ML20151Y422

Amend 121 to License DPR-57,revising Tech Specs to Support Installation of Analog Transmitter Trip Sys

ML20151Y422
Person / Time
Site:	Hatch
Issue date:	01/17/1986
From:	Muller D Office of Nuclear Reactor Regulation
To:	City of Dalton, GA, Georgia Power Co, Municipal Electric Authority of Georgia, Oglethorpe Power Corp
Shared Package
ML20151Y426	List: ML20151Y422 ML20151Y432 ML20236E785
References
DPR-57-A-121, TAC 59309 NUDOCS 8602120692
Download: ML20151Y422 (54)

Text

.

[f UNITED STAT Es o,,

+

NUCLEAR REGULATORY COMMISSION g

W ASHINGTON, D. C. 20555 5

p

\\...../

GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.121 License No. DPR-57 1.

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

A.

The application for amendment by Georgia Power Company, et al.,

(the licensee) dated July 24, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

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

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

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

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby amended to read as follows:

9602120692 860117 DR ADOCK 0500 1

. s Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.121, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days.

FOR THE NUCLEAR REGULATORY COMISSION Y

A A

4 Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: January 17, 1986 i

h f

k i

l 1

I t

I f

ATTACHMENT TO LICENSE AMENDMENT NO.121 f

FACILITY OPERATING LICENSE NO. DPR-57 DOCKET NO. 50-321 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.

The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.

Remove Insert 1.1-3 1.1-3 1.1-5 1.1-5 Fig. 2.1-1 Fig. 2.1-1 1.2-2 1.2-2 1.2-6 1.2-6 3.1-4 3.1-4 3.1-7 3.1-7 3.1-12 3.1-12 3.2-2 3.2-2 3.2-3 3.2-3 3.2-5 "3.2-5 3.2-6 3.2-6 3.2-8 3.2-8 3.2-9 3.2-9 3.2-10 3.2-10 3.2-11 3.2-11 3.2-12 3.2-12 3.2-14 3.2-14 3.2-22 3.2-22 3.2-24 3.2-24 3.2-25 3.2-25 3.2-27 3.2-27 3.2-28 3.2-28 3.2-30 3.2-30 3.2-31 3.2-32 3.2-32 3.2-32 3.2-33 3.2-33 3.2-34 3.2-34 3.2-35 3.2-35 i

3.2-36 3.2-36 3.2-38 3.2-38 3.2-39 3.2-39 i

3.2-48 3.2-48 3.2-49 3.2-49 3.2-50 3.2-50 3.2-50a 3.2-50a 3.2-51 3.2-51 3.2-52 3.2-52 3.2-53 3.2-53 3.2-54 3.2-54 t

i

-g.

6

]

Remove Insert f

3.2-55 3.2-55 3.2-56 3.2-56 l

3.2-57 3.2-57 3.2-58 3.2-58 l

3.2-59 3.2-59 i

3.2-60 3.2-60 i

3.2-61 3.2-61 4

3.2-62 3.2-62 3.2-63 3.2-63 i

3.7-19 3.7-19 i

1 5

3 l

l s

1 i

i I

k i

i t

9 4

i t

I i

k 4

i i

I t

i

,m p

_ =~

w

.--e w

w--

-r,gp y-e-+

.m y__

-.a-v q 9-

=*g.y

-g

.=

SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1.A.1.d ApRM Rod Block Trip Setting This section deleted I

i i

i i

i l

i i

(1 2.1.A.2.

Reactor Vessel Water Low Level Scram Trip Setting (Level 3) 4 Reactor vessel water low level scram trip setting (Level 3) shall be 210.0 incnes (narrow range scale).

3.

Turbine Stop Valve Closure Scram j

i

-Turbine stop valve closure scram trip setting shall be s 10 percent valve closure from full open. This scram is only effective when tur-l bine steam flow is above that corres-ponding to 30% of rated core thernal

~ l(.

power, as measured by turbine first i

stage pressure.

1

.i 1 2-3 Amendment No. 58, 73, 193, 185, 121

SAFETY LIMITS LIMITING SAFETY SYSTEM SETTINGS 2.1.B.

Reactor Vessel Water Level Trip Settings Which Initiate Core Standby Cool-ing Systems (CSCS)

Reactor vessel water level trip settings i

which initiate core standby cocl-ing systems shall be as shown in Tables 3.2-2 thru 3.2-6 at normal operating conditions.

