Text

Amends 40 & 33 to Licenses NPF-35 & NPF-52,respectively, Modifying Tech Specs to Accommodate Removal of Resistance Temp Device (RTD) Bypass Manifold Sys & Installation of in Line RTDs
	ML20147D160
Person / Time
Site:	Catawba
Issue date:	02/17/1988
From:	Crocker L; Office of Nuclear Reactor Regulation
To:	;
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References
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- _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _

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NUCLEAR REGULATORY COMMISSION

'l WASHINGTON, D. C. 20555

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DUKE PCWER COMPANY NORTH CAROLINA ELECTRIC MEMBERSHIP CORPORATION SALUDA RIVER ELECTRIC COOPERATIVE, INC.

DOCKET NO. 50-413 CATAWBA NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 40 License No. NPF-35 1.

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

A.

The application for amendment to the Catawba Nuclear Station, Unit 1 (the facility) Facility Operating License No. NPF-35 filed by the Duke Power Company acting for itself, North Carolina Electric Membership Corporation and Saluda River Electric Cooperative, Inc., (licensees) dated July 22, 1987, and supplemented May 26, August 31, October 1, October 30, November 19 and December 14, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisiens of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) 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 Cennission's regulations set forth in 10 CFR Chapter I; D.

The issuance of this amendtrent will not be inimical to the comon l

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 Comission's regulations and all applicable requirerents have been satisfied.

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

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(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 40, and the Environmental Protection Plan contained in Appendix B both of which are attached hereto, are hereby incorporated into the license.

Duke Power Company shall operate the facility in accordance with sha Technical Specifications and the Environmental Protection Plan.

3.

This license amendnent is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION w

W Lawrence P. Crocker

, Acting Director Project Directorate 11-3 Division of Reactor Projects I/II

Attachment:

Technical Specification Changes 4

Date of Issuance: February 17, 1988 l

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UNITED STATES 6

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NUCLEAR FIEGULATORY COMMISSION 5-

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DUKE POWER COMPANY NORTH CAROL MA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET NO. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Arendment No. 33 License No NPF-52 1.

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

A.

The application for amendment to the Catawba Nuclear Station, Unit 2 (the facility) Facility Operating License No. NPF-52 filed by the Duke Power Company acting for itself, North Carolina Municipal Power Agency No. 1 and Piedmont Municipal Power Agency, (licensees) dated July 22, 1987, as supplemented May 26. August 31, October 1, October 30, November 19, and December 14, 1987, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (1) 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 Comission's regulations set forth in 10 CFP, Chapter I; D.

The issuance of this amendment will net be inimical to the comen defense and security or to the health and safety of the public; and E.

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

2.

Accordingly, the license is hereby arended by page changes +,o the Technical Specifications as indicated in the attachments to this license amendmerit, i

and Paragraph 2.C.(2) of Facility Operating License No. NPF-52 is hereby amended to read as follows:

2 (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Arrendment No. 33, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license. Duke Power Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This ifcense amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGt!LATORY COMMISSION Lawrence P. crocker

. Acting Director Project Directorate 11-3 Division of Reactor Projects I/II

Attachment:

Technical Specification Changes Date of Issuance:

February 17, 1988 9

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. (2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No, 40, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into the license.

Duke Power Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

, Acting Director Project Directorate II-3 Division of Reactor Pro,iects I/II

Attachment:

Technical Specification Changes Date of Issuance:

February 17, 1988 sc/ttsflk L. f. c y

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ATTACHMENT TO LICENSE AMENDMENT NO.

40 FACILITY OPERATING LICENSE NO. NPF-35 00CXET N0. 50-413 E

TO LICENSE AMENDMENT NO. 33

, FACILITY OPERATING LICENSE NO. NPF-52 COCKET NO. 50-414 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed paces. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.

i Amended Overleaf fini Page l

24 P-3 2-7 2-8 2-10 2-9 8 2-5 B 2-6 3/4 3-1 3/4 3-7 3/4 3-29 3/4 3-30 3/4 3-35 3/4 3-36

--r

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 4

2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.3-1.

