Trips Block) interlock.

(continued)

(i)

Not used.

0)

Above the P-9 (Power Range Neutron Flux) interlock.

Catawba Units 1 and 2 3.3.1-18 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 8)

Reactor Trip System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

16. Reactor Trip System Interlocks

a.

Intermediate 2(d) 2 R

SR 3.3.1.11

<: 6E-11 1E-10 Range Neutron SR 3.3.1.13 amp"*

amp'"

Flux, P-6

?6.6E.a%

1E-5% RTP RTP

b.

Low Power 1 per train S

SR3.3.1.5 NA NA Reactor Trips Block, P-7

c.

Power Range 4

S SR 3.3.1.11 s; 50.2% RTP 48% RTP Neutron Flux, SR 3.3.1.13 P-8

d.

Power Range 4

S SR3.3.1.11 s; 70% RTP 69%RTP Neutron Flux, SR3.3.1.13 P-9

e.

Power Range 1,2 4

R SR 3.3.1.11

<: 7.8% RTP 10% RTP Neutron Flux, P-10 SR 3.3.1.13 and s; 12.2%

RTP

f.

Turbine Impulse Pressure, P-13 2

S SR3.3.1.12 SR3.3.1.13 S; 12.2% RTP turbine impulse pressure equivalent 10% RTP turbine impulse pressure equivalent

17. Reactor Trip 1,2 2 trains Q,U SR 3.3.1.4 NA NA Breakers(k) 3(a), 4(a), 5(a) 2 trains C

SR 3.3.1.4 NA NA

18. Reactor Trip Breaker 1.2 1 each per T

SR 3.3.1.4 NA NA Undervoltage and RTB Shunt Trip Mechanisms 3(a). 4(a). 5(a) 1 each per C

SR 3.3.1.4 NA NA RTB

19. Automatic Trip Logic 1,2 2 trains P,U SR 3.3.1.5 NA NA 3(a), 4(a), 5(a) 2 trains C

SR 3.3.1.5 NA NA (continued)

The? 6E-11 amp Allowable Value and the 1 E-1 0 amp NOMINAL TRIP SETPOINT value apply to the Westinghouse-supplied compensated ion chamber Intermediate Range neutron detectors. The compensated ion chamber neutron detectors are being replaced with Thermo Scientific-supplied fission chamber neutron detectors. The? 6.6E-6% RTP Allowable Value and the 1 E-5% RTP NOMINAL TRIP SETPOINT value apply to the replacement fission chamber Intermediate Range neutron detectors.

(a) With RTBs closed and Rod Control System capable of rod withdrawal.

(d)

Below the P-6 (Intermediate Range Neutron Flux) interlocks.

(k)

Including any reactor trip bypass breakers that are racked in and closed for bypassing an RTB.

Catawba Units 1 and 2 3.3.1-19 Amendment Nos. 263, 2591

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 8)

Reactor Trip System Instrumentation Note 1: Overtemperature LlT The Overtemperature LlT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 4.3% (Unit 1) and 4.5% (Unit 2) of RTP.

Where: LlT is the measured RCS LlT by loop narrow range RTDs, OF.

LlTo is the indicated LlT at RTP, OF.

s is the Laplace transform operator, sec*

1 T is the measured RCS average temperature, OF.

T' is the nominal Tavg at RTP (allowed by Safety Analysis), ~ the values specified in the COLR.

P is the measured pressurizer pressure, psig p' is the nominal RCS operating pressure, = the value specified in the COLR K1

= Overtemperature LlT reactor NOMINAL TRIP SETPOINT, as presented in the COLR, K2

= Overtemperature LlT reactor trip heatup setpoint penalty coefficient, as presented in the COLR, K3

= Overtemperature LlT reactor trip depressurization setpoint penalty coefficient, as presented in the COLR, L1, L2 = Time constants utilized in the lead-lag compensator for LlT, as presented in the COLR, L3

= Time constant utilized in the lag compensator for LlT, as presented in the

COLR, L4, L5

= Time constants utilized in the lead-lag compensator for Tavg, as presented in the COLR, L6

= Time constant utilized in the measured T avg lag compensator, as presented in the COLR, and fl(LlI) = a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for qt - qb between the "positive" and "negative" f1(LlI) breakpoints as presented in the COLR; fl(LlI) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent Lli that the magnitude of % - qb is more negative than the f1(LlI) "negative" breakpoint presented in the COLR, the L1T Trip Setpoint shall be automatically reduced by the f1(LlI) "negative" slope presented in the COLR; and Catawba Units 1 and 2 3.3.1-20 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 7 of 8)

Reactor Trip System Instrumentation (iii) for each percent l!1 that the magnitude of qt - qb is more positive than the f1(l!I) "positive" breakpoint presented in the COLR, the l!T Trip Setpoint shall be automatically reduced by the f1(l!I) "positive" slope presented in the COLR.

Note 2: Overpower l!T The Overpower l!T Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 2.6% (Unit 1) and 3.1% (Unit 2) of RTP.

t,r(1 + 1'1 s) (

1

)

t,T. {K K 1'7 s

(

1 Jr K [r 1

(1 + 1'2 s) 1 + 1'3 s:O; 0

4 51 + 1'7 S 1 + 1'6 S 6

1 + 1'6 S Where:

l!T is the measured RCS l!T by loop narrow range RTDs, of.

l!T0 is the indicated l!T at RTP, of.

s is the Laplace transform operator, sec*1*

T is the measured RCS average temperature, of.

T" is the nominal Tavg at RTP (calibration temperature for l!T instrumentation),

~ the values specified in the COLR.

~ = Overpower l!T reactor NOMINAL TRIP SETPOINT as presented in the

COLR, Ks

= the value specified in the COLR for increasing average temperature and the value specified in the COLR for decreasing average temperature, Ks

= Overpower l!T reactor trip heatup setpoint penalty coefficient as presented in the COLR for T > T" and K6 = the value specified in the COLR for T ~ T",

11,12

= Time constants utilized in the lead-lag compensator for l!T, as presented in the COLR, 13

= Time constant utilized in the lag compensator for l!T, as presented in the

COLR, 16

= Time constant utilized in the measured Tavg lag compensator, as presented in the COLR, 17

= Time constant utilized in the rate-lag controller for T avg, as presented in the COLR, and f2(l!I) = a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for qt - qb between the "positive" and "negative" f2(l!I) breakpoints as presented in the COLR; f2(l!I) =0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (continued)

Catawba Units 1 and 2 3.3.1-21 Amendment Nos. 263, 259

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 8 of 8)

Reactor Trip System Instrumentation (ii) for each percent AI that the magnitude of qt - qb is more negative than the f2(AI) "negative" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f2(AI) "negative" slope presented in the COLR; and (iii) for each percent AI that the magnitude of qt - qb is more positive than the f2(AI) "positive" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f2(AI) "positive" slope presented in the COLR.

Catawba Units 1 and 2 3.3.1-22 Amendment Nos. 263, 259

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.

SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.2 Perform ACTUATION LOGIC TEST.

In accordance with the Surveillance Frequency Control Program S R 3.3.2.3

NO T E ---------**-----------------

Final actuation of pumps or valves not required.

Perform T ADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.4 Perform MASTER RELAY TEST.

In accordance with the Surveillance SR 3.3.2.5 Perform COT.

SR 3.3.2.6 Perform SLAVE RELAY TEST.

Frequency Control Program In accordance with the Surveillance Frequency Control Program I n accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.3.2-10 Amendment Nos. 263, 259

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.7 Perform COT.

In accordance with the Surveillance Frequency Control Program S R 3.3.2.8

NOT E -------------------------

Verification of setpoint not required for manual initiation functions.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program S R 3.3.2. 9

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

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.1 a -------------------------------NOTE------------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after SG pressure is ::: 600 psig.

Verify ESFAS RESPONSE TIMES are within limit.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.3.2-11 Amendment Nos. 263, 259

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.11 Perform COT.

In accordance with the Surveillance Frequency Control Program SR 3.3.2.12 Perform ACTUATION LOGIC TEST.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.3.2-12 Amendment Nos. 263, 259

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 1 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

1.

Safety Injection(b)

a.

Manual initiation 1,2,3,4 2

B SR 3.3.2.8 NA NA

b.

Automatic Actuation Logic and Actuation Relays 1,2,3,4 2 trains C

SR 3.3.2.2 SR 3.3.2.4 SR 3.3.2.6 NA NA

c.

Containment Pressure - High 1,2,3 3

D SR 3.3.2.1 SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

~ 1.4 psig 1.2 psig

d.

Pressurizer Pressure - Low 1,2,3(a) 4 D

SR 3.3.2.1 SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

1839 psig 1845 psig

2.

Containment Spray*

a.

Manual Initiation 1,2,3,4 1 per train, 2 trains B

SR 3.3.2.8 NA NA

b.

Automatic Actuation Logic and Actuation Relays 1,2,3,4 2 trains C

SR 3.3.2.2 SR 3.3.2.4 SR 3.3.2.6 NA NA

c.

Containment Pressure High High 1,2,3 4

E SR 3.3.2.1 SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

~ 3.2 psig 3.0 psig

3.

Containment Isolation(b)

a.

Phase A Isolation (1 ) Manual Initiation 1,2,3,4 2

B SR 3.3.2.8 NA NA (2) Automatic Actuation Logic and Actuation Relays 1,2,3,4 2 trains C

SR 3.3.2.2 SR 3.3.2.4 SR 3.3.2.6 NA NA (3) Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

(continued)

The requirements of this Function are not applicable for entry into the applicable MODES following implementation of the modifications associated with ECCS Water Management on the respective unit.

(a) Above the P-11 (Pressurizer Pressure) interlock.

(b) The requirements of this Function are not applicable to Containment Purge Ventilation System and Hydrogen Purge System components, since the system containment isolation valves are sealed closed in MODES 1, 2, 3, and 4.

Catawba Units 1 and 2 3.3.2-13 Amendment Nos. 263, 259

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 2 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

3.

Containment Isolation (continued)

b.

Phase B Isolation (1 ) Manual Initiation 1,2,3,4 1 per train, 2 trains B

SR 3.3.2.8 NA NA (2) Automatic Actuation Logic and Actuation Relays 1,2,3,4 2 trains C

SR 3.3.2.2 SR 3.3.2.4 SR 3.3.2.6 NA NA (3) Containment Pressure High High 1,2,3 4

E SR 3.3.2.1 SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

3.2 psig 3.0 psig

4.

Steam Line Isolation

a.

Manual Initiation (1) System 1,2(b),3(b) 2 trains F

SR 3.3.2.8 NA NA (2) Individual 1,2(b),3(b) 1 per tine G

SR 3.3.2.8 NA NA

b.

Automatic Actuation Logic and Actuation Relays 1,2(b),3(b) 2 trains H

SR 3.3.2.2 SR 3.3.2.4 SR 3.3.2.6 NA NA

c.

Containment Pressure - High High 1,2(b),3(b) 4 E

SR 3.3.2.1 SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

3.2 psig 3.0 psig

d.

Steam Line Pressure (1) Low 1,2(b),3(a)(b) 3 per steam line D

SR 3.3.2.1 SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

~ 744 psig 775 psig (continued)

(a)Above the P-11 (Pressurizer Pressure) interlock.

(b) Except when all MSIVs are closed and de-activated.

Catawba Units 1 and 2 3.3.2-14 Amendment Nos. 263, 259

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 3 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

4. Steam Line Isolation (continued)

(2) Negative 3(b)(c}

3 per steam D

SR 3.3.2.1

~ 122.8(d) 100(d) psi Rate High line SR 3.3.2.5 psi SR 3.3.2.9 SR 3.3.2.10

5.

Turbine Trip and Feedwater Isolation

a.

Turbine Trip (1) Automatic 1,2 2 trains SR 3.3.2.2 NA NA Actuation SR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (2) SG Water 1,2 4 perSG J

SR 3.3.2.1

~85.6%

83.9%

Level SR 3.3.2.2 (Unit 1)

(Unit 1)

High-High SR 3.3.2.4

~ 78.9%

77.1%

(P-14)

SR 3.3.2.5 (Unit 2)

(Unit 2)

SR 3.3.2.6 SR 3.3.2.9 SR 3.3.2.10 (3) Safety Refer to Function 1 (Safety Injection) for all initiation functions and requirements. See Injection Item 5.a.(1) for Applicable MODES.

b.

Feedwater Isolation 2 trains H

SR 3.3.2.2 NA NA Actuation (1) Automatic SR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (continued)

(b) Except when all MSIVs are closed and de-activated.

(c) Trip function automatically blocked above P-11 (Pressurizer Pressure) interlock and may be blocked below P-11 when Steam Line Isolation Steam Line Pressure - Low is not blocked.

(d) Time constant utilized in the ratellag controlier is ~ 50 seconds.

(e) Except when all MFIVs, MFCVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve.

Catawba Units 1 and 2 3.3.2-15 Amendment Nos. 263, 259

3.3.2 ESFAS Instrumentation Table 3.3.2-1 (page 4 of 5)

Engineered Safety Feature Actuation System Instrumentation APPLICABLE MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT (2)

SG Water 4 per SG D

SR 3.3.2.1 585.6%

83.9%

Level-High SR 3.3.2.2 (Unit 1)

(Unit 1)

High (P-14)

SR 3.3.2.4

~78.9%

77.1%

SR 3.3.2.5 (Unit 2)

(Unit 2)

SR 3.3.2.6 SR 3.3.2.9 SR 3.3.2.10 (3) Safety Refer to Function 1 (Safety Injection) for all initiation functions and requirements. See Injection Item 5.b.(I) for Applicable MODES.

(4)

Tavg-Low 1,2(e) 4 J

SR 3.3.2.1

~ 561°F 564°F SR 3.3.2.5 SR 3.3.2.9 coincident with Refer to Function 8.a (Reactor Trip, P-4) for all initiation functions and requirements.

Reactor Trip, P-4 (5) Doghouse 1,2(e)

(111 logic)

L (111 logic)

$ 12 inches 11 inches WaterLevel -

2 per SR3.3.2.8 above 577 ft above 577 High High doghouse floor Jevel fI floor level (213 logic)

(213 logic) 3 per train SR 3.3.2.8 per SR 3.3.2.9 doghouse SR 3.3.2.12

6. AUXiliary Feedwater

a.

Automatic 1.2,3 2 trains H

SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays

b.

SG Water Level 1,2,3 4 perSG 0

SR 3.3.2.1

~9%

10.7%

-Low Low SR 3.3.2.5 (Unit 1)

(Unit 1)

SR 3.3.2.9

~ 35.1%

36.8%

SR 3.3.2.10 (Unit 2)

(Unit 2)

c.

Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

d.

Loss of Offsite 1,2,3 3 per bus D

SR 3.3.2.3

~3242V 3500 V Power SR 3.3.2.9 SR 3.3.2.10

e.

Trip of all Main 1,2 3 per pump K

SR 3.3.2.8 NA NA Feedwater SR 3.3.2.10 Pumps

f.

Auxiliary 1,2,3 3 per train M

SR 3.3.2.8 A) ~ 9.5 psig A) 10.5 Feedwater Pump SR 3.3.2.10 psig Train A and Train B Suction

8) ~ 5.2 psig

8) 6.2 psig Transfer on (Unit 1)

(Unit 1)

Suction

~ 5.0 psig 6.0 psig Pressure - Low (Unit 2)

(Unit 2)

(continued)

(e)

Except when all MFIVs, MFCVs, and associated bypass valves are closed and de-activated or isolated by a closed manual valve.

Catawba Units 1 and 2 3.3.2-16 Amendment Nos. 263, 259

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 5 of 5)

Engineered Safety Feature Actuation System Instrumentation MODES OR OTHER NOMINAL SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE TRIP FUNCTION CONDITIONS CHANNELS CONDITIONS REQUIREMENTS VALUE SETPOINT

7. Automatic Switch over to Containment Sump

a.

Automatic Actuation Logic and Actuation Relays 1,2,3,4 2 trains C

SR 3.3.2.2 SR 3.3.2.4 SR 3.3.2.6 NA NA

b.

Refueling Water Storage Tank (RWST) Level Low 1,2,3,4 4

N SR 3.3.2.1 SR 3.3.2.7(*)(b)

SR 3.3.2.9(a)(b)

SR 3.3.2.10

2: 162.4 inches*

177.15 inches*

Coincident with Safety Injection Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

8.

ESFAS Interlocks

a.

Reactor Trip, P-4 1,2,3 1 per train, 2 trains F

SR 3.3.2.8 NA NA

b.

Pressurizer Pressure, P-11 1,2,3 3

0 SR 3.3.2.5 SR 3.3.2.9

2: 1944 and s 1966 psig 1955 psig

c.

Tavg - Low Low, P-12 1,2,3 1 per loop 0

SR 3.3.2.5 SR 3.3.2.9

~ 550°F 553°F

9.

Containment Pressure Control System

a.

Start Permissive 1,2,3,4 4 per train P

SR 3.3.2.1 SR 3.3.2.7 SR 3.3.2.9 s: 1.0 psid 0.9 psid

b.

Termination 1,2,3,4 4 per train P

SR 3.3.2.1 SR 3.3.2.7 SR 3.3.2.9

~ 0.25 psid 0.35 psid

10. Nuclear Service Water Suction Transfer - Low Pit Level 1,2,3,4 3 per pit Q,R SR 3.3.2.1 SR 3.3.2.9 SR 3.3.2.11 SR 3.3.2.12

?; EI. 555.4 ft EI. 557.5 ft Following implementation of the modifications associated with ECCS Water Management on the respective unit, the Allowable Value for this Function shall be ~ 91.9 inches and the Nominal Trip Setpoint for this Function shall be 95 inches.

(a) If the as-found channel setpoint is outside its predefined as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(b) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Nominal Trip Setpoint (NTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the NTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the Surveillance procedures (field setting) to confirm channel performance. The methodologies used to determine the as-found and the as-left tolerances are specified in the UFSAR.

Catawba Units 1 and 2 3.3.2-17 Amendment Nos. 263, 259

PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS

NOT E ------------------------------------------------------------

SR 3.3.3.1 and SR 3.3.3.3 apply to each PAM instrumentation Function in Table 3.3.3-1.

SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.

In accordance with the Surveillance Frequency Control Program SR 3.3.3.2 Not Used S R 3.3.3.3

N 0 TE S-----------------------------

1.

Neutron detectors are excluded from CHANNEL CALIBRATION.

2.

CHANNEL CALIBRATION may consist of an electronic calibration of the Containment Area High Range Radiation Monitor, not including the detector, for range decades above 10 Rlh and a one point calibration check of the detector below 10 R/h with an installed or portable gamma source.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.3.3-3 Amendment Nos. 263, 259

Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized.

In accordance with the Surveillance Frequency Control Program S R 3.3.4.2

N 0 TE----------------------------

Not applicable to Reactor Trip Breaker Position.

Perform CHANNEL CALIBRATION for each required In accordance with instrumentation channel.

the Surveillance

Frequency Control I Program Catawba Units 1 and 2 3.3.4-2 Amendment Nos. 263, 259

LOP DG Start Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY S R 3.3.5.1

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

Testing shall consist of voltage sensor relay testing excfuding actuation of load shedding diesel start, and time delay times.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.2 Perform CHANNEL CALIBRATION with NOMINAL TRIP SETPOINT and Allowable Value as follows:

a.

Loss of voltage Allowable Value> 3242 V.

Loss of voltage NOMINAL TRIP SETPOINT =

3500 V.

b.

Degraded voltage Allowable Value.::: 3738 V.

Degraded voltage NOMINAL TRIP SETPOINT =

3766 V.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.3.5-2 Amendment Nos. 263, 259

Containment Air Release and Addition Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS

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

Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Air Release and Addition Isolation Function.

SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform ACTUATION LOGIC TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.2 Perform MASTER RELAY TEST.

In accordance with the Surveillance Frequency Control Program SR 3.3.6.3 Perform SLAVE RELAY TEST.

In accordance with the Surveillance Frequency Control Program S R 3.3.6.4

NaTE---------------------------

Verification of setpoint is not required.

Perform TADOT.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.3.6-2 Amendment Nos. 263, 259

BDMS 3.3.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.9.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.9.2 Perform COT.

In accordance with the Surveillance Frequency Control Program SR 3.3.9.3 Verify each automatic valve moves to the correct position and Reactor Makeup Water pumps stop upon receipt of an actual or simulated actuation signal.

I n accordance with the Surveillance Frequency Control Program SR 3.3.9.4

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

Only required to be performed when used to satisfy Required Action A.3 or B.3.

Perform CHANNEL CHECK on the Source Range Neutron Flux Monitors.

In accordance with the Surveillance Frequency Control Program S R 3.3.9.5

N OTE ------------------------------

Only required to be performed when used to satisfy Required Action A.3 or B.3.

Verify combined f10wrates from both Reactor Makeup Water Pumps are:::: the value in the COLR.

In accordance with the Surveillance Frequency Control Program

( continued)

Catawba Units 1 and 2 3.3.9-3 Amendment Nos. 263, 259

BDMS 3.3.9 SURVEILLANCE REQUIREMENTS (continued)

SU RVEI LLANCE SR 3.3.9.6

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

Only required to be performed when used to satisfy Required Action A.3 or B.3.

Perform COT on the Source Range Neutron Flux Monitors.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.3.9-4 Amendment Nos. 263, 259

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is within limits.

In accordance with the Surveillance Frequency Control Program SR 3.4.1.2 Verify RCS average temperature is within limits.

In accordance with the Surveillance Frequency Control Program SR 3.4.1.3 Verify RCS total flow rate is within limits.

In accordance with the Surveillance Frequency Control Program SR 3.4.1.4 Perform CHANNEL CAUBRA TION for each RCS total flow indicator.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.1-3 Amendment Nos. 263, 259

RCS PfT Limits 3.4.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

NOTE----------

Required Action C.2 shall be completed whenever this Condition is entered.

C.1 Initiate action to restore parameter(s) to within limits.

Immediately Requirements of LCO not met any time in other than MODE 1, 2, 3, or 4.

C.2 Determine RCS is acceptable for continued operation.

Prior to entering MODE 4 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.3.1. ------------------------------NOTE--------------------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and RCS heatup and cooldown rates are within limits.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.3-2 Amendment Nos. 263, 259

RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops -

MODES 1 and 2 LCO 3.4.4 Four RCS loops shall be OPERABLE and in operation.

APPLICABILITY:

MODES 1 and 2.

ACTIONS CONDITION I

REQUIRED ACTION COMPLETION TIME A.

Requirements of LCO not met.

A.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each RCS loop is in operation.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.4-1 Amendment Nos. 263, 259

RCS Loops - MODES 3 3.4.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation.

In accordance with the Surveillance Frequency Control Program SR 3.4.5.2 Verify steam generator secondary side water levels are

12% narrow range for required RCS loops.

In accordance with the Surveillance Frequency Control Program SR 3.4.5.3 Verify correct breaker alignment and indicated power are In accordance with available to the required pumps that are not in operation.

the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.5-3 Amendment Nos. 263, 259

RCS Loops - MODES 4 3.4.6 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

One RHR loop OPERABLE.

AND ALL RCS loops inoperable.

B.1 Be in MODE 5.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C.

Both required RCS or RHR loops inoperable.

OR No RCS or RHR loop in operation.

C.1 AND C.2 Suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1 and maintain kelt < 0.99.

Initiate action to restore one loop to OPERABLE status and operation.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation.

In accordance with the Surveillance Frequency Control Program SR 3.4.6.2 Verify SG secondary side water levels are =:: 12% narrow range for required RCS loops.

In accordance with the Surveillance Frequency Control Program

( continued)

Catawba Units 1 and 2 3.4.6-2 Amendment Nos. 263, 259

RCS Loops - MODES 4 3.4.6 SURVEILLANCE SR 3.4.6.3 Verify correct breaker alignment and indicated power are available to the required pump that is not in operation.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.6-3 Amendment Nos. 263, 259

RCS Loops - MODES 5, Loops Filled 3.4.7 ACTIONS A.

8.

CONDITION One RHR loop inoperable.

AND Required SGs secondary side water levels not within limits.

Required RHR loops inoperable.

OR No RH R loop in operation.

REQUIRED ACTION A.1 Initiate action to restore a second RHR loop to OPERABLE status.

OR A.2 Initiate action to restore required SG secondary side water levels to within limits.

8.1 Suspend operations that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.

AND B.2 Initiate action to restore one RHR loop to OPERABLE status and operation.

COMPLETION TIME Immediately Immediately Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify one RHR loop is in operation.

In accordance with the Surveillance Frequency Control Program SR 3.4.7.2 Verify SG secondary side water level is.::: 12% narrow range in required SGs.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.4.7-2 Amendment Nos. 263. 259

RCS Loops - MODES 5, Loops Filled 3.4.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE SR 3.4.7.3 Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation.

FREQUENCY I n accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.7-3 Amendment Nos. 263, 259

RCS Loops - MODES 5, Loops Not Filled 3.4.8 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

Required RHR loops inoperable.

B.1 No RHR loop in operation.

AND B.2 SURVEILLANCE REQUIREMENTS Suspend operations that Immediately would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.

Initiate action to restore Immediately one RH R loop to OPERABLE status and operation.

SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one RHR loop is in operation.

In accordance with the Surveillance Frequency Control Program SR 3.4.8.2 Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.8-2 Amendment Nos. 263, 259

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.9.1 Verify pressurizer water level is.:::. 92% (1656 ft\\

In accordance with the Surveillance Frequency Control Program SR 3.4.9.2 Verify capacity of each required group of pressurizer heaters is.::: 150 kW.

In accordance with the Surveillance Frequency Control Program SR 3.4.9.3 Verify required pressurizer heaters are capable of being powered from an emergency power supply.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.9-2 Amendment Nos. 263, 259

Pressurizer PORVs 3.4.11 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME F.

(continued)

F.2 AND F.3 Restore one block valve to OPERABLE status if three block valves are inoperable.

Restore remaining block valve(s) to OPERABLE status.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 72 hours G.

Required Action and associated Completion Time of Condition F not met.

G.1 AND G.2 Be in MODE 3.

Be in MODE 4.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 12 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.4.11.1 -----------------------------NOTE--------------------------------

Not required to be met with block valve closed in accordance with the Required Action of Condition B or E.

Perform a complete cycle of each block valve.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.4.11-3 Amendment Nos. 263, 259

Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.11.2 ----------------------------NOTE------------------------------

Required to be performed in MODE 3 or MODE 4 when the temperature of all RCS cold legs is > 200QF.

Perform a complete cycle of each PORV.

In accordance with the Surveillance Freq uency Control Program SR 3.4.11.3 ---------------------------NOTE-----------------------------

This SR is not applicable to valve NC-36B.

Verify the nitrogen supply for each PORV is OPERABLE by:

a.

Manually transferring motive power from the air supply to the nitrogen supply,

b.

Isolating and venting the air supply, and

c.

Operating the PORV through one complete cycle.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.11-4 Amendment Nos. 263, 259

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of two pumps (charging, safety injection, or charging and safety injection) are capable of injecting into the RCS.

In accordance with the Surveillance Frequency Control Program SR 3.4.12.2 Verify each accumulator is isolated.

In accordance with the Surveillance Frequency Control Program SR 3.4.12.3 Verify RHR suction isolation valves are open for each required RHR suction relief valve.

In accordance with the Surveillance Frequency Control Program SR 3.4.12.4 Verify PORV block valve is open for each required PORV.

In accordance with the Surveillance Frequency Control Program SR 3.4.12.5 ------------------------------N OT E-------------------------------

Not required to be met until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing RCS cold leg temperature to.::: 210°F.

Perform a COT on each required PORV, excluding actuation.

In accordance with the Surveillance Frequency Control Program SR 3.4.12.6 Perform CHANNEL CALIBRATION for each required PORV actuation channel.

In accordance with the Surveillance Frequency Control Program

( continued)

Catawba Units 1 and 2 3.4.12-5 Amendment Nos. 263, 259

SURVEILLANCE SR 3.4.12.7 Verify associated RHR suction isolation valves are open, with operator power removed and locked in removed position, for each required RHR suction relief valve.

LTOP System 3.4.12 FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.12-6 Amendment Nos. 263, 259

LTOP System 3.4.12 Table3.4.12~1 (Page 1 of1)

(UNIT 1 ONLY)

Reactor Coolant Pump Operating Restrictions for Low Temperature Overpressure Protection Reactor Coolant System Cold Leg Maximum Number of Pumps Allowed in Temperature Operation 2

4 Catawba Units 1 and 2 3.4.12-7 Amendment Nos. 263, 259

LTOP System 3.4.12 Table 3.4.12-1 (Page 1 of 1)

(UNIT 2 ONLY)

Reactor Coolant Pump Operating Restrictions for Low Temperature Overpressure Protection Reactor Coolant System Cold Leg Maximum Number of Pumps Allowed in Temperature Operation 1

4 Catawba Units 1 and 2 3.4.12-8 Amendment Nos. 263, 259 I

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1

NOTES------------------------------

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

NOTE----

Only required to be performed during steady state operation Verify RCS Operational LEAKAGE within limits by performance of RCS water inventory balance.