1.

HPCI Actuation (Level 2)

HPCI actuation (Level 2) shall occur at a water level 2 -47 l

inches.

+

2.

Core Spray and LPCI Actuation 4

(Level 1)

~

1 Core Spray and LPCI actuation (Level 1) shall occur at a water level 2 -113 inches.

l

(

l 4

4 2

t 4

4

'l i

i i

I l

?

f

. Amendment No. 793,121 1.1-5 4

'" T NOT E: SCALE IN INCHES ABOVE VESSEL 2ERO t

WATER LEVEL NOMENCLATURE HEIGHT ABOVE VESSEL ZERO

g. IINCHES)

READING INSTRUMENT 800 - -

(8) 573 5

+56.5 SARTON/ROSEMOUNT 750 -

17) 559

+42 GE/MAC (4) 549

+32 GEMAC

"" 723.56 FLANGE (3) 527

+10.0 SARTON/ROSEMOUNT (2)

, 470

--47 BARTON/ROSEMOUNT 700 - -

(1) 404

-113 SARTON/ROSEMOUNT i

(Of 315

-202 BARTON/ROSEMOUNT 650 - -

MAIN STE AM

<- 640 -

LINE f

600 - -

. 577

+60

+60 - -

40 -

+40 -

56.5

- 573.5 (8)

(g).

(7) 42 Hi ALARM 9

- 55917)

HPCI8, 14) 32 LO ALARM

$50 - = 54 9141 RC C TRIPS (3) 10.0 LOW (LEVEL 3).

- 52713)

-ORYER SKIRT-- 517-0 0--

0 -- RE ACTOR BOTTOM OF STE AM

[ 470 (2)

- 47 LOW LOW (LEVEL 21 FEED - 4835 17 - -

CORE WATER

.~. " ~ SPRAY g

INITIATE HPCI, RCIC, 450 --

~ 404 (1}

-113 LOW LOW LOW (LEVEL 1) 400 -

C.

INITIATE RHR, C.S.,

-15C --

- 367 START DIESEL AND 352.56 CONTRIBUTE TO A.D.S.

353 -6 CLOSE MSIV'S 2/3 COR E

-202 HElJHT 315 (0) 300 --

PERMISSIVE ACTIVE (LEVEL O)

FUEL 250--

200. = 208.56 REClRC

-317 - -

d

- 178.56 -OISCH A R G E RECIRC NO22LE SUCTION - 161.5 NOZZLE 150 im m

50 --

l FIGURE 2.1-1

(

REACTOR VESSEL WATER LEVEL l

\\

Amendment No. J$l, 103, 121 I

(

LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS 2.2.A Nuclear System pressure (cont.)

The allowable setpoint relief error for each valve shall be + 1%.

In the event that an instalTed safety-relief valve requires replacement, a spare valve whose setpoint is lower than that of the failed valve may be substituted for the failed valve until the first refueling outage following such substitution.

No more than two valves with lower setpoints may be substituted in place of valves with higher setpoints.

Spare valves which are used as substitutes under the abovementioned provisions shall have a setpoint equal to 1080 psig 21%

or 1090 psig 31%,

1.2.A.2.

When Ooerating The RHR Sys-2.1.A.2. When Operating The RHR System in the Shutdown Cooling Mode tem in the Shutdown Cooling The reactor pressure trip set-(

ting which closes (on increas-The reactor vessel steam dome pressure shall not exceed 162 ing pressure) or permits open-ing (on decreasing pressure) of psig at any time whe'i operat-the shutdown cooling isola-ing the RHR system in the Shut-tion valves shall be 5 145 psig.

l down Cooling Mode.

i 1

4 i

i l

t t

t l

('

1 I

I Amendment No. Jp3,121 1,2-2

BASES FOR LIMITING SAFETY SYSTEM SETTINGS 2.2 REACTOR COOLANT SYSTEM INTEGRITY A.

Nuclear System Pressure 1.

When Irradiated Fuel is in the Reactor The 11 relief / safety valves are sized and set point pressures are estab-I lished in accordance with the following requirements of Section III of the ASME Code:

The lowest relief / safety valve must be set to open at or below a.

vessel design pressure and the highest relief / safety valve must be set to open at or below 105% of design pressure.

The valves must limit the reactor pressure to no more than 110% of b.

design pressure.

The primary system relief / safety valves are sized to limit the primary system pressure, including transients, to the limits expressed in the ASME Boiler and Pressure Vessel Code,Section III, Nuclear Vessels.

