Text

Proposed Technical Specifications and Bases Amendment 2.0, Safety Limits; 3.3, Instrumentation; 3.4, Reactor Coolant System; 5.6, Reporting Requirements
	ML032130497
Person / Time
Site:	Mcguire, Catawba, McGuire
Issue date:	06/25/2003
From:	Mccollum W; Duke Power Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML032130497 (55)
	v • d • e

Duke EMPower.

A Duke Ened Company Duke Power Company A Dwke Energ Company Energy Center P.O. Box 1006 Charlotte, NC 28201-1006 June 25, 2003 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

Subject:

Duke Energy Corporation McGuire Nuclear Station, Units 1 and 2 Docket Numbers 50-369 and 50-370 Catawba Nuclear Station, Units 1 and 2 Docket Numbers 50-413, 50-414 Proposed Technical Specifications and Bases Amendment 2.0, Safety Limits; 3.3, Instrumentation; 3.4, Reactor Coolant System; 5.6, Reporting Requirements Duke License Amendment Request of March 24, 2003 On March 24, 2003 Duke Energy Corporation (Duke) submitted proposed amendments to the McGuire and Catawba Nuclear Station Facility Operating Licenses and Technical Specifications (TS) and Bases.

On June 9, 2003 Duke personnel met with members of the NRC staff to discuss issues related to that submittal.

This letter transmits a revised TS amendment package based on the results of that public meeting.

The proposed amendment reduces the required minimum measured Reactor Coolant System (RCS) flow rate from 390,000 gallons per minute (gpm) to a previously approved value of 382,000 gpm for McGuire Units 1 and 2, and Catawba Unit 1; and reduces the required minimum measured RCS flow rate from 390,000 gpm to the originally licensed value of 385,000 gpm for Catawba Unit 2. Additionally, the proposed change relocates RCS related cycle-specific parameter limits from the TS to, and thus expands, the Core Operating Limits Reports (COLR) for the McGuire and Catawba Nuclear Stations.

These proposed changes will allow Duke the (d

U.S. Nuclear Regulatory Commission Page 2 June 25, 2003 flexibility of enhancing operating and core design margins without the need for cycle-specific license amendment requests.

The proposed amendment changes TS 2.2.1, "Reactor Core Safety limits," and associated Bases, by relocating the reactor core safety limit figure to the COLRs and replacing it with specific fuel Departure from Nucleate Boiling Ratio (DNBR) and peak fuel centerline temperature safety limit requirements.

The amendment proposes to change TS Table 3.3.1-1, "Reactor Trip System Instrumentation,f by relocating Overtemperature AT and Overpower AT nominal RCS operating pressure, nominal Tr and constant (K) values to the respective station COLR.

The amendment also revises TS 3.4.1, "RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,n and associated Bases, by relocating the pressurizer pressure, RCS average temperature, and RCS total flow rate values to the respective station COLR.

The minimum RCS total flow rates are retained in TS Table 3.4.1-1.

TS 5.6.5, "Core Operating Limits Report (COLR)," would be modified to reflect the above relocations to the COLR.

The requested changes are based on NRC approval of Westinghouse Owners Group (WOG) Technical Specifications Task Force (TSTF) TSTF-339, "Relocate TS Parameters to the COLR Consistent with WCAP-14483," Revision 2, and Westinghouse WCAP-14483-A, "Generic Methodology for Expanding Core Operating Limits Report."

The return to a lower RCS minimum flow rate, previously reviewed and approved by the NRC, is similar to that submitted by the Comanche Peak Nuclear Station on May 24, 1999, and approved by the NRC in the Safety Evaluation Report dated August 30, 1999.

During the June 9, 2003 meeting, Duke requested approval of the proposed changes by November 2003.

Upon reconsideration, Duke would prefer that this license amendment request be approved in time to support the commencement of McGuire Unit 2, Cycle 16 operations currently scheduled to begin October 6, 2003.

Duke has

U.S. Nuclear Regulatory Commission Page 3 June 25, 2003 determined that the NRC's standard 30 day grace period will be sufficient for the implementation of this amendment.

The contents of this revised amendment package are as follows: A provides revised marked copies of the affected TS and Bases pages for McGuire showing the proposed changes.

Attachment B provides revised marked copies of the affected TS and Bases pages for Catawba showing the proposed changes. Attachment 2A, containing reprinted pages of the affected TS and Bases pages for McGuire, will be provided to the NRC upon issuance of the approved amendment. B, containing reprinted pages of the affected TS and Bases pages for Catawba, will be provided to the NRC upon issuance of the approved amendment. Attachment 3 provides a revised description of the proposed changes and technical justification.

The revisions provided in this letter do not change the conclusions reached in the original March 24, 2003 No Significant Hazards determination.

Additionally, the basis for the categorical exclusion from the performance of an Environmental Assessment/Impact Review in the March 24, 2003 is unaffected by this revision.

Implementation of this amendment will impact the McGuire and Catawba Updated Final Safety Analysis Reports (UFSAR).

Changes to the UFSARs will be submitted in accordance with 10CFR50.71(e) requirements.

Pursuant to 10CFR50.91, a copy of this proposed amendment is being sent to the appropriate state officials.

Inquiries on this matter should be directed to J. A.

Effinger at (704) 382-8688.

U.S. Nuclear Regulatory Commission Page 4 June 25, 2003 Very truly yours, W. R. Mc CollumX r.

Senior Vice Pr sident Nuclear Support

U.S. Nuclear Regulatory Commission Page 5 June 25, 2003 W. R. Mc Collum, Jr., being duly sworn, affirms that he is the person who subscribed his name to the foregoing statement, and that all matters and facts set forth herein are true and correct to the best of his knowledge.

