Text

License Amendment, Issuance of Amendments Regarding Technical Specification Change to Allow Manual Operation of the Containment Spray System (TAC Nos. MD9752 and MD9753)
	ML092530088
Person / Time
Site:	Catawba
Issue date:	06/28/2010
From:	Jacqueline Thompson; Plant Licensing Branch II
To:	Morris J; Duke Energy Carolinas
	Thompson Jon, NRR/DORL/LPL 2-1, 415-1119
References
	TAC MD9752, TAC MD9753
	Download: ML092530088 (40)
	v • d • e

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555*0001 June 28, 2010 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 TECHNICAL SPECIFICATION CHANGES TO ALLOW MANUAL OPERATION OF THE CSS (TAC NOS. MD9752 AND MD9753)

Dear Mr. Morris:

The Nuclear Regulatory Commission has issued the enclosed Amendment No. 257 to Renewed Facility Operating License NPF-35 and Amendment No. 252 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 September 2,2008, as supplemented by letters dated June 18,2009, July 8, 2009, August 13, 2009, September 8, 2009, November 10, 2009, and March 8, 2010.

The amendments revise the technical specifications to allow manual operation of the containment spray system and to revise the upper and lower limits of the refueling water storage tank.

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.

J. Morris -2 If you have any questions, please call me at 301-415-1119.

~n\~ <<

JJ:ompson, 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. 257 to NPF-35

2. Amendment No. 252 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. 257 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 September 2, 2008, as supplemented by letters dated June 18,2009, July 8,2009, August 13, 2009, September 8,2009, November 10, 2009, and March 8,2010, 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 (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.

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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. 257 ,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 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION oria 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: June 28, 2010

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. 252 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 September 2, 2008, as supplemented by letters dated June 18, 2009, July 8, 2009, August 13, 2009, September 8, 2009, November 10, 2009, and March 8, 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 (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.

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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. 252 ,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 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION oria 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: June 28, 2010

ATTACHMENT TO LICENSE AMENDMENT NO. 257 RENEWED FACILITY OPERATING LICENSE NO. NPF-35 DOCKET NO. 50-413 AND LICENSE AMENDMENT NO. 252 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 Insert Licenses Licenses NPF-35, page 4 NPF-35, page 4 NPF-52, page 4 NPF-52, page 4 TSs TSs 1.1-4 1.1-4 1.1-5 1.1-5 1.1-6 1.1-6 3.3.2-12 3.3.2-12 3.3.2-16 3.3.2-16 3.5.4-2 3.5.4-2 3.6.6-1 3.6.6-1 3.6.6-2 3.6.6-2

-4 (2) Tech~ical SPecifications i . .

The,TBch~ical:SP.edfir.8tions coriiain~ in Appendix A, ~s n~vised through "

,4Vnendment ,No

257 which are,attached"hereto. are h8reby IncDrp(>fBted Inlo this renewed operallrig liceJ'!s8. Duke Energy Corolir'8s, LlC shfitt ope,.te'the facility 11'1 ~ordance with the Technical Specifications. " ,',

'(3) Updated Final Safety Ana!Vsi5' ReDyrl

,The Updated Finel Safety Analysis Report supplement submitted puriuent to 10 CfR 54.21(d), as revised 9n December 16, 2002,*descnbes ~rt8ln fUture actlvtties to be completed before the period of extended operation. Duke shllll

, complete theM actlYitles no later than February 24, 2026, and shall notify the NRC In writing When Implementation of these actlvlties 'Is, compfete and can be

'verified by NRC inspection.

The Updated Final Safety Analysls'Repor1 supplem~Ot a8 revISed on December 16, 2002, d~ above. shall'be Included In the* neXt scheduled' update to the Updated Final safety AnalySis ReP9rt req~irecf by 10 CFR 50.71(8)(4), following issuance 'of this renewed openiting liCenSe. Unlil that update Is co'mptet', Duke may 'm~k,e changes to the prOgrams described !nsuch supplem~nt without prior Commjssion approval, provi~ed thet Duke evaluates each such ch;lflge' pursuant to the criteria set forth In 10 CFR 50.59 and otherwise compRes with the requirements in that section.

(4) anU~tust Conditions Duke ~nergy Caroflnas. LLC sl:lall compfy with the antitrust conditions deOneated 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 shaH implement and mal~ain In effect all pro~ions

, of thtt approved fire protection program as ~8Cribed in the UI)d8ted Final S8fety Analysis Report, B$ amended; for ~he facility and 'as approved in the SER through SUpplement 5, subject to the following provision: '

The license., may make ChllliguS to the apprgved fire protoctiOn program without prior approval of the Commission' only If th9se changes would not adversely affect the ability to achieve and maintain safe shutdown In the event of a fire.

"The parenthetical ng'ation following the title of this renewed operating license conditjgn denotes the lection of tho safety EvalUEItion Report and/or its supplements wherein' this renewed license condition is discussed.

Renewed License No. NPF-35 Amendment Ng. 257

-4 (2) Technical . SPeclflesttons

.The TechnicarSpecifieations contained ilJ App.endlx*A, as ~vtsed through '. .'

Amendment .No~ 252 which are..attached*hereto, are hereby in~rBted Inlo 'I.

thi~ renewed operiitmg I~e. DUke Energy C.-oli~as, LLC sh~N operate the

. fl~Ilty I~ .ccordance with the Tectl"tical ~'pecifications,' . . . ' .'

.(3)' . UPdated Finalsafetv.A~s'Ri;port'*

,.The Updated Anal Safety Analysis RepOrt supplement submitted purlU8nt to 10 C~R 54.21(d). as revtled ~ D8cember 16, 2002"describes Certain f'uture activities to be completed before the period Qf extended ,oPeration'. Duke shall

. complete these actiVities no later than February 24, 2026, and shalt notify the NRC In writing wtJen Implementation of these activities Is. complete and C.-l be

'V8~18d by NRC insptt~lQn. '

The Updated ,Finel S~fety Analyels'.Repol1 supplem;nt as revised on "

December 16.2002, descrlQed above, shall be InclUded In the next scheduled' update to the Updated Final Safe'ty AnalySis ReP9rt requi'ed by 10 CFR 5O.71(eX4), following Issuance'of this renewed openitlng license.. Until that ,

update Is comptet** Du~e may ~k. changes to the programs described In liuch supplement wlthout prior Comm!sslon approval, proviCjed that Duke.evaluates each such change' pursuant to the criteria set for1h In 10 CFR 50,59 and otherwise complie~ with tbe requirements in that section.

(4) Antlt[,u§t Conditions Duke !=nergy Carolinas, llC st)all compty wlth the antitrust condlUon$ delineated in Appendix C to this renewed operating license.

(~) FIre protection Program (~ection 9.5.1, SER, SSER.'#2, SSER #3, ~SER #4, SS,ER #5).

D,uke Energy Carottn8s,LlC shall ImPlement end mainJain in effec1 all provisions of the approved fire protection program as d~scribed in the UpdBted FInaI.Safely AnalysiS ~epO~. as' amended; for the facility' and'as approved In the SER through Supplement 5, subject to thefoll~wtng provision: " ,

. The 'lC8ns~' may make changes to the e~proV8d' fire pr:otBC:tion PIOgram without prtor approval of the Commission only If those changes would not adversely affect the ability to achieve and maintain safe shutdown In the event of a fire,

-The parenthetical notaten following the tltl8 of this renewed operating license conditiOn denotes the sectloh of the Safety Evaluation Report andfor Its supPlements wherein' this renewed ncerise condition is discussed, Renewed lIcen5e No. NPF*5~

Amendment No, 252

Definitions 1.1 1.1 Definitions (continued)

MASTER RELAY TEST A MASTER RELAY TEST shall consist of energizing each master relay and verifying the OPERABILITY of each relay.

The MASTER RELAY TEST shall include a continuity check of each associated slave relay.

MODE A MODE shall correspond to anyone inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

NOMINAL TRIP SETPOINT The NOMINAL TRIP SETPOINT shall be the design value of a setpoint. The trip setpoint implemented in plant hardware may be less or more conservative than the NOMINAL TRIP SETPOINT by a calibration tolerance. Unless otherwise specified, if plant conditions warrant, the trip setpoint implemented in plant hardware may be set outside the NOMINAL TRIP SETPOINT calibration tolerance band as long as the trip setpoint is conservative with respect to the NOMINAL TRIP SETPOINT.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a. Described in Chapter 14 of the UFSAR;

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

(continued)

Catawba Units 1 and 2 1.1-4 Amendment Nos. 257, 2.52

Definitions 1.1 1.1 Definitions (continued)

QUADRANT POWER TILT QPTR shall be the ratio of the maximum upper excore RATIO (QPTR) detector calibrated output to the average of the upper excore detector calibrated outputs, or the ratio of the maximum lower excore detector calibrated output to the average of the lower excore detector calibrated outputs, whichever is greater.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer rate to the (RTP) reactor coolant of 3411 MWt.

REACTOR TRIP The RTS RESPONSE TIME shall be that time interval from SYSTEM (RTS) RESPONSE when the monitored parameter exceeds its RTS trip setpoint TIME at the channel sensor until/oss of stationary gripper coil voltage. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. In lieu of measurement, response time may be verified for selected components provided that the components and the methodology for verification have been previously reviewed and approved by the NRC.

SHUTDOWN MARGIN (SDM) SDM shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subcritical from its present condition assuming:

a. All rod cluster control assemblies (RCCAs) are fully inserted except for the single RCCA of highest reactivity worth, which is assumed to be fUlly withdrawn. With any RCCA not capable of being fully inserted, the reactivity worth of the RCCA must be accounted for in the determination of SDM; and

b. In MODES 1 and 2, the fuel and moderator temperatures are changed to the nominal zero power design level.

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing each slave relay and verifying the OPERABILITY of each slave relay. The SLAVE RELAY TEST shall include, as a minimum, a continuity check of associated testable actuation devices.

(continued)

Catawba Units 1 and 2 1.1-5 Amendment Nos. 257, 252

Definitions 1.1 1.1 Definitions (continued)

STAGGERED TEST BASIS A STAGGERED TEST BASIS shall consist of the testing of one of the systems, subsystems, channels, or other designated components during the interval specified by the Surveillance Frequency, so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency intervals, where n is the total number of systems, subsystems, channels, or other designated components in the associated function.

