Text

Amends 13 & 5 to Licenses NPF-35 & NPF-52,respectively, Modifying Tech Specs to Lower lo-lo Reactor Trip for Steam Generator Level When Reactor Operating Above 30% Power Level
	ML20215D064
Person / Time
Site:	Catawba
Issue date:	09/30/1986
From:	Jabbour K; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20215D067	List: ML20215D064; ML20215D077;
References
	NUDOCS 8610100622
	Download: ML20215D064 (15)
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UNITED STATES 8

NUCLEAR REGULATORY COMMISSION o

r 8-WASHINGTON, D. C. 20555 5,.,.....j l

DUKE POWER 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. 13 License No. NPF-35 1.

The Nuclear Regulatory Comission (the Comission) 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) 4 dated September 10, 1985, as supplemented November 27, 1985, January 7, and July 31, 1986, complies with the standards and requirements of the 4

Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations as set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance: (i)thattheactivitiesauthorizedby this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted l

in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.

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

E.

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

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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:

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

\\9 Kahtan Jabbour, Project Manager -

PWR Project Directorate No. 4 Division of PWR Licensing-A

Attachment:

Technical Specification Changes Date of Issuance:

September 30, 1986 N

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UNITED STATES

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DUKE POWER COMPANY NORTH' CAROLINA MUNICIPAL POWER AGENCY NO. 1 PIEDMONT MUNICIPAL POWER AGENCY DOCKET N0. 50-414 CATAWBA NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 5 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. I and Piedmont Municipal Power Agency (licensees) dated September 10, 1985, as supplemented November 27, 1985, January 7, and July 31, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's regulations as set forth in 10 CFR Chapter I; B.

The facility will operate in confonnity with the application, as amended, the provisions of the Act, and the 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 Comission's regulations set forth in 10 CFR Chapter I; D.

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

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

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. 2.

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

(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through /cendment No. 5, and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. Duke Power Contpany 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 b

Kahtan Jabbour, Project Manager PWR Project Directorate No. 4 Division of PWR Licensing-A

Attachment:

Technical Specification Changes Date of Issuance: September 30, 1986 1

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ATTACHMENT TO LICENSE AMENDMENT NO. 13 FACILITY OPERATING LICENSE N0. NPF-35 DOCKET NO. 50-413 AND TO LICENSE AMENDMENT NO. 5 FACILITY OPERATING LICENSE NO. NPF-52 DOCKET N0. 50-414 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change. The corresponding overleaf pa~ges are also provided to maintain document completeness.

Amended Overleaf Page Page 2-5 2-6 3/4 3-8 3/4 3-7 3/4 3-10 3/4 3-9 3/4 3-12 3/4 3-11 3/4 3-31 3/4 3-32

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4 TABLE 2.2-1 (Continued) n h

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E**

TOTAL SENSOR ALLOWANCE ERROR g

FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE d

13. Steam Generator Water Level Low-Low g

a.

Unit 1 17 14.2 1.5

>17% of span

>15.3% of span from

~% to 30% RTP*

u from 0% to 30%

0 RTP* increasing increasing linearly linearly to to >38.3% of span

> 40.0% of span froiii 30% to 100% RTP*

from 30% to 100%

RTP*

b.

Unit 2 17 14.2

1. 5

>17% of narrow

>15.3% of narrow range span range span i

14.

Undervoltage - Reactor 8.57 0

1.0

>77% of bus

>76% (5016 volts)

Coolant Pumps voltage (5082 volts) with a 0.7s response time i

15.

Underfrequency - Reactor 4.0 0

1.0

>56.4 Hz with a

->55.9 Hz Coolant Pumps 6.2s response time

16. Turbine Trip 1

yy a.1 Low Control Valve EH N.A.

N.A.

>550 psig

>500 psig i

gg Pressure-Low (Unit 1)

$k

'.2 Stop Valve EH N.A.

N.A.

>550 psig

>500 psig a

3g Pressure-Low (Unit 2) gg b.

Turbine Stop Valve N.A.

N.A.

>1% open

>1% open Closure g s 17.

Safety Injection Input N.A.