W. R. Mc Collum, Jr n ior Vice President, Nuclear Support Subscribed and sworn to me:

~~~

AimT iByMe.

"ZS)

Gvc3 Date

., Notary Public

4,j 2 2 2ob 6 My commission expires:

SEAL

U.S. Nuclear Regulatory Commission Page 6 June 25, 2003 xc (w/attachments):

L. A. Reyes U. S. Nuclear Regulatory Commission Regional Administrator, Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, GA 30303 R. E. Martin NRC Project Manager (MNS) (CNS)

U. S. Nuclear Regulatory Commission Mail Stop 0-8 H12 Washington, DC 20555-0001 S. M. Shaeffer Senior Resident Inspector (MNS)

U. S. Nuclear Regulatory Commission McGuire Nuclear Site E. F. Guthrie Senior Resident Inspector (CNS)

U. S. Nuclear Regulatory Commission Catawba Nuclear Site B. 0. Hall, Section Chief Radiation Protection Section 1645 Mail Service Center Raleigh, NC 27699-1645 R. Wingard, Director Division of Radioactive Waste Management Bureau of Land and Waste Management Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201

U.S. Nuclear Regulatory Commission Page 7 June 25, 2003 bxc wattachments:

C. J. Thomas G. D. Gilbert M. T. Cash G. B. Swindlehurst R. A. Hight L. J. Rudy K. E. Nicholson J. M. Ferguson (RGC data file)

K. L. Crane R. L. Gill Catawba Owners: NCMPA-l, SREC, PMPA, NCEMC McGuire Master File (MGO1DM)

Catawba Document Control File 801.01 (CN04DM)

ELL

ATTACHMENT 1A McGUIRE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS AND TECHNICAL SPECIFICATION BASES MARKED COPY

INSERT 1 the COLR for four loop operation; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.14 for the WRB-2M CHF correlation.

2.1.1.2 The peak fuel centerline temperature shall be maintained <5080 degrees F, decreasing 58 degrees F for every 10,000 MWd/mtU of fuel burnup.

INSERT 2 The reactor core SLs are established to preclude violation of the following fuel design criteria:

a.There must be at least a 95% probability at a 95%

confidence level (the 95/95 DNB criteria) that the hot fuel rod in the core does not experience DNB; and b.There must be at least a 95% probability at a 95%

confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.

The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Over Temperature and Overpower LT reactor trip functions.

That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation.

Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and I that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.

1

INSERT 3 The numerical limits of these variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle.

However, the minimum RCS flow, based on previously analyzed maximum steam generator tube plugging, is retained in the TS LCO.

2

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed theeilfied i

2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained < 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

McGuire Units and 2 2.0-1 Amendment Nos. JPV6

SLs 2.0 Pi 660 650 640 I--IL 0

(U (I) 0 aIx 630

DO NOT OPERATE IN THIS OPl2280 psiN

\\45 pi OPERATION\\

620 610 600 590 580 0.2 0.4 0.6 0.8 1.0 Fraction of Rated Thermal Power Figure 2.1.1-1 Reactor Core Safety Limits -

Four Loops in Operation McGuire Units 1 and 2 2.0-2 Amendment Nos. y,

Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and A0Os. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a.

There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b.

The hot fuel pellet in the core must not experience centerline fuel melting.

The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient r TICS Fewuj IMTE,KX conditions for Reactor Coolant System (RCS) temperature ressure, and THERMAL POWER level tat wouna eactra departure rom nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

Automatic enforcement of these reactor core SLs is provided by the A

/

Ao£ uripq peeCpr-u T

ti p)

I. /AM etPT1Ke~prir/

,e. / v~r3r, ersl;udrrF,;

/ g~r6#4r e

it n a e

-ereha yifth Iotle beesth or

,dyltheht~ly Upure Ieridass ens~sStf as fedf j meratio usK i~h PS cesi a

measke of ore ower, pr p0 na 0 e ow r.

The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits," or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

McGuire Units I and 2 B 2.1.1-2 Revision No.K

Reactor Core SLs B 2.1.1 FSAFTS The 0en0s provided in i(,2r,<

f the loci of points of b J /FA

oJ C S RCS Pressure, and average temperature for which the minimum DNBR is I

-C p

r£ m VW not less than the safety analyses limit, that fuel centerline temperature 4

oi

m.

remains below melting, that the average enthalpy in the hot leg is less

)

than or equal to the enthalpy of saturated liquid, and that the exit quality is within the itsfd by the DNBR correlation.

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, Reactor Trip System (RTS) Instrumentation.

In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT If SL 2.1.1 is violated, the requirement to go to MODE 3 VIOLATIONS places the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

REFERENCES

1.

10 CFR 50, Appendix A, GDC 10.

2.

UFSAR, Section 7.2.

X /A67-I PUnwe0ptv 4%

CZ McGuire Units 1 and 2 B 2.1.1-3 Revision No.

Reactor Core SLs B 2.1.1 Lr

It I

0.

I I

I 7va (f)

OT&T CORE Figure B 2.1.1-1 Illustration of Overtemperature and Overpower AT Protection McGuire Units 1 and 2 B 2.1.1-4 Revision NoA

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 7)

Reactor Trip System Instrumentation Note 1: Overtemroerature AT The Overtemperature AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 4.4% of RTP.

aT (1 +s (v 1

) sAT0 K, -K2 (1 + r 4 s) 1

-T s)(P-P]f 1

T(I +rs) 1

[,~

(

(AI)}(1+T Where:

AT is measured RCS AT by loop narrow range RTDs, F.

AT0 is the indicated AT at RTP, F.

s is the Laplace transform operator, sec-1.

T is the measured RCS average temperature, F.

' Is the norinal Tavg at RTP, <Z. W

=

3 P is the measured pressurizer pressure, psig P' is the nominal RCS operating pressure, =

=

K,

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

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

= Overternperature AT reactor trip depressurization setpoint penalty coefficient, as presented In the COLR,

,rux2

= Time constants utilized in the lead-ag controller for AT, as presented in the COLK S3

Time constants utlized in the lag compensator for AT, as presented in the COLR x4. xs

nTime constants utilized In the lead-ag controler for T., as presented in the

COLR, e

= Tine constants utilized In the measured Tsw lag compensator, as presented in the COLR, and.

f,(aI) = a function of the hidcated difference between top and bottom detectors of the power-range nuclear Ion chambers; with gains to be selected based on measured nstrument response during plant startup tests such tat (i) for qt - q, between the 'positive' and 'negative' fi(A) breakpoints as presented In the COLR; fj(AQ) = 0, where qt and q, are percent RATED THERMAL POWER In the top and bottom halves of the core respectively, and q. + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (continued)

McGuire Units 1 and 2 3.3.1-18 Amendment Nos.

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 7)

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

Note 2: Overoower AT The Overpower AT Function Allowable Value shall not exceed the following NOMINAL TRIP SETPOINT by more than 3.0% of RTP.

A&T (1_6) 5 (1 + rs (

)5s ATs {K K 1r7t JT6

(

I 1 K

1+ l T

f (Ij Where:

AT is measured RCS AT by loop narrow range RTDs, F.

ATo is the indicated ET at RTP, F.

s is the Laplace transform operator, sec'.

T is the measured RCS average temperature Is the noninal T, at RTP, s

AL.=

K14

=

Overpower AT reactor NOMINAL TRIP SETPOINT as presented in the

COL, K6

=

mok orncreasing ra r

ov veage temperature, K.

= Overpower AT reactor tp hey setol coefficient as presented in the COLR fo afor T s vu 2 = Tr e edn ed te lagultler for AT, as presented in the COLR T3

= Thie constants utilized n the tag compensatorfor AT. as presented the

COLR, Ire

= Time constants utilized in the measured To lag corpensator, as presented In the COLR, a7

= Tune constant utilized in the rate-ag controller for T,,, as presented In the COLR, and f2(AI) = a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that (continued)

McGuire Units and 2 3.3.1-19 Amendment Nos.

4

RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued) conditions and take corrective actions. Additionally, low temperature overpressure protection systems provide overpressure protection when below MODE 4.

9.

Pressurizer Water Level-High The Pressurizer Water Level-High trip Function provides a backup signal for the Pressurizer Pressure-High trip and also provides protection against water relief through the pressurizer safety valves.

These valves are designed to pass steam in order to achieve their design energy removal rate. A reactor trip is actuated prior to the pressurizer becoming water solid. The setpoints are based on percent of instrument span. The LCO requires three channels of Pressurizer Water Level-High to be OPERABLE. The pressurizer level channels are used as input to the Pressurizer Level Control System. A fourth channel is not required to address contro/protection interaction concems. The level channels do not actuate the safety valves, and the high pressure reactor trip is set below the safety valve setting. Therefore, with the slow rate of charging available, pressure overshoot due to level channel failure cannot cause the safety valve to lift before reactor high pressure trip.

In MODE 1, when there is a potential for overfilling the pressurizer, the Pressurizer Water Level-High trip must be OPERABLE. This trip Function is automatically enabled on increasing power by the P-7 interlock. On decreasing power, this trip Function is automatically blocked below P-7. Below the P-7 setpoint, transients that could raise the pressurizer water level will be slow and the operator will have sufficient time to evaluate unit conditions and take corrective actions.