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

TRIP ACTUATING DEVICE A TADOT shall consist of operating the trip actuating device OPERATIONAL TEST and verifying the OPERABILITY of required alarm, interlock, (TADOT) 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.257,~* 252

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 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays

c. Containment 1,2,3 3 D SR 3.3.2.1 ,.; 1.4 psig 1.2 psig Pressure - High SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

d. Pressurizer 1,2,3(a) 4 D SR 3.3.2.1 ~ 1839 psig 1845 psig Pressure - Low SR 3.3.2.5 SR 3.3.2.9 SR 3.3.2.10

2. Containment Spray'

a. Manual Initiation 1,2,3,4 1 per train, B SR 3.3.2.8 NA NA 2 trains

b. Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays

c. Containment 1,2,3 4 E SR 3.3.2.1 ,.; 3.2 psig 3.0 psig Pressure - SR 3.3.2.5 High High SR 3.3.2.9 SR 3.3.2.10

3. Containment Isolation(b)

a. Phase A Isolation (1 ) Manual 1,2,3,4 2 B SR 3.3.2.8 NA NA Initiation (2) Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation SR 3.3.2.4 Logic and SR 3.3.2.6 Actuation Relays (3) Safety Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Injection (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 3.3.2-12 Amendment Nos. 257, 252

ESFAS Instrumentation 3.3.2 Table 3.3.2-1 (page 5 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

7. Automatic Switchover to Containment Sump

a. Automatic 1,2,3,4 2 trains C SR 3.3.2.2 NA NA Actuation Logic SR 3.3.2.4 and Actuation SR 3.3.2.6 Relays

b. Refueling Water 1,2,3,4 4 N SR 3.3.2.1 ~ 162.4 177.15 Storage Tank SR 3.3.2. -r<aXb) inches' inches' (RWST) Level SR 3.3.2.9(a Xb)

Low SR 3.3.2.10 Coincident with Refer to Function 1 (Safety Injection) for all initiation functions and requirements.

Safety Injection

8. ESFAS Interlocks

a. Reactor Trip, P-4 1,2,3 1 per train, F SR 3.3.2.8 NA NA 2 trains

b. Pressurizer 1,2,3 3 0 SR 3.3.2.5 ~ 1944 and 1955 psig Pressure, P-11 SR 3.3.2.9 ~ 1966 psig

c. Tavg - Low Low, 1,2,3 1 per loop 0 SR 3.3.2.5 ~ 550°F 553°F P-12 SR 3.3.2.9

9. Containment Pressure Control System

a. Start Permissive 1,2,3,4 4 per train P SR 3.3.2.1 ~ 1.0 psid 0.9 psid SR 3.3.2.7 SR 3.3.2.9

b. Termination 1,2,3,4 4 per train P SR 3.3.2.1 ~ 0.25 psid 0.35 psid SR 3.3.2.7 SR 3.3.2.9

10. Nuclear Service 1,2,3,4 3 per pit a,R SR 3.3.2.1 ~ EI. 555.4 ft EI. 557.5 ft Water Suction SR 3.3.2.9 Transfer - Low Pit SR 3.3.2.11 Level SR 3.3.2.12 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 3.3.2-16 Amendment Nos. 257, 252

RWST 3.5.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.4.1 Verify RWST borated water temperature is ~ 70°F and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

~ 100°F.

SR 3.5.4.2 Verify RWST borated water volume is ~ 363,513 7 days gallons.*

SR 3.5.4.3 Verify RWST boron concentration is within the limits 7 days specified in the COLR.

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. 257, 252

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 A.1 Restore containment spray 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months train inoperable. train to OPERABLE status.

B. Required Action and B.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. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power operated, 31 days and automatic* valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

(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. 257, 252

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 18 months path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal....

SR 3.6.6.4 Verify each containment spray pump starts automatically 18 months on an actual or simulated actuation signal....

SR 3.6.6.5 Verify that each spray pump is de-energized and 18 months 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).

SR 3.6.6.6 Verify that each spray pump discharge valve closes or is 18 months 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).

SR 3.6.6.7 Verify each spray nozzle is unobstructed. 10 years

... 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. 257, 252

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 257 TO RENEWED FACILITY OPERATING LICENSE NPF-35 AND AMENDMENT NO. 252 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 September 2, 2008, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082490094), as supplemented by letters dated June 18, 2009 (ADAMS Accession No. ML091750057), July 8, 2009 (ADAMS Accession No. ML091960176), August 13, 2009 (ADAMS Accession No. ML092260586), September 8, 2009 (ADAMS Accession No. ML092590042), November 10, 2009, (ADAMS Accession No. ML093170217), and March 8, 2010 (ADAMS Accession No. ML100700687), 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 supplements dated June 18, 2009, July 8, 2009, August 13, 2009, September 8, 2009, November 10, 2009, and March 8, 2010, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the NRC staff's original proposed no significant hazards consideration determination as published the Federal Register on April 7, 2009 (74 FR 15767).

The proposed changes would revise TS Table 3.3.2-1, Functions 2b and 2c, to delete automatic actuation logic for containment spray and revise TS Table 3.3.2-1, Function 7b, to lower the allowable value and nominal trip setpoint for the refueling water storage tank (RWST) level. In addition, the proposed TS change will revise the surveillance requirement (SR) of TS 3.5.4 to raise the RWST volume requirement.

The objectives of this amendment request are to maximize the amount of water available for emergency core cooling (ECC) from the RWST, to reduce the probability of transfer to containment sump recirculation, to increase operator response time before the transfer to containment sump recirculation conditions are satisfied, and to eliminate a current Catawba 1 and 2 non-conforming item. These proposed amendments are based on the Nuclear Energy

- 2 Institute (NEI) and the Pressurized-Water Reactor Owners Group (PWROG) initiative to extend the post-LOCA (Ioss-of-coolant accident) injection phase and to delay the onset of the containment sump recirculation phase. These TS changes are being implemented to improve overall plant safety. The initiative to submit these proposed amendments was discussed in a letter from the licensee dated September 13, 2006 (ADAMS Accession No. ML062640514).

The licensee later met with the NRC staff on February 26, 2007, to discuss the scope of the amendment. Catawba 1 and 2 is serving as the lead ice condenser plant for this initiative.

To accomplish these TS changes, the licensee will make the following hardware modifications:

Deletion or disabling of the containment spray automatic actuation circuitry

Adjustment of the RWST low and low-low level alarm setpoints

Installation of a new nonsafety-related dual channel narrow range RWST level instrument loop

Installation of a new redundant nonsafety-related wide range RWST level annunciator alarm.

In 2004, Catawba 1 and 2 identified a condition whereby postulated breaks in either a bottom mounted instrumentation nozzle, a reactor vessel nozzle, or certain reactor vessel head locations may lead to the accumulation of water in the incore instrumentation room. The inventory of water lost from the containment sump could potentially result in a containment sump level indicating less than required at the low-level RWST setpoint. This could result in a condition where inadequate sump inventory exists to support the operation of the residual heat removal (RHR) pumps. As a result of this concern, Duke evaluated a spectrum of small break LOCA mass and energy releases. The evaluation indicated that there are certain small break LOCA scenarios for which alternate emergency procedure coping strategies had to be developed that are different than what is described in the "Updated Final Safety Analysis Report" (UFSAR). This amendment request will resolve this operable but nonconforming condition. In addition, Duke's submittal seeks to support their resolution of Generic Safety Issue (GSI) 191, consistent with the intent of NRC Bulletin 2003-01, and is responsive to issues identified by Generic Letter (GL) 2004-02. It also reduces diesel generator automatic loading conditions.

2.0 REGULATORY EVALUATION

The Commission's regulatory requirements related to the contents of the TSs are set forth in Title 10 of the Code of Federal Regulations (10 CFR), Part 50, and its various sections are of particular importance in the evaluation of the subject amendment.

The emergency core cooling system (ECCS) must be designed so that its calculated cooling performance following postulated LOCAs conforms to 10 CFR 50.46(b)(5) requirements.

The regulation at 10 CFR 50.36 assures the TS limiting conditions for operations are consistent with assumed values of the initial conditions in the licensee's safety analyses. In accordance

-3 with 10 CFR 50.36, the NRC staff and the Nuclear Steam Supply System Owner's Group developed improved standard technical specifications (ISTSs) that meet the criteria set out in 10 CFR 50.36(c)(2).

The regulation at 10 CFR 50.49, "Environmental qualification of electric equipment important to safety for nuclear power plants," requires that the safety-related electrical equipment which is relied upon to remain functional during and following design-basis events be qualified for accident (harsh) environment. This provides assurance that the equipment needed in the event of an accident will perform its intended function.

The regulation at 10 CFR 50.65, "Requirements for monitoring the effectiveness of maintenance at nuclear power plants," requires that preventative maintenance activities must not reduce the overall availability of the systems, structures, or components.

The regulation at 10 CFR 50.67, "Accident source term," applies to all licensees who seek to revise the current accident source term used in their design-basis radiological analyses. This Safety Evaluation (SE) describes the impact of the proposed changes on analyzed design-basis accident (DBA) radiological consequences. By letter dated September 30,2005 (ADAMS Accession No. ML052730312), Catawba 1 and 2 implemented a full-scope alternative source term (AST) in accordance with 10 CFR 50.67 and the guidance of Regulatory Guide (RG) 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors." Therefore, this SE has been conducted to verify that the results of the licensee's affected DBA radiological dose consequence analyses continue to meet the dose acceptance criteria given in 10 CFR 50.67. The applicable acceptance criteria are 5 rem total effective dose equivalent (TEDE) in the control room (CR), 25 rem TEDE at the exclusion area boundary (EAB), and 25 rem TEDE at the outer boundary of the low population zone (LPZ).

The dose acceptance criterion in the Technical Support Center (TSC) is accepted to be 5 rem TEDE for the duration of the accident to show compliance with the regulatory requirements of NUREG-0737, "Clarification of TMI [Three Mile Island] Action Plan Requirements," and Paragraph IV.E.8 of Appendix E to 10 CFR Part 50.

The regulation at 10 CFR 50.120, "Training and qualification of nuclear power plant personnel,"

provides requirements for training and qualification of personnel, including human factors considerations.

The regulation at 10 CFR Part 50, Appendix A, "General Design Criteria for Nuclear Power Plants," establishes minimum requirements for the principal design criteria for water-cooled nuclear power plants similar in design and location to plants for which construction permits have been issued by the NRC. The General Design Criteria (GDC) are also considered to be generally applicable to other types of nuclear power units and are intended to provide guidance in establishing the principal design criteria for such other units. Chapter 3, Section 3.1, of the Catawba 1 and 2 UFSAR describes the GDC applicable to Catawba 1 and 2, including:

GDC 13, "Instrumentation and control,"

GDC 16, "Containment design,"

-4 GOC 16 relates to the containment and associated systems establishing a leak-tight barrier against the uncontrolled release of radioactivity to the environment and assuring that the containment design conditions important to safety are not exceeded for as long as the postulated accident conditions require: The proposed change of the containment spray system (CSS) from automatic start to a manual start does not impact the overall effectiveness of the containment in serving as a barrier to the fission product release following an accident. The technical evaluation of the containment analysis performed in support of this change shows that acceptable containment performance is maintained during post-accident conditions.