N.A.

L S.

from ESF ee NH

^ RIP = RATED THERMAL POWER

TABLE 2.2-1 (Continued) n h

REACTOR. TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

TOTAL SENSOR ALLOWANCE ERROR E

FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOINT ALLOWABLE VALUE d

18.

Reactor Trip System Interlocks w

e.

m a.

Intermediate Range N.A.

N.A.

>l x 10 10 amps

>6 x 10 11 amps Neutron Flux, P-6 b.

Low Power Reactor Trips Block, P-7

1),P-10 input N.A.

N.A.

$10% of RTP*

$12.2% of RTP*

2) P-13 input N.A.

N.A.

<10% R.TP* Turbine

<12.2% RTP* Turbine Impulse Pressure Impulse Pressure Equivalent Equivalent m

E c.

Power Range Netitron N.A.

N.A.

148% of RTP*

$50.2% of RTP*

Flux, P-8 d.

Power Range Neutron N.A.

N.A.

169% of RTP*

$70% of RTP*

Flux, P-9 e.

Power Range Neutron N.A.

N.A.

N.A

->10% of RTP*

->7.8% of RTP*

Flux, P-10 f.

Power Range Neutron N.A.

N. A.

N.A.

110% of RTP*

$12.2% of RTP*

Flux, Not P-10 g.

Turbine Impulse Chamber N.A.

N.A.

<10% RTP* Turbine

<12.2% RTP* Turbine Pressure, P-13 Impulse Pressure Impulse Pressure Equivalent Equivalent 19.

Reactor Trip Breakers N.A.

N.A.

N.A N.A.

N.A.

20. Automatic Trip and N.A.

N.A.

Interlock Logic "RTP = RATED THERMAL POWER W

r, s

TABLE 3.3-2 n

D REACTOR TRIP SYSTEM INSTRtMENTATION RESPONSE TIMES

>5 e

FUNCTIONAL UNIT RES'ONSE TIME c

1.

Manual Reactor Trip N.A.

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2.

Power Range, Neutron Flux 5 0.5 second*

o.

m 3.

Power Range, Neutron Flux, High Positive Rate N.A.

4.

Power Range, Neutron Flux, High Negative Rate 5 0.5 second*

5.

Intermediate Range, Neutron Flux N.A.

6.

Source Range, Neutron Flux N.A.

w 7.

Overtenperature AT 5 4 seconds

8.

Overpower AT

$ 4 seconds 9.

Pressurizer Pressure-Low

$ 2 seconds 10.

Pressurizer Pressure-High 5 2 seconds 11.

Pressurizer Water Level-High N.A.

i,

Neutron det.ectors are exempt from response time testing.

Response time of the neutron fit:x signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

.I TABLE 3.3-2 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES l

I FUNCTIONAL UNIT RESPONSE TIME C

i 5

12.

Low Reactor Coolant Flow l

a.

Single Loop (Above P-8) 5 1 second i

g b.

Two Loops (Above P-7 and below P-8) 5 1 second l

13.

Steam Generator Water Level-Low-Low a.

Unit 1

< 3.5 seconds I

b.

Unit 2

-7 2.0 seconds i

14.

Undervoltage-Reactor Coolant Pumps 5 1.5 seconds 15.

Underfrequency-Reactor Coolant Pumps 5 0.6 second mh 16.

Turbine Trip w

a.1 Control Valve EH Pressure-Low (Unit 1)

N.A.

a.2 Stop Valve EH Pressure-Low (Unit 2)

N.A.

b.

Turbine Stop Valve Closure N.A.

17.

Safety Injection Input from ESF N.A.

18.

Reactor Trip System Interlocks N.A.

F.

gg 19.

Reactor Trip Breakers N.A.

S$

gg

20. Automatic Trip and Interlock Logic N.A.

%5 L. ;-.

S 3C

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e i

TABLE 4.3-1 kk.

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 1

h TRIP ACTUATING MODES FOR ANALOG CHANNEL DEVICE WHICH c:

55 CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE Ut FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED 1.

Manual Reactor Trip N.A.

N.A.