10.

Reactor Coolant Flow-Low

a.

Reactor Coolant Row-Low (Single Loop)

The Reactor Coolant Flow-Low (Single Loop) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, while avoiding reactor trips due to normal variations in loop flow.

Above the P-8 setpoint, which is approximately 48% RTP, a loss of flow in any RCS loop will actuate a reactor trip. The setpoints are based onsyminimum PS flow McGuire Units 1 and 2 B 3.3.1-17 Revision No.,jA

RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICABILITY (continued)

.,Fs X

ae. co m

e,, jgW. Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE in MODE 1 above P-8.

In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-8 setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 1.b) because of the lower power level and the greater margin to the design limit DNBR.

b.

Reactor Coolant Flow-Low (Two Loons)

The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow.

Above the P-7 setpoint and below the P-8 setpoint, a loss of flow in two or more loops will initiate a reactor trip. The setpoints are based oninium W rwflow o1f4AW

-..a, ',r>ia Co 4.

}

> a. Each loop has three flow detectors to monitor flow.

The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE.

In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since power distributions that would cause a DNB concern at this low power level are unlikely. Above the P-7 setpoint, the reactor trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR.

McGuire Units 1 and 2 B 3.3.1-18 Revision No.x

RCS Pressure, Temperature, and Flow DNS Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 APPLICABILITY:

RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in Table 3.4.1 -1.

MODE 1.

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

Pressurizer pressure limit does not apply during:

a.

THERMAL POWER ramp > 5% RTP per minute; or

b.

THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressurizer pressure or A.1 Restore DNB parameter(s) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months RCS average to within limit.

temperature DNB parameters not within limits.

B.

RCS total flow rate

-,AJ -pr~Irc Zp 13.1 Reduce THERMAL POWER to < 98% RTP.

AND B.2 Reduce the Power Range Neutron Flux - High Trip Setpoint below the nominal setpoint by 2% RTP.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 6 hours (continued)

McGuire Units 1 and 2 3.4.1 -1 Amendment Nos. 423a

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

RCS total flow rate f

qS Oc By s

VA~

,.I71-le C.1 Restore RCS total flow rate 1

7 9'/oSa^

C.2.1 Reduce THERMAL POWER to < 50% RTP.

AND C.2.2 Reduce the Power Range Neutron Flux - High Trip Setpoint to < 55% RTP.

AND C.2.3 Restore RCS total flow rate 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 hours 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 24 hours C14%, OF7iffV44uz-

'C'o '--j rA..- e'oc-K D.

Required Action and D. 1 Be in MODE 2.

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

McGuire Units 1 and 2 3.4.1 -2 Amendment Nos. 19U P-T01i-)

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

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.2 Verify RCS average temperature is within limits.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.3 Verify RCS total flow rate is within limits.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.4 Perform CHANNEL CALIBRATION for each RCS total 18 months flow indicator.

McGuire Units 1 and 2 3.4.1-3 Amendment NosIl

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Table 3.4.1-1 (page 1 of RCS DNB Parameters 1)

No. OPERABLE LIMITS CHANNELS PARAMETER INDICATION

1.

Indicated RCS Average Temperature meter meter 4

3 computer computer 4

3

< 01.2At 4 _

< 5$1fr X

1

< FR761fv

Ž:>

??5

'pMgT

> z WX dev 2,r7/5szpereg

2.

Indicated Pressurizer Pressure meter meter 4

3 computer computer 4

3

3.

RCS Total Flow Rate

>%0aKn t

A_

40 oOS

,0 p_~lK or MaZiA.Sp -TV 6R A

{~~~~~W e Q0 1tSn 4,Z McGuire Units 1 and 2 3.4.1-4 Amendment Nos.

4P

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued) assessed for their impact on the acceptance criteria. A key assumption for the analysis of these events is that the core power distribution is within the limits of LCO 3.1.6, Control Bank Insertion Limits"; LCO 3.2.3, NAXIAL FLUX DIFFERENCE (AFD)"; and LCO 3.2.4, QU NT POWER TILT RATIO (QPTR).'

The pressurizer pressure limits and the RCS average temperature limits correspond to analytical limits used in the safety analyses, with allowance for measurement uncertainty.

The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).

LCO This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analss.

Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion.

RCS total flow rate contains a measurement error4/X/&/ based on the performance of past precision heat balances and using the result to calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCS flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow measurement. Potential fouling of the feedwater venturi, which might not have been detected, could have biased the result from these past precision heat balances in a nonconservative manner.

Therefore, a penalty e'WWfor undetected fouling of the feedwater venturi raises the nominal flow measurement aIlowancelfa4%'for no fouling.

71 The Wnumerical values j6,0/for pressure and average temperature are given for the measurement location with adjustments for the indication instruments.

APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern.

McGuire Units 1 and 2 B 3.4.1-2 Revision No '<

-1

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued)

A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pres r

At i

I64 is provided in SL 2.1.1, d

a e A8 t

Ls.wu are less restrictive than the limits of this T~ Wet I,*i....

H LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.

ACTIONS A.1 Pressurizer pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within LCO limits, action must be taken to restore parameter(s).

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.

B.1 and B.2 RCS total flow rate is not a controllable parameter and is not expected to vary during steady state operation. If the indicated RCS total flow rate is

,gOffigprf(yWg3R~srgft then THERMAL POWER may not exceed 98% RTP. THERMAL POWER must be reduced within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The completion time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with Required Action A.1. In addition, the Power Range Neutron Flux - High Trip Setpoint must be reduced from the nominal setpoint by 2% RTP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpoints recognizes that, with power reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System.

/q 4/o, BT t 4

< boos oF

Zac, SC§FE 3

A_~~~7 J I

I McGuire Units 1 and 2 B 3.4.1-3 Revision No. C

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS (continued) 4lass

°114-/J br. -nK VAZJ.

C.1. C.2.1. C.2.2. and C.2.3 If the indicated RCS total flow rate is/M0 r

then RCS total flow must be restored to ;

,,1OV74wfithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or power must be

~-~rr~f~. TW-4 a reduced to less than 50% RTP. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is

__S923Z\\.C.

Ad <°q oconsistent with Required Action A.1. If THERMAL POWER is reduced to A

n less than 50% RTP, the Power Range Neutron Flux - High Trip Setpoint 7?-. am~ ag'.'e2 SZF,~

must also be reduced to

55% RTP. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to (4

Ad Casing

> reset the trip setpoints is consistent with Required Action B.2. This is a sensitive operation that may inadvertently trip the Reactor Protection S

System. 0peration is permitted to continue provided the RCS total flow is restored to 4~frvit thin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint.

D.1 If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable to reach the required plant conditions in an orderly manner.

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS pressure and related equipment status. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.

McGuire Units 1 and 2 B 3.4.1-4 Revision No.*

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.2 Annual Radiological Environmental Operating Report (continued)

The Annual Radiological Environmental Operating Report shall include summarized and tabulated results of the analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event that some individual results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible.

5.6.3 Radioactive Effluent Release Report NOTE--

A single submittal may be made for a multiple unit station. The submittal should combine sections common to all units at the station; however, for units with separate radwaste systems, the submittal shall specify the releases of radioactive material from each unit.

The Radioactive Effluent Release Report covering the operation of the unit in the previous year shall be submitted prior to May 1 of each year in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit.

The material provided shall be consistent with the objectives outlined in Chapter 16 of the UFSAR and in conformance with 10 CFR 50.36a and 10 CFR Part 50, Appendix 1,Section IV.B.1.

5.6.4 Monthly Operating Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

5.6.5 CORE OPERATING LIMITS REPORT (COLR)

a.

Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

7. ?.

x Moderator Temperature Coefficient BOL and EOL limits and 300 eillance limit for Specification 3.1.3,

!E StE=Zfe-Ttao

%sl>