ACTION:

a.

With a Reactor Trip System Instrumentation or Interlock Setpoint

.less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value Column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.

b.

With the Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 2.2-1 and determine within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its Setpoint adjusted consistent with the Trip Setpoint value.

Equation 2.2-1 Z+R+S i TA i

Where:

Z=

The value from Column Z of Table 2.2-1 for the affected channel, R=

The "as measured" value (in percent span) of rack error for the affected channel, S=

Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Table 2.2-1 for the affected channel, and l

TA = The value from Column TA (Total Allowance) of Table 2.2-1 for the affected channel.

i

{

t CATAWBA - UNITS 1 & 2 2-3 I

,~.

I

.g TABLE 2.2.-1 y

REACTOR TRIP SYSTEM INSTRUNENTATION TRIP SETPOINTS oTi TOTAL SENSOR ALLOWANCE ERROR

=

g FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE 3

1.

Manual Reactor Trip M.A.

N.A.

w 2.

Power Range, Neutron Flux e-a.

High Setpoint

7. 5 4.56 0

$109% of RTP*

$111.1% of RTP*

b.

Low Setpoint 8.3 4.56 0

125% of RTP*

$27.1%'of RTP*

3.

Power Range, Neutron Flux,

1. 6 0.5 0

$5% of RTP* with 16.3% of RTP" with-High Positive Rate a time constant a time constant

> 2 seconds I

> 2 seconds 4.

Power Range, Neutron flux,

'6 0.5 0

15% of RTP* with 16.3% of RTP* with High Negative Rate a time constant a time constant

>2 seconds

>2 seconds 2

5.

Intermediate Range, 17.0 8.4 0

125% of RTP*

$31% of RTP*

Neutron Flux-i 6.

Source Range, Neutron Flux 17.0 10 0

$105 cps

$1.4 x 10 cps 5

7.

Overtemperature AT 7.2(8.9 )

4.47(5.41 )

2.03(2.65 ) -See Note 1 See Note 2 8.

Overpower AT 4.3(4.9 )

1.3(1.24 )

1.2(1.7 )

See Note 3 See Note 4 9.

Pressurizer Pressure-Low 4.0 2.21 1.5

>1945 psig

>1938 psig***

10.

Pressurizer Pressure-High 7.5

4. %

0.5 52385 psig

$2399 psig

((

11.

Pressurizer Water Level-High 5.0 2.18

1. 5

$92% of instrument $93.8% of instrument gg span span-

??

~>90% of loop

>89.2%(88.M)ofloog!

zr 12.

Reactor Coolant Flow-Low 2.5 1.77(1.41 )

0.6 I

design. flow **

i$esign flow **

U$

22 "RIP = RAltp THERMAL POWER 11

- Loop design flow = %,900 gpa

- - Time constants utilized in the lead-lag controller for Pressurizer Pressure-Low are 2 seconds for 'ead 00 and 1 second for lag.

Channel calibration shall ensure that these time constants are adjusted to'these j

values.

l

Applicable upon deletion of RTD Bypass System.

1 s

~....... -

.m..

TABLE 2.2-1 (Continued) n h

TABLE NOTAIl0NS E

NOTE 1: OVERTEMPERATURE AT 1

{t Il + T S} I ^ o (Kr - K AT

[T (3 g) - T'] + K (P - P') - f (AI)}

2 3

3 d

Where:

AT Measured AT by RTD Manifold Instrumentation;

=

1+t S Lead-lag compensator on measured AT; e

=

1+T S 2

ro 3 = 8(12 ) s, l

ti, r2 Time constants utilized in lead-lag compensator for AT, 1

=

12 = 3 s; 1

Lag compensator on measured AT;

=

y, 3

T3 Time constant utilized in the lag compensator for AT, r3 = 0;

=

ro,',

AT, Indicated AT at RATED THERMAL POWER;

=

1.411(1.38 );

K

=

2 0.02401/*F; K

=

f{

The function generated by the lead-lag compensator for T,yg

=

dynamic compensation; "k"

Time constants utilized in the lead-lag compensator for T,yg,1 4 = 28(22 ) s,

=

T4, is ts = 4 S; 55 T

' Average temperature, *F;

=

zz

.o.o Lag compensator on measured T,yg; gg

=

y, 3

22 Time constant utilized in the measured T lag compensator, is = 0;

=

Is er avg 00

.e.-.