GOC 17, "Electric power systems,"

GOC 17 requires, in part, that "An onsite electric power system and an offsite electric power system shall be provided to permit functioning of structures, systems, and components important to safety..." and "The onsite electric power supplies, including the batteries, and the onsite electric distribution system, shall have sufficient independence, redundancy, and testability to perform their safety functions assuming a single failure." It also states that "Electric power from the transmission network to the onsite electric distribution system shall be supplied by two physically independent circuits (not necessarily on separate rights of way) designed and located so as to minimize to the extent practical the likelihood of their simultaneous failure under operating and postulated accident and environmental conditions ..." and "Provisions shall be included to minimize the probability of losing electric power from any of the remaining supplies as a result of, or coincident with, the loss of power generated by the nuclear power unit, the loss of power from the transmission network, or the loss of power from the onsite electric power supplies."

GOC 18, "Inspection and testing of electric power systems,"

GOC 18 requires that "Electric power systems important to safety shall be designed to permit appropriate periodic inspection and testing of important areas and features ..."

GOC 19, "Control Room,"

GOC 20, "Protection System Functions,"

GOC 21, "Protection System Reliability and Testability,"

GOC 35, "Emergency Core Cooling,"

GOC 38, "Containment heat removal,"

The proposed change of the CSS from automatic start to a manual start does not impact the safety function of the containment heat removal system to reduce the containment pressure and temperature following a LOCA and maintain them at acceptably low levels.

The technical evaluation of the supporting analyses demonstrate that automatic start

-5 capability of this system is not required. With the functioning and support of the ice condenser system, the CSS will continue to be able to perform its design function in response to accident conditions.

Note that even though the CSS would not start automatically for the LOCA event, it remains available to the operator and can be manually initiated. Also, the auxiliary containment spray from the RHR pumps will be retained for use as a contingency for beyond design-basis events, such as loss of both trains of normal containment spray. In addition, the ice condenser serves primarily as a large heat sink to readily reduce the containment temperature and pressure to condense the steam. Therefore, the NRC staff considers that GOC 38 remains satisfied for the LOCA.

GOC 39, "Inspection of containment heat removal system,"

GOC 40, "Testing of Containment Heat Removal Systems,"

GOC 41, "Containment Atmosphere Cleanup,"

Note that even though the CSS would not start automatically for the LOCA event, it remains available to the operator and can be manually initiated to control fission products released into the reactor containment. Also, the auxiliary containment spray from the RHR pumps will be retained for use as a contingency for beyond design-basis events, such as loss of both trains of normal containment spray. In addition, the design basis of the ice condenser as an iodine removal system is to use the chemical and physical properties of ice to reduce the fission product iodine concentration in the post LOCA containment atmosphere. Therefore, the NRC staff considers that GOC 41 remains satisfied for containment atmosphere cleanup.

GOC 42, "Inspection of Containment Atmosphere Cleanup Systems,"

. GOC 43, "Testing of Containment Atmosphere Cleanup Systems,"

Except where the licensee proposes a suitable alternative, the I\IRC staff used the regulatory guidance provided in applicable sections of RG 1.183 and NUREG-0800, "Standard Review Plan" (SRP), Chapter 15, for design-basis accidents, and SRP Chapter 6.4, for control room habitability, in performing this review. Also, the NRC staff referenced the Catawba 1 and 2 Technical Specifications (TSs) and UFSAR for verification of design-basis input, as the licensee relied on these documents for input to their analyses.

GOC 50, "Containment design basis,"

The proposed change of the CSS from automatic start to a manual start does not impact the safety function of the containment heat removal system designed so that the containment structure and its internal components can accommodate, without exceeding, the design leakage rate and with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA. The technical evaluation of the supporting analyses shows that containment performance will remain acceptable

-6 following the design-basis LOCA because the existing limits regarding post-accident containment pressure and temperature will not be exceeded. In addition, the containment design leakage rate as specified in the TSs is not exceeded and has sufficient margin. The containment will continue to be inspected and tested as specified in 10 CFR Part 50, Appendix J, TS requirements and the applicable portions of the American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code (Code).

The regulation at 10 CFR Part 50, Appendix E, "Emergency Planning and Preparedness for Production and Utilization Facilities," was considered in the review.

The regulation at 10 CFR Part 55, "Operators Licenses," was considered in the review.

Other regulatory documents considered in this review included:

NUREG-0800, "Standard Review Plan,"

NUREG-0711, "Human Factors Engineering Program Review Model," Revision 2; NUREG-1764, "Guidance for the Review of Changes to Human Actions;"

GSI 191, "Assessment of Debris Accumulation on PWR Sump Performance,"

RG 1.105, Revision 3, "Setpoints for Safety-Related Instrumentation," issued December 1999, describes a method acceptable to the NRC staff for complying with the U.S.

Nuclear Regulatory Commission (NRC) regulations for ensuring that setpoints for safety related instrumentation are initially within, and remain within the TS limits.

GL 2004-02, "Potential Impact of Debris Blockage on Emergency Recirculation during Design Basis Accidents at Pressurized-Water Reactors,"

Regulatory Issue Summary (RIS) 2006-17, "NRC Staff Position on the Requirements of 10 CFR 50.36, 'Technical Specifications,' Regarding Limiting Safety System Settings during Periodic Testing and Calibration of Instrument Channels," dated August 24,2006, described an acceptable method to meet the requirements of (10 CFR) 50.36."

Information Notice (IN) 97-78, "Crediting of Operator Actions in Place of Automatic Actions and Modifications of Operator Actions, Including Response Times,"

GL 82-33, "Supplement 1 to NUREG-0737 - Requirements for Emergency Response Capability"

3.0 TECHNICAL EVALUATION

3.1 Description of proposed change The RWST supplies borated water to the chemical and volume control system (CVCS) during abnormal operating conditions, to the refueling pool during refueling and make-up operations,

-7 and to the ECCS during accident conditions. The RWST supplies both trains of the ECCS through separate supply headers during the injection phase of a LOCA recovery. A motor operated isolation valve is provided in each header to isolate the RWST once the system has been transferred to the recirculation mode. The recirculation mode is entered when pump suction is transferred to the containment sump following receipt of the RWST - Low Level signal.

Currently, TS 3.3.2 and TS 3.6.6 require automatic CSS operation to reduce containment pressure and temperature following a high energy line break inside containment following a containment pressure high-high signal. Following the actuation signal, both trains of containment spray pumps start to transfer water from the RWST to the upper containment spray nozzles to reduce (1) the containment pressure and temperature and (2) the radioactive fission product airborne concentration. The licensee proposed to revise TS 3.3.2 and TS 3.6.6 to remove the automatic start signal for the containment spray system. The CSS will continue to be operated manually after the RHR pumps have completed the injection phase of the accident.

After RHR suction is aligned to the containment sump, one containment spray pump may be manually started after adequate sump level is verified and if containment pressure remains greater than three pounds per square inch gauge (psig).

The licensee performed a reanalysis to determine the containment response, crediting the ice condenser safety systems and the AST methodology. The results of the reanalysis concluded that automatic containment spray operation is not required during the injection phase of accident mitigation and that it can be manually initiated later in the event once the ECCS has been realigned to the recirculation mode of operation. The ice condenser system will continue to be able to function in response to accident conditions. It is a passive system which does not rely on the availability of electric power in order to perform its function.

The RWST low-level setpoint is established such that adequate time is provided for completion of all required manual actions to complete the switchover to cold leg recirculation prior to the loss of usable RWST volume and loss of suction to the ECCS pumps. The manual actions are performed in two parts, ensuring the RHR pumps have auto-swapped to the sump, and transferring the intermediate and high-head safety injection pumps to the RHR pump discharge.

The RWST low-low level alarm setpoint is defined such that the second part of the sequence is completed before the RWST is depleted.

The new wide range RWST level annunciator alarm is an RWST pre-low-level alarm that originates from the existing wide range level instrumentation. This new alarm will alert the operator that RWST level is approaching the RWST low-level alarm setpoint and provides time to prepare for the transfer from cold leg injection to cold leg recirculation and to verify adequate containment sump level. This alarm is added as an enhancement and is not required for operability or accident mitigation.

The proposed TS change to remove the automatic start signal for the CSS will maximize the amount of water available for ECCS from the RWST and it will also increase the operator response time. In addition, the licensee proposed to lower the TS 3.3.2 allowable value and nominal trip setpoint for the RWST Level - Low function. This change is based on the reduced tank depletion rate following the removal of the automatic CSS operation and changes in vortexing allowance based on testing and analytical refinements. Removal of the automatic

- 8 start of the containment spray pumps reduces the outflow from the RWST during switchover.

This reduction in the RWST draindown rate permits a longer response time to transfer the ECCS pump suction to the containment sump, and it is also decreases the RWST level needed to preclude air entrainment due to vortexing. Therefore, the RWST low-level setpoint will be reduced to 95 inches with an allowable value of 91.9 inches.

The licensee also proposed to raise the TS 3.5.4 RWST minimum volume limit. This change is based on a plant modification to install two new redundant narrow-range level instruments. The new level intruments will reduce the amount of uncertainty currently applied to the level measurement. These narrow-range instruments are only used for surveillances and do not actuate any automatic signal or operator actions for accident mitigation. The current wide range RWST level instrument loops are used for both the RWST low and RWST low-low level alarm setpoints. The new narrow range level RWST instrument loops are used to monitor the pre accident initial conditions of the RWST volume per the TSs. The TS minimum volume of the RWST is 363,513 gallons. By addition of redundant narrow range level instrumentation with significant reduction in uncertainty level measurement, the minimum RWST volume will increase to 377,537 gallons.

The licensee performed an evaluation of UFSAR, Chapter 15, category III and IV events, to determine the impact of these TS changes and concluded that the plant response remained within the current design and licensing limits. The licensee also evaluated the impact of these proposed changes on ECCS pump net positive suction head (NPSH) analyses and concluded that there is sufficient NPSH margin available.

3.2 Instrumentation and Control The licensee has proposed this amendment to maximize the amount of water available for the ECCS from the RWST and to increase operator response time before the transfer to containment sump recirculation. For this amendment request, the licensee installed two narrow range level transmitters to increase the minimum level of the RWST volume. In addition to this measure, a combination of system alignment and procedural guidance will also preclude containment spray pump operation. Procedural guidance will allow one containment spray pump to be operated manually only when it is aligned to the containment spray sump. This will minimize water draindown from the RWST and allow a longer response time to transfer the ECCS pump suction to the containment sump, thus allowing a reduction in the RWST low and low-low level setpoints.