No credit is taken from a scre initiated directly from the isolation event, or for power operated relief / safety valves, sprays, or other i

power operated pressure relieving devices. Thus, the probability of failure of the turbine generator trip SCRAM or main steam isolation valve closure SCRAM is conservatively assumed to be unity.

Credit is taken for subsequent indirect protection system action such as neutron flux SCRAM and reactor high pressure SCRAM, as allowed by the ASME C

Code. Credit is also taken for the dual relief / safety valves in their ASME Code qualified mode of safety operation.

Sizing on this basis is applied to the most severe pressurization transient, which 's the main steam isolation valves closure, starting from operation at 105 percent of the reactor warranted steamflow condition.

Reference 2, Figure 4 shows peak, vessel bottom pressures attained when the main steam isolation valve closure transients are terminated i

by various modes of reactor scram, other than that which would be initiated diractly from the isolation event (trip scram).

Relief /

safety valve capacities for this analysis are 84.0 percent, Tepresen-tative of the 11 relief / safety valves.

The relief / safety valve settings satisfy the Code requirements for relief / safety valves that the lowest valve set point be at or below the vessel design pressure of 1250 psig. These settings are also sufficiently above the normal operating pressure range to prevent unnecessary cycling caused by minor transients. The results of postu-i lated transients where inherent relief / safety valve actuation is 1

required are given in Section 14.3 of the FSAR.

i 2.

When Operating the RHR System in the Shutdown Cooling Mode L

An interlock exists in the logic for the RHR shutdown cooling valves, which are normally closed during power operation, to prevent open the valves above a preset pressure setpoint of 145 psig. This setpoint I

is selected to assure that pressure integrity of the RHR system is main-

{(-,

tained. Administrative operating procedures require the operator to i

t Amendment No. M3,121 1.2-6

^

.q 1

Table 3.1-1 (Cont'd) g R

2 Operable

= Mumber Source of Scram Trip Signal Channels Scram Trip Setting Source of Scram Signal is g Scram Required to be Operable Required Per Except as Indicated Below (a)

Trip System (b) l i

2

< 1.92 psig Not required to be operable when 5

Drywell Pressure-High primary containment integrity is f

not required.

May be bypassed j

when necessary during purging for wo i

containment inerting or deinerting.

i yy l

2

> 10.0 inches

w "

6 Reactor Vessel Water Level -

Low (Level 3)

g Permissible to bypass (initiates

!l

)^

7 Scram Discharge Volume liigh control rod block) in order to High Level reset RPS when the Mode Switch is in the REFUEL or SHUTOOWN position.

a.

Float Switches 2

171 gallons b.

Themal level Sensors 2

171 gallons i

8 APitM Flow Referenced Simulated 2

S < 0.66W+62%

i i

Themal Power Monitor (Not to exceed 117%)(1)

Tech Spec 2.1.A.1.c f

2 5 < 120% Power j

Fixed High-ifigh Neutron Tech Spec 2.1. A.1.c (2)

Flux l

An APRM is inoperable if there 2

Not Applicable Inoperative are less than two LPRM inputs per level or there are less than 11 LPRM inputs to the i

i APRM channel, i

l.

O' A

O, Table 4.1-1 Reactor Protection System (RPS) Instrumentation functional Test, Functional Test Minimum Frequency, and Calibration Minimum Frequency l

g Instrument Functional Test Instrument Calibration i

s l

Scram l

g Number Source of Scram Trip Signal Group Minimum Frequency Minimum Frequency l

<+

(a)

(b)

(c) 5 1

Mode Switch in SHUTDOWN A

Once/ Operating Cycle Not Applicable E

2 Manual Scram A

Every 3 months Not Applicable w?

3 IRM High High Flux C

Once/ Week during refueling Once/ Week and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of E

Startup (e) rNo Inoperative C

Once/ week during refueling Once/ Week and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of

["

Startup (e) 4 Reactor Vessel Steam Dome D

Once/ Month Once/ operating cycle y

Pressure - High a

5 Drywell Pressure-High D

Once/ Month Once/ operating cycle 6

Reactor Vessel Water Level -

D Once/ Month (g)

Once/ Operating Cycle Low (Level 3) 7 Scram Discharge Volume High High Level a.

Float Switches A

Once/ Month (f)

(h) b.