In accordance with the Surveillance Frequency Control Program SR 3.4. 13.2 --------------------------------N OTE ----------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

NOTE------

Only required to be performed during steady state operation Verify primary to secondary LEAKAGE is ~ 150 gallons per day through anyone SG.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.13-2 Amendment Nos. 263, 259

RCS PIV Leakage 3.4.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE S R 3.4. 14.1

N OT ES------------------------------

1.

Not required to be performed in MODES 3 and 4.

2.

Not required to be performed on the RCS PIVs located in the RHR 'flow path when in the shutdown cooling mode of operation.

3.

RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

Verify leakage from each RCS PIV is equivalent to :: 0.5 gpm per nominal inch of valve size up to a maximum of 5 gpm at an RCS pressure.::: 2215 psig and:: 2255 psig.

FREQUENCY In accordance with the Inservice Testing Program, and in accordance with the Surveillance Frequency Control Program Prior to entering MODE 2 whenever the unit has been in MODE 5 for 7 days or more, if leakage testing has not been performed in the previous 9 months Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following valve actuation due to automatic or manual action or flow through the valve (continued)

Catawba Units 1 and 2 3.4.14-3 Amendment Nos. 263, 259

SURVEILLANCE SR 3.4.14.2 Verify RHR system interlock prevents the valves from being opened with a simulated or actual RCS pressure signal =::. 425 psig.

RCS PIV Leakage 3.4.14 FREQUENCY I n accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.14-4 Amendment Nos. 263, 259

RCS Leakage Detection instrumentation 3.4.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of the containment atmosphere particulate radioactivity monitor.

In accordance with the Surveillance Frequency Control Program SR 3.4.15.2 Perform COT of the containment atmosphere particulate radioactivity monitor.

In accordance with the Surveillance Frequency Control Program SR 3.4.15.3 Perform CHANNEL CALIBRATION of the containment floor and equipment sump level monitors.

In accordance with the Surveillance Frequency Control Program SR 3.4.15.4 Perform CHANNEL CALIBRATION of the containment atmosphere particulate radioactivity monitor.

In accordance with the Surveillance Frequency Control Program SR 3.4.15.5 Perform CHANNEL CALIBRATION of the containment ventilation unit condensate drain tank level monitor.

In accordance with the Surveillance Frequency Control Program SR 3.4.15.6 Perform CHANNEL CALIBRATION of the incore instrument sump level alarm.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.15-4 Amendment Nos. 263, 259

RCS Specific Activity 3.4.16 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and associated Completion Time of Condition A not met.

C.1 Be in MODE 3 with Tavg < 500°F.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months OR DOSE EQUIVALENT 1-131 in the unacceptable region of Figure 3.4.16-1.

SURVEILLANCE REQUIREMENTS SURVEILLANCE UENCY SR 3.4.16.1 Verify reactor coolant gross specific activity::: 100iE I n accordance with

~Ci/gm.

the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.4.16-2 Amendment Nos. 263, 259

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.16.2 --------------------------NOTE---------------------------------

Only required to be performed in MODE 1.

Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity::: 1.0 J..lCi/gm.

In accordance with the Surveillance Frequency Control Program Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change of ;:: 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period S R 3.4. 16.3 ------------------------------NOTE------------------------------

Not required to be performed until 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for> 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Determine E from a sample taken in MODE 1 after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for;:: 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.4.16-3 Amendment Nos. 263, 259

RCS Loops - Test Exceptions 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 RCS Loops -

Test Exceptions LCO 3.4.17 The requirements of LCO 3.4.4, "RCS Loops -

MODES 1 and 2," may be suspended, with THERMAL POWER < P-7.

APPLICABILITY:

MODES 1 and 2 during startup and PHYSICS TESTS.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

THERMAL POWER A.1 Open reactor trip breakers.

Immediately

~ P-7.

SURVEILLANCE REQUIREMENTS SLI RVEILLANCE FREQUENCY SR 3.4.17.1 Verify THERMAL POWER is< P-7.

In accordance with the Surveillance Frequency Control Program SR 3.4.17.2 Perform a COT for each power range neutron flux-low and intermediate range neutron flux channel, P-10, and P-13.

Prior to initiation of startup and PHYSICS TESTS Catawba Units 1 and 2 3.4.17-1 Amendment Nos. 263, 259

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.2 Verify borated water volume in each accumulator is

~ 7630 gallons and.:: 8079 gallons.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is

~ 585 psig and.:: 678 psig.

In accordance with the Surveillance Frequency Control Program SR 3.5.1.4 Verify boron concentration in each accumulator is within the limits specified in the COLR.

In accordance with the Surveillance Frequency Control Program AND

NOTE-----

Only required to be performed for affected accumulators Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each solution volume increase of

~ 75 gallons that is not the result of addition from the refueling water storage tank (continued)

Catawba Units 1 and 2 3.5.1-2 Amendment Nos. 263, 259

SURVEILLANCE SR 3.5.1.5 Verify power is removed from *each accumulator isolation valve operator when RCS pressure is > 1000 psig.

Accumulators 3.5.1 FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.5.1-3 Amendment Nos. 263, 259

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed.

Number Position Function NI162A Open SI Cold Leg Injection NI121A Closed SI Hot Leg Injection NI152B Closed SI Hot Leg Injection NI183B Closed RHR Hot Leg Injection NI173A Open RHR Cold Leg Injection NI178B.

Open RHR Cold Leg Injection NI100B Open SI Pump Suction from RWST NI147B Open SI Pump Mini-Flow In accordance with the Surveillance Frequency Control Program SR 3.5.2.2 Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

I n accordance with the Surveillance Frequency Control Program SR 3.5.2.3 Verify ECCS piping is full of water.

In accordance with the Surveillance Frequency Control Program SR 3.5.2.4 Verify each ECCS pump's developed head at the test flow point is greater than or equal to the required developed head.

In accordance with the Inservice Testing Program (continued)

Catawba Units 1 and 2 3.5.2-2 Amendment Nos. 263, 259

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

I n accordance with the Surveillance Frequency Control Program SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each position stop is in the correct position.

Centrifugal Charging Pump Injection Throttle Valve Number NI14 NI16 NI18 NI20 Safety Injection Pump Throttle Valve Number NI164 NI166 NI168 NI170 In accordance with the Surveillance Frequency Control Program SR 3.5.2.8 Verify, by visual inspection, that the ECCS containment sump strainer assembly is not restricted by debris and shows no evidence of structural distress or abnormal corrosion.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.5.2-3 Amendment Nos. 263, 259

RWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify RWST borated water temperature is.::: 70°F and

< 100°F.

In accordance with the Surveillance Frequency Control Program SR 3.5.4.2 Verify RWST borated water volume is > 363,513 gallons.*

In accordance with the Surveillance Frequency Control Program SR 3.5.4.3 Verify RWST boron concentration is within the limits specified in the COLR.

In accordance with the Surveillance Frequency Control Program

Following implementation of the modifications associated with ECCS Water Management on the respective unit, the RWST borated water volume for this SR shall be.::: 377,537 gallons.

Catawba Units 1 and 2 3.5.4-2 Amendment Nos. 263, 259

Seal Injection Flow 3.5.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE S R 3.5.5. 1

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

Not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at

> 2215 psig and::: 2255 psig.

Verify manual seal injection throttle valves are adjusted to give a flow within limit with centrifugal charging pump operating and the charging flow control valve full open.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.5.5-2 Amendment Nos. 263, 259

Containment Air Locks 3.6.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2. 1

NOTES---------------------------

1.

An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

2.

Results shall be evaluated against acceptance criteria applicable to SR 3.6.1.1.

Perform required air lock leakage rate testing in accordance with the Containment Leakage Rate Testing Program.

In accordance with the Containment Leakage Rate Testing Program SR 3.6.2.2 Perform a pressure test on each inflatable air lock door seal and verify door seal leakage is < 15 sccm.

In accordance with the Surveillance Frequency Control Program SR 3.6.2.3 Verify only one door in the air lock can be opened at a time.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.2-5 Amendment Nos. 263, 259

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.3.1 Verify each containment purge supply and exhaust isolation valves for the lower compartment and the upper compartment, instrument room, and the Hydrogen Purge System is sealed closed, except for one purge valve in a penetration flow path while in Condition E of this LCO.

In accordance with the Surveillance Frequency Control Program SR 3.6.3.2 Verify each Containment Air Release and Addition System isolation valve is closed, except when the valves are open for pressure control, ALARA or air quality considerations for personnel entry, or for Surveillances that require the valves to be open.

In accordance with the Surveillance Frequency Control Program S R 3.6.3.3

NO TE --------------------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative controls.

Verify each containment isolation manual valve and blind flange that is located outside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.6.3-5 Amendment Nos. 263, 259

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.3.~

N()-rE-----------------------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual valve and blind flange that is located inside containment or annulus and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

Prior to entering M()DE 4 from M()DE 5 if not performed within the previous 92 days SR 3.6.3.5 Verify the isolation time of automatic power operated containment isolation valve is within limits.

In accordance with the Inservice Testing Program SR 3.6.3.6 Perform leakage rate testing for Containment Purge System, Hydrogen Purge System, and Containment Air Release and Addition System valves with resilient seals.

In accordance with the Containment Leakage Rate Testing Program SR 3.6.3.7 Verify each automatic containment isolation valve that is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.6.3-6 Amendment Nos. 263, 259

Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 Containment pressure shall be.::: -0.1 psig and:::. +0.3 psig.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Containment pressure I A.1 not within limits.

Restore containment pressure to within limits.

1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B.

Required Action and 8.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

AND B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.4.1 Verify containment pressure is within limits.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.4-1 Amendment Nos. 263, 259

Containment Air Temperature 3.6.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment upper compartment average air temperature is within limits.

In accordance with the Surveillance Frequency Control Program SR 3.6.5.2 Verify containment lower compartment average air temperature is within limits.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.5-2 Amendment Nos. 263, 259

Containment Spray System 3.6.6 3.6 CONTAINMENT SYSTEMS 3.6.6 Containment Spray System LCO 3.6.6 Two containment spray trains shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One containment spray train inoperable.

A.1 Restore containment spray train to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.

Required Action and associated Completion Time not met.

B.1 AND B.2 Be in MODE 3.

Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months I

84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.6.1 Verify each containment spray manual, power operated, and automatic* valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Following implementation of the modifications associated with ECCS Water Management on the respective unit, there will be no automatic valves in the Containment Spray System.

Catawba Units 1 and 2 3.6.6-1 Amendment Nos. 263, 259

Containment Spray System 3.6.6 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.6.2 Verify each containment spray pump's developed head at In accordance with the flow test point is greater than or equal to the required the Inservice developed head.

Testing Program SR 3.6.6.3 Verify each automatic containment spray valve in the flow In accordance with path that is not locked, sealed, or otherwise secured in the Surveillance position, actuates to the correct position on an actual or Frequency Control simulated actuation signal....

Program SR 3.6.6.4 Verify each containment spray pump starts automatically In accordance with on an actual or simulated actuation signal.'"

the Surveillance Frequency Control Program SR 3.6.6.5 Verify that each spray pump is de-energized and prevented from starting upon receipt of a terminate signal and is allowed to manually...... start upon receipt of a start permissive from the Containment Pressure Control System (CPCS).

In accordance with the Surveillance Frequency Control Program SR 3.6.6.6 Verify that each spray pump discharge valve closes or is prevented from opening upon receipt of a terminate signal and is allowed to manually"" open upon receipt of a start permissive from the Containment Pressure Control System (CPCS).

In accordance with the Surveillance Frequency Control Program SR 3.6.6.7 Verify each spray nozzle is unobstructed.

Following activities which could result in nozzle blockage

... Following implementation of the modifications associated with ECCS Water Management on the respective unit, the requirements of SR 3.6.6.3 and SR 3.6.6.4 shall no longer be applicable.

...... Following implementation of the modifications associated with ECCS Water Management on the respective unit, spray pump starting and spray pump discharge valve opening are manual functions.

Catawba Units 1 and 2 3.6.6-2 Amendment Nos. 263, 259

HSS 3.6.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.8.1 Operate each HSS train for.::: 15 minutes.

In accordance with the Surveillance Frequency Control Program SR 3.6.8.2 Verify the fan motor current is.:: 69 amps when the fan speed is.::: 3560 rpm and.:: 3600 rpm with the hydrogen skimmer fan operating and the motor operated suction valve closed.

In accordance with the Surveillance Frequency Control Program SR 3.6.8.3 Verify the motor operated suction valve opens automatically and the fans receive a start permissive signal.

In accordance with the Surveillance Frequency Control Program SR 3.6.8.4 Verify each HSS train starts on an actual or simulated actuation signal after a delay of ~ 8 minutes and < 10 minutes.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.8-2 Amendment Nos. 263, 259

HIS 3.6.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.9.1 Energize each HIS train power supply breaker and verify

34 ignitors are energized in each train.

In accordance with the Surveillance Frequency Control Program SR 3.6.9.2 Verify at least one hydrogen ignitor is OPERABLE in each containment region.

In accordance with the Surveillance Frequency Control Program SR 3.6.9.3 Energize each hydrogen ignitor and verify temperature is

1700°F.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.9-2 Amendment Nos. 263, 259

AVS 3.6.10 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.10.1 Operate each AVS train for.::: 10 continuous hours with heaters operating.

In accordance with the Surveillance Frequency Control Program SR 3.6.10.2 Perform required AVS filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.6.10.3 Verify each AVS train actuates on an actual or simulated actuation signal.

I n accordance with the Surveillance Frequency Control Program SR 3.6.10.4 Verify each AVS filter cooling bypass valve can be opened.

In accordance with the Surveillance Frequency Control Program SR 3.6.10.5 Verify each AVS train flow rate is.::: 8100 cfm and::::. 9900 cfm.

I n accordance with the Surveillance Frequency Control Program SR 3.6.10.6 Verify each AVS train produces a pressure equal to or more negative than -0.88 inch water gauge when corrected to elevation 564 feet.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.10-2 Amendment Nos. 263, 259

ARS 3.6.11 3.6 CONTAINMENT SYSTEMS 3.6.11 Air Return System (ARS)

LCO 3.6.11 Two ARS trains shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One ARS train inoperable.

A.1 Restore ARS train to OPERABLE status.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months B.

Required Action and associated Completion Time not met.

B.1 AND 8.2 Be in MODE 3.

Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.11.1 Verify each ARS fan starts on an actual or simulated actuation signal, after a delay of.:::. 8.0 minutes and

. 10.0 minutes, and operates for.:::. 15 minutes.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.6.11-1 Amendment Nos. 263, 259

ARS 3.6.11 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.11.2 Verify, with the ARS air return fan damper closed and with the bypass dampers open, each ARS fan motor current is < 59.0 amps when the fan speed is ~ 1174 rpm and.::£ 1200 rpm.

In accordance with the Surveillance Frequency Control Program SR 3.6.11.3 Verify, with the ARS fan not operating, each ARS motor operated damper opens automatically on an actual or simulated actuation signal after a delay of ~ 9 seconds and.::£ 11 seconds.

In accordance with the Surveillance Frequency Control Program SR 3.6.11.4 Verify the check damper is open with the ARS fan operating.

In accordance with the Surveillance Frequency Control Program SR 3.6.11.5 Verify the check damper is closed with the ARS fan not operating.

In accordance with the Surveillance Frequency Control Program SR 3.6.11.6 Verify that each ARS fan is de-energized or is prevented from starting upon receipt of a terminate signal from the Containment Pressure Control System (CPCS) and is allowed to start upon receipt of a start permissive from the CPCS.

I n accordance with the Surveillance Frequency Control Program SR 3.6.11.7 Verify that each ARS fan motor-operated damper is prevented from opening in the absence of a start permissive from the Containment Pressure Control System (CPCS) and is allowed to open upon receipt of a start permissive from the CPCS.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.11-2 Amendment Nos. 263, 259

Ice Bed 3.6.12 3.6 CONTAINMENT SYSTEMS 3.6.12 Ice Bed LCO 3.6.12 The ice bed shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Ice bed inoperable.

A.1 Restore ice bed to OPERABLE status.

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months B.

Required Action and 8.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

AND B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.12.1 Verify maximum ice bed temperature is.::: 27°F.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.6.12-1 Amendment Nos. 263, 259

Ice Bed 3.6.12 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6. 12.2

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

The chemical analysis may be performed on either the liquid solution or on the resulting ice.

Verify, by chemical analysis, that ice added to the ice condenser meets the boron concentration and pH requirements of SR 3.6.12.7.

Each ice addition SR 3.6.12.3 Verify, by visual inspection, accumulation of ice on structural members comprising flow channels through the ice bed is ~ 15 percent blockage of the total flow area for each safety analysis section.

In accordance with the Surveillance Frequency Control Program SR 3.6.12.4 Verify total mass of stored ice is.::: 2,132,000 Ibs by calculating the mass of stored ice, at a 95 percent confidence, in each of three Radial Zones as defined below, by selecting a random sample of.::: 30 ice baskets in each Radial Zone, and Verify:

1.

Zone A (radial rows 8, 9), has a total mass of In accordance with the Surveillance Frequency Control Program

.::: 324,000 Ibs

2. Zone B (radial rows 4, 5, 6, 7), has a total mass of

.::: 1,033,100 Ibs

3. Zone C (radial rows 1, 2, 3), has a total mass of

> 774,900 Ibs SR 3.6.12.5 Verify that the ice mass of each basket sampled in SR I n accordance with 3.6.12.4 is.::: 600 Ibs.

the Surveillance Frequency Control Program

( continued)

Catawba Units 1 and 2 3.6.12-2 Amendment Nos. 263, 259

Ice Bed 3.6.12 SURVEILLANCE REQUIREMENTS (continued)

SR 3.6.12.6 SURVEILLANCE Visually inspect, for detrimental structural wear, cracks, corrosion, or other damage, two ice baskets from each group of bays as defined below:

a.

Group 1 - bays 1 through 8;

b.

Group 2 bays 9 through 16; and

c.

Group 3 - bays 17 through 24.

FREQUENCY In accordance with the Surveillance Frequency Control Program SR 3.6.12.7 ---------------------------- NOTE ------------------------------

The requirements of this SR are satisfied if the boron concentration and pH values obtained from averaging the individual sample results are within the limits specified below.

Verify, by chemical analysis of the stored ice in at least one randomly selected ice basket from each ice condenser bay, that ice bed:

a. Boron concentration is > 1800 ppm and :s. 2330 ppm; and

b.

pH is ~ 9.0 and :s. 9.5.

In accordance with the SUrveillance Frequency Control Program Catawba Units 1 and 2 3.6.12-3 Amendment Nos. 263, 259

Ice Condenser Doors 3.6.13 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

Required Action and associated Completion Time of Condition B not met.

C.1 Restore ice condenser door to OPERABLE status and closed positions.

48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months D.

Required Action and associated Completion Time of Condition A or C not met.

D.1 AND D.2 Be in MODE 3.

Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.13.1 Verify all lower inlet doors indicate closed by the Inlet Door Position Monitoring System.

In accordance with the Surveillance Frequency Control Program SR 3.6.13.2 Verify, by visual inspection, each intermediate deck door is closed and not impaired by ice, frost, or debris.

In accordance with the Surveillance Frequency Control Program SR 3.6.13.3 Verify, by visual inspection, each top deck door:

a.

Is in place; and

b.

Has no condensation, frost, or ice formed on the door that would restrict its opening.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.6.13-2 Amendment Nos. 263, 259

Ice Condenser Doors 3.6.13 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.13.4 Verify, by visual inspection, each lower inlet door is not impaired by ice, frost, or debris.

In accordance with the Surveillance Frequency Control Program SR 3.6.13.5 Verify torque required to cause each lower inlet door to begin to open is.:: 675 in-Ib and verify free movement of the door.

In accordance with the Surveillance Frequency Control Program SR 3.6.13.6 Deleted.

SR 3.6.13.7 Verify for each intermediate deck door:

a.

No visual evidence of structural deterioration;

b.

Free movement of the vent assemblies; and

c.

Free movement of the door.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.13-3 Amendment Nos. 263, 259

Divider Barrier Integrity 3.6.14 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Required Action and 0.1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

0.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.14.1 Verify, by visual inspection, all personnel access doors and equipment hatches between upper and lower containment compartments are closed.

Prior to entering MODE 4 from MODES SR 3.6.14.2 Verify, by visual inspection, that the seals and sealing surfaces of each personnel access door and equipment hatch have:

a.

No detrimental misalignments;

b.

No cracks or defects in the sealing surfaces; and

c.

No apparent deterioration of the seal material.

Prior to final closure after each opening AND

NOTE------

Only required for seals made of resilient materials In accordance with the Surveillance Frequency Control Program SR 3.6.14.3 Verify, by visual inspection, each personnel access door or equipment hatch that has been opened for personnel transit entry is closed.

After each opening (continued)

Catawba Units 1 and 2 3.6.14-2 Amendment Nos. 263, 259

Divider Barrier Integrity 3.6.14 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.14.4 Remove two divider barrier seal test coupons and verify both test coupons' tensile strength is.:: 39.7 psi.

In accordance with the Surveillance Frequency Control Program SR 3.6.14.5 Visually inspect..:: 95% of the divider barrier seal length, and verify:

a.

Seal and seal mounting bolts are properly installed; and

b.

Seal material shows no evidence of deterioration due to holes, ruptures, chemical attack, abrasion, radiation damage, or changes in physical appearance.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.14-3 Amendment Nos.

263, 259

Containment Recirculation Drains 3.6.15 SURVEILLANCE REQUIREMENTS SURVEILLANCE REQUIREMENTS FREQUENCY SR 3.6.15.1 Verify, by visual inspection, that:

a.

Each refueling canal drain valve is locked open; and

b.

Each refueling canal drain is not obstructed by debris.

Prior to entering MODE 4 from MODE 5 after each partial or complete fill of the canal SR 3.6.15.2 Verify, by visual inspection that no debris is present in the upper compartment or refueling canal that could obstruct the refueling canal drain.

I n accordance with the Surveillance Frequency Control Program SR 3.6.15.3 Verify for each ice condenser floor drain that the:

a.

Valve opening is not impaired by ice, frost, or debris;

b.

Valve seat shows no evidence of damage;

c.

Valve opening force is ~ 66 Ib; and

d.

Drain line from the ice condenser floor to the lower compartment is unrestricted.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.15-2 Amendment Nos. 263, 259

Reactor Building 3.6.16 3.6 CONTAINMENT SYSTEMS 3.6.16 Reactor Building LCO 3.6.16 The reactor building shall be OPERABLE.

APPLICABILITY:

MODES 1,2,3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Reactor building inoperable.

A.1 Restore reactor building to OPERABLE status.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months B.

Required Action and associated Completion Time not met.

B.1 AND B.2 Be in MODE 3.

Be in MODE 5.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 36 hours SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.6.16.1 Verify the door in each access opening is closed, except when the access opening is being used for normal transit entry and exit.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.6.16-1 Amendment Nos. 263, 259

Reactor Building 3.6.16 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.6.16.2 Verify that during the annulus vacuum decay test, the vacuum decay time is > 87 seconds.

In accordance with the Surveillance Frequency Control Program SR 3.6.16.3 Verify reactor building structural integrity by performing a visual inspection of the exposed interior and exterior surfaces of the reactor building.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.6.16-2 Amendment Nos. 263, 259

SG PORVs 3.7.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify one of the nitrogen bottles on each SG PORV is pressurized ~ 2100 psig.

In accordance with the SUNeillance Frequency Control Program SR 3.7.4.2 Verify one complete cycle of each SG PORV.

In accordance with the SUNeillance Frequency Control Program SR 3.7.4.3 Verify one complete cycle of each SG PORV block valve.

In accordance with the SUNeiliance Frequency Control Program Catawba Units 1 and 2 3.7.4-2 Amendment Nos. 263, 259

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY S R 3.7.5. 1

N 0 TE --------------------------

Not applicable to automatic valves when THERMAL POWER is ~ 10% RTP.

Verify each AFW manual, power operated, and automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

In accordance with the Surveillance Frequency Control Program S R 3.7.5.2

N 0 TE --------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after ~ 600 psig in the steam generator.

Verify the developed head of each AFW pump at the flow test point is greater than or equal to the required developed head.

In accordance with the I nservice Testing Program S R 3.7.5.3

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

Not applicable in MODE 4 when steam generator is relied upon for heat removal.

Verify each AFW automatic valve that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.7.5-3 Amendment Nos. 263, 259

AFW System 3.7.5 SURVEILLANCE REQUIREMENTS (continueci<---_______----.-______

SURVEILLANCE FREQUENCY S R 3.7.5.4

NOTES--------------------------

1.

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after ~ 600 psig in the steam generator.

2.

Not applicable in MODE 4 when steam generator is relied upon for heat removal.

Verify each AFW pump starts automatically on an actual or simulated actuation signal.

SR 3.7.5.5 Verify proper alignment of the required AFW flow paths by verifying flow from the condensate storage system to each steam generator.

In accordance with the Surveillance Frequency Control Program Prior to entering MODE 2, whenever unit has been in MODE 5 or 6 for

> 30 days Catawba Units 1 and 2 3.7.5-4 Amendment Nos. 263, 259

CSS 3.7.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.6.1 Verify the CSS inventory is ~ 225,000 gal.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.6-2 Amendment Nos. 263, 259

CCW System 3.7.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1

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

Isolation of CCW flow to individual components does not render the CCW System inoperable.

Verify each CCW manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

In accordance with the Surveillance Frequency Control Program SR 3.7.7.2 Verify each CCW automatic valve in the flow path servicing safety related equipment that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.7.7.3 Verify each CCW pump starts automatically on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.7-2 Amendment Nos. 263, 259

NSWS 3.7.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY S R 3.7.8.1

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

Isolation of NSWS flow to individual components does not render the NSWS inoperable.

Verify each NSWS manual, power operated, and automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

I n accordance with the Surveillance Frequency Control Program SR 3.7.8.2

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

Not required to be met for valves that are maintained in position to support NSWS single supply header operation.

Verify each NSWS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.7.8.3 Verify each NSWS pump starts automatically on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.8-3 Amendment Nos. 263, 259

SNSWP 3.7.9 3.7 PLANT SYSTEMS 3.7.9 Standby Nuclear Service Water Pond (SNSWP)

LCO 3.7.9 The SNSWP shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A

SNSWP inoperable.

Ai Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months A2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.9.1 Verify water level of SNSWP is 2: 571 ft mean sea level.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.7.9-1 Amendment Nos. 263, 259

SNSWP 3.7.9 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE S R 3.7.9.2

N OTE --------------------------

Only required to be performed during the months of July, August, and September.

Verify average water temperature of SNSWP is.:s. 95°F at an elevation of 568 ft. in SNSWP.

SR 3.7.9.3 Verify, by visual inspection, no abnormal degradation, erosion, or excessive seepage of the SNSWP dam.

FREQUENCY In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.9-2 Amendment Nos. 263, 259

CRAVS 3.7.10 REQUIRED ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME G.

One or more CRA VS train(s) heater inoperable.

G.1 Restore CRAVS train(s}

heater to OPERABLE status.

7 days G.2 Initiate action in accordance with Specification 5.6.6.

7 days SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CRA VS train for ~ 10 continuous hours with the heaters operating.

In accordance with the Surveillance Frequency Control Program SR 3.7.10.2 Perform required CRA VS filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with VFTP SR 3.7.10.3 Verify each CRAVS train actuates on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.7.10.4 Perform required CRE unfiltered air inleakage testing in accordance with the Control Room Envelope Habitability Program.

In accordance with the Control Room Envelope Habitability Program Catawba Units 1 and 2 3.7.10-3 Amendment Nos. 263, 259

CRACWS 3.7.11 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

Two CRACWS trains inoperable in MODE 5 or 6, or during movement of recently irradiated fuel assemblies.

D.1 Suspend movement of recently irradiated fuel assemblies.

Immediately E.

Two CRACWS trains inoperable in MODE 1, 2,3, or 4.

E.1 Enter LCO 3.0.3.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.11.1 Verify the control room temperature is < gO°F.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.11-2 Amendment Nos. 263, 259

ABFVES 3.7.12 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.12.1 Operate each ABFVES train for ~ 10 continuous hours with the heaters operating.

In accordance with the Surveillance

. Frequency Control

. Program SR 3.7.12.2 Perform required ABFVES filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP SR 3.7.12.3 Verify each ABFVES train actuates on an actual or simulated actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.7.12.4 Verify one ABFVES train can maintain the ECCS pump rooms at negative pressure relative to adjacent areas.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.12-2 Amendment Nos. 263, 259

FHVES 3.7.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.13.1 Verify required FHVES train in operation.

In accordance with the Surveillance Frequency Control Program SR 3.7.13.2 Operate required FHVES train for.::: 10 continuous hours with the heaters operating.