3.2.1 TS Table 3.3.2-1, Function 7b, Automatic Switchover to Containment Sump, RWST Level-Low The licensee revised the allowable value (AV) and normal trip setpoint (NTSP) from greater than or equal to 162.4 inches and 177.15 inches, respectively, to greater than or equal to 91.9 inches and 95 inches, respectively.

The licensee's submittal of September 2, 2008, did not provide the setpoint calculation for the proposed changes. In its letters dated June 18, 2009, and August 13, 2009, the licensee provided summary calculations for the proposed setpoint changes. For the RWST low level, the licensee established the analytical limit (AL) of 74.03 inches, and the total loop uncertainty was

- 9 calculated at 8.51 inches. Based on this, the licensee established the NTSP of 95 inches, which includes a margin of more than 12 inches. The licensee calculated the AV based on the uncertainty associated with the portion of the loop tested for the desired interval and also on the uncertainty of the loop not being tested. The more conservatively calculated value was then used as the AV. Therefore, the licensee established the AV for low-level RWST as 91.8 inches.

The licensee calculated as-found tolerance (AFT) based on the uncertainties, which include reference accuracy, drift, setting tolerance, and measurement and test equipment uncertainties.

As-left tolerance (ALT) was calculated using the reference accuracy, measurement and test equipment uncertainty, and reading resolution uncertainty. These values were calculated around the NTSP and the licensee's method used to calculate these terms meets NRC staff guidance provided in RIS 2006-17. Therefore, the licensee has demonstrated that these values meet the requirements of 10 CFR 50.36(c)(1 )(ii)(A). The calculated values of AFT and ALT were determined to be +/-2.49 inches and +/-1.77 inches, respectively. However, the licensee stated that in the plant procedure the value of ALT will be less than or equal to the calculated ALT, which is more conservative and, therefore, is acceptable to the NRC staff. The licensee also added two notes to function 7.b in TS Table 3.3.2-1 to define the operability of the instrument loop in the TSs. These notes meet the NRC staff guidance provided in RIS 2006-17 and, therefore, are acceptable to the NRC staff.

3.2.2 TS SR 3.5.4.2 The licensee revised TS SR 3.5.4.2 to increase the minimum water volume in the RWST from greater than or equal to 363,513 gallons to greater than or equal to 377,537 gallons.

The licensee added two narrow range level transmitters to RWST level monitoring to increase the accuracy of the level instrumentation. These level transmitters are nonsafety-related and have an accuracy of 3.5 inches compared to the existing accuracy of approximately 14.5 inches.

The licensee justified the use of nonsafety-related level transmitters because these instruments are used only for surveillance and do not actuate any automatic or operator actions for accident mitigation. The NRC staff was concerned, however, that these instruments do not have operability requirements. It was also not clear how the accuracy of these instruments was maintained. In its letter dated August 13, 2009, the licensee noted that since these instruments are required for the fulfillment of SR 3.5.4.2, they fall within the scope of Catawba 1 and 2 Site Directive 2.3.2, "Control of Process Instrumentation Used as Test and Measurement Equipment." This directive specifies controls that ensure that required accuracies are periodically checked and adjustments are made. Since these instruments are used for the fulfillment of the SR and their accuracies are controlled by the site directive, the NRC staff finds the licensee's response acceptable.

3.3 Containment The licensee requested changes to the Catawba 1 and 2 TSs to modify the CSS actuation logic by changing the automatic start of the CSS to a manual start during a design-basis accident. A revised containment analysis was performed in order to demonstrate containment integrity with the proposed change. The analysis evaluated peak containment pressure, maximum sump liquid temperature, and maximum containment vapor temperature for various break scenarios.

The peak containment pressure and maximum sump liquid temperature were obtained as a result of a large-break LOCA located on a cold leg pump discharge pipe. The maximum

- 10 containment vapor temperature resulted from a large steam line break (SLB). The NRC staffs technical evaluation of the various containment analyses and the licensee's evaluation supporting the change are given in the following subsections.

3.3.1 Large-Break LOCA Containment Analysis For the large-break LOCA containment analysis the licensee used the NRC-approved ice condenser containment analysis methodology described in the licensee's amendment request (LAR). The methodology uses the GOTHIC computer code for the containment analysis and the RELAP5 computer code for the mass and energy (M&E) release calculation. The licensee states that Catawba 1 is equipped with feedring steam generators (FSGs) and that Catawba 2 is equipped with Model 05 steam generators (SGs). The licensee selected Catawba 1 for analysis because the FSGs are limiting for containment response due to their higher initial primary and secondary fluid mass. The analysis evaluated changes from the current licensing basis analysis including: (a) higher initial water inventory in the RWST, (b) a lower value of RWST low-level alarm setpoint, (c) a lower value of RWST low-low alarm setpoint, and (d) deletion of the automatic initiation of containment spray system. The analysis also evaluated changes in operator actions resulting from the proposed revisions to the emergency procedures including: (a) the suction transfer of high head and intermediate head safety injection pumps to the residual heat removal (RHR) pump discharge done upon receiving the RWST low-low level alarm signal; (b) manual start of one containment spray pump after verification of adequate sump level, if containment pressure is greater than three psig, after the suction transfer of the RHR pump to the containment sump at the RWST low-level alarm setpoint. The licensee stated that the timing of this operator action depends on the single failure which for most cases would occur prior to the RWST reaching the low-low level. In case the single failure affects the valves that automatically transfer RHR suction to the sump, this operator action would be after RWST reaches low-low level; and (c) not to use auxiliary containment spray.

By letter dated June 18, 2009, the licensee described the major input assumptions listed in UFSAR pages 6.2-11 and 6.2-12 used in the current GOTHIC long-term analysis (with the exception of spray pump flow from the RWST in assumptions 9 and 13) for pump discharge pipe rupture, including any differences introduced by the proposed change. These differences included: (a) the service water temperature is decreased from 100 of to 96 of, which is more conservative than its TS value of 95 of, (b) the RWST water temperature is increased from 105 of to 110°F after considering an increase resulting from its uncertainty calculation; its TS value being 100 of, (c) the air-return fan flow rate is decreased from 40,000 cubic feet per minute (cfm) to 39,200 cfm, after incorporating a decrease in its performance attributed to variation in emergency power frequency, (d) cooling water to the RHR heat exchanger is decreased from 5,000 gallons per minute (gpm) to 4,900 gpm, after incorporating a decrease in its performance attributed to variation in emergency power frequency consistent with current UFSAR analysis. The NRC staff considers the above differences in assumptions acceptable because they are conservative in comparison to the TS values.

By letter dated June 18, 2009, the licensee stated that certain features of the proposed analysis did not represent a change from current assumptions. These included that (a) there is no change in the heat transfer coefficient or the method, (b) set-points, time delays and the signal used for actuation of air-return fans are not changed, (c) the proposed analysis does not change

- 11 the functioning of the dampers. The pressure difference between the lower and the upper compartment does not exceed 0.5 pounds per square inch differential (psid) when the analysis assumes operation of the air return fans.

By letter dated December 18, 2000 (ADAMS Accession No. ML003780723), the NRC staff accepted Revision 1 of Topical Report (TR) DPR-NE-3004-PA. By letter dated June 18, 2009, the licensee stated that a revised analysis was performed using the NRC-approved methodology documented in TR DPR-NE-3004-PA, as amended by Attachment 4 of the LAR.

The methodology documented in TR DPR-NE-3004-PA, Revision 1, uses the RELAP5 computer code for M&E release and the GOTHIC computer code for containment analysis. The NRC staff has reviewed Attachment 4 to the LAR and finds the amendments to TR DPR-NE 3004-PA, Revision 1, acceptable as documented in Section 3.3.9 of this SE.

By letter dated June 18, 2009, the licensee explained the changes in the trend at various times in the containment pressure profile contained in Figure 1 of Attachment 1 to the proposed amendment. The NRC staff reviewed the licensee's explanations and considers them acceptable.

The limiting peak containment pressure resulted from the assumption of a cold-leg pump discharge pipe break LOCA and a single failure of one engineered safety feature (ESF) train which minimized the spray flow and the heat exchangers available for containment heat removal. The analysis gave a peak containment pressure of 14.18 psig compared to the current UFSAR peak pressure of 13.65 psjg. The current UFSAR containment design pressure is 15.0 psig. The increase in peak pressure is due to the decrease in the integrated containment spray flow over the transient period.

The licensee performed a sensitivity study to evaluate the containment pressure response to a potential failure of an operating containment spray pump. In the analysis, the operating pump is assumed to fail at the time of peak pressure during the transient, which is the limiting scenario.

The result of the study showed that if the spray flow is reinitiated up to 20 minutes from the time of failure of the operating pump, the peak containment pressure will remain below the containment design pressure of 15 psig. The NRC staff considers the analysis acceptable because 20 minutes is a reasonable time for an operator to start the standby pump in order to reinitiate the containment spray flow.

The limiting sump liquid temperature resulted from the assumption of a hot-leg break LOCA and a single failure of one ESF train, which minimized the spray flow and the heat exchangers available for containment heat removal. The licensee analyzed the hot-leg break LOCA for maximum and minimum ECCS flows and found that the case with the maximum ECCS flow temperature resulted in maximum sump liquid temperature. The maximum sump liquid temperature was calculated to be 198.7 of which occurs at the time of suction transfer to the sump. This temperature is greater than the current UFSAR peak value of 190 of for approximately 18.8 minutes. The higher temperature is due to absence of containment spray flow mixing with the sump fluid until the time of suction transfer to the sump. The NRC staff considers the analysis acceptable because the licensee's analyses still showed acceptable results despite this increase in maximum sump liquid temperature.

- 12 3.3.2 Large Steam line Break Containment Analysis The current large steamline break analysis predicts that the average lower containment vapor temperature peaks within the first 30 seconds and then drops to between 240 of and 250 of after 60 seconds. Due to the timing of spray initiation, the containment sprays do not affect the lower containment temperature. In addition, the analysis did not include the actuation of the containment air return fans because they would not automatically start during the time of interest for this event. Therefore, the licensee concluded that the current steam line break analysis bounds the containment response with the proposed modification which eliminates the automatic actuation of the CSS. The NRC staff has reviewed the licensee's evaluation of the large steam line break containment analysis and finds it acceptable.

3.3.3 ECCS Pumps NPSH Evaluation The licensee stated that the available NPSH for the current ECCS pumps does not take credit for the containment pressure above the atmospheric pressure or the static head of the sump fluid above the elevation of the containment sump screen structure during a design-basis accident. Similarly, the proposed NPSH analysis also does not take credit for the same.