Thermal Level Sensors B

Once/ Month (f)

Once/ operating cycle 8

APRM Fixed High-High Flux B

Once/ Week (e)

TwiceNeek Inoperable B

OnceNeek (e)

TwiceNeek Downscale B

Once/ Week (e)

Twice/ Week Flow Reference Simulated B

Once/ Week (f)

Once/ Operating Cycle Thermal Power Monitor 15% Flux C

Within 24 Hours of Startup (e)

Once/ Week

3.1.A.4.

Reactor Vessel Steam Dome Pressure - High (Continued) setting also protects the core from exceeding thermal hydraulic limits as

(,

a result of pressure increases from some events that occur when the reactor r

is operating at less than rated power and flow, f

5.

Drywell Pressure - High High pressure in the drywell could indicate a break in the primary pressure The reactor is tripped to minimize the possibility of fuel boundary system.

The damage and reduce the amount of energy being added to the coolant.

trip setting was selected as low as possible without causing spurious trips.

6.

Reactor Vessel Water Level - Low (Level 3)

The bases for the Reactor Vessel Water Level-Low Scram Trip Setting (Level

3) are discussed in the bases for Specification 2.1.A.2.

7.

Scram Discharge Volume High High Level The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping. A part of this piping is an instrument volume which is the low point in the piping. No credit was taken for this volume in the design of the discharge piping as concerns the amount of water which must be accommodated during a scram. During normal operation the discharge volume is empty; however, should the discharge volume fill i

j with water, the water discharged to the piping from the reactor could not be accommodated which would result in a slow scram time or partial or no control rod insertion. To preclude this occurrence, level switches have been provided in the instrument volume which scram the reactor when the volume of water reaches 71 gallons. As indicated above, there is suffi-cient volume in the piping to accommodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accommodate the dis-charged water and precludes the situation in which a scram would be required 4

out not able to perform its function adequately.

8.

APRM Three APRM instrument channels are provided for each protection trip sys-APRM's A and E operate contacts in one trip logic and APRM's C and tem.

E operate contacts in the other trip logic. APRM's B, D and F are arranged l

l similarly in the other protection trip system. Each protection trip sys-i tem has one more APRM than is necessary to meet the minimum number re-quired per channel. This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration.

Flow Referenced Simulated Thermal Power Monitor and Fixed High-High a.

i Neutron Flux The bases for the APRM Flow Referenced Simulated Thermal Power Monitor and Fixed High-High Neutron Flux Scram Trip Settings are discussed in the bases for Specification 2.1.A.I.c.

i 3.1-12 Amendment No. 69, 193, 195, 121

n' o

E m

Table 3.2-1 a"

d INSTRUMENTATION WillCll INITIATES REAC10R VESSEL AND PRIMARY S

CONTAINHENT ISOLATION 5

h

. ired Action to be taken if w

f.

Trip 0, rable Instrument Condition Channels Trip Setting number of channels is

%}io.

not met for both trip Remarks (d)

Nomenclature per Trip systems (c)

Mo)

System (b) 1 Reactor Vessel Low (Level 3) 2 1 10.0 inches Initiate an orderly Initiates Group 2 & 6 shutdown and achieve isolation.

Water Level Narrow Range the Cold Shutdown Condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or isolate the shutdown cooling system.

Initiate an orderly Starts the SGTS, l

Low Low 2

1-47 inches shutdown and achieve initiates Group 5 (Level 2) the Cold Shutdown isolation, and w

'm Condition within 24 ini tiates secondary containment a

hours.

isolation.

Low Low Low 2

1-113 inches Initiate an orderly Initiates Group 1 l

shutdown and achieve isolation.

(Level 1) the Cold Shutdown Con-dition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Isolates the shutdown l 2

Reactor Vessel Steam Low Permissive 1

-<145 psig Isolate shutdown cooling suction valves cooling.

of the RHR system.

Dome Pressure (Shut-down Cooling Mode)

Initiate an orderly Starts the standby l

2 5 1.92 psig shutdown and achieve gas treatment system, 1

Drywell Pressure liigh initiates Group 2 the Cold Shutdown Condition within 24 isolation and second-ary containment hours isolation.

l n.

.m F

e Table 3.2-1 (Cont.)

I 1

l Required j{ Rsf.

Trip Operable Action to be taken if No.