In accordance with the Surveillance Frequency Control Program SR 3.7.13.3 Perform required FHVES filter testing in accordance with In accordance with the Ventilation Filter Testing Program (VFTP).

the VFTP SR 3.7.13.4 Verify one FHVES train can maintain a pressure

-0.25 inches water gauge with respect to atmospheric pressure during operation at a flow rate.::: 36,443 cfm.

I n accordance with the Surveillance Frequency Control Program SR 3.7.13.5 Verify each FHVES filter bypass damper can be closed.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.13-2 Amendment Nos. 263, 259

Spent Fuel Pool Water Level 3.7.14 3.7 PLANT SYSTEMS 3.7.14 Spent Fuel Pool Water Level LCO 3.7.14 The spent fuel pool water level shall be.::: 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY:

During movement of irradiated fuel assemblies in the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Spent fuel pool water level not within limit.

A.1

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

LCO 3.0.3 is not applicable.

Suspend movement of irradiated fuel assemblies in the spent fuel pool.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.14.1 Verify the spent fuel pool water level is.::: 23 ft above the top of the irradiated fuel assemblies seated in the storage racks.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.14-1 Amendment Nos. 263, 259

Spent Fuel Pool Boron Concentration 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Boron Concentration LCO 3.7.15 The spent fuel pool boron concentration shall be within the limit specified in the COLR.

APPLICABILITY:

When fuel assemblies are stored in the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Spent fuel pool boron concentration not within limit.

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

LCO 3.0.3 is not applicable.

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

A.2 Initiate action to restore spent fuel pool boron concentration to within limit.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the spent fuel pool boron concentration is within limit.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.15-1 Amendment Nos. 263, 259

Secondary Specific Activity 3.7.17 3.7 PLANT SYSTEMS 3.7.17 Secondary Specific Activity LCO 3.7.17 The specific activity of the secondary coolant shall be.:s. 0.10 I-lCi/gm DOSE EQUIVALENT 1-131.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A

Specific activity not A1 Be in MODE 3.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months within limit.

AND A2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.7.17.1 Verify the specific activity of the secondary coolant is

.:s. 0.10 IJCi/gm DOSE EQUIVALENT 1-131.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.7.17-1 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power availability for each offsite circuit.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.2

NOTES--------------------------

1.

Performance of SR 3.8.1.7 satisfies this SR.

2.

All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.

3.

A modified DG start involving idling and gradual acceleration to synchronous speed may be used for this SR as recommended by the manufacturer.

When modified start procedures are not used, the time, voltage, and frequency tolerances of SR 3.8.1.7 must be met.

Verify each DG starts from standby conditions and achieves steady state voltage.::: 3950 V and::: 4580 V, and frequency.::: 58.8 Hz and::: 61.2 Hz.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.8.1-5 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE SR 3.8.1.3

NOTES------------------------

1.

DG loadings may include gradual loading as recommended by the manufacturer.

2.

Momentary transients outside the load range do not invalidate this test.

3.

This Surveillance shall be conducted on only one DG at a time.

4.

This SR shall be preceded by and immediately follow without shutdown a successful performance of SR 3.8.1.2 or SR 3.8.1.7.

Verify each DG is synchronized and loaded and operates for.:::. 60 minutes at a load.:::. 5600 kW and.::: 5750 kW.

SR 3.8.1.4 Verify each day tank contains.:::. 470 gal of fuel oil.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.5 Check for and remove accumulated water from each day tank.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.6 Verify the fuel oil transfer system operates to transfer fuel oil from storage system to the day tank.

In accordance with the Surveillance Frequency Control Program FREQUENCY In accordance with the Surveillance Frequency Control

! Program

( continued)

Catawba Units 1 and 2 3.8.1-6 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.7

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

All DG starts may be preceded by an engine prelube period.

Verify each DG starts from standby condition and achieves in ::: 11 seconds voltage of.::: 3950 V and frequency of.::: 57 Hz and maintains steady-state voltage

.::: 3950 V and::: 4580 V, and frequency.::: 58.8 Hz and

61.2 Hz.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.8 Verify automatic and manual transfer of AC power sources from the normal offsite circuit to each alternate offsite circuit.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.8.1-7 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.9

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

If performed with the DG synchronized with offsite power, it shall be performed at a power factor ~ 0.9.

Verify each DG rejects a load greater than or equal to its associated single largest post-accident load, and:

a.

Following load rejection, the frequency is ~ 63 Hz;

b.

Within 3 seconds following load rejection, the voltage is.::: 3950 V and ~ 4580 V; and

c.

Within 3 seconds following load rejection, the frequency is.::: 58.8 Hz and ~ 61.2 Hz.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.10 Verify each DG does not trip and generator speed is maintained ~ 500 rpm during and following a load rejection of.::: 5600 kW and ~ 5750 kW.

I n accordance with the Surveillance Frequency Control Program

( continued)

Catawba Units 1 and 2 3.8.1-8 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE SR 3.8.1.11

NOTES----------------------------

1.

All DG starts may be preceded by an engine prelube period.

2.

This Surveillance shall not be performed in MODE 1, 2, 3, or4.

Verify on an actual or simulated loss of offsite power signal:

a.

De-energization of emergency buses;

b.

Load shedding from emergency buses;

c.

DG auto-starts from standby condition and:

1.

energizes the emergency bus in

.s 11 seconds,

2.

energizes auto-connected shutdown loads through automatic load sequencer,

3.

maintains steady state voltage

.::: 3950 V and.s 4580 V,

4.

maintains steady state frequency

.::: 58.8 Hz and.s. 61.2 Hz, and

5.

supplies auto-connected shutdown loads for> 5 minutes.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.8.1-9 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE SR 3.8.1.12 ----------------------------------NOTE--------------------------

All DG starts may be preceded by prelube period.

Verify on an actual or simulated Engineered Safety Feature (ESF) actuation signal each DG auto-starts from standby condition and:

a.

In ::: 11 seconds after auto-start and during tests, achieves voltage.::: 3950 V and::: 4580 V;

b.

In ::: 11 seconds after auto-start and during tests, achieves frequency.::: 58.8 Hz and::: 61.2 Hz;

c.

Operates for.::: 5 minutes; and

d.

The emergency bus remains energized from the offsite power system.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.8.1-10 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.13 Verify each DG's non-emergency automatic trips are bypassed on actual or simulated loss of voltage signal on the emergency bus concurrent with an actual or simulated ESF actuation signal.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.14 -----------------------------NOTE-------------------------

Momentary transients outside the load and power factor ranges do not invalidate this test.

Verify each DG operating at a power factor ~ 0.9 operates for.::: 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months loaded.::: 5600 kW and

~5750 kW.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.8.1-11 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE FREQUENCY SR 3.8.1.15 --------------------------------NOTES-------------------------

1.

This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated:::. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months loaded> 5600 kW and

.:::. 5750 kW or until operating temperature is stabilized.

Momentary transients outside of load range do not invalidate this test.

2.

All DG starts may be preceded by an engine prelube period.

Verify each DG starts and achieves, in.:::. 11 seconds, voltage:::. 3950 V, and frequency:::. 57 Hz and maintains steady state voltage:::. 3950 V and.:::. 4580 V and frequency:::. 58.8 Hz and.:::. 61.2 Hz.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.16 -------------------------------NOTE------------------------------

This Surveillance shall not be performed in MODE 1, 2, 3,or4.

Verify each DG:

a.

b.

loaded with emergency loads upon a simulated restoration of offsite power; Transfers loads to offsite power source; and the Surveillance Frequency Control Progr~m

c.

Returns to standby operation.

Synchronizes with offsite power source while In accordance with

( continued)

Catawba Units 1 and 2 3.8.1-12 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.17

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

This Surveillance shall not be performed in MODE 1, 2, 3, or 4.

Verify, with a DG operating in test mode and connected to its bus, an actual or simulated ESF actuation signal overrides the test mode by:

a.

Returning DG to standby operation; and

b.

Automatically energizing the emergency load from offsite power.

In accordance with the Surveillance Frequency Control Program SR 3.8.1.18 Verify interval between each sequenced load block is within the design interval for each automatic load sequencer.

In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.8.1-13 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE SR 3.8.1.19 -----------------------------NOTES-----------------------------

1.

All DG starts may be preceded by an engine prelube period.

2.

This Surveillance shall not be performed in MODE 1, 2, 3, or4.

Verify on an actual or simulated loss of offsite power signal in conjunction with an actual or simulated ESF actuation signal:

a.

De-energization of emergency buses;

b.

Load shedding from emergency buses; and

c.

DG auto-starts from standby condition and:

1.

energizes the emergency bus in S 11 seconds,

2.

energizes auto-connected emergency loads through load sequencer,

3.

achieves steady state voltage> 3950 V and S 4580 V,

4.

achieves steady state frequency.::. 58.8 Hz and S 61.2 Hz, and

5.

supplies auto-connected emergency loads for.::. 5 minutes.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.8.1-14 Amendment Nos. 263, 259

AC Sources - Operating 3.8.1 SURVEILLANCE SR 3.8.1.20 ---------------------------------NOTE---------------------------

All DG starts may be preceded by an engine prelube period.

Verify when started simultaneously from standby condition, each DG achieves, in ::: 11 seconds, voltage of

~ 3950 V and frequency of ~ 57 Hz and maintains steady state voltage ~ 3950 V and::: 4580 V, and frequency

~ 58.8 Hz and::: 61.2 Hz.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.8.1-15 Amendment Nos. 263, 259

Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 ACTIONS (continued)

D.

E.

F.

CONDITION One or more DGs with new fuel oil properties not within limits.

One or more DGs with starting air receiver pressure < 210 psig and

150 psig.

Required Action and associated Completion Time not met.

OR One or more DGs diesel fuel oil, lube oil, or starting air subsystem not within limits for reasons other than Condition A, B, C, D, or D.1 E.1 F.1 REQUIRED ACTION Restore stored fuel oil properties to within limits.

Restore starting air receiver pressure to

210 psig.

Declare associated DG inoperable.

NTIME 30 days 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.3.1 Verify the fuel oil storage system contains> 77,100 gal of fuel for each DG.

FREQUENCY In accordance with the Surveillance Frequency Control Program (continued)

Catawba Units 1 and 2 3.8.3-2 Amendment Nos. 263, 259

Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.3.2 Verify lubricating oil inventory is 2: 400 gal.

In accordance with the Surveillance Frequency Control Program SR 3.8.3.3 Verify fuel oil properties of new and stored fuel oil are tested in accordance with, and maintained within the limits of, the Diesel Fuel Oil Testing Program.

In accordance with the Diesel Fuel Oil Testing Program SR 3.8.3.4 Verify each DG air start receiver pressure is 2: 210 psig.

In accordance with the Surveillance Frequency Control Program SR 3.8.3.5 Check for and remove accumulated water from each fuel oil storage tank.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.8.3-3 Amendment Nos. 263, 259

DC Sources - Operating 3.8.4 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

A and/or 0 channel of DC electrical power subsystem inoperable.

Associated train of DG DC electrical power subsystem inoperable.

0.1 Enter applicable Condition(s) and Required Action(s) of LCO 3.8.9, "Distribution Systems-Operating". for the associated train of DC electrical power distribution subsystem made inoperable.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify DC channel and DG battery terminal voltage is

~ 125 V on float charge.

In accordance with the Surveillance Frequency Control Program SR 3.8.4.2 Not used.

SR 3.8.4.3 Verify no visible corrosion at the DC channel and DG battery terminals and connectors.

Verify battery connection resistance of specific connection(s) meets Table 3.8.4-1 limit.

In accordance with the Surveillance Frequency Control Program

( continued)

Catawba Units 1 and 2 3.8.4-2 Amendment Nos. 263, 259

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.4.4 Verify DC channel and DG battery cells, cell plates, and racks show no visual indication of physical damage or abnormal deterioration that could degrade battery performance.

In accordance with the Surveillance Frequency Control Program SR 3.8.4.5 Remove visible terminal corrosion, verify DC channel and DG battery cell to cell and terminal connections are clean and tight, and are coated with anti-corrosion material.

In accordance with the Surveillance Frequency Control Program SR 3.8.4.6 Verify all DC channel and DG battery connection resistance values meet Table 3.8.4-1 limits.

In accordance with the Surveillance Frequency Control Program SR 3.8.4.7 Verify each DC channel battery charger supplies

.::: 200 amps and the DG battery charger supplies.::: 75 amps with each charger at.::: 125 V for.::: 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

In accordance with the Surveillance Frequency Control Program SR 3.8.4.8

N0 TES------------------------

1.

The modified performance discharge test in SR 3.8.4.9 may be performed in lieu of the service test in SR 3.8.4.8.

2.

This Surveillance shall not be performed for the DG batteries in MODE 1, 2, 3, or 4.

Verify DC channel and DG battery capacity is adequate In accordance with to supply, and maintain in OPERABLE status, the the Surveillance required emergency loads for the design duty cycle when Frequency Control subjected to a battery service test.

Program (continued)

Catawba Units 1 and 2 3.8.4-3 Amendment Nos. 263, 259

DC Sources - Operating 3.8.4 SURVEILLANCE SR 3.8.4.~

NOlrE---------------------------

lrhis Surveillance shall not be performed for the DG batteries in MODE 1, 2, 3, or 4.

Verify DC channel and DG battery capacity is ~ 80% of the manufacturers rating when subjected to a performance discharge test or a modified performance discharge test.

FREQUENCY In accordance with the Surveillance Frequency Control Program 18 months when battery shows degradation or has reached 85% of expected life with capacity < 100%

of manufacturer's rating AND

NOlrE-------

Not applicable to DG batteries 24 months when battery has reached 85% of the expected life with capacity ~

100% of manufacturer's rating Catawba Units 1 and 2 3.8.4-4 Amendment Nos. 263, 259

Battery Cell Parameters 3.8.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.6.1 Verify battery cell parameters of the channels of DC and DG batteries meet Table 3.8.6-1 Category A limits.

In accordance with the Surveillance Frequency Control Program SR 3.8.6.2 Not used.

SR 3.8.6.3 Verify battery cell parameters of the channels of DC and DG batteries meet Table 3.8.6-1 Category B limits.

In accordance with the Surveillance Frequency Control Program Once within 7 days after a battery discharge

< 110 V Once within 7 days after a battery overcharge

> 150V SR 3.8.6.4 Verify average electrolyte temperature for the channels of DC and DG batteries of representative cells is ~ 60°F.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.8.6-4 Amendment Nos. 263, 259

Inverters* Operating 3.8.7 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.7.1 Verify correct inverter voltage and alignment to required AC vital buses.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.8.7-2 Amendment Nos. 263, 259

Inverters - Shutdown 3.8.8 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2.3 A.2.4 Suspend operations involving positive reactivity additions that could result in loss of required SDM or required boron concentration.