R N.A.

1, 2, 3*, 4*, 5*

ro 2.

Power Range, Neutron Flux a.

High Setpoint 5

D(2,4),

M N.A.

N.A.

1, 2 M(3,4),

Q(4,6),

R(4,5) b.

Low Setpoint S

R(4)

M N.A.

N.A.

1###, 2 3.

Power Range, Neutron Flux, N.A.

R(4)

M N.A.

N.A.

1, 2 wi High Positive Rate

)l 4.

Power Range, Neutron Flux, N.A.

R(4)

M N.A.

N.A.

1, 2 High Negative Rate 5.

Intermediate Range, S

R(4,5)

S/U(1),M N.A.

N.A.

1###, 2 Neutron Flux 6.

Source Range, Neutron Flux S

R(4,5)

S/U(1),M(9)

N.A.

2##, 3, 4, 5 7.

Overtemperature AT S

R(12)

M N.A.

N.A.

1, 2

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8.

Overpower AT S

R M

N.A.

1, 2 9.

Pressurizer Pressure-Low S

R M

N.A.

1 10.

Pressurizer Pressure-High S

R M

N.A.

1, 2 11.

Pressurizer Water Level-High R

M N.A.

N.A.

1 12.

Reactor Coolant Flow-Low S

R M

N.A.

1

TABLE 4.3-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS s

TRIP i

ANALOG ACTUATING MODES FOR i

e CHANNEL DEVICE WHICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE d

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED w

e 13.

Steam Generator Water Level-S R(13)

M N.A.

N.A.

1, 2 l

m Low-Low 14.

Undervoltage - Reactor Coolant N.A.

R N.A.

M N.A.

1 Pumps t-l 15.

Underfrequency - Peactor N.A.

R N.A.

M N.A.

I Coolant Pumps R

16. Turbine Trip a.1 Control Valve EH N.A.

R N.A.

S/U(1,10)

N.A.

N Y

Pressure - Low (Unit 1)

O a.2 Stop Valve EH Pressure N.A.

R N.A.

S/U(1,10)

N.A.

N Low (Unit 2) s-b.

Turbine Stop Valve Closure N.A.

R N.A.

S/U(1,10)

N.A.

U i

l 17.

Safety Injection Input from N.A.

N.A.

R**

N.A.

1, 2 ESF ff18.

Reactor Trip System Interlocks l

g&

a.

Intermediate Range l

gg Neutron Flux, P-6 N.A.

R(4)

M N.A.

N.A.

2#

xz b.

Low Power Reactor P-Trips Block, P-7 N.A.

R(4)

M(8)

N.A.

1 vi r mX c.

Power Range Neutron EE Flux, P-8 N.A.

R(4)

M(8)

N.A.

1 33 d.

Low Power Range Neutron 4

3s Flux, P-9 N.A.

R(4)

M(8)

N.A.

1

- This surveillance need not be performed until prior to entering STARTUP following the Unit 1 first refueling.

l

p e

s TABLE 4.3-1 (Continued) bk REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

5 TRIP ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH c:

25 CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE DI FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED 18.

Reactor Trip System Interlocks (Continued) ro e.

Power Range Neutron Flux, P-10 N.A.

R(4)

M(8)

N.A.

I f.

Power Range Neutron Flux, Not P-10 N.A.

R(4)

M(8)

N.A.

1, 2 g.

Turbine Impulse Chamber t'

Pressure, P-13 N.A.

R M(8)

N.A.

1 i'

19.

Reactor Trip Breaker N.A.

N.A.

M(7,11)

N.A.

1, 2, 3*, 4*, 5*

U 20.

Automatic Trip and Interlock N.A.

N.A.

M(7) 1, 2, 3*, 4*, 5*

Logic

c.

TABLE 4.3-1 (Continued) l TABLE NOTATIONS Only if the Reactor Trip System breakers happen to be closed and the Control Rod Drive System is capable of rod withdrawal.

Above P-9 (Reactor Trip on Turbine Trip Interlock) Setpoint.

Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. 1. Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

(1) If not performed in previous 7 days.