~~~~~~~~~~~~~(continued)

McGuire Units 1 and 2 5.6-2 Amendment Nosh14W

5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

3.

Shutdown Bank Insertion Limit for Specification 3.1.5, Control Bank Insertion Limits for Specification 3.1.6, Axial Flux Difference limits for Specification 3.2.3.

4.A.

Heat Flux Hot Channel Factor for Specification 3.2.1, Nudear Enthalpy Rise Hot Channel Factor limits for Specification 3.2.2, P. A.

Overtemperature and Overpower Delta T setpoint parameter values for Specification 3.3.1,

/J$.

Accumulator and Refueling Water Storage Tank boron concentration limits for Specification 3.5.1 and 3.5.4,

// Zl, Reactor Coolant System and refueling canal boron concentration limits for Specification 3.9.1,

/l.

Spent fuel pool boron concentration limits for Specification 3.7.14, 13, X.

SHUTDOWN MARGIN for Specification 3.1.1.

b.

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1.

WCAP-9272-P-A, "WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY,' (

Proprietary).

2.

WCAP-1 0266-P-A, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE,' M Proprietary).

3.

BAW-10168P-A, B&W Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants," (B&W ropnietary).

'?. gEAc~T

<4 n

Sy sash,~

e~S4

/ t' AzJ'T t.,

He FL~

A (continued)

McGuire Units 1 and 2 5.6-3 Amendment Nos.

A0318ZL

ATTACHMENT 1B CATAWBA UNITS 1 AND 2 TECHNICAL SPECIFICATIONS AND TECHNICAL SPECIFICATION BASES MARKED COPY

INSERT 1 the COLR for four loop operation; and the following SLs shall not be exceeded:

2.1.1.1 The departure from nucleate boiling ratio (DNBR) shall be maintained 1.14 for the WRB-2M CHF correlation.

2.1.1.2 The peak fuel centerline temperature shall be maintained <5080 degrees F, decreasing 58 degrees F for every 10,000 MWd/mtU of fuel burnup.

INSERT 2 The reactor core SLs are established to preclude violation of the following fuel design criteria:

a. There must be at least a 95% probability at a 95%

confidence level (the 95/95 DNB criteria) that the hot fuel rod in the core does not experience DNB; and b.There must be at least a 95% probability at a 95%

confidence level that the hot fuel pellet in the core does not experience centerline fuel melting.

The reactor core SLs are used to define the various RPS functions such that the above criteria are satisfied during steady state operation, normal operational transients, and anticipated operational occurrences (AOOs).

To ensure that the RPS precludes the violation of the above criteria, additional criteria are applied to the Over Temperature and Overpower AT reactor trip functions.

That is, it must be demonstrated that the average enthalpy in the hot leg is less than or equal to the saturation enthalpy and that the core exit quality is within the limits defined by the DNBR correlation. Appropriate functioning of the RPS ensures that for variations in the THERMAL POWER, RCS Pressure, RCS average temperature, RCS flow rate, and AI that the reactor core SLs will be satisfied during steady state operation, normal operational transients, and AOOs.

1

INSERT 3 The numerical limits of these variables are contained in the COLR to provide operating and analysis flexibility from cycle to cycle.

However, the minimum RCS flow, based on previously analyzed maximum steam generator tube plugging for Unit 1 and the original licensed value for Unit 2, is retained in the TS LCO.

2

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs In MODES 1 and 2, the combination of THERMAL POWER, Reactor Coolant System (RCS) highest loop average temperature, and pressurizer pressure shall not exceed the specified in R

,o 6

,e 2.1.2 RCS Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained < 2735 psig.

2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

Catawba Units 1 and 2 2.0-1 Amendment Nos.)' V

SLs 2.0 6719\\/

DO NOT OPERATE IN THIS A A

660 650 640 ii\\:-

2400 psia

\\

630 F-620

/

a 610 600

/

a 590 ACCEPTABLE OPERATION 580 0.0 0.2 0.4 0.6 0.8 1.0 1.2 Fraction of Rated Thermal Power Figure 2.1.1-1 Reactor Core Safety Limits Four Loops in Operation Catawba Units 1 and 2 2.0-2 Amendment No.

Reactor Core SLs B 2.1.1 BASES APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and A0Os. The reactor core SLs are established to preclude violation of the following fuel design criteria:

a.

There must be at least 95% probability at a 95% confidence level (the 95/95 DNB criterion) that the hot fuel rod in the core does not experience DNB; and

b.

The hot fuel pellet in the core must not experience centerline fuel melting.