Instrument Condition Channels Trip Setting number of channels is i

jf(a)

Nomenclature per Trip not met for both trip Remarks (d) 3 System (b) systems (c) 1 2

53 times normal Initiate an orderly load Initiates Group 1 et

[y 4 Main Steam Line High full power back-reduction and close MSIVs isolation.

Radiation l

ground within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

• • 3

5 Main Steam Line Low 2

1825 psig Initiate an orderly load Initiates Group 1 reduction and close isolation. Only Pressure MSIVs within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

required in RUN mode, therefore activated when Mode Switch is in RUN position.

6 Main Steam Line High 2

5138% rated flow Initiate an orderly load Initiates Group 1 Flow (1115 psid) reduction and close MSIVs isolation.

within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

7 7

Main Steam Line High 2

1194*F Initiate an orderly load Initiates Group 1 l

reduction and close MSIVs isolation Tunnel Temperature w

within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

8 Reactor Water High 1

20-80 gpm Isolate reactor water Final trip setting cleanup system.

will be determined Cleanup System during startup test Differential Flow program.

9 Reactor Water High 2

$124*F Isolate reactor water cleanup system.

Cleanup Area Temperature 10 Reactor Water High 2

567 F Isolate reactor water cleanup system.

Cleanup Area Ventilation Differential Temperature 11 Condenser Vacuum Low 2

17" Hg. vacuum ~

Initiate an orderly load Initiate Group'I reduction and close MSIVs isolation within 8 hrs.

m

^

N

• =,

f $

Table 3.2-2 t a 5

INSTRUMENTATION WHICH INITIATES OR CONTROLS HPCI Dbf.

Instrument Trip Required Trip Setting Remarks Condition Operable

• No.

Nomenclature Channels

. Ma) per Trip System (b)

!3 1.

Reactor Vessel Water Level Low Low 2

2,-47 inches Initiates HPCI; Also initiates l

RCIC.

(Level 2) 2.

Drywell Pressure High 2

51.92 psig Initiates HPCl; Also initiates l

LPCI and Core Spray and pro-vides a permissive signal to ADS.

3.

HPCI Turbine Overspeed Mechanical 1

5 5000 rpm Trips HPCI turbine

( 4.

HPCI Turbine Exhaust Pressure High 1

$ 146 psig Trips HPCI turbine l

l

• 5.

HPCI Pump Suction Pressure Low I

< 12.6 inches Trips HPCI turbine 8

Hg vacuum 6.

Reactor Vessel Water level High 2

5 +56.5 inches Trips HPCI turbine (Level 8) 7.

HPCI Pump Discharge Flow High 1

2870 gpm Closes HPCI minimum flow bypass

(> 9.04 inches) line to suppression chamber.

Low I

< 605 gpm Opens HPCI minimum flow bypass

[14.36 inches) line if pressure permissive is present.

8.

HPCI Emergency Area High 1

5 169'F Closes isolation valves in Cooler Ambient Temperature turbine.

~

O.

n l

$I CL 5

Table 3.2-2 (Cont.)

?+

!?

Raf.

Instrument Trip Required Trip Setting Remarks 33 ' N 3.

. Condition Operable Nomenclature Channels (c).

per Trip g{

System (b)

U

~

9.

HPCI Steam Supply Pressure Low 2

1100 psig Closes isolation valves in l

HPCI system, trips HPCI turbine.

10.

HPCI Steam Line AP (Flow)

High 1

$303% rated Close isolation valves in flow HPCI system, trips HPCI turbine.

11.

HPCI Turbine Exhaust High 1

520 psig Close isolation valves in l

HPCI system, trips HPCI Diaphragm Pressure turbine.

y 12.

Suppression Chamber Area High 1

5169 F Close isolation valves in HPCI system, trips HPCI Ambient Temperature turbine.

m 13.

Suppression Chamber Area High 1

542 F Close isolation valves in l

HPCI system, trips HPCI Differential Air temperature turbine.

14.

Condensate Storage Tank Low 2

10 inches Automatic interlock switches suction from CST to Level suppression chamber.

15.

Suppression Chamber Water High 2

5154.2 inches Automatic interlock switches with respect to suction from CST to Level torus invert suppression chamber.

1 Not Applicable Monitors availability of 16.

HPCI Logic Power Failure power to logic system.

Monitor The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be established a.

between items in Table 3.2-2 and items in fable 4.2-2.

R' F

.R E

o Table 3.2-3 r*

5 INSTRUMENTATION WHICH INITIATES OR CONTROLS RCIC D

Raf.