..r

IABLE 2.2-1 (Continued) n D

TABLE NOTATIONS (Continued)

E E

NOTE 1:

(Continued)

I' z

- 590.8 F (Nominal T allowed by Safety Analysis);

avg d

K

=

3 0.001189; Pressurizer pressure, psig; P

=

p.

~

P' 2235 psig (Nominal RCS operating pressure);

=

Laplace transform operator, s 1; 5

=

and f (AI) is 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 q between -22.5% and -6.5%,

t b

tr f (AI) = 0, where q and g are percent RATED THERMAL POWER in the top and bottom 3

i halves of the core respectively, and q + g is total THERMAL P0hTR in percent of RATED THERMAL POWER; (ii)

For each percent that the magnitude of q q is a re negative than -22.5%, the b

gg AT Trip Setpoint shall be automatically reduced by 3.151% of its value at RATED THERMAL POWER; and sa kk g3 (iii)

For each percent that the magnitude of q q is a re positive than -6.5%, the AT Trip g

b gg Setpoint shall be automatically reduced by 1.641% of its value at RATED THERMAL POWER.

U$

The channel's maxim Trip Setpoint shall not exceed its computed Trip Setpoint by gg NOTE 2:

more than 2.4%(3.0%p).

l 1a ee

TABLE 2.2-1 (Continued) n h

TABLE N01ATIONS (Continued) r$

NOTE 3: OVERPOWER AT g

AT (1 + t,5) (

1

)

(1 + r25) (1 + r35) < 37 gg* _ g* (1 + 1 (1 + tsS) T - Ks [T ((1 + te )) - T"] - f (al)I 5

) (

1

)

1 12 5) s 2

0

(

7 v.

As defined in Note 1, s.

Where:

AT

=

1*

As defined in Note 1,

=

7 As defined in Note 1,

=

ti, 12 I

As defined in Note 1,

=

y g

As defined in Note 1,

=

7 r3 e

As defined in Note 1, AT,

=

1.070',,

K

=

4 3

0,'J2/*F for increasing average temperature and 0 for decreasing average F

=

temperature, D

The function generated by the rate-lag controller for T,,g dynamic

=

y S

compensation, Time Constant utilized in the rate-lag Controller for T,yg, 17= 10 s,

=

17 I

As defined in Note 1,

=

y, Ts3 As defined in Note 1,

=

Is

i TABLE 2.2-1 (Continued) r3 l

y TABLE NOTATIONS (Continued) 2 NOTE 3:

(Cont'nued) e 0.001707/ F for T > 590.8 F and Ks = 0 for T $ 590.8*F, Ks

=

c:3 Dl T

As defined in Note 1,

=

T" Indicated T,,g at RATED THERMAL POWER (Calibration temperature for AT

=

^*

instrumentation, 5 590.8'F),

S

=

As defined in Note 1, and f (AI) 0 for all AI.

=

2 NOTE 4:

The channel's maxi y Trip Setpoint shall not exceed its computed Trip Setpoint by S'

more than 2.6%(2. W ).

[.

o

$.5 kk 55 Y.f U$

22

3. "

<* e

Applicable upon deletion of RTD Bypass System.

l m,

vv

a LIMITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core pro-tection during reactor startup to mitigate the consequences of an uncontrolled i

rod cluster control assembly bank withdrawal from a suberitical condition.

These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux channels.

The Source Range channels will initiate a Reactor trip at about 105 counts per second unless manually blocked when P-6 becomes active or automatically blocked when P-10 becomes active.

The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Overtemperature AT The Overtemperature AT trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to (1) (with l

the RTO Bypass System installed) piping transit delays from the core to the temperature detectors (about 4 seconds), or (2) (with the RTO Bypass System removed) thermal delays associated with the RTDs mounted in thermowells (about 5 seconds)) and pressure is within the range between the Pressurizer High and t

Low Pressure trips.