The Reactor Trip System setpoints (Ref. 2), in combination with all the LCOs, are designed to prevent any anticipated combination of transient P PC S FPI.OU fj.~Ir 5

coditions for Reactor Coolant System RCS) temperature ressure, and

' THE/A POWER level that would result in a departure rom nucleate boiling ratio (DNBR) of less than the DNBR limit and preclude the existence of flow instabilities.

of these reactor core SLs is provided by the a

v

~~A4f per1 A h 0 ej S; k c}a T:M Tt FS A4s ITg Strums;e x5^S y BOR64-ES.

The SLs represent a design requirement for establishing the RPS trip setpoints identified previously. LCO 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits,' or the assumed initial conditions of the safety analyses (as indicated in the UFSAR, Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

Catawba Units 1 and 2 B 2.1.1-2 Revision No.3t

Reactor Core SLs B 2.1.1 BASES

=

F oL 2-S AFETY LIMITS The 9(p~ provided in P fegWXA/T"show the loci of points ofI,&

RCS Pressure, and average temperature for which the minimum DNBR is F cro,

e of not less than the safety analyses limit, that fuel centerline temperature So 4 FOD A

W ex. ^

remains below melting, that the average enthalpy in the hot leg is less pow_

_J than or equal to the enthalpy of saturated liquid, and that the exit quality is within the limits defined by the DNBR correlation.

ThIcuesi Fire 1.11 eb eoe thqlpy ris h ch nel act (FH),

ref rece I

Z Ir;hae F x/4 tl a _

adpr

~ed HF orreatio adth Tch Il JSefiq tiM inihur$i flq~v te. The efor,,th se uspovi e is]

t alse

, ass t Z inal edattie abvefefren vaue vilb 0

. th cuesin tig re 2.1 1-1 ustgte the aus de gd to pr yend the nit rom/reachinb thg IinJ it.

APPLICABILITY SL 2.1.1 only applies in MODES 1 and 2 because these are the only MODES in which the reactor is critical. Automatic protection functions are required to be OPERABLE during MODES 1 and 2 to ensure operation within the reactor core SLs. The steam generator safety valves or automatic protection actions serve to prevent RCS heatup to the reactor core SL conditions or to initiate a reactor trip function, which forces the unit into MODE 3. Setpoints for the reactor trip functions are specified in LCO 3.3.1, Reactor Trip System (RTS) Instrumentation.

In MODES 3, 4, 5, and 6, Applicability is not required since the reactor is not generating significant THERMAL POWER.

SAFETY LIMIT If SL 2.1.1 is violated, the requirement to go to MODE 3 places VIOLATIONS the unit in a MODE in which this SL is not applicable.

The allowed Completion Time of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months recognizes the importance of bringing the unit to a MODE of operation where this SL is not applicable, and reduces the probability of fuel damage.

REFERENCES

1.

10 CFR 50, Appendix A, GDC 10.

2.

UFSAR, Section 7.2.

96/91EtMEl{

"Al VaO1t096 41 I

'4 Catawba Units 1 and 2 B 2.1.1-3 Revision No.Ai

Reactor Core SLs B 2.1.1 0.

0-I I

I I

S=

SafeV TAvc (of)

OTAT

ORE Figure B 2.1.1-1 Illustration of Overtemperature and Overpower AT Protection Catawba Units 1 and 2 B 2.1.1-4 Revision No ko

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 5 of 7)

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

AT (I +VIS)I

)

ATO JK1 -K2 0

[

4S T

I

]'+

K3(P-P) f(A)1

( + z S ( 1 + r3

)

IT r-K (a + '5 )I[T(1

_ Tf ]

K(PpiI Where:

AT is the measured RCS AT by loop narrow range RTDs, F.

ATo is the indicated AT at RTP, F.

s is the Laplace transform operator, sec'.

T.

VAIJS4.

T is the measured RCS average temperature, F.

T is the nominal T., at RTP (allowed by Safety Analysis),

lwbSfyys)a P is the measured pressurizer pressure, psig ve-Co P is the nominal RCS operating pressure, = Ml K,

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

=

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

=

Overtemperature AT reactor trip depressurization setpoint penalty coefficient, as presented in the COLR,

'r, 12

=

Time constants utilized in the lead-lag compensator for AT, as presented in the COLR, 13

=

Time constant utilized in the lag compensator for AT, as presented in the

COLR, T4, '

=

Time constants utilized in the lead-lag compensator for T., as presented in the COLR, 16

=

Time constant utilized in the measured T., lag compensator, as presented in the COLR, and f1(Al) = a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for q - qb between the "positive" and negative" f(Al) breakpoints as presented in the COLR; f(Al) = 0, where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + qb is total THERMAL POWER in percent of RATED THERMAL POWER; (ii) for each percent Al that the magnitude of q - q is more negative than the f(AI) "negative" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f(AI) "negative" slope presented in the COLR; and (continued)

Catawba Units and 2 3.3.1-18 Amendment Nos.M8A

RTS Instrumentation 3.3.1 Table 3.3.1-1 (page 6 of 7)

Reactor Trip System Instrumentation (iii) for each percent Al that the magnitude of q - qb is more positive than the f(Al) "positive" breakpoint presented in the COLR, the AT Trip Setpoint shall be automatically reduced by the f(Al) "positive" slope presented in the COLR.

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

T (r

~)<aTJ K 4-K 5 7

(____F I

AT (I+ V S.

_ AT(-Ksi)

-1

- K 6 [T T f 2 (AI)}

(1 +

2 S) 1+3s

+V7 s

l+r6 s

I+ +6 s

J 2

Where:

AT is the measured RCS AT by loop narrow range RTDs, F.

ATo is the indicated AT at RTP, F.

s is the Laplace transform operator, sec1.

T is the measured RCS average temperature, F.

T is the nominal Ta, at RTP (calibratiot urt n entation)

K4

= Overpower AT reactor NOMINAL TRIP SETPOINT as presented in the

COLR, K5

= kiaw for increasn average m

for decreasing average temperature,

,kF-v1sj_ 5 a-K6

= Overpower AT reactor trip heatup set cient as presented in the COLR for T > T and K6 =,for < T, T., 2

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

=

Time constant utilized in the lag compensator for AT, as presented in the

COLR, US

=

Time constant utilized in the measured Ta, lag compensator, as presented in the COLR, I,

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

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

Catawba Units 1 and 2 3.3.1-19 Amendment Nos.

RTS Instrumentation B 3.3.1 BASES APPLICABLE SAFETY ANALYSES, LCO, and APPLICAIWTY (continued) setpoints are based on Ar minimum R4d flow 416

.sps -,A, œ gi'o

'c OL?

A19irf*M~

Each RCS loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE in MODE 1 above P-B.