Instrument Trip Required Trip Setting Remarks Condition Operable

{l No.

(c)

Nomenclature Channels per Trip

g System (b) sd A

1.

Reactor Vessel Water Level Low Low 2

>-47 inches Initiates RCIC; also initiates l

~~

HPCI.

(Level 2) 2.

RCIC Turbine Overspeed Electrical 1

5110% rated Trips RCIC turbine.

Mechanical 1

5125% rated Trips RCIC turbine.

3.

RCIC Turbine Exhaust High 1

s+45 psig Trips RCIC turbine.

l Pressure k' 4.

RCIC Pump Suction Pressure Low 1

512.6 inches Trips RCIC turbine.

l Hg Vacuum m

5.

Reactor Vessel Water Level High 2

<+56.5 inches Trips RCIC; automatically resets (Level 8) when water drops below level 8, 4

system automatically restarts at level 2.

6.

RCIC Pump Discharge Flow High 1

> 87 gpm Closes RCIC minimum flow

(>10.6 inches) bypass line.to suppression chamber.

4 Low 1

553 gpm Opens RCIC minimum flow l~<3.87 inches) bypass line if pressure permissive is present.

~

7.

RCIC Emergency Area High 1

5169'F Closes isolation valves in RCIC system, trips RCIC Cooler Ambient Temperature turbine.

s

r

.O '

,A 1

Table 3.2-3 (cont.)

5 Raf.

Instrument Trip Required Trip Setting Remarks i

= No.

Condition Operable (a)

Nomenclature Channels per Trip w

System (b)

"s l

8.

RCIC Steam Supply Pressure Low 2

>60 psig Closes isolation valves in l

i%d RCIC system, trips RCIC l

r"o turbine.

9.

RCIC Steam Line AP (Flow)

High 1

1306% rated Closes isolation valves in flow RCIC system, trips RCIC turbine.

L 10.

RCIC Turbine Exhaust High 1

120 psig Closes isolation valves in l

i RCIC system, trips RCIC Diaphragm Pressure turbine.

w 11.

Suppression Chanber Area High 1

1169*F Closes isolation valves in RCIC system, trips RCIC l

'u Ambient Temperature turbine.

I e

12.

Suppression Chanber Area High 1

142 F Closes isolation valves in l

RCIC' system, trips RCIC Differential Air turbine.

Temperature 13.

RCIC Logic Power Failure 1

Not Applicable Monitors availability of l

power to logic system.

Monitor 14.

Condensate Storage Tank Low 2

-> 0" Transfers suction from CST l

to suppression pool l

Water Level

< 0" Transfers suction from CST l

f 15.

Suppression Pool Water High 2

to suppression pool l

Level i

i l

m i

'g.

Table 3.2-4 m

'El N

INSTRUMENTATION WillCll INITIATES OR CONTROLS ADS 5'

Remarks

,if Trip Required Trip Setting 1 - Raf.

Instrument Condition Operable

g No.

Nomenclature Channels

}* (a) per Trip System (b)

N-Confirms low level, ADS permissive 1.

Reactor Vessel Water Level Low (Level 3) 1 3 10.0 inches Reactor Vessel Water Level Low Low Low 2

3-113 inches Permissive signal to ADS timer (Level 1)

Initiates HPCl; also initiates LPCI liigh 2

< 1.92 psig and core spray and provides a 2.

Drywell Pressure permissive signal to ADS timer Permissive signal to ADS timer

-l P

3.

RHR Pump Discharge High 2

2112 psig

y Permissive signal to ADS timer l

Pressure

e High 2

1137 psig ca 4.

CS Pump Discharge Pressure seconds With Level 3 and Level 1 and high 1

120 1 12 5.

Auto Depressurization drywell pressure and CS or RHR pump at pressure, timing sequence begins.

Timer If the ADS timer is not reset it will initiate ADS.

1 Not appilcable Monitors availability of power to J

Automatic Blowdown Control logic system 6.

Power failure Monitor The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be established between items in Table 3.2-4 and items in Table 4.2-4.

a.

ble Whenever any CCCS subsystem is required to be operable by Section 3.5, th b.

ithin that system shall be repaired or the reactor shall be placed in the Cold Shutdown Condition w trip systems.

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after this trip systems is made or found to be inoperable, o

o m

^

Table 3.2-5 s

lI INSTRUMENTAT10N WHICH INillATES OR CONTROLS THE LPCI MODE OF RHR o"

Remarks 5

Trip Required Trip Setting

[Raf.