The Setpoint is automatically varied with:

(1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer pressure, and (3) axial power distribution.

With normal axial power distribution, this Reactor trip limit is always below the core Safety Limit as shown in Figure 2.2-1.

If axial peaks are greater than design, as indicated by the difference between top and 4

bottom power range nuclear detectors, the Reactor trip is automatically reduced

{

according to the notations in Table 2.2-1.

Overpower aT The Overpower AT trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible 4

overpower conditions, limits the required range for Ovsetemperature AT trip, and provides a backup to the High Neutron Flux trip.

The Setpoint is auto-matically varied with:

(1) coolant temperature to correct for temperature-induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for either piping delays from the core to the loop temperature detectors (with the RTO Bypass System installed), or instrumentation delay associated with the loop temperature detectors (with the i

RTO Bypass System removed), to ensure that the allowable heat generation rate (kW/ft) is not exceeded.

The Overpower AT trip provides protection to mitigate 1

the consequences of various size steam breaks as reported in WCAP-9226, "Reac-l tor Core Response to Excessive Secondary Steam Releases."

CATAWBA - UNITS 1 & 2 B 2-5 Amendment No. 40 (Unit 1)

Amendment No. 33 (Unit 2) 1

LIMITING SAFETY SYSTEM SETTINGS BASES Pressurizer Pressure In each of the pressurizer pressure channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against low pressure which could lead to DNS by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, is automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves.

On decreasing power, the Pressurizer High Water Level trip is automatically blocked by P-7 (a level of approxi-mately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, is automatically reinstated by P-7.

Reactor Coolant flow The Low Reactor Coolant Flow trips provide core protection and prevents DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%

of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow.

Above P-8 (a power level of approximately 48% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.

Conversely, on decreasing power between P-8 and P-7 an automatic Reactor trip will occur on low reactor coolant flow in more than one loop and below P-7 the trip function is automatically blocked.

CATAWBA - UNITS 1 & 2 8 2-6

3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a 91nimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months."

Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.

The response time of RTDs associated with the Reactor Trip System shall be demonstrated to be within their limits at least once per 18 months.

t

This surveillance need not be performed for the primary RTD response time testing portion of items 7 and 8 from Table 3.3-2 for Unit 2 until prior to entering STARTUP following the Unit 2 first refueling.

CATAWBA - UNITS 1 & 2 3/4 3-1 Amendment No. 40 (Unit 1)

Amendment No. 33 (Unit 2)

1 1

TABLE 3.3-2 l

REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES s

n-FUNCTIONAL UNIT RESPONSE TIME 1.

Manual Reactor Trip N.A.

~

2.

Power Range, Neutron Flux

< 0.5 second*

e-3.

Power Range, Neutron Flux, m

High Positive Rate N.A.

4.

Power Range, Neutron Flux, High Negative Rate

< 0.5 second*

5.

Intermediate Range, Neutron Flux N.A.

R 6.

Source Range, Neutron Flux N.A.

T 7.

Overtemperature AT

$ 4(8 ) seconds

l w

8.

Overpower AT

$ 4(8 ) seconds l

9.

Pressurizer Pressure-Low

< 2 seconds 10.

Pressurizer Pressure-High FF

-< 2 seconds 3@

11.

Pressurizer Water Level-High N.A.

kk 55 55 U$

22

Neutron detectors are exempt from response time testing.

Response time of the neutron flux signal portion

3. 3.

of the channel shall be measured from detector output or input of first electronic component in channel.

Applicable upon deletion of RTD Bypass System.

00 g

TABLE 3.3-4 (Continued) b ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETPOINT ALLOWABLE VALUE gg 1

4. Steam Line Isolation v

a.

Manual Initiation N.A.

N.A.

b.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays c.

Containment Pressure-High-High 12.7 0.71

1. 5 5 3 psig

$ 3.2 psig d.

Steam Line Pressure - Low 4.6 1.31 1.5 1 725 psig 1 694 psig*

e.