In MODE 1 above the P-8 setpoint, a loss of flow in one RCS loop could result in DNB conditions in the core. In MODE 1 below the P-B setpoint, a loss of flow in two or more loops is required to actuate a reactor trip (Function 1O.b) because of the lower power level and the greater margin to the design limit DNBR.

b.

Reactor Coolant Row-Low (Two Loops)

The Reactor Coolant Flow-Low (Two Loops) trip Function ensures that protection is provided against violating the DNBR limit due to low flow in two or more RCS loops while avoiding reactor trips due to normal variations in loop flow.

Above the P-7 setpoint and below the P-B setpoint, a loss of flow in two or more loops will initiate a reactor tri.

The 6

)

zJ Co'2WO setpoints are based oru~Xninimum pW.Af4v flow Each loop has three flow detectors to monitor flow. The flow signals are not used for any control system input.

The LCO requires three Reactor Coolant Flow-Low channels per loop to be OPERABLE.

In MODE 1 above the P-7 setpoint and below the P-8 setpoint, the Reactor Coolant Flow-Low (Two Loops) trip must be OPERABLE. Below the P-7 setpoint, all reactor trips on low flow are automatically blocked since power distributions that would cause a DNB concern at this low power level are unlikely. Above the P-7 setpoint, the reactor trip on low flow in two or more RCS loops is automatically enabled. Above the P-8 setpoint, a loss of flow in any one loop will actuate a reactor trip because of the higher power level and the reduced margin to the design limit DNBR.

Catawba Units 1 and 2B B 3.3.1-18 Revision No.,(

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 APPLICABILITY:

RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified in Table 3.4.1-1.

MODE 1.

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

Pressurizer pressure limit does not apply during:

a.

THERMAL POWER ramp > 5% RTP per minute; or

b.

THERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

Pressurizer pressure or A.1 Restore DNB parameter(s) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months RCS average to within limit.

temperature DNB parameters not within limits.

B.

RCS total flow rate

> cl1)e Ba Be 1 la3_i7 4 10 9c3 oF '?40.

t'.es t

< 5pgc IS,@ As4E~

CQL.R B.1 Reduce THERMAL POWER to s 98% RTP.

AND B.2 Reduce the Power Range Neutron Flux - High Trip Setpoint below the nominal setpoint by 2% RTP.

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 6 hours (continued)

Catawba Units 1 and 2 3.4.1 -1 Amendment Nos.MM

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C.

RCS total flow rate

/'4 S%

of JC I/ALJE synca F

74 o-T

\\=CL-9 C.1 Restore RCS total flow rate tot OR S1e:

VAI.-JP C.2.1 e uce RMAL POWER to < 50% RTP.

AND C.2.2 Reduce the Power Range Neutron Flux - High Trip Setpoint to < 55% RTP.

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

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 6 hours 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months I

3 i

D.

D.1 Be in MODE 2.

Required Action and associated Completion Time not met.

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Catawba Units 1 and 2 3.4.1-2 Amendment Nos.

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

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.2 Verify RCS average temperature is within limits.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.3 Verify RCS total flow rate is within limits.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.4 Perform CHANNEL CALIBRATION for each RCS total 18 months flow indicator.

Catawba Units 1 and 2 3.4.1-3 Amnendmnent Nos.

W

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 Table 3.4.1-1 (page 1 of 1)

RCS DNB Parameters

.' 4 rdf -

C O01.)g ERABLE 4NELS LIMITS No. OP PARAMETER INDICATION CHAl'

1.

Indicated RCS Average Temperature - Unit 1 meter meter 4

3 computer computer 4

3

<ad

< Wlir4d

&.0

< W; &-

6b YWV*

' 2m{yA1v I

I computer computer 4.