Initiate action to restore required inverters to OPERABLE status.

Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.8.1 Verify correct voltage and alignment to required AC vital bus.

I n accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.8.8-2 Amendment Nos. 263, 259

Distribution Systems - Operating 3.8.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.9.1 Verify correct breaker alignments and voltage to required AC. DC channel, DC train. and AC vital bus electrical power distribution subsystems.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.8.9-3 Amendment Nos. 263, 259

Distribution Systems - Shutdown 3.8.10 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.2.4 Initiate actions to restore required AC, channels of DC, DC trains, and AC vital bus electrical power distribution subsystems to OPERABLE status.

Immediately A.2.5 Declare associated required residual heat removal subsystem(s) inoperable and not in operation.

Immediately A.2.6 Declare affected Low Temperature Overpressure Protection feature(s) inoperable.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.8.10.1 Verify correct breaker alignments and voltage to required AC, DC channel, DC train, and AC vital bus electrical power distribution subsystems.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.8.10-2 Amendment Nos. 263, 259

Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.9.1 Boron Concentration LCO 3.9.1 Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall be maintained within the limit specified in the COLR.

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

Only applicable to the refueling canal and refueling cavity when connected to the RCS.

APPLICABILITY:

MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Boron concentration not within limit.

A.1 A.2 AND A.3 Suspend CORE ALTERATIONS.

Suspend positive reactivity additions.

Initiate action to restore boron concentration to within limit.

Immediately Immediately Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit specified in COLR.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.9.1-1 Amendment Nos. 263, 259

Nuclear Instrumentation 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.2.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program S R 3.9.2.2

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

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.9.2-2 Amendment Nos. 263, 259

Containment Penetrations 3.9.3 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B.

One or more CPES train(s) heater inoperable.

B.1 Restore CPES train( s) heater to OPERABLE status.

7 days B.2 Initiate action in accordance with Specification 5.6.6.

7 days SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required containment penetration is in the required status.

In accordance with the Surveillance Frequency Control Program SR 3.9.3.2 Operate each CPES for ~ 10 continuous hours with the heaters operating.

In accordance with the Surveillance Frequency Control Program SR 3.9.3.3 Perform required CPES filter testing in accordance with the Ventilation Filter Testing Program (VFTP).

In accordance with the VFTP Catawba Units 1 and 2 3.9.3-2 Amendment Nos. 263. 259

RHR and Coolant Circulation - High Water Level 3.9.4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

(continued)

A.4 Close all containment penetrations providing direct access from containment atmosphere to outside atmosphere.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.9.4.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of ~ 1000 gpm and RCS temperature is:::. 140°F.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.9.4-2 Amendment Nos. 263, 259

RHR and Coolant Circulation - Low Water Level 3.9.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME

6.

(continued)

B.2 6.3 Initiate action to restore one RHR loop to operation.

Close all containment penetrations providing direct access from containment atmosphere to outside atmosphere.

Immediately 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 Verify one RHR loop is in operation and circulating reactor coolant at a flow rate of ~ 1000 gpm and RCS temperature is :::. 140°F.

In accordance with the Surveillance Frequency Control Program SR 3.9.5.2 Verify correct breaker alignment and indicated power available to the required RHR pump that is not in In accordance with the Surveillance operation.

I Frequency Control Program Catawba Units 1 and 2 3.9.5-2 Amendment Nos. 263, 259

Refueling Cavity Water Level 3.9.6 3.9 REFUELING OPERATIONS 3.9.6 Refueling Cavity Water Level LCO 3.9.6 Refueling cavity water level shall be maintained ~ 23 ft above the top of reactor vessel flange.

APPLICABILITY:

During CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Refueling cavity water level not within limit.

A.1 Suspend CORE ALTERATIONS.

Immediately A.2 Suspend movement of irradiated fuel assemblies within containment.

Immediately SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify refueling cavity water level is ~ 23 ft above the top In accordance with of reactor vessel flange.

the Surveillance I Frequency Control Program Catawba Units 1 and 2 3.9.6-1 Amendment Nos. 263, 259

Unborated Water Source Isolation Valves 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Unborated Water Source Isolation Valves LCO 3.9.7 Each valve used to isolate unborated water sources shall be secured in the closed position.

APPLICABILITY:

MODE 6.

ACTIONS

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

Separate Condition entry is allowed for each unborated water source isolation valve.

CONDITION REQUIRED ACTION COMPLETION TIME A.

NOTE----------

A.1 Suspend CORE Immediately Required Action A.3 ALTERATIONS.

must be completed whenever Condition A is AND entered.

A.2 Initiate actions to secure Immediately valve in closed position.

One or more valves not secured in closed AND position.

A.3 Perform SR 3.9.1.1.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.9.7.1 Verify each valve that isolates unborated water sources is secured in the closed position.

FREQUENCY In accordance with the Surveillance Frequency Control Program Catawba Units 1 and 2 3.9.7-1 Amendment Nos. 263, 259

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.16 Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Area Ventilation System (CRAVS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a.

The definition of the CRE and the CRE boundary.

b.

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

c.

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

d.

Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CRAVS, operating at a makeup flow rate of:5 4000 cfm, at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.

e.

The quantitative limits on unfiltered air inleakage into the CRE. These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c.

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

Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.

f.

The provisions of SR 3.0.2 are applicable to the Frequencies for asseSSing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

( continued)

Catawba Units 1 and 2 5.5-15 Amendment Nos. 263, 259

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.17 Surveillance Frequency Control Program This Program provides controls for Surveillance Frequencies. The program shall ensure that the Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operations are met.

a.

The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency is controlled by the program.

b.

Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c.

The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

Catawba Units 1 and 2 5.5-16 Amendment Nos. 263, 259

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 263 TO RENEWED FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO. 259 TO RENEWED FACILITY OPERATING LICENSE NPF-52 DUKE ENERGY CAROLINAS, LLC CATAWBA NUCLEAR STATION, UNITS 1 AND 2 DOCKET NOS. 50-413 AND 50-414

1.0 INTRODUCTION

By application dated March 31, 2010 (Agencywide Documents Access and Management System (ADAMS), Accession No. ML100920160), as supplemented by letter dated November 30,2010, (ADAMS Accession No. ML103370241), Duke Energy Carolinas, LLC (Duke, the licensee), requested changes to the Technical Specifications (TSs) for the Catawba Nuclear Station, Units 1 and 2 (Catawba 1 and 2). The supplement dated November 30, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published the Federal Register on November 16,2010 (75 FR 70034).

The amendments would revise the TSs by relocating specific surveillance frequencies to a licensee-controlled document using a risk-informed justification. The proposed changes would adopt the Nuclear Regulatory Commission (NRC, the Commission) staff-approved TS Task Force (TSTF) traveler TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control - RITSTF [Risk-Informed TSTF11nitiative 5b" (Reference 1). When implemented, TSTF-425 relocates most periodic frequencies of TS surveillances to a licensee-controlled program, the Surveillance Frequency Control Program (SFCP), and provides requirements for the new program in the Administrative Controls Section of the TSs. All surveillance frequencies can be relocated except:

frequencies that reference other approved programs for the specific interval (such as the Inservice Testing Program or the Primary Containment Leakage Rate Testing Program);

-2 frequencies that are purely event-driven (e.g., "each time the control rod is withdrawn to the 'full out' position");

frequencies that are event-driven, but have a time component for performing the surveillance on a one-time basis once the event occurs (e.g., "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after thermal power reaching ~ 95% RTP [Rated Thermal Power]"); and frequencies that are related to specific conditions (e.g., battery degradation, age and capacity) or conditions for the performance of a surveillance requirement (e.g., "drywell to suppression chamber differential pressure decrease").

A new program would be added to the Administrative Controls in TS Section 5.5.17. The new program is called the SFCP and describes the requirements for the program to control changes to the relocated surveillance frequencies. The proposed licensee changes to the Administrative Controls of the TSs to incorporate the SFCP include a specific reference to Nuclear Energy Institute (NEI) 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk Informed Method for Control of Surveillance Frequencies," (Reference 2). as the basis for making any changes to the surveillance frequencies once they are relocated out of the TSs.

By letter dated September 19, 2007, (Reference 3), the NRC staff approved NEI 04-10, Revision 1, as acceptable for referencing by licensees proposing to amend their TSs to establish an SFCP. This acceptance was limited as specified in NEI 04-10, Revision 1, and Reference 3.

The NRC staff issued a "Notice of Availability" for TSTF-425. Revision 3. in the Federal Register on July 6,2009 (74 FR 31996). The notice included a model Safety Evaluation (SE). In its application dated March 31, 2010, the licensee stated that "Duke Energy has concluded that the justifications presented in the TSTF-425 proposal and the safety evaluation prepared by the NRC staff is applicable to Catawba Units 1 and 2, and justify this amendment to incorporate the changes to the Catawba TS." The SE that follows is based, in large part, on the model SE for TSTF-425.

2.0 REGULATORY EVALUATION

In the "Final Policy Statement on Technical Specifications Improvements for Nuclear Power Reactors," published in the Federal Register on July 22, 1993 (58 FR 39132) the NRC staff addressed the use of Probabilistic Safety Analysis (PSA, currently referred to as Probabilistic Risk Assessment (PRA>> in determining the content of the TSs. On page 39135 of this Federal Register publication, the Commission states, in part, that:

The Commission believes that it would be inappropriate at this time to allow requirements which meet one or more of the first three criteria [in Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Section 50.36(c)(2)(ii)] to be deleted from Technical specifications based solely on PSA (Criterion 4). However. if the results of PSA indicate that Technical Specifications can be relaxed or removed, a deterministic review will be performed. * * *

- 3 The Commission Policy in this regard is consistent with its Policy Statement on "Safety Goals for the Operation of Nuclear Power Plants," 51 FR 30028, published on August 21,1986. The Policy Statement on Safety Goals states in part, "* *

probabilistic results should also be reasonably balanced and supported through use of deterministic arguments. In this way, judgments can be made" *

about the degree of confidence to be given these [probabilistic] estimates and assumptions. This is a key part of the process of determining the degree of regulatory conservatism that may be warranted for particular decisions. This defense-in-depth approach is expected to continue to ensure the protection of public health and safety." * *..

The Commission will continue to use PSA, consistent with its policy on Safety Goals, as a tool in evaluating specific line-item improvements to Technical Specifications, new requirements, and industry proposals for risk-based Technical Specification changes.

Approximately two years later, the NRC provided additional detail concerning the use of PRA in the "Use of Probabilistic Risk Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement," published in the Federal Register on August 16,1995 (60 FR 42622). On page 42627 of this FR publication, the Commission states, in part, that:

PRA addresses a broad spectrum of initiating events by assessing the event frequency.

Mitigating system reliability is then assessed, including the potential for multiple and common-cause failures. The treatment therefore goes beyond the single failure requirements in the deterministic approach. The probabilistic approach to regulation is, therefore, considered an extension and enhancement of traditional regulation by considering risk in a more coherent and complete manner.

On pages 42628 and 42629 of this Federal Register publication, the Commission provided its policy on use of PRA which states:

Although PRA methods and information have thus far been used successfully in nuclear regulatory activities, there have been concerns that PRA methods are not consistently applied throughout the agency, that sufficient agency PRAlstatistics expertise is not available, and that the Commission is not deriving full benefit from the large agency and industry investment in the developed risk assessment methods. Therefore, the Commission believes that an overall policy on the use of PRA in nuclear regulatory activities should be established so that the many potential applications of PRA can be implemented in a consistent and predictable manner that promotes regulatory stability and efficiency. This policy statement sets forth the Commission's intention to encourage the use of PRA and to expand the scope of PRA applications in all nuclear regulatory matters to the extent supported by the state-of-the-art in terms of methods and data.

Implementation of the policy statement will improve the regulatory process in three areas: Foremost, through safety decision making enhanced by the use of PRA insights; through more efficient use of agency resources; and through a reduction in unnecessary burdens on licensees.

Therefore, the Commission adopts the following policy statement regarding the expanded NRC use of PRA:

-4 (1)

The use of PRA technology should be increased in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data and in a manner that complements the NRC's deterministic approach and supports the NRC's traditional defense-in-depth philosophy.

(2)

PRA and associated analyses (e.g., sensitivity studies, uncertainty analyses, and importance measures) should be used in regulatory matters, where practical within the bounds of the state-of-the-art, to reduce unnecessary conservatism associated with current regulatory requirements, regulatory guides, license commitments, and staff practices. Where appropriate, PRA should be used to support the proposal for additional regulatory requirements in accordance with 10 CFR 50.109 (8ackfit Rule). Appropriate procedures for including PRA in the process for changing regulatory requirements should be developed and followed.

It is, of course, understood that the intent of this policy is that existing rules and regulations shall be complied with unless these rules and regulations are revised.

(3)

PRA evaluations in support of regulatory decisions should be as realistic as practicable and appropriate supporting data should be publicly available for review.

(4)

The Commission's safety goals for nuclear power plants and subsidiary numerical objectives are to be used with appropriate consideration of uncertainties in making regulatory judgments on the need for proposing and backfitting new generic requirements on nuclear power plant licensees.

The Commission's regulatory requirements related to the content of the TSs are set forth in 10 CFR 50.36, "Technical specifications." This regulation requires that the TSs include items in the following five specific categories: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) surveillance requirements; (4) design features; and (5) administrative controls. The regulation does not specify the particular requirements to be included in a plant's TSs.

As stated in 10 CFR 50.36(c)(3), "Surveillance reqUirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." To meet this requirement, the surveillance requirement must specify an adequate test, calibration, or inspection, and an appropriate frequency of performance. The licensee has proposed to implement changes to surveillance frequencies in the SFCP using the methodology in NEI 04-10, which includes qualitative considerations, results of risk analyses, sensitivity studies and any bounding analyses, recommended monitoring of structures, systems, and components (SSCs), and documentation of the evaluation. Furthermore, changes to frequencies are subject to regulatory review and oversight of the SFCP implementation through the rigorous NRC review of safety-related SSC performance provided by the reactor oversight program.

The licensee's SFCP is intended to ensure that surveillance requirements specified in the TSs are performed at intervals sufficient to assure the above regulatory requirements are met.