Steam Line Pressure-8.0 0.5 0

< 100 psi

$ 122.8 psi **

Negative Rate - High S. Feedwater Isolation u

u, a.

Automatic Actuation Logic N.A.

N.A.

4, Actuation Relays

<a b.

Steam Generator Water Level-High-High (P-14) 1.

Unit 1 S.4 2.18 1.5

< 82.4% of

< 84.2% of narrow iiarrow range range instrument instrument span

.BI.$I Span

" 8.

2.

Unit 2 9.7 2.18

1. 5 1

8. EE of 5 79.9% of narrow 7

k2 narrow range range instrument 3a instrument span 2e 2e span

(,$,

T,yg-Low 4.0(6.0 )

1.12(0.71 )

1.2(0.8 ) 1 564*F 1 562*F(561*F )

l c.

d.

Doghouse Water Level-High

1. 0 0

0.5 11 inches 12 inches above 577' above 577' EL EL floor level floor level e.

Safety Injection See Item 1. above for all Safety Inject on Setpoints and Allowable Values.

i

}

=- -

TABLE 3.3-4 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATIDN TRIP SETPOINIS r

52 SENSOR TOTAL ERROR 25 (S)

TRIP SETPOINT ALLOWABLE VALUE FUNCTIONAL UNIT ALI.0WANCE (TA) Z c:

d

6. Turbine Trip a.

Manual Initiation N.A.

N.A.

e b.

Automatic Actuation N.A.

N.A.

na Logic and Actuation Relays c.

Steam Generator Water Level-High-High (P-14) 1.

Unit 1 S.4 2.18

1. 5

< 82.4% of

< 84.Z% of narrow harrow range range instrument instrument span span

}{

2.

Unit 2 9.7 2.18 1.5

<78.1% of

<79.9% of narrow u,

J, harrow range range instrument instrument span span d.

Trip of All Main M.A.

N.A.

M.A.

Feedwater Pumps e.

Reactor Trip (P-4)

N.A.

M.A.

N.A.

f.

Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.

7. Containment Pressure Control System a.

Start Permissive N.A.

N.A.

< 0.4 psid

< 0.45 psiu b.

Termination N.A.

N.A.

M.A.

> 0.3 psid

> 0.25 psid

8. Auxiliary feedwater a.

Manual Initiation N.A H.A.

N.A.

N.A N.A.

b.

Automatic Actuation Logic M.A.

N.A.

and Actuation Relays

l TABLE 3.3-4 (Continued)

Sg ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS SENSOR TOTAL ERROR FUNCTIONAL UNIT ALLOWANCE (TA)

Z_

(S)

TRIP SETPOINT ALLOWABLE VALUE g

U

18. Engineered Safety Features Actuation System Interlocks e-m a.

Pressurizer Pressure, P-11 N.A.

N.A.

1955 psig 11944 psig l

b.

Pressurizer Pressure, not P-11 N.A.

N.A.

1955 psig

<1%6 psig Low-Low T,yg, P-12 N.A.

N.A.

>SS3*F 1SS1*F(550*F )

c.

d.

Reactor Trip, P-4 N.A.

N.A.

e.

Steam Generator Level, P-14 See Item S. above for all Steam Generator Water Level Trip Setpoints w

and Allowable Values.

O CC II as

&.N 22 hk 3C L.,_._.

-. - ~.

w t

TABLE 3.3-4 (Continued)

TABLE NOTATIONS

Time constants utilized in the lead-lag controller for Steam Line Pressure-Low are 12 > 50 seconds and T2 5 5 seconds.

Channel calibration shall ensure that these time constants are adjusted to these values.

- The time constant utilized in the rate-lag controller for Steam Line Pressure-Negative Rate-High is greater than or equal to 50 seconds.

Channel calibration shall ensure that this time constant is adjusted to this value.

Applicable upon deletion of RTO Bypass System.

t f

b L

f i

CATAWBA - UNITS 1 & 2 3/4 3-36 Amendment No. 40 (Unit 1)

Amendment No. 33 (Unit 2)

_

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