3

3.
RCS Total Flow Rate
/Alp6d1w <-
.738S,10Qf, A
G6zATVJ Tt4 "rt Tapi va'1
/
Is IFE Le CoJ nil,,r Catawba Units 1 and 2 3.4.1-4 Amendment Nos4
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABLE SAFETY ANALYSES (continued)
/
'" =o The pressurizer pressure limits and the RCS average temperature limits correspond to analytical l AIW A~0-used in the safety analyses, with allowance for measurement uncertainty.
The RCS DNB parameters satisfy Criterion 2 of 10 CFR 50.36 (Ref. 2).
LCO This LCO specifies limits on the monitored process variables-pressurizer pressure, RCS average temperature, and RCS total flow rate-to ensure the core operates within the limits assumed in the safety analyses.
CJ Operating within these limits will result in meeting the acceptance criteria, including the DNBR criterion.
RCS total flow rate contains a measurement error A1Y4)V9 based on the performance of past precision heat balances and using the result to calibrate the RCS flow rate indicators. Sets of elbow tap coefficients, as determined during these heat balances, were averaged for each elbow tap to provide a single set of elbow tap coefficients for use in calculating RCS flow. This set of coefficients establishes the calibration of the RCS flow rate indicators and becomes the set of elbow tap coefficients used for RCS flow measurement. Potential fouling of the feedwater venturi, which might not have been detected, could have biased the result from these past precision heat balances in a nonconservative manner. Therefore, a penalty 0&,W/ for undetected fouling of the feedwater venturi raises the nominal flow measurement allowance W,9for no fouling.
The numerical valuesp 1WA44for pressure and average P5 c X-~
f s
temperaturfre given for the measurement location with adjustments for c - XW the indication instruments.
APPLICABILITY In MODE 1, the limits on pressurizer pressure, RCS coolant average temperature, and RCS flow rate must be maintained during steady state operation in order to ensure DNBR criteria will be met in the event of an unplanned loss of forced coolant flow or other DNB limited transient. In all other MODES, the power level is low enough that DNB is not a concern.
A Note has been added to indicate the limit on pressurizer pressure is not applicable during short term operational transients such as a THERMAL POWER ramp increase > 5% RTP per minute or a THERMAL POWER step increase > 10% RTP. These conditions represent short term perturbations where actions to control pressure variations might be counterproductive. Also, since they represent transients initiated from power levels < 100% RTP, an increased DNBR margin exists to offset the temporary pressure variations.
Catawba Units 1 and 2 B 3.4.1-2 Revision No.>,
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES APPLICABILITY (continued)
I eiW JKt~IXIaJw~dR is provided in SL 2.1.1, T Reactor Core SLs." Those limits are less restrictive than the limits of this LCO, but violation of a Safety Limit (SL) merits a stricter, more severe Required Action. Should a violation of this LCO occur, the operator must check whether or not an SL may have been exceeded.
ACTIONS A.1 Pressurizer pressure and RCS average temperature are controllable and measurable parameters. With one or both of these parameters not within 0 D LCO limits, action must be taken to restore parameter(s).
L J The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for restoration of the parameters provides I z sufficient time to adjust plant parameters, to determine the cause for the off normal condition, and to restore the readings within limits, and is based on plant operating experience.
_ X B.1 and B.2 RCS total flow rate is not a controllable parameter and is not expected to v
i vary during steady state operation. If the indicated RCS total flow rate is 4
t l
,m~d-dy!8380!,,
then THERMAL POWER may not exceed 98% RTP. THERMAL POWER must be reduced within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .
The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is consistent with Required Action A.1.
In addition, the Power Range Neutron Flux - High Trip Setpoint must be t
l
/
reduced from the nominal setpoint by 2% RTP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to reset the trip setpoints recognizes that, with power reduced, the safety analysis assumptions are satisfied and there is no urgent need to reduce the trip setpoints. This is a sensitive operation that may inadvertently trip the Reactor Protection System.
~~~~~~C.1, C.2.1. C.2.2, and C.2.3 ktiv AG-atA
&6R47MZ TV a If the indicated RCS total flow rate is,1 then RCS total flow
?i4JA.,- 'C G
9 4
must be restored t
,A
.Pwithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or power must be
]
Te './A4V-freduced to less than 50% RTP. The Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is I P9 e-I P ASO
J y consistent with Required Action A.1. If THERMAL POWER is reduced to less than 50% RTP, the Power Range Neutron Flux - High Trip Setpoint Catawba Units 1 and 2 B 3.4.1-3 Revision No.3l
RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES ACTIONS (continued) must also be reduced to < 55% RTP. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months to
,i6SZ. T ILc)Ot~
<reset the trip setpoints is consistent with Required Action B.2. This is a lE4QJA L
-x) sensitive operation that may inadvertently trip the Reactor Protection
\\1)V~Ific
~System.
Operation is permitted to continue provided the RCS total flow is 7*--
-rsr tct se si within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The Completion Time of 24
,.~.J 7<Z COL 4?.
hours is reasonable considering the increased margin to DNB at power levels below 50% and the fact that power increases associated with a transient are limited by the reduced trip setpoint.
D.1 If the Required Actions are not met within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 2 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable to reach the required plant conditions in an orderly manner.
SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This surveillance demonstrates that the pressurizer pressure remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS pressure and related equipment status. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.
SR 3.4.1.2 This surveillance demonstrates that the average RCS temperature remains within the required limits. Alarms and other indications are available to alert operators if this limit is approached or exceeded. The frequency of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, considering the other indications available to the operator in the control room for monitoring the RCS. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operating practice to be sufficient to regularly assess for potential degradation and to verify operation is within safety analysis assumptions.
Catawba Units 1 and 2 B 3.4.1-4 Revision Noy\\
Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)
I. TL,*.
7 -,.)
~
~
Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3.1.3, 5e~ny Ace As
\\;-e Shutdown Bank Insertion Limit for Specification 3.1.5, L7
).
Control Bank Insertion Limits for Specification 3.1.6, S.
Axial Flux Difference limits for Specification 3.2.3, Heat Flux Hot Channel Factor for Specification 3.2.1,
7. E.
Nuclear Enthalpy Rise Hot Channel Factor for Specification 3.2.2, Overtemperature and Overpower Delta T setpoint parameter values for Specification 3.3.1,
/ ~
{° Accumulator and Refueling Water Storage Tank boron concentration limits for Specification 3.5.1 and 3.5.4, Reactor Coolant System and refueling canal boron concentration limits for Specification 3.9.1, ZO.
Spent fuel pool boron concentration limits for Specification 3.7.15, 13, t SHUTDOWN MARGIN for Specification 3.1.1.
b.
The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:
1.
WCAP-9272-P-A, WESTINGHOUSE RELOAD SAFETY EVALUATION METHODOLOGY" &~P Proprietary).
2.
WCAP-10266-P-A, "THE 1981 VERSION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE' (W Proprietary).
(continued)
Catawba Units 1 and 2 5.6-3 Amendment Nos. 14W14T
ATTACHMENT 2A McGUIRE UNITS 1 AND 2 TECHNICAL SPECIFICATIONS AND TECHNICAL SPECIFICATION BASES (TO BE PROVIDED TO THE NRC UPON ISSUANCE OF APPROVED AMENDMENT)
ATTACHMENT 2B CATAWBA UNITS 1 AND 2 TECHNICAL SPECIFICATIONS AND TECHNICAL SPECIFICATION BASES (TO BE PROVIDED TO THE NRC UPON ISSUANCE OF APPROVED AMENDMENT)
ATTACHMENT 3 DESCRIPTION OF PROPOSED CHANGES AND TECHNICAL JUSTIFICATION
DESCRIPTION OF PROPOSED CHANGES AND TECHNICAL JUSTIFICATON Proposed Changes The proposed changes reduce the required minimum measured Reactor Coolant System (RCS) flow rate from 390,000 gpm to a previously reviewed and approved value of 382,000 gpm for McGuire Units 1 and 2 and Catawba Unit 1; reduces the required minimum measured Reactor Coolant System (RCS) flow rate from 390,000 gpm to the originally licensed value of 385,000 gpm for Catawba Unit 2; and relocate RCS related cycle-specific parameter limits from the Technical Specifications (TS) to, and thus expand, the Core Operating Limits Reports (COLR) for the McGuire and Catawba Nuclear Stations.