Existing regulatory requirements, such as 10 CFR 50.65, "Requirements for monitoring the

-5 effectiveness of maintenance at nuclear power plants," and Appendix B to 10 CFR Part 50, require licensee monitoring of surveillance test failures and implementation of corrective actions to address such failures. One of these actions may be to consider increasing the frequency at which a surveillance test is performed. In addition, the SFCP implementation guidance in NEI 04-10 requires monitoring the performance of SSCs for which surveillance frequencies are decreased to assure reduced testing does not adversely impact the SSCs. These requirements, and the monitoring required by NEI 04-10, are intended to ensure that surveillance frequencies are sufficient to assure that the requirements of 10 CFR 50.36 are satisfied and that any performance deficiencies will be identified and appropriate corrective actions taken.

Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant-Specific Changes to the Licensing Basis" (Reference 4), describes a risk-informed approach, acceptable to the NRC, for assessing the nature and impact of proposed licensing-basis changes by considering engineering issues and applying risk insights.

This RG also provides risk acceptance guidelines for evaluating the results of such evaluations.

RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications" (Reference 5), describes an acceptable risk-informed approach specifically for assessing proposed TS changes.

RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" (Reference 6), describes an acceptable approach for determining whether the quality of the PRA, in total or the parts that are used to support an application, is sufficient to provide confidence in the results, such that the PRA can be used in regulatory decision making for light-water reactors.

3.0 TECHNICAL EVALUATION

The licensee's adoption of TSTF-425 for Catawba 1 and 2 provides for administrative relocation of applicable surveillance frequencies, and provides for the addition of the SFCP to the Administrative Controls Section of the TSs. TSTF-425 also requires the application of NEI 04-10 for any changes to surveillance frequencies within the SFCP. The licensee's application for the changes proposed in TSTF-425 included documentation regarding the PRA technical adequacy consistent with the requirements of RG 1.200. In accordance with NEI 04-10, PRA methods are used, in combination with plant performance data and other considerations, to identify and justify modifications to the surveillance frequencies of equipment at nuclear power plants. This is consistent with the guidance provided in RG 1.174 and RG 1.177.

3.1 RG 1.177, Five Key Safety Principles RG 1.177 identifies five key safety principles required for risk-informed changes to the TSs.

Each of these principles is addressed by the industry methodology document, NEI 04-10, and is evaluated below in SE Sections 3.1.1 through 3.1.5 with respect to the proposed amendment.

-6 3.1.1 The Proposed Change Meets Current Regulations Paragraph (c)(3) in 10 CFR 50.36 requires that TSs will include surveillance requirements which are "requirements relating to test, calibration, or inspection to assure that necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met." The proposed amendment would relocate most periodic surveillance requirement frequencies, currently shown in the Catawba 1 and 2 TSs, to a licensee-controlled program (i.e., the SFCP). The surveillance requirements themselves would remain in the TSs, as required by 10 CFR 50.36(c)(3). The requirements for the SFCP would be added to a new subsection in TS Section 5.0. In accordance with TS Section 5.0, any changes to the surveillance requirement frequencies would be made in accordance with NEI 04-10, Revision 1. By letter dated September 19, 2007 (Reference 3), the NRC staff found that the methodology in NEI 04-10, Revision 1, met NRC regulations, specifically 10 CFR 50.36(c)(3), and was an acceptable program for controlling changes to surveillance requirement frequencies.

Based on the above considerations, the NRC staff concludes that the proposed change is consistent with the requirements in 10 CFR 50.36(c)(3). Therefore, the proposed change meets the first key safety principle of RG 1.177.

3.1.2 The Proposed Change Is Consistent With the Defense-in-Depth Philosophy Consistency with the defense-in-depth philosophy, the second key safety principle of RG 1.177, is met if:

A reasonable balance is preserved among prevention of core damage, prevention of containment failure, and consequence mitigation.

Over-reliance on programmatic activities to compensate for weaknesses in plant design is avoided.

System redundancy, independence, and diversity are preserved commensurate with the expected frequency, consequences of challenges to the system, and uncertainties (e.g.,

no risk outliers).

Defenses against potential common cause failures are preserved, and the potential for the introduction of new common cause failure mechanisms is assessed.

Independence of barriers is not degraded.

Defenses against human errors are preserved.

The intent of the General Design Criteria in 10 CFR Part 50, Appendix A, is maintained.

TSTF-425 requires the application of NEI 04-10 for any changes to surveillance requirement frequencies within the SFCP. NEI 04-10 uses both the core damage frequency (CDF) and the large early release frequency (LERF) metrics to evaluate the impact of proposed changes to surveillance frequencies. The guidance of RG 1.174 and RG 1.177 for changes to the CDF and

- 7 the LERF is achieved by evaluation using a comprehensive risk analysis, which assesses the impact of proposed changes including contributions from human errors and common cause failures. Defense-in-depth is also included in the methodology explicitly as a qualitative consideration outside of the risk analysis, as is the potential impact on detection of component degradation that could lead to an increased likelihood of common cause failures. The NRC staff concludes that both the quantitative risk analysis and the qualitative considerations assure that a reasonable balance of defense-in-depth is maintained. Therefore, the proposed change meets the second key safety principle of RG 1.177.

3.1.3 The Proposed Change Maintains Sufficient Safety Margins The engineering evaluation that will be conducted by the licensee under the SFCP, when surveillance requirement frequencies are revised, will assess the impact of the proposed frequency change in accordance with the principle that sufficient safety margins are maintained.

The guidelines used for making that assessment will include ensuring the proposed surveillance test frequency change is not in conflict with approved industry codes and standards or adversely affects any assumptions or inputs to the safety analysis, or. if such inputs are affected, justification is provided to ensure sufficient safety margin will continue to exist.

The design, operation, testing methods, and acceptance criteria for SSCs, specified in applicable codes and standards (or alternatives approved for use by the NRC) will continue to be met as described in the plant licensing basis (including the Final Safety Analysis Report and Bases to the TSs), since these are not affected by changes to the surveillance requirement frequencies. Similarly, there is no impact to safety analysis acceptance criteria as described in the plant licensing basis.

Based on the above considerations, the NRC staff concludes that there is reasonable assurance that safety margins will be maintained through the use of the SFCP methodology.

Therefore, the proposed change meets the third key safety principle of RG 1.177.

3.1.4 When Proposed Changes Result in an Increase in Core Damage Frequency or Risk, the Increases Should Be Small and Consistent With the Intent of the Commission's Safety Goal Policy Statement RG 1.177 provides a framework for risk evaluation of proposed changes to surveillance frequencies. This requires the identification of the risk contribution from impacted surveillances, determination of the risk impact from the change to the proposed surveillance frequency. and performance of sensitivity and uncertainty evaluations. TSTF-425 requires application of NEI 04-10 in the SFCP. NEI 04-10 satisfies the intent of RG 1.177 requirements for evaluating the change in risk, and for assuring that such changes are small.

3.1.4.1 Quality of the PRA The quality of the Catawba 1 and 2 PRA is compatible with the safety implications of the proposed TS change and the role the PRA plays in justifying the change. That is, the more the potential change in risk or the greater the uncertainty in that risk from the requested TS change, or both, the more rigor that must go into ensuring the quality of the PRA.

- 8 The licensee used RG 1.200 to address the Catawba 1 and 2 PRA technical adequacy.

RG 1.200 is NRC's developed regulatory guidance which, in Revision 1, endorsed with comments and qualifications the use of "ASME [American Society of Mechanical Engineers]

PRA Standard RA-Sb-2005, 'Addenda B to ASME RA-S-2002 Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications,'" (Reference 7), NEI 00-02, Revision 1, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," (Reference 8), and NEI 05-04, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard" (Reference 9). The licensee has performed an assessment of the PRA models used to support the SFCP against the requirements of RG 1.200 to assure that the PRA models are capable of determining the change in risk due to changes to surveillance requirement frequencies of SSCs, using plant-specific data and models. Capability category II of ASME RA Sb-2005 was applied as the standard, and any identified deficiencies to those requirements are assessed further to determine any impacts to proposed decreases to surveillance frequencies, including by the use of sensitivity studies where appropriate. The NRC staff notes that in RG 1.200, Revision 2, the NRC staff endorsed with comments and qualifications an updated combined standard which includes requirements for fire, seismic, and other external events PRA models. The existing internal events standard was subsumed into the combined standard, but the technical requirements are essentially unchanged. Since NEI 04-10 specifically identified the use of RG 1.200, Revision 1, to assess the internal events standard, the licensee's approach is reasonable and consistent with the approved methodology.

The NRC staff reviewed the licensee's assessment of the Catawba 1 and 2 PRA and the remaining open deficiencies that do not conform to capability category II of the ASME PRA standard (Table 2-1 of Attachment 2 of the licensee amendment request). The NRC staffs assessment of these open "gaps," to assure that they may be addressed and dispositioned for each surveillance frequency evaluation per the NEI 04-10 methodology, is provided below.

Gap #1: Accident sequence notebooks and system model notebooks should document the phenomenological conditions created by the accident sequence progression. In response to the request for additional information (RAI), the licensee stated that for each surveillance frequency change evaluation, any phenomenological conditions created by the accident sequence progression will be identified, included and documented in the analysis.

Gap #2: SSC boundaries, SSC failure modes and success criteria definitions should be established for failure rates and common cause failure parameters. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will use definitions for SSC boundary, unavailability boundary, failure mode, and success criteria consistently across the systems and data analyses.

Gap #3: Data calculations should be revised to group standby and operating component data.

Group components by service condition to the extent supported by the data. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will include sensitivity studies to consider the impact of grouping data into operating vs. standby failure rates and by service condition.

- 9 Gap #4: As part of the Bayesian update process, checks are performed to assure that the posterior distribution is reasonable given the prior distribution and plant experience. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will verify that the Bayesian update process produces a reasonable posterior distribution.

Gap #5: Comparisons should be done of the component boundaries assumed for the generic common cause failure (CCF) estimates to those assumed in the PRA to ensure that these boundaries are consistent. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will ensure that CCF probabilities are consistent with component boundaries and plant experience.

Gap #6: Human reliability analysis should consider the potential for calibration errors. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will identify and consider the impact that equipment calibration errors could have on the results and conclusions.

Gap #7: Maintenance and calibration activities that could simultaneously affect equipment in either different trains of a redundant system or diverse systems should be identified. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will identify and consider the impact that equipment calibration errors could have on the results and conclusions.

Gap #8, #12: Mean values should be developed for pre-and post-initiator human error probabilities (HEPs). In response to the RAI, the licensee stated that each surveillance frequency change evaluation will use mean values for pre-and post-initiator HEPs.

Gap #9: When estimating HEPs, the impact of plant-specific and scenario-specific performance shaping factors should be considered and documented. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will use HEP values that have been quantified with consideration of plant-specific and scenario-specific performance shaping factors.

Gap #10, #11, #13: Human reliability analysis documentation should be enhanced to include time available to complete actions, a review of Human Failure Events (HFEs) and their final HEPs relative to each other, and appropriate credit if given for operator recovery actions. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will use HEP events with time available inputs based on plant-specific thermal hydrauliC analyses; post-initiator HEPs will be reviewed against each other to check their reasonableness given the scenario context, plant procedures, operating practices and experience; and operator actions will only be credited if they are feasible.

Gap #14: The licensee identified twelve initiating event gaps to the supporting requirements for capability category II of the PRA standard. In response to the RAI, the licensee confirmed that no technical issues were identified for any of these supporting requirements but there remained a need to enhance the documentation. The licensee stated that the Catawba 1 and 2 initiating events analysis is revised with each PRA update to ensure that it remains consistent with industry operating experience as well as current plant deSign, operation and experience.

- 10 Furthermore, the licensee noted that a calculation was performed to address the initiating events supporting requirements. Each surveillance frequency change evaluation will review this calculation for potential impacts on the analysis. In addition, each surveillance frequency change will include a sensitivity analysis to determine the impact of the assumptions and sources of model uncertainty on the 5b analysis result.

Gap #15: Six internal flooding supporting requirements are not met in the Catawba 1 and 2 PRA. In response to the RAI, the licensee stated that a plan and schedule are in place for addressing internal flood issues related to the PRA standard for Catawba 1 and 2. In the interim, for each surveillance frequency change, all supporting requirements not meeting capability category II will be evaluated with sensitivity studies.

Gap #16: In crediting HFEs that support the accident progression analysis, explicitly model reactor coolant system depressurization for smaliloss-of-coolant accidents (LOCAs) and perform the dependency analysis on the HEPs. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will include a sensitivity study to assess the importance of explicitly modeling RCS depressurization for small LOCAs.

Gap #17, #20, #23, #25, #29: Collectively, these gaps identify deficiencies in the documentation process that do not directly affect the technical adequacy of the PRA model.

Gap #18, #19: Enhancement to the uncertainty analysis by use of a documented, systematic process to identify significant assumptions is recommended. In response to the RAI, the licensee stated that use of this application will include a sensitivity analysis for these gaps per NEI 04-10 if applicable to the specific surveillance test interval evaluation.

Gap #21: Documentation should include thermal hydraulic bases for all safety function success criteria for all initiating events. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will ensure that the success criteria address all initiators.

Gap #22: The acceptability of the results should be shown for the thermal hydraulic, structural, or other supporting engineering bases used to support the success criteria. In response to the RAI, the licensee stated that each surveillance frequency change evaluation will check and ensure the reasonableness and acceptability of the thermal hydraulic analyses result used to support the success criteria.

Gap #24, #27: System documentation should be enhanced to include an up-to-date system walkdown checklist and system engineer review for each system. In response to the RAI, the licensee stated that until each system notebook is updated, the impact of these gaps will be evaluated for each surveillance frequency change.

Gap #26: Quantitative evaluations should be provided for screening criteria associated with system unavailability and unreliability. In response to the RAI, the licensee stated that for each surveillance frequency change, the component and failure mode screening performed in the system analysis will be verified to meet the qlJantitative requirements provided in SY-A14.

Gap #28: A consideration of potential SSC failures due to adverse environmental conditions should be identified and documented. In response to the RAI, the licensee stated that for each

- 11 surveillance frequency change, potential SSC failures due to adverse environmental conditions will be identified, included and documented in the analysis.

Based on the licensee's assessment using the applicable PRA standard and RG 1.200, the level of PRA quality, combined with the proposed evaluation and disposition of gaps, is sufficient to support the evaluation of changes proposed to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1 of RG 1.177.

3.1.4.2 Scope of the PRA The licensee is required to evaluate each proposed change to a relocated surveillance frequency using the guidance contained in NEI 04-10 to determine its potential impact on risk, due to impacts from internal events, fires, seismic, other external events, and from shutdown conditions. Consideration is made of both CDF and LERF metrics. In cases where a PRA of sufficient scope or where quantitative risk models were unavailable, the licensee uses bounding analyses, or other conservative quantitative evaluations. A qualitative screening analysis may be used when the surveillance frequency impact on plant risk is shown to be negligible or zero.