However, in the proposed analysis, the higher sump liquid temperature reduced the NPSH margin (Le., available NPSH minus required NPSH) for a period of 18.8 minutes during which the sump liquid temperature is higher than its current value. The licensee stated that the NPSH margin is still sufficient in this duration and that the sump water temperature would be lower than the current maximum temperature of 190 of after approximately one containment pool turnover.

By letter dated June 18, 2009, the licensee stated that by implementing the proposed changes the available NPSH margin for the RHR pump would be reduced by 5 feet of water and the containment spray pump NPSH margin would be reduced by 2.5 feet of water. After implementation of the proposed changes, the new margins will be 6.72 feet of water and 2.22 feet of water for the RHR and containment spray pumps, respectively. The licensee calculated these margins using assumptions of worst debris loading of the suction strainers with maximum flow rates and with the minimum sump water level.

By letter dated June 18, 2009, the licensee also stated that the proposed change provided a benefit because it resulted in a reduction in containment sump debris and loading on the containment sump strainers. The licensee provided qualitative reasons for the reduction in debris loading on the suction strainer which are (a) the containment spray actuation is delayed, which allows time for more debris to settle in the sump rather than being transported to the strainer; (b) the ECCS water management relies on only one containment spray train, and therefore, the reduced turbulence will allow more debris to settle and less debris to be transported to the strainer.

Based on the information provided by the licensee, the NRC staff considers the proposed changes to the NPSH evaluation to be acceptable. The NRC staff agrees with the qualitative reasons that the requested license amendment would reduce the debris loading on the suction strainer. For this reason, as well as the limited duration of the increased sump fluid temperature, the NRC staff expects that the effect of implementing the water management initiative would either be neutral or salutary with respect to the performance of the suction

- 13 strainer. The NRC is also aware of the licensee's efforts to respond to GL 2004-02 and will ensure that the resolution of these issues is acceptable in light of the NPSH margins resulting from this license amendment.

3.3.4 Minimum Containment Pressure Analysis for ECCS Evaluation The licensee stated that the proposed change to containment spray does not adversely impact the minimum containment pressure analysis for ECCS evaluation because the absence of containment spray would be expected to increase the minimum containment pressure as a function of time. The NRC staff accepts the licensee's evaluation because the analysis in Section 3.3.1 of this SE shows that the containment pressure is higher in the absence of sprays.

3.3.5 Anticipated Transient Without Scram Event During an anticipated transient without scram (ATWS) event, the pressurizer relief valve lift will cause blowdown of the pressurizer into the pressurizer relief tank (PRT) eventually causing the PRT rupture disks to break, releasing steam inside the containment. The licensee stated that the resulting mass and energy (M&E) release into the containment would not produce a limiting containment pressure response. The licensee. also stated that if the containment pressure were to increase to the high-high containment pressure setpoint, then the operation of the containment air return fans is sufficient to ensure an acceptable containment pressure response. The NRC staff agrees with the licensee's evaluation of an ATWS.

3.3.6 Ice Condenser Lower Inlet Door Evaluation The ice condenser lower inlet doors are required to open during a DBA in order to allow the flow of hot steam and air mixture into the ice condenser to remove its heat and thereby limit the containment peak pressure and temperature. The doors are designed to fully open at a one pound-per-square-foot differential (psfd) pressure. During a small break accident, the design requirement for each ice condenser door is to partially open by approximately the same amount.

In the event of a DBA, the lower inlet doors of the ice condenser open due to the pressure rise in the lower containment allowing air and steam to enter into the ice condenser. The pressure increase in the ice condenser causes the intermediate and top deck doors to open allowing the air to enter the containment upper compartment. The licensee stated that the proposed amendment has no adverse impact on the functioning of the ice condenser doors because the pressure differential which causes the doors to open is not affected by delaying the initiation of sprays. The NRC staff agrees with the licensee's evaluation.

In the event of a small break accident, the licensee states it is expected that absence of sprays will not impact the performance of the doors as designed. In the current TSs, the automatic containment spray initiation occurs at three psid (432 psfd) pressure across the doors. The NRC staff agrees with the licensee's evaluation because the differential pressure during the accident that results in partial opening of the doors is not reduced in absence of sprays.

3.3.7 Early Containment Air Return Fan Operation The licensee stated that in the current analysis for small break LOCA mitigation by early operation of the air return fan, the containment pressure does not reach the spray initiation set

- 14 point, and therefore, the analysis did not include sprays. Therefore, the proposed change does not impact the early air return fan operation during small break LOCAs. The NRC staff agrees with the licensee's evaluation.

3.3.8 Minimum Containment Sump Level Evaluation The proposed delay of containment sprays will reduce the upper containment holdup volume of liquid prior to the commencement of the sump recirculation and will, therefore, result in additional liquid inventory in the containment sump. The sump level will be relatively higher and will, therefore, be an enhancement towards accident mitigation.

3.3.9 Evaluation of Proposed Change in Topical Report DPC-NE-3004-PA, Revision 1 In Attachment 4 of the LAR, the licensee proposes to change TR DPC-NE-3004-PA, Revision 1, by adding a new Section 3.6 and renumbering the existing Section 3.6 to 3.7.

The proposed Subsection 3.6.1 of the TR describes the changes in the ECCS alignments and an overview of its operation with the proposed changes. The NRC staff considers the description acceptable because it is consistent with the proposed changes in this LAR.

The proposed Subsection 3.6.2 of the TR describes the changes in the RELAP5 computer model for M&E release analysis. The model change consists of including nitrogen in the containment atmosphere. The licensee proposes two alternate approaches for the model changes. The licensee's preferred approach is to specify nitrogen as a containment boundary condition. The alternate approach is to introduce nitrogen from an additional node located at the upstream of the break based upon a pressure differential. For both approaches, the licensee proposed to assume a conservative boundary volume temperature. The NRC staff accepts both approaches because they are realistic modeling approaches including nitrogen volume as a node or a boundary condition to the containment volume.

The proposed Subsection 3.6.3.1 of the TR explains the "reverse break flow" phenomena in which saturated steam from the containment is predicted to be drawn into the reactor coolant system by the steam condensed on sub-cooled ECCS injection. The licensee stated that it introduces "an embedded conservatism in the M&E release calculation as it artificially increases the M&E release." The licensee stated that in the proposed modeling approach described above, the nitrogen introduced in the containment will conservatively mitigate this phenomenon and reduce the magnitude of penalty caused by the reverse break flow phenomena. The NRC staff considers the explanation of reverse break flow phenomena and its conservative mitigation by the revised modeling approach acceptable.

The proposed Subsection 3.6.3.1 of the TR explained another phenomenon described as "refilling of the intact reactor coolant loop pump seals for cold leg breaks." The licensee stated that (a) for hot leg breaks, this phenomenon has been previously observed and does not affect the analytical results; (b) for cold leg breaks, this phenomenon has the potential to directly affect the peak containment pressure response. By letter dated June 18, 2009, the licensee stated that for hot leg breaks, the refilling of the intact loop seals does not alter the vapor release to the containment because the majority of the ECCS flow passes through the core absorbing decay heat and exits through the break into the containment. It does not change the break flow

- 15 characteristics and therefore does not affect the containment pressure response. For the limiting cold leg break, the licensee's analysis shows that the intact loop seal fills in about 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months and 40 minutes (Figure 1 in Attachment 1 of the LAR). The steam condensation rate due to the ECCS flow is therefore reduced due to the refilling of the intact loop seal, thereby increasing the steam flow into the containment through the break which causes a higher containment pressure response. The peak containment pressure occurs after the intact loop pump seal has filled. The NRC staff agrees with the licensee's reasoning and explanation of the phenomenon.

3.3.10 Containment Sump Water pH The NRC staff reviewed Section 3.2.7 of Attachment 1 of the LAR regarding the licensee's analysis for maintaining an alkaline sump pH for 30 days following a LOCA. According to RG 1.183, maintaining pH basic will minimize re-evolution of iodine from the suppression pool water. The proposed LAR impacts the licensee's current sump pH calculation by increasing the volume of water available from the RWST. The licensee stated that they calculated the post LOCA sump pH using the proposed upper bound for RWST volume and assuming no transport of boric acid from the RWST to the sump via containment spray. All other inputs to the calculation remained the same as the licensee's current licensing basis, which was previously approved by the NRC staff by letter dated September 30,2005, as part of an AST amendment.

The calculation showed a slightly higher initial pH and a slightly lower long-term pH when compared to the current licensing basis; however, the pH at reference temperature remained greater than 7 at all times. Based on the information provided by the licensee the NRC staff finds that the post-LOCA sump pH will remain basic following a LOCA, and the proposed changes are acceptable in regard to the post-LOCA sump pH.

Although the pH remains greater than 7 under the proposed conditions, it changes slightly from the current licensing basis. Specifically the pH is slightly higher early in the event and slightly lower in the end of the event. The licensee's chemical effects analysis will be impacted by the change in the post-LOCA pH profile. Specifically, an increase in the initial pH may increase the rate of aluminum corrosion. In addition the slightly lower final pH may decrease the solubility of aluminum, resulting in increased chemical precipitation. The increase in maximum sump pool temperature will also impact the analysis of aluminum corrosion and precipitation. Impacts from the efforts to resolve issues associated with GL 2004-02 will also be evaluated to ensure current analyses remain valid or that changes are reviewed and found acceptable.

3.3.11 Equipment Qualification Evaluation The NRC staff reviewed the application for changes to Environmental Qualification (EQ). In Section 3.2.4, "Equipment Qualification," of the application, the licensee states that "As a result of the proposed modifications, the environmental accident profiles based on a LOCA scenario for areas located inside containment and the annulus were revised." The NRC staff requested additional information regarding the environmental accident profiles (before and after the modifications) for inside containment and the annulus, as well as a discussion on any differences in the profiles. Furthermore, the NRC staff requested a discussion of how the existing EQ profiles are bounding. In a letter dated November 10, 2009, the licensee provided the "EQ Equip Evaluation Curve" in the form of Figure B-6A, "Long Term EQ Reactor BUilding Temperature Response EQ Evaluation Curve (Logarithmic Time Scale)," and Figure B-6B,

- 16 "Long Term EQ Reactor Building Temperature Response EQ Evaluation Curve (Linear Time Scale)." The EQ Equip Evaluation Curve serves as the profile for equipment qualifications and bounds the temperature responses of the lower containment, upper containment, annulus, and sump, as shown in Figure B-6A.