McGuire and Catawba Units 1 and 2 were originally licensed with a minimum reactor coolant system (RCS) flow rate of 385,000 gpm.
Due to steam generator tube plugging issues, analyses were performed for McGuire and Catawba Units 1 and 2 assuming a bounding tube plugging percentage and a license amendment to reduce the RCS minimum flow rate from 385,000 gpm to 382,000 gpm was requested for McGuire Units 1 and 2 and Catawba Unit 1. Although these analyses were applicable to Catawba Unit 2, a license amendment to lower the RCS minimum flow rate was not requested for this unit because Catawba Unit 2's steam generators had not experienced the same rate of tube plugging as had Catawba Unit 1. This license amendment request was approved by the NRC staff in Safety Evaluation Reports dated December 17, 1993 (for Catawba Unit 1)1 and March 22, 1994 (for McGuire Units 1 and 2)2.
Subsequent to the replacement of steam generators, McGuire Unit 1 and 2 and Catawba Unit 1 were successfully operated with an RCS minimum total flow rate of 382,000 gpm in the technical specifications until those values were changed to 390,000 gpm through NRC Safety Evaluation Reports of March 2, 2000 for McGuire, and March 1, 2000 for Catawba.
The NRC Safety Evaluation Report of March 1, 2000, also revised 1 Reference Duke submittal of October 25, 1993, as supplemented by letters of December 3 and 6, 1993 2 Reference Duke submittal of October 25, 1993, as supplemented by letters of December 3, 1993, and February 14, 1994 1
the Catawba Unit 2 RCS minimum total flow rate from 385,000 gpm to 390,000 gpm in the technical specifications.
The analyses supporting the RCS minimum total flow rate of 390,000 gpm assumed a minimal steam generator tube plugging percentage.
The RCS minimum total flow rates for McGuire and Catawba Units 1 and 2 were increased to make more effective use of available operating and analytical margins.
This 390,000 gpm RCS total flow rate should be considered a cycle-specific minimum value, reflecting the condition of the McGuire and Catawba steam generators at the time the license amendment request was made.
Values for the RCS pressure, temperature, and flow are specified in TS 3.4.1, RCS Pressure, Temperature, and Flow Departure for Nucleate Boiling (DNB) Limits."
The proposed change reduces the required minimum measured RCS flow rate for McGuire Units 1 and 2, and Catawba Unit 1 from 390,000 gpm to 382,000 gpm; and reduces the required minimum measured RCS flow rate for Catawba Unit 2 from 390,000 gpm to the originally licensed value of 385,000 gpm, and further replaces the current minimum RCS flow rate of 390,000 gpm with a reference to the COLR.
The previously reviewed and approved accident evaluation of the replacement steam generators at a minimum RCS flow rate of 382,000 gpm was included as Attachment 1 to the McGuire Unit 1 and 23, and Catawba Unit 14 steam generator replacement license amendment requests of September 30, 1994.
The amendment also revises TS 3.4.1, RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)
Limits," and associated Bases, by relocating the pressurizer pressure, RCS average temperature, and RCS total flow rate values to the respective station COLR.
The 3 Reference Duke submittal of September 30, 1994, as supplemented by letters of September 18, 1995, and March 15, April 29, May 16, September 23, and October 28, 1996, and January 16, April 22, and May 2, 1997, and the NRC Safety Evaluation Report dated May 5, 1997, for McGuire Units 1 and 2
' Reference Duke submittal of September 30, 1994, as supplemented by letters of September 18, 1995, January 19, March 15, May 16, and August 27, 1996; and the NRC Safety Evaluation Report of August 29, 1996, for Catawba Unit 1 2
minimum RCS total flow rates of 382,000 gpm for McGuire Units 1 and 2, and Catawba Unit 1; and 385,000 gpm for Catawba Unit 2, will be maintained in TS Table 3.4.1-1 so as to assure that a lower flow rate will not be used without prior NRC approval.
TS 5.6.5, "Core Operating Limits Report (COLR),N has been modified to reflect the above relocations to the COLR.
The proposed amendment also relocates TS Figure 2.1.1-1, "Reactor Core Safety Limits," to the COLRs, replacing it with more specific requirements regarding the safety limits (i.e. fuel DNB design basis and the fuel centerline melt design basis) conforming with WCAP-14483-A.
As discussed in the Safety Evaluation Reports to WCAP-14483-A, it is necessary to relocate TS Figure 2.1.1-1 to the COLR since cycle-dependent changes to parameters upon which TS Figure 2.1.1-1 is based would require a license amendment request to revise the figure.
The amendment also revises TS Table 3.3.1-1, "Reactor Trip System Instrumentation,, by relocating numerical values pertaining to Overtemperature AT and Overpower AT nominal RCS operating pressure, nominal T, and constant (K) values to the COLR. Additionally, the basis for the RCS low flow reactor trip setpoints has been revised by referring to the COLR in the associated Bases document.
TS 5.6.5, "Core Operating Limits Report (COLR),f will be modified to reflect the above relocations to the COLR.
The proposed changes will allow Duke the flexibility of enhancing operating and core design margins without the need for cycle-specific license amendment requests.
The relocation of these cycle-specific TS values to the COLR will result in a more complete COLR containing cycle-specific operating conditions and core reload related parameters.
The safety and quality of operations at Duke's McGuire and Catawba nuclear stations will not be compromised by the implementation of this amendment request as TS 5.6.5(c) requires that all applicable limits of the safety analyses be met when generating cycle-specific requirements in the COLR.
3
Basis for Proposed Change NRC Generic Letter 88-16, "Removal of Cycle-Specific Parameters From Technical Specifications," dated October 4, 1988, provides guidance to licensees for the removal of cycle-dependent variables from the TS provided that these values are included in a COLR and are determined with NRC-approved methodologies referenced in the TS.
Westinghouse Electric Company (Westinghouse) subsequently developed WCAP-14483, "Generic Methodology for Expanding Core Operating Limits Report," describing how cycle-specific parameters may be relocated to the COLR.
WCAP-14483 was accepted for referencing by the NRC on January 19, 1999.
The Safety Evaluation Report, contained in the January 19, 1999 NRC letter approving WCAP-14483-A, concludes that additional information contained in the TS may be relocated to the COLR.
The limits on the parameters which are removed from the Technical Specifications and added to the COLRs must be developed or justified using NRC-approved methodologies.
All accident analyses, performed in accordance with these methodologies, must meet the applicable NRC-approved limits of the safety analysis.
The removal of parameter limits from the Technical Specifications and their addition to the COLRs does not obviate the requirement to operate within these limits.
Furthermore, any changes to those limits must be performed in accordance with TS 5.6.5(c).
If any of the applicable limits of the safety analyses are not met, prior NRC approval of the change is required, just as is the case for a license amendment request.
For more routine modifications, where NRC-approved methodologies and limits of the safety analysis remain applicable, the potentially burdensome and lengthy process of amending the Technical Specifications may be avoided.
The requested changes are essentially administrative in nature; therefore, the required level of safety will be maintained.
The requested changes are based upon NRC approved Westinghouse Owners Group (WOG) Technical Specifications Task Force (TSTF) TSTF-339, "Relocate TS Parameters to the COLR Consistent with WCAP-14483,0 Revision 2, and Westinghouse WCAP-14483-A.
In accordance with these documents, previously approved RCS minimum total flow rates are retained in the TS to preclude the use of lower flow 4
rates without prior NRC approval.
The return to a lower RCS minimum flow rate, previously reviewed and approved by the NRC, is similar to that submitted by the Comanche Peak Nuclear Station on May 24, 1999, and approved by the NRC in the Safety Evaluation Report dated August 30, 1999.
Consistent with the Comanche Peak submittal is a generic statement in TS 2.1.1 requiring compliance with the DNBR limit for the correlation(s) used for a specific core design.
The same approach has also been made with respect to the CFM limits.
Both of these statements more clearly address the requirements of 10CFR §50.36 by stating the actual safety limits of DNB and CFM.
Future changes to TS 2.1.1 are also minimized.
5

v t e Letter Sequence Request
TAC:MB8359, Relocate TS Parameters to Colr (Open)
Initiation Request, Request Acceptance... Administration Questions 2 October 2003 Answers Results Approval... Other: ML033140352, ML033570127, ML033580393
MONTHYEAR2003-11-06 [Table View]

ML032130497

Text

Subject:

Time constants utlized in the lag compensator for AT, as presented in the COLR x4. xs

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