Catawba 1 and 2's PRA includes a plant-specific seismic analysis and fire model. The current Catawba 1 and 2 seismic PRA model of record utilizes Seismic Margins Methodology and was recently updated as part of a revision. The fire PRA model is integrated into the overall PRA model, therefore; quantitative fire risk insights can be obtained. Both seismic and fire models use the same analysis and methodology as described in the original Individual Plant Examination for External Events (IPEEE). Furthermore, the licensee is planning to perform a self-assessment against the supporting requirements for both fire and seismic events of ASME/ANS PRA standard RA-Sa-2009, Addendum A to RA-S-2008 for Catawba 1 and 2 fire and seismic PRA. The licensee states that any deviations from ASME Standard Capability Category II requirements for each application of initiative 5b will be addressed.

The Catawba 1 and 2 PRA does not include an approved quantitative shutdown PRA model; therefore the licensee states that it will either 1) utilize the plant shutdown safety assessment tool developed to support implementation of NUMARC 91-06, or 2) perform an alternate qualitative risk evaluation process to assess the proposed surveillance frequency change.

The licensee's evaluation methodology is sufficient to ensure the scope of the risk contribution of each surveillance frequency change is properly identified for evaluation, and is consistent with Regulatory Position 2.3.2 of RG 1.177.

3.1.4.3 PRA Modeling The licensee will determine whether the SSCs affected by a proposed change to a surveillance frequency are modeled in the PRA. Where the SSC is directly or implicitly modeled, a quantitative evaluation of the risk impact may be carried out. The methodology adjusts the failure probability of the impacted SSCs, including any impacted common cause failure modes, based on the proposed change to the surveillance frequency. Where the SSC is not modeled in the PRA, bounding analyses are performed to characterize the impact of the proposed change to the surveillance frequency. Potential impacts on the risk analyses due to screening criteria

- 12 and truncation levels are addressed by the requirements for PRA technical adequacy consistent with guidance contained in RG 1.200, and by sensitivity studies identified in NEI 04-10.

The licensee will perform quantitative evaluations of the impact of selected testing strategy (Le.,

staggered testing or sequential testing) consistent with the guidance of NUREG/CR-6141 and NUREG/CR-5497, as discussed in NEI 04-10.

Thus, through the application of NEI 04-10 the Catawba 1 and 2 PRA modeling is sufficient to ensure an acceptable evaluation of risk for the proposed changes in surveillance frequency, and is consistent with Regulatory Position 2.3.3 of RG 1.177.

3.1.4.4 Assumptions for Time-Related Failure Contributions The failure probabilities of SSCs modeled in the Catawba 1 and 2 PRA include a standby time related contribution and a cyclic demand-related contribution. NEI 04-10 criteria adjust the time related failure contribution of SSCs affected by the proposed change to surveillance frequency.

This is consistent with RG 1.177, Section 2.3.3, which permits separation of the failure rate contributions into demand and standby for evaluation of surveillance requirements. If the available data do not support distinguishing between the time-related failures and demand failures, then the change to surveillance frequency is conservatively assumed to impact the total failure probability of the SSC, including both standby and demand contributions. The SSC failure rate (per unit time) is assumed to be unaffected by the change in test frequency, and will be confirmed by the required monitoring and feedback implemented after the change in surveillance frequency is implemented. The process requires consideration of qualitative sources of information with regard to potential impacts of test frequency on SSC performance, including industry and plant-specific operating experience, vendor recommendations, industry standards, and code-specified test intervals. Thus the process is not reliant upon risk analyses as the sole basis for the proposed changes.

The potential beneficial risk impacts of reduced surveillance frequency, including reduced downtime, lesser potential for restoration errors, reduction of potential for test caused transients, and reduced test-caused wear of equipment, are identified qualitatively, but are conservatively not required to be quantitatively assessed. Thus, through the application of NEI 04-10, the licensee has employed reasonable assumptions with regard to extensions of surveillance test intervals, and its approach is consistent with Regulatory Position 2.3.4 of RG 1.177.

3.1.4.5 Sensitivity and Uncertainty Analyses NEI 04-10 requires sensitivity studies to assess the impact of uncertainties from key assumptions of the PRA, uncertainty in the failure probabilities of the affected SSCs, impact to the frequency of initiating events, and of any identified deviations from capability category II of ASME PRA Standard ASME RA-Sb-2005. Where the sensitivity analyses identify a potential impact on the proposed change, revised surveillance frequencies are considered, along with any qualitative considerations that may bear on the results of such sensitivity studies. Required monitoring and feedback of SSC performance, once the revised surveillance frequencies are implemented, will also be performed. Thus, through the application of NEI 04-10, the licensee has appropriately considered the possible impact of PRA model uncertainty and sensitivity to

- 13 key assumptions and model limitations, and is consistent with Regulatory Position 2.3.5 of RG 1.177.

3.1.4.6 Acceptance Guidelines The licensee will quantitatively evaluate the change in total risk (including internal and external events contributions) in terms of CDF and LERF for both the individual risk impact of a proposed change in surveillance frequency and the cumulative impact from all individual changes to surveillance frequencies using the guidance contained in NRC-approved NEI 04-10 in accordance with the TS SFCP. Each individual change to a surveillance frequency must show a risk impact below 1 E-6 per year for a change to the CDF, and below 1 E-7 per year for a change to the LERF. These criteria are consistent with the limits of RG 1.174 for very small changes in risk. Where the RG 1.174 limits are not met, the process either considers revised surveillance frequencies which are consistent with RG 1.174 or the process terminates without permitting the proposed changes. Where quantitative results are unavailable to permit comparison to acceptance guidelines, appropriate qualitative analyses are required to demonstrate that the associated risk impact of a proposed change to surveillance frequency is negligible or zero. Otherwise, bounding quantitative analyses are required which demonstrate the risk impact is at least one order of magnitude lower than the RG 1.174 acceptance guidelines for very small changes in risk. In addition to assessing each individual SSC surveillance frequency change, the cumulative impact of all changes must result in a risk impact below 'I E-5 per year for a change to the CDF, and below 1 E-6 per year for a change to the LERF, and the total CDF and the total LERF must be reasonably shown to be less than 1 E-4 per year and 1E-5 per year, respectively. These are consistent with the limits of RG 1.174 for acceptable changes in risk, as referenced by RG 1.177 for changes to surveillance frequencies.

The NRC staff interprets this assessment of cumulative risk as a requirement to calculate the change in risk from a baseline model utilizing failure probabilities based on the surveillance frequencies prior to implementation of the SFCP, compared to a revised model with failure probabilities based on changed surveillance frequencies. The NRC staff further notes that the licensee includes a provision to exclude the contribution to cumulative risk from individual changes to surveillance frequencies associated with inSignificant risk increases (less than 5E-8 CDF and 5E-9 LERF) once the baseline PRA models are updated to include the effects of the revised surveillance frequencies.

The quantitative acceptance guidance of RG 1.174 is supplemented by qualitative information to evaluate the proposed changes to surveillance frequencies, including industry and plant-specific operating experience, vendor recommendations, industry standards, the results of sensitivity studies, and SSC performance data and test history.

The final acceptability of the proposed change is based on all of these considerations and not solely on the PRA results compared to numerical acceptance guidelines. Post implementation performance monitoring and feedback are also required to assure continued reliability of the components. The NRC staff concludes that the licensee's application of NEI 04-10 provides reasonable acceptance guidelines and methods for evaluating the risk increase of proposed changes to surveillance frequencies, consistent with Regulatory Position 2.4 of RG 1.177.

Therefore, the proposed change satisfies the fourth key safety principle of RG 1.177 by assuring that any increase in risk is small and consistent with the intent of "Use of Probabilistic Risk

- 14 Assessment Methods in Nuclear Regulatory Activities; Final Policy Statement," published in the Federal Register on August 16, 1995 (60 FR 42622).

3.1.5 The Impact of the Proposed Change Should Be Monitored Using Performance Measurement Strategies The licensee's adoption of TSTF-425 requires application of NEI 04-10 in the SFCP. NE104-10 requires performance monitoring of SSCs whose surveillance frequency has been revised as part of a feedback process to assure that the change in test frequency has not resulted in degradation of equipment performance and operational safety. The monitoring and feedback includes consideration of maintenance rule monitoring of equipment performance. In the event of degradation of SSC performance, the surveillance frequency will be reassessed in accordance with the methodology, in addition to any corrective actions which may apply as part of the maintenance rule requirements. The NRC staff concludes that the performance monitoring and feedback specified in NEI 04-10 is sufficient to reasonably assure acceptable SSC performance and is consistent with Regulatory Position 3.2 of RG 1.177. Therefore, the proposed change meets the fifth key safety principle of RG 1.177.

3.2 Addition of Surveillance Frequency Control Program to TS Section 5 The proposed amendment would add the SFCP into the Administrative Controls Section of the Catawba 1 and 2 TSs. Specifically, new TS Section 5.5.17, "Surveillance Frequency Control Program," would read as follows:

This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure that the associated Limiting Conditions for Operation are met.

a.

The Surveillance Frequency Control Program shall contain a list of Frequencies of the Surveillance Requirements for which the Frequency is controlled by the program.

b.

Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c.

The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

The NRC staff concludes that the proposed addition to the Administrative Controls Section of the TSs adequately identifies the scope of the SFCP and defines the methodology to be used in a revision of surveillance frequencies. Therefore, the proposed TS change is acceptable.

3.3 Technical Evaluation Conclusion

The NRC staff has reviewed the licensee's proposed relocation of some surveillance frequencies to a new licensee-controlled program, the SFCP, and its proposal to control

- 15 changes to surveillance frequencies in accordance with the new program. Based on the above considerations, the NRC staff concludes that the proposed amendment is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the South Carolina State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to the installation or use of facility components located within the restricted area as defined in 10 CFR Part 20 and change surveillance requirements. The NRC staff has determined that the amendments involve no significant increase in the amounts and no significant change in the types of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public comment on such finding (75 FR 70034). The amendments also relate to changes in record keeping, reporting, or administrative procedures or requirements. Accordingly, the amendments meet the eligibility criteria for categorical exclusions set forth in 10 CFR 51.22(c)(9) and (c)(10).

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

6.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

7.0 REFERENCES

1.

TSTF-425, Revision 3, "Relocate Surveillance Frequencies to Licensee Control-RITSTF Initiative 5b," March 18,2009 (ADAMS Accession No. ML090850642).

2.

NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," April 2007 (ADAMS Accession No. ML071360456).

3.

Letter, H. K. Nieh, NRC, to B. Bradley, NEI, "Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report (TR) 04-10, Revision 1, 'Risk-Informed Technical Specification Initiative 5B, Risk-Informed Method for Control of Surveillance Frequencies' (TAC No.

MD6111)," September 19,2007 (ADAMS Accession No. ML072570267).

4.

RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," NRC, July 1998 (ADAMS Accession No. ML003740133).

- 16

5.

RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking: Technical Specifications," NRC, August 1998 (ADAMS Accession No. ML003740176).

6.

RG 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," NRC, January 2007 (ADAMS Accession No. ML070240001).

7.

ASME PRA Standard ASME RA-Sb-2005, "Addenda B to ASME RA-S-2002, 'Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications,'" ASME, New York, New York, December 30,2005.

8.

NEI 00-02, Revision 1, "Probabilistic Risk Assessment (PRA) Peer Review Process Guidance," May 19, 2006 (ADAMS Accession No. ML061510621).

9.

NEI 05-04, Revision 0, "Process for Performing Follow-On PRA Peer Reviews Using the ASME PRA Standard," NEI, Washington, DC, January 2005.

Principal Contributor: J. Patel, NRR Date: March 29, 2011

March 29, 2011 Mr. J. R. Morris Site Vice President Catawba Nuclear Station Duke Energy Carolinas, LLC 4800 Concord Road York, SC 29745

SUBJECT:

CATAWBA NUCLEAR STATION, UNITS 1 AND 2, ISSUANCE OF AMENDMENTS REGARDING REVISION OF THE TECHNICAL SPECIFICATIONS TO RELOCATE SPECIFIC SURVEILLANCE FREQUENCIES TO A LICENSEE-CONTROLLED PROGRAM USING A RISK-INFORMED JUSTIFICATION (TSTF-425) (TAC NOS. ME3722 AND ME3723)

Dear Mr. Morris:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 263 to Renewed Facility Operating License NPF-35 and Amendment No. 259 to Renewed Facility Operating License NPF-52 for the Catawba Nuclear Station, Units 1 and 2, respectively. The amendments consist of changes to the Technical Specifications (TSs) in response to your application dated March 31,2010, as supplemented by letter dated November 30,2010.

The amendments revise the Technical Specifications by relocating specific surveillance frequencies to a licensee-controlled document using a risk-informed justification.

A copy of the related Safety Evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

If you have any questions, please call me at 301-415-1119.

Sincerely, IRA!

Jon Thompson, Project Manager Plant Licensing Branch 11-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. 50-413 and 50-414

Enclosures:

1. Amendment No. 263 to NPF-35

2. Amendment No. 259 to NPF-52

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

PUBLIC lPl2-1 R/F RidsAcrsAcnw_MailCTR Resource RidsNrrDirsltsb Resource RidsNrrDor1lpI2-1 Resource RidsRgn2MailCenter Resource RidsNrrDorlDpr Resource RidsNrrDraApla Resource RidsNrrPMCatawba-Resource (hard copy)

RidsOgcRp Resource J. Patel, NRR RidsNrrLAMOBrien Resource (hard copy)

DHarrision, NRR ADAMS Accession No. ML110670536 'SE transmitted by memo dated 2123/11, Ml110340130 OFFICE NRRflPl2-1/PM NRRflPl2-lILA DIRS/ITSBIBC DRAIAPLAI BC OGC NRRflPl2-1/BC NRRflPl2-lIPM NAME JThompson MO'Brien RElliott w/comments DHarrison*

DRoth NlO GKulesa JThompson DATE 03/14/11 03114111 03/15/11 02/23111 03123111 03/28/11 03128111 OFFICIAL RECORD COpy

v t e Letter Sequence Approval
TAC:ME3722, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Approved, Closed) TAC:ME3723, Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5B (Approved, Closed)
Initiation Request, Request Acceptance Supplement Administration Questions 21 October 2010 Answers Results Approval Other:
MONTHYEAR2010-05-17 [Table View]

ML110670536

Text

SUBJECT:

Dear Mr. Morris:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

3.3 Technical Evaluation Conclusion

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

7.0 REFERENCES

SUBJECT:

Dear Mr. Morris:

Enclosures:

Navigation menu

Search