Figure B-6B, "Long-Term EQ Reactor Building Temperature Response (Linear Time Scale),"

indicates that the lower containment bounding temperature is greater than the EQ Equip Evaluation Curve from day one to approximately day eight or nine. The NRC staff requested additional information regarding how the EQ equipment in lower containment remains qualified and will be able to perform its safety function when exposed to a temperature greater than the qualification curve. The licensee stated in its letter dated March 8, 2010, that Figure B-6B shows the temperature curves on a linear scale and that the lower containment subcurve consists of only t~lree data points to form a bounding straight-line approximation. However, the licensee further stated that all evaluations were performed using the EQ Equip Evaluation Curve on a logarithmic time scale and that all EQ equipment in containment and the annulus remain qualified. Figure B-6A shows, on a logarithmic scale, that the EQ Equip Evaluation Curve is bounding. The NRC staff reviewed the environmental accident profiles and finds that the EQ Equip Evaluation Curve is bounding. In addition, the NRC staff finds that all equipment remains qualified and able to perform its safety function, since the licensee performed evaluations using the environmental profiles and EQ Equip Evaluation Curve on a logarithmic scale.

In its letter dated November 10, 2009, the licensee stated that for the purpose of equipment qualifications, the EQ Equip Evaluation Curve consists of the highest temperature value from each of the area subcurves for each time instance with an added 5 degrees Fahrenheit CF) margin. The NRC staff requested additional information regarding the justification of the added margin. The licensee responded in its March 8, 2010, letter that 1) the resulting margins met the recommendations of The Institute for Electrical and Electronics Engineers (IEEE) Standard (Std.) 323-1974, "IEEE Standard for Qualifying Class 1E Equipment for Nuclear Power Stations," 2) the evaluations performed remained consistent with IEEE Std. 323-1974, and

3) there were no adverse impacts on the existing qualifications for EQ equipment.

The licensee stated in Section 3.2.4 of the application that the current steamline break containment analysis bounds the plant response with proposed modifications based on the use of a computer code. The NRC staff requested additional information regarding why post steamline break temperatures in the annulus were not calculated assuming that the proposed modifications were in place. In addition, the licensee stated in Section 3.2.1 of the application that the current steamline break analysis bounds the plant response with the proposed modification to remove the automatic containment spray actuation logic. The NRC staff requested additional information on the basis of this conclusion. The licensee provided additional information by letter dated November 10, 2009, stating that the current steam line break analysis is a short-term analysis that examines the period of time during which superheated conditions may exist in the lower containment. Furthermore, the licensee stated that the lower containment temperature response is affected by 1) the steam released from the break, 2) ice condenser drain flow, and 3) the passive heat sinks. In addition, the current steamline break analysis does not credit liquid from containment spray in upper containment to enter lower containment. Therefore, the analysis does not include the timing of the containment spray actuation, equipment time delays, and containment spray header fill time. The NRC staff finds that, based on the fact that liquid entering upper containment due to containment spray

- 17 actuation does not affect the lower containment analyses, the proposed modification does not impact the current steamline break analysis and that the current steam line break analysis remains bounding.

In Section 3.2.9 of the application, the licensee stated that for ice condenser plants, the increase in containment pressure resulting from the elimination of containment spray would be limited.

The NRC staff requested clarification on what is meant by a limited increase in containment pressure and requested the values of containment pressure before and after the modifications.

In addition, the NRC staff requested additional information on the impact of increased containment pressure as a result of the modifications. By letter dated November 10, 2009, the licensee stated that a minimum containment pressure response is used in the peak clad temperature LOCA analysis since the maximum flow from two trains of containment spray minimizes the steam enthalpy credited in the analysis, which results in a conservative peak clad temperature. The licensee stated that there are no plans to incorporate the proposed modification into the minimum containment pressure analysis and the peak clad temperature LOCA analysis since the result would be a credit in the peak clad temperature analysis (a decreased fuel clad temperature). In addition, the licensee stated that the ability of the containment spray to condense steam in the lower containment is limited during the first few minutes of the accident and the containment pressure difference would be small. Thus, the NRC staff finds that the current minimum containment pressure and peak clad temperature LOCA analyses remain bounding, as the analyses ensure a conservative peak clad temperature.

Based on the above review, the NRC staff finds that there will not be any significant impact on the EO program at Catawba 1 and 2 due to the proposed amendment, and that the EO analyses continue to meet the requirements of 10 CFR Part 50, Appendix A, GDC 17 and GDC 18,10 CFR 50.49, and 10 CFR 50.65 for this application.

3.4. Human Factors 3.4.1 Emergency and Abnormal Operating Procedures By letter dated July 8, 2009, Duke identified the need for new operator procedural guidance and additional licensed operator training associated with the proposed LAR. These changes are scheduled for development as part of the plant modification process. The most significant changes are the changes to the cold leg recirculation sequence to reflect that containment spray will no longer take suction from the RWST, and the changes to the RWST setpoints to take advantage of the greater accuracy of the RWST level instrumentation. Duke has identified that the time available for these procedural actions is greater than the time required to perform the actions. All procedure changes and training will be scheduled for completion prior to implementation of the proposed LAR. Duke will use current administrative processes to develop new and revised plant procedures.

3.4.2 Operator Actions The manual start of the CSS is very similar to the original manual start except that the sequence is changed slightly, only one train is started, and the suction will be pre-aligned to the sump because the RWST will no longer be used as a source for CSS. The cue is the RWST low-level

- 18 alarm that initiates the alignment of RHR and CSS to the sump per procedure steps. This action is considered time-critical by Duke and will be monitored to ensure that time available exceeds time required.

3.4.3 Control Room Controls, Displays, and Alarms Chart recorders and level gauges will be relabeled. Setpoints associated with swapover, RWST low level, low-low-level, high level, makeup level, and pre-low level will be adjusted and/or reassigned as appropriate. The Operator Aid Computer (OAC) will require some point descriptions to be relabeled as "Wide Range" or added "Narrow Range". Some graphics and banding may also change in the OAC. These changes will be tracked by the Operations Training group so that associated changes to training materials and lesson plans can be implemented and used to train and test operators prior to implementation of the LAR.

3.4.4 Safety Parameter Display System (SPDS)

Duke has identified impacts to the Containment status tree on SPDS. The logic of this status tree will be changed to better reflect the conditional statements now required, i.e., pumps running and flow to CSS heat exchanger exists.

3.4.5 Control Room Plant Reference Simulator and Operator Training The type of training selected by the licensee is "Immediate Classroom" and "Simulator Training" as described below:

Classroom training typically includes:

The purpose of the proposed modification.

A summary of the engineering modification packages being installed in Catawba 2.

Summary descriptions of the type of accident scenarios where the Refueling Water, Containment Spray, and ESFAS System changes will impact operator responses.

A general walkthrough of the affected procedures with an explanation of any new or modified operator tasks.

Simulator training typically includes:

Additional or repeat information from classroom phase of training.

Accident scenarios to exercise the procedure changes, new system operation, and any new or modified skill/task in the form of simulator training guides.

The above training will be developed and presented to all affected licensed operators and non licensed operators from both units. Following completion, the information will be incorporated into the existing training materials and simulator guides similar to any other plant change. The simulator software will be modified after the outage in which plant modifications are made and when the normal simulator core upgrade for the next cycle is performed. Initial training will be presented just prior to, or during, this outage time. In any case, the training should be

- 19 completed prior to the beginning of the startup of the next cycle. Continuing training, as outlined in Catawba 1 and 2 Operator Training procedures, will be performed as normally scheduled, with these subjects incorporated into the existing training materials.

3.4.6 NRC Staff Evaluation of Human Factors The NRC staff has reviewed the licensee's proposed changes related to the human factors area and concludes that the licensee has adequately considered the impact of the proposed LAR on changes to operator actions, procedures, control room components, plant simulator and operator training programs to ensure that operators' performance is not adversely affected by the proposed changes to Technical Specifications. The NRC staff concludes that the licensee will continue to meet the requirements of GDC 19, 10 CFR 50.120, 10 CFR Part 55, and GL 82-33.

3.5 DBA Radiological Consequences The NRC staff reviewed the licensee's regulatory and technical assessments as they relate to the radiological consequences of DBA analyses in support of the proposed changes to the Catawba 1 and 2 licensing basis. The licensee determined, and the NRC staff agrees, that the only DBA analysis which is affected by the manual initiation of CSS to mitigate post-accident radiation doses is the design-basis LOCA. Further, this SE focuses on the specific effect of the proposed license changes and acknowledges Catawba 1 and 2's current AST design-basis assumptions. Information regarding the effect of the licensee's proposed changes on the design-basis LOCA dose consequence analysis was provided by the licensee in Attachment 1 of the LAR. The findings of this SE are based upon the descriptions and results of the licensee's assessment and other supporting information docketed by the licensee.

The original AST analysis of the LOCA was approved by the NRC on September 30, 2005. In the current licensing basis analysis for the design- basis LOCA, the licensee simulated the start of containment air return fans and containment spray operation beginning at 10 minutes. The dose analysis also takes credit for auxiliary containment spray from the RHR pumps. In addition, for some design-basis LOCA scenarios, credit has been taken for operation of both containment spray pumps. However, the revised Catawba 1 and 2 LOCA analysis, as submitted with the current LAR, incorporates information learned since the current analysis was performed. As part of the proposed modifications, the operators will start one containment spray pump only after the RHR system has been aligned to the containment sump for post LOCA ECCS recirculation. The RHR system will not be credited in the dose analysis for auxiliary containment spray for design-basis events. The dose consequence analysis, following the design-basis LOCA at Catawba 1 and 2, was recalculated to simulate these and other effects of the proposed modifications.

The reanalysis of the containment response to a large break LOCA has been determined using the Duke ice condenser containment response methodology described in TR DPC-NE-3004-PA.

Catawba 1 is equipped with feedring steam generators (FSGs). Catawba 2 is equipped with Model D5 SGs. Since the FSGs are limiting for containment response due to higher initial primary containment and secondary containment fluid mass, CataWba 1 was selected for reanalysis. The revised analysis was used to establish a new baseline before performing the sensitivity study to support the proposed modifications.

- 20 The radiological consequence analysis of the design-basis LOCA assumes full core melt with release of the radioactive material to the reactor coolant system and then to the containment over a total period of 1.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The source term used by the licensee assumes full power operation, including calorimetric uncertainty, to give 102 percent of rated power. The licensee has determined, and the NRC staff agrees, that there are four (4) affected components of the DBA analysis that have been modified since the NRC SE dated September 30, 2005 (AST):

Containment airborne radioactivity

Containment leakage

ESF system leakage

Control Room Habitability and Modeling The following sections address the impact of the licensee's proposed changes on these four affected activity release paths.

3.5.1 Containment Airborne Radioactivity Currently, Duke does not assume credit for fission product removal by sprays until the start of the air return fans, which is consistent with the assumption that the fission products are initially released in the lower compartment which has no spray. The licensee takes credit for elemental iodine removal until a decontamination factor (DF) of 200 is reached. The licensee also assumes that the particulate spray removal is reduced to 10 percent of the calculated value for the duration of the spray, considering that a DF of 50 would be reached very soon after the initiation of the air return fans. The licensee's current modeling of fission product removal in containment follows the guidance in RG 1.183, and is acceptable to the NRC staff.