(2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(3) Single point comparison of incore to excore axial flux difference above 15% of RATED THERMAL POWER.

Recalibrate if the absolute difference is greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(4) Neutron-detectors may be excluded from CHANNEL CALIBRATION.

(5) Detector plateau curves shall be obtained, evaluated and compared to manufacturer's data.

For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicat:1e for entry into MODE 2 or 1.

(6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

- (8) With power greater than or equal to the interlock setpoint the required ANALOG CHANNEL OPERATIONAL TEST shall consist of verifying that the interlock is in the required state by observing the permissive status light.

(9) Monthly surveillance in MODES 3*, 4*, and 5* shall also include verifi-cation that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive status light.

Monthly surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to one-half decade (square root of 10) above background.

(10) Setpoint verification is not applicable.

(11) At least once per 18 months and following maintenance or adju-tment of the Reactor trip breakers, the TRIP ACTUATING DEVICE OPERATIONAL TEST shall include independent verification of the Undervoltage and Shunt trips.

(12) CHANNEL CALIBRATION shall include the RTD bypass loops flow rate.

(13) For Unit 1, CHANNEL CALIBRATION shall ensure that the filter time constant associated with Steam Generator Water Level Low-Low is adjusted to a value less than or equal to 1.5 seconds.

CATAWBA - UNITS 1 & 2 3/4 3-12 Amendment No.13 (Unit 1)

Amendment No. 5 (Unit 2)-

TABLE 3.3-4 (Continued) 9 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS Y

SENSOR h

TOTAL ERROR

, FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETPOINT ALLOWABLE VALUE

@. Auxiliary Feedwater (Continued) h c.

Steam Generator Water Level - Low-Low g

8'

1) Unit 1 17 14.2 1.5

> 17% of span

> 15.3% of N

from 0% to ipan from 0% to 30% RTP 30% RTP increasing increasing linearly to i

linearly to

> 38.3% of span

> 40.0% of from 30% to 100%

ipan from 30%

RTP to 100% RTP

2) Unit 2 17 14.2 1.5

> 17% of

> 15.3% of narrow w) narrow range range instrument w

span span d.

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

e.

Loss-of-Offsite Power N.A.

N.A.

> 3500 V

> 3200 V f.

Trip of All Main Feedwater Pumps N.A.

N.A.

g.

Auxiliary Feedwater Suction Pressure-Low kk

1) CAPS 5220, 5221, 5222 N.A.

N.A.

> 10.5 psig

> 9.5 psig hh

2) CAPS 5230, 5231, 5232 N.A.

N.A.

H.A.

> 6.2 psig

> 5.2 psig gg a.

Unit 1 N.A.

N.A.

> 6.2 psig

> 5.2 psig i

b.

Unit 2 N.A.

N.A.

> 6.0 psig

> 5.0 psig

[f9.ContainmentSumpRecirculation

,y a.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays

_((

b.

Refueling Water Storage N.A.

N.A.

> 177.15 inches > 162.4 inches Tank Level-Low Sc Coincident With Safety Injection See Item 1. above for all Safety Injection Setpoints and Allowable Values.

i s

' TABLE 3.3-4 (Continued) 9 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

SENSOR TOTAL ERROR E

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S)

TRIP SETPOINT ALLOWABLE VALUE i

G 10.

Loss of Power s

a.

4 kV Bus Undervoltage-Loss N.A.

N.A.

1 3500 V 1 3200 V l

of Voltage b.

4 kV Bus Undervoltage-N.A.

N.A.

1 3685 V 1 3611 V Grid Degraded Voltage

11. Control Room Area Ventilation Operation 5

a.

Automatic Actuation Logic and Actuation Relays N.A.

N.A.

y b.

Loss-of-Offsite Power N.A.

N.A.

1 3500 V 1 3200 V c.

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

i 12.

Containment Air Return and Hydrogen Skimmer Operation a.

Nanual Initiation N.A.

N.A.

b.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays 1

c.

Containment Pressure-12.7 0.71 1.5 1 3 psig 5 3.2 psig i

High-High k

E M

l

ML20215D064

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