The Catawba 1 and 2 CSS has two trains of safety-related pumps, heat exchangers, upper containment spray nozzles, and associated valves and piping. Duke's analysis of record takes no credit for the scrubbing of radioactive materials in the ice condenser; however, as a design basis Catawba 1 and 2 credits the chemical and physical properties of the ice condenser ice as an iodine removal system to reduce the fission product iodine concentration in the post-LOCA containment atmosphere. Following approval of this LAR, the containment spray automatic start signal would be removed and spray operation would be manually controlled when pump suction is aligned to the containment sump. Thus the licensee will take credit for the CSS for reducing containment pressure, temperature, and radioactivity in the upper containment compartment for removal of fission products other than organic iodine and noble gases by the CSS during the cold leg recirculation and hot leg recirculation phases of the accident.

The entire reactor coolant system inventory was assumed to be at TS levels and released to the containment at T = 0 hours0 days 0 hours 0 weeks 0 months . It is also assumed, in accordance with the guidance of RG 1.183, that 100 percent of the airborne activity is instantaneously and homogeneously mixed in the containment atmosphere. The time constants and DF cutoff times for washout of fission products with the Catawba 1 and 2 CSS were revised to account for the proposed modifications.

These included delay of start of containment spray for 4800 seconds (80 min.), no simulation of RHR auxiliary containment spray, and start of only one containment spray pump regardless of the design-basis LOCA scenario. The time constants and cutoff times, representative of the proposed modifications for the baseline analysis, are compared in Attachment 1 to the LAR in Table 1 for diatomic iodine and Table 2 for fission product particulates.

- 21 The methodologies used by Duke to calculate the time constants for elemental iodine removal and particulate removal by sprays follow the guidance in RG 1.183 and more specific guidance in SRP 6.5.2, Rev. 2, "Containment Spray as a Fission Product Cleanup System." The tables indicate that in the baseline analysis, it is assumed that the operators stop the containment spray pumps at 3000 seconds (50 min.) for the design-basis LOCA with minimum safeguards (one containment spray pump in operation) and at 1530 seconds (25.5 min.) for all other design basis LOCA scenarios. It is also assumed that for all design-basis LOCA scenarios at 3000 seconds, the operators start auxiliary containment spray from one RHR pump. Now with the proposed modifications in place, it is assumed that the operators complete the alignment of the CSS to the containment sump and start only one containment spray pump for post-LOCA recirculation at 4800 seconds. The operators also do not align the RHR system to the auxiliary containment spray headers.

Consequently, when the time constants for washout of particulates are applied to spray washout, the OF for spray removal of particulates is calculated to reach 50 for the design-basis LOCA with one and/or two containment spray pumps in operation. Thus, with the proposed modifications in place, the OF for spray removal of particulates is calculated to reach 50 for all design-basis LOCA scenarios, which justifies the licensee's proposal for the operator to manually initiate only one containment spray pump to successfully mitigate a LOCA.

3.5.2 Containment Leakage Each unit at the Catawba Nuclear Station has an ice condenser containment. This containment is divided into a lower compartment, an upper compartment and the ice condenser. The reactor and reactor coolant system are located in the lower compartment. The only connections between the lower and upper compartments are the containment air return fans, which blow air from the upper to the lower containment, and the ice condenser, through which the LOCA blowdown passes from the lower to the upper containment. However, since the gap release from the fuel is not assumed to occur until 30 seconds after the initiation of the event, the blowdown phase is not assumed to include the LOCA fission product source term. Also, Duke assumed that the fission product source term is released to and mixes homogeneously in the lower compartment volume only. The licensee's modeling of the LOCA release into containment and subsequent containment mixing is in accordance with the guidance in RG 1.183.

The secondary containment or annulus is provided with the Annulus Ventilation System (AVS) to filter primary containment leakage before release to the environment. The AVS is activated by a safety injection (SI) signal, and draws the annulus to a negative pressure. Duke reanalyzed the post-LOCA conditions in the annulus and the AVS response with the containment pressure and compartment temperatures associated with the proposed modifications for the three following DBA LOCA scenarios:

1) DBA LOCA with minimum safeguards - only one of the two trains responds as designed.

2) No AVS failures, with failure of a CSS or residual heat removal system heat exchanger.

3) Failure of one AVS pressure transmitter - one train operates in exhaust mode until operators secure it 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months after the initiating event.

However, for the reanalysis of the DBA LOCA, Duke analyzed the partitioning of containment leakage by source. The current analysis partitions the containment leakage in proportion to the

- 22 volumes of the lower and upper compartments. The respective volumes were changed to 60 percent from the lower compartment and 40 percent from the upper compartment. As a complementary measure, containment leakage to the annulus was assumed to mix with 50 percent of the air in the annulus, then filtered for release by the AVS.

The post-LOCA annulus drawdown times and AVS exhaust and recirculation airflow rates for the DBA LOCA scenarios limiting for offsite and control room radiation doses, with the proposed modifications in place, were all analyzed by Duke. The results of the DBA LOCA with RHR System or CSS heat exchanger failure, and DBA LOCA with initially closed Control Room Area Ventilation System (CRAVS) outside air intake, are displayed in Table 5 of Attachment 1 of the licensee's application dated September 2,2008. The time for the AVS to draw the annulus pressure to -0.25-inch water gauge everywhere inside the annulus, along with the AVS exhaust and recirculation airflow rates, were used as inputs to the AST analysis of the design-bases LOCA in support of the proposed modifications. The licensee has shown that, after implementing a new containment analysis, based on the CSS and/or RHR auxiliary CSS not being actuated for the LOCA, the containment pressure still decreases enough by 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the accident initiation to assume the 50-percent reduction in leak rate. These assumptions on the annulus mixing and percentage filtered by the AVS are acceptable and consistent with the current licensing basis of the Catawba 1 and 2 facilities.

3.5.3 ESF System Leakage In addition to the leakage of radioactive iodine in molecular form directly from the containment atmosphere to the outside, Duke identified a possible leakage path through the ESF in its docketed AST analysis. The containment sump water can leak either to the RWST or to the auxiliary building. Water can leak to the RWST, mostly through the valves that isolate the ESF recirculating system from the RWST. Since, after an accident, this water will contain radioactive iodine and since the RWST is vented to the atmosphere, this leakage will constitute a path for the radioactive iodine to leak to the outside and contribute to the external dose rates. In its submittal, Duke reanalyzed this source of iodine release for a LOCA with the proposed modifications assumed.

The licensee recalculated the iodine partition fractions for ESF leakage in the auxiliary building based on the calculated changes in the post-LOCA containment sump pH with the proposed modifications. The increased iodine partition fractions of ESF backleakage to the RWST were calculated with the initial water volume set to correspond to the RWST ECCS outlet elevation and vortex allowance. As input to this partitioning calculation, the licensee assumed a backleakage rate of 10 gpm, in contrast to the 20 gpm backleakage rate which is currently assumed. This reduction in assumed backleakage is still acceptable to the NRC staff, as it continues to result in a very conservative assumption when compared to other pressurized water reactor ESF system leakage assumptions. The source term for the ECCS leakage is the radioactivity in the containment sump water and it is assumed to consist of only the iodine isotopes and their precursors. The iodine species in the ECCS leakage release are assumed to be 97-percent elemental iodine and 3-percent organic iodine compounds. These assumptions are in accordance with guidance in RG 1.183. Duke's method for calculating Catawba 1 and 2 site-specific iodine release fractions (partitioning) from the ECCS leakage was previously approved in the issuance of the September 30,2005, AST amendment.

- 23 Duke based the calculation of the iodine partitioning on the methodology in NUREG/CR-5950, "Iodine Evolution and pH ControL" Using this methodology, the licensee calculated the formation of volatile elemental iodine based on the concentrations of iodine and iodide ions in the leakage. The total amount of iodine in the containment sump water, including stable 1-127 and long-lived 1-129, was used in the pH calculations. Post-LOCA containment sump pH was calculated with the proposed modifications assumed to be in place. The water inventory of the RWST was set to its upper bound value adjusted only for the elevation of the outlet piping. In particular, no flow from the CSS was simulated. Flow from the ECCS was used to simulate the transport of boric acid solution from the RWST to the containment sump. The CSS will not take suction from the RWST with the proposed modifications. In addition, the iodine partition fractions for ESF backleakage to the RWST were recalculated based on changes to the containment sump pH and RWST low-low level setpoint. Duke also included some conservatism in regard to its ECCS leakage analysis. For each ESF leakage scenario and time interval, the recalculated value was compared to the baseline values. The higher of the two values was taken for the iodine partition fraction. Based on the docketed information, all other input to the calculation of post-LOCA containment sump pH remained unchanged from the current licensing basis.

Although in the post-LOCA conditions the sump water pH is maintained at the value of equal to or higher than seven, when it mixes with the water remaining in the RWST after the injection phase, its pH could drop to a value significantly below seven. Low pH favors conversion of the dissolved iodine from a soluble ionic form to the scarcely soluble elemental form. Some of this iodine will be, therefore, released to the RWST air space from which it could leak to the atmosphere. The licensee assumed that the sump water leaks into the RWST at a rate of 10 gpm. As more of this water leaks into the RWST, both its pH and the concentration of iodine increase.

The NRC staff evaluated the licensee's pH calculation, as discussed above, and as it pertains to post-LOCA particulate release to the containment environment. The analysis of the licensee's containment sump water pH by the NRC is contained in Section 3.3.10 of this SE. The licensee calculated time-dependent water temperatures, iodine concentrations in the liquid and vapor, and the water pH for both the ECCS fluid and RWST. Using these calculated values, Duke determined iodine release rates from the RWST atmosphere to the environment and from the ECCS leakage areas to the environment. The limiting iodine partition fractions for the post LOCA ESF leakage are shown in Tables 3 and 4 in Attachment 1 of the LAR. The values associated with the proposed modifications were compared to the baseline values in these tables, and the time to begin recirculation with the proposed modifications in place was set to 2160 sec (0.6 hrs). Specifically, Table 3 in Attachment 1 of the LAR indicates that the iodine partition fractions for post-LOCA ESF leakage presented correspond to the LOCA scenarios limiting for radiation doses at offsite locations and in the control room. Separate values are presented for leakage inside and outside the ESF pump rooms. The licensee indicated that the filters of the auxiliary building filtered ventilation exhaust system (ABFVES) are aligned to the ESF pump rooms. This is assumed to occur after 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The values presented under the heading "In the ESF pump Rooms" in Table 6 in Attachment 1 of the LAR are associated with the LOCA scenarios which are limiting for radiation doses at offsite locations and in the control rooms. From the 72-hour time interval, indicating ABFVES filter alignment to the ESF pump rooms, baseline comparison with that of the proposed

- 24 modifications indicate an increased iodine partition fraction of 0.003 for offsite locations located both inside and outside the ESF pump rooms (modeled LOCA with failure of cooling water flow through an RHR or Containment Spray Heat Exchanger). Therefore, the analysis verified a minimal change from the established baseline values for post-LOCA ESF leakage, and the indicated leakage does not substantially increase with the proposed modifications and assumptions in place. The results in Table 3 of Attachment 1 of the LAR also indicate no change in the control room partition fraCtion for the LOCA scenarios which are limiting for radiation doses, with an initially closed control room outside air intake.

The NRC staff reviewed the licensee's assumptions, methodology, and conclusions regarding the pH of sump water and the corresponding fraction of the iodine dissolved in sump water converted into the elemental form. From the results of these calculations, the NRC staff concluded that the proposed modifications result in a nE;lgligible fraction of the dissolved iodine being converted into elemental form and low release of radioactive iodine to the environment.

The NRC staff also concluded that the amount of iodine that would be released from the RWST as a result of backleakage would be negligible because of the low concentration of iodine when the tank pH is below seven.

3.5.4 Control Room Habitability and Modeling The design-basis LOCA dose consequence analysis modeling of the Catawba 1 and 2 control room design and operation changed as a result of the proposed modifications. The calculation of post-LOCA radiation dose accounts for release of activity, transport of effluent to the CRAVS outside air intakes, and buildup of activity in the control room. To model the control room response to the LOCA with loss of offsite power, Duke assumed that the CRAVS initiates at the time of the core damage, and begins pressurization and filtration of the control room. Duke also assumed that one train of the CRAVS is immediately secured by the operators; therefore, only one train of CRAVS is credited in the dose analysis. As previously approved for the Catawba 1 and 2 AST amendments by letter dated September 30, 2005, the limiting value for asymmetry in the airflow into the two CRAVS outside air intakes was determined to be 60/40, and the control room atmospheric dispersion factors (x/Q ) were adjusted to account for the imbalance. Also consistent with its docketed AST analysis, the licensee assumed the lower limits for the CRAVS filter train recirculation flow rates. It also calculated doses for both the lower limit and upper limit for the CRAVS intake flow rate. The licensee assumed that the unfiltered inleakage into the control room was 100 cfm for the duration of the event.

The NRC staff also assessed other contributors to the post-LOCA control room radiation doses for the AST amendment SE. In particular, the direct radiation dose to the control room from fission products outside the control room was calculated consistent with the regulatory positions for full scope implementation of AST methodology. According to the licensee's analysis, this direct constituent to the radiation dose in the control room was determined to be 0.75 Rem.

This value was added to the TEDE from post-LOCA transport activity to the CRAVS outside air intakes to yield the total post-LOCA TEDE in the control room. The above noted changes yielded a moderate increase in the post-LOCA TEDE at the EAB. The cumulative impact of these changes on the LPZ dose is very small (less than 0.1 rem TEDE). The addition of the direct constituent from external source to the effluent constituent yielded an increase in the post LOCA control room dose by 0.75 rem TEDE. The other changes combined yielded a very small increase in the post-LOCA control room dose (less than 0.1 rem TEDE).

- 25 The baseline dose at the EAB and LPZ and in the control room for the limiting LOCA scenarios are presented in Table 5 of Attachment 1 to the LAR. Duke then took these baseline values, applied the proposed modifications (modeled as the design-basis LOCA with failure of cooling water flow through a heat exchanger of either the RHR system or the CSS), and compared the dose results of the proposed modifications to that of the baseline analysis. The latter comparison is also presented in the tower section of Table 5 of Attachment 1 to the LAR. Either the RHR system or the CSS was verified to be limiting for dose at the offsite locations, and the design-basis LOCA, with an initially closed CRAVS outside air intake, was verified to be limiting for the control room dose, based on Duke's analysis. For an assumed 2-hour duration, the total dose of the proposed modifications increased minimally for the EAB, LPZ, and control room.

Specifically, the total EAB dose increased by 1.91 to 8.79 rem TEDE; the total LPZ dose increased by 0.52 to 3.78 rem TEDE; and the total control room dose increased by 0.31 to 3.31 rem TEDE, all of which are acceptable to the NRC staff and remain consistent with the applicable dose acceptance criteria described in RG. 1.183.

3.6 Piping Design Temperature Evaluation The peak sump liquid temperature for the proposed changes is 198.7 of which is greater than the current UFSAR piping design temperature of 190 of. The licensee therefore proposes to increase the piping design temperature to 200 of. The NRC staff has reviewed Section 3.2.3 "Piping Analysis," of Attachment 1 to the LAR, and has determined that the licensee has adequately addressed the impacts of the delay of the containment spray actuation, after swapping to the containment sump. The licensee has adequately addressed the effects of the associated increase in temperature on the piping stresses and the support loads. The NRC staff agrees with the licensee that the resulting impacts of the existing piping qualification are negligible due to the following reasons: (1) there will be a minimal increase in temperature (approximately 10 OF) for a short duration of time (18.8 minutes); (2) the affected piping segment length is short, approximately 20 feet, for each ECCS train; and (3) the thermal expansion of the piping is unrestrained. Therefore, the NRC staff concludes that the impact of the delay of the containment spray actuation, after swapping to the containment sump, has a negligible effect on the piping qualification, and is, therefore, acceptable.

3.7 Technical Evaluation Summary The NRC staff has reviewed the proposed TS changes to allow manual operation of the containment spray system, to revise the RWST level and to increase the RWST water volume.

These changes will increase the ECCS inventory, increase the operator response time, and extend the post-LOCA injection phase. The LAR has demonstrated that it complies with 10 CFR 50.36(c)(2) and 10 CFR 50.46(b)(5) and addresses GL 2004-02.

The NRC staff has concluded, based on the evaluation of the proposed changes discussed above, that (1) the containment and associated systems will maintain a leak-tight barrier against the uncontrolled release of radioactivity to the environment and the containment design conditions important to safety are not exceeded; (2) the containment heat removal system will continue to perform its safety function in response to the DBAs; and (3) the containment heat removal system will maintain the containment structure and its internal compartments without exceeding the design leakage rate, and with sufficient margin, the calculated pressure and the temperature conditions resulting from any DBA.

- 26 As noted above, the NRC staff expects that this license amendment will have a neutral or salutary effect on sump strainer performance. The proposed resolution of any issues related to GL 2004-02 will be evaluated to verify that impacts on sump strainer performance are acceptable.

The NRC staff has reviewed the licensee's proposed changes related to the human factors area and concludes that the licensee has adequately considered the impact of the proposed LAR on changes to operator actions, procedures, control room components, plant simulator and operator training programs to ensure that operators' performance is not adversely affected by the proposed changes to the TSs. Thus, the NRC staff concludes that the licensee will continue to meet the requirements of GDC 19, 10 CFR 50.120, 10 CFR Part 55, and GL 82-33.

As described above, the NRC staff reviewed the licensee's assumptions, inputs, and methods to assess the impact of the proposed licensing basis changes on the postulated design-basis LOCA. The NRC staff finds that Duke's revised design-basis LOCA analysis conservatively accounts for the impact of the proposed changes. As a result, the licensee will continue to meet the applicable dose acceptance criteria, as identified in Section 2.0 of this evaluation, following implementation of these changes. The NRC staff further finds reasonable assurance that Catawba 1 and 2, as modified by this license amendment, will continue to provide sufficient safety margin, with adequate defense-in-depth, to mitigate unanticipated events and to compensate for uncertainties in accident progression, analysis assumptions, and input parameters. Therefore, the proposed license amendment is acceptable with respect to the radiological dose consequences of the DBAs.

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 (74 FR 15767). Accordingly, the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). 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.

- 27

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.

Principal Contributors: H. Garg, NRR A. Sallman, NRR K. Desai, NRR G. Lapinsky, NRR L. Benton, NRR A. Boatright, NRR S. Ray, NRR Date: June 28, 2010

J. Morris -2 If you have any questions, please call me at 301-415-1119.

Sincerely, IRA by JStang fori 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. 257 to NPF-35

2. Amendment No. 252 to NPF-52

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

PUBLIC RidsNrrDirslolb Resource RidsNrrDirsltsb Resource LPL2-1 R/F RidsNrrDraAadb Resource RidsNrrDssSrxb Resource RidsNrrDorlLpl2-1 Resource RidsNrrDeEicb Resource RidsNrrDssScvb Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDorlDpr Resource RidsOgcRp Resource RidsNrrLAMOBrien Resource (hard copy) RidsRgn2MailCenter Resource A. Sallman, NRR/DSS RidsNrrPMCatawba Resource (hard copy) H. Garg, NRR/DE K. Desai, NRR/DSS G. Lapinsky, NRR/DIRS L. Benton, NRR/DRA A. Boatright, NRRIDRA G. Wilson, NRR/DE RidsNrrDeEeebResource ADAMS Accession No. ML092530088

no change to SE input sent by memo ML092110593 **no change to SE input sent by memo ML092180536

- - no change to SE input sent by memo ML092320453 ****no change to SE input sent by memo ML092150891 no c hanQe to SE'm put sent bJY memo ML092170494 ******no c hange t 0 SE'mput sent bJy memo ML101160167 OFFICE NRR/LPL2-11 PM NRR/LPL2-11 LA DIRS/ITSBI BC DSS/SRXBI BC DSS/SCVBI BC DSS/EICBI BC JThompson MO'Brien RElliott GCranston* RDennig** WKemper***

NAME (JStang for)

DATE 06/28/10 06/02/10 06/09/10 07/31/09 08/20109 08/25/09 OFFICE DSS/IOLBI (A)BC DSS/AADBI BC DE/EEEB/BC OGC NRR/LPL2-11 BC NRR/LPL2-1/PM JMunro**** RTaylor***** GWilson****** MDreher GKulesa JThompson NAME (VSreenivas for) (JStang for)

DATE 08/12/09 08/26/09 04/28/10 06/21/10 06/25/10 06/28/10 OFFICIAL RECORD COpy

v t e Letter Sequence Approval
TAC:MD9752, (Approved, Closed) TAC:MD9753, (Approved, Closed)
Initiation Request, Request, Request Acceptance... Administration Questions Answers Results Approval Other: ML102070470
MONTHYEAR2010-06-28 [Table View]

ML092530088

Text

SUBJECT:

Dear Mr. Morris:

Enclosures:

Attachment:

Attachment:

1.0 INTRODUCTION

2.0 REGULATORY EVALUATION

3.0 TECHNICAL EVALUATION

4.0 STATE CONSULTATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 CONCLUSION

Enclosures:

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