Instrument Condition Operable

!*No.

Ncmenclature Channels

.y(c) per Trip System (b) 1.

Reactor Vessel Water Level Low Low Low 2

1 -113 inches initiates LPCI mode of RHR l

a D3 (Level 1)

Initiates LPCI mode of RHR. Also High 2

1 1.92 psig initiates HPCI and Core Spray 2.

Drywell Pressure and provides a permissive signal to ADS Low Permissive

]

$ 145 psig With primary containment isola-4 Reactor Vessel Steam tion signal, closes RHR (LPCI) 3.

Dome Pressure inboard motor operated injection Y

valves Permissive to ctese Recirculation l low 2

1335 psig Discharge Valve and Bypass Valve Low 2

1422 psig*

Permissive to open LPCI injection

. valves e

4.

Reactor Shroud Water Level Low 1

3-202 inches

. Acts as permissive to divert some LPCI flow to containment (Level 0) spray Initiates annunciator when valve N/A 1

Valve not 5.

LPCI Cross Connect closed is not closed Valve Open Annunciator

• This trin function shall bc <500 psio

A' A

.A.

"o.

S Table 3.2-5 (Cont.)

INSTRUMENTATION WHICH INITIATES OR CONTROLS THE LPCI MODE OF RHR 2:

33 af.

Instrument Trip Required Trip Setting Remarks i

R Condition Operable No.

Nomenclature Channels

(a) per Trip System (b) l 6

RHR (LPCI) Pump Flow Low I

>1670 gpm Opens LPCI minimum flow line upon

[4.7 inches) receipt of low flow signal from both pumps and closes LPCI l

minimum flow line when signal from either pump is not present 1

0<t<1 seconds With loss of normal power, and

'l l

w l

Ia 7 RHR(LPCI) Pump Start Timers upon receipt of emergency power, l

j, I

9<t<11 seconds one RHR pump starts immediately,-

the other three follow in 10 l

w seconds 1

>10 minutes Cancels LPCI injection valve l

l 8

Valve Selection Timers initiation signal l

1 Not Applicable Monitors availability of power l-9 RHR Relay Logic Power to logic system Failure Monitor I

i l

~

t

S

^

O Table 3.2-6 F

INSTRUMENTATION WHICH INITIATES OR CONTROLS CORE SPRAY Remarks

@ Rmf.

Instrument Trip Required Trip Setting Condition Operable

" No.

Nomenclature Channels 5 (a) per Trip System (b)

• E 1.

Reactor Vessel Water Level Low Low Low

> -113 inches Initiates CS.

l (Level 1) 2 2.

Drywell Pressure High 2

51.92 psig initiates CS.

Also initiates HPCI and LPCI Mode of RHR and provides a permissive signal to ADS 3.

Reactor Vessel Steam Dome Low 2

1422 psig

• Permissive to open CS l

injection valves.

Pressure To be determined Monitors integrity of CS 1

4.

Core Spray Sparger during startup piping inside vessel and core w

Differential Pressure testing shroud.

7 Minimum flow bypass line is I

Low 1

2 61D gpm e^

5.

CS Pump Discharge Flow

(> 4.13 inches) closed when low flow signal is not present.

1 Not Applicable Monitors availability of 6

Core Spray Logic Power power to logic system.

Failure Monitor

• This trip function shall be <500 psig The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be establ a.

between items in Table 3.2-6 and items in Table 4.2-6.

Whenever any CCCS subsystem is required to be operable by Section 3.5, there shall be two operable If the required number of operable channels cannot be met for one of the trip systems, that b.

system shall be repaired or the reactor shall be placed in the Cold Shutdown Condition within 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> systems.

d70cv 0219 trirs spstem is made or found to be inoperable.

O.

A A

Table 3.2-11 INSTRUMENTATION WHICH PROVIDES SURVEILLANCE INFORMATION a

Required a

Operable

" Raf.

5 No.

Instrument Instrument Type and Channels Range Action Remarks (a)

(b) 1 Recorder -150" to +60" (c)

(d)

D 1

Reactor Vessel Water Level 2

Indicator -150" to.+60" (c)

(d) 3 1

Recorder -317" to -17" (c)

(d) l 2

Shroud Water Level

,3 1

Indicator -317" to -17" (c)

(d) 1 Recorder 0 to 1500 psig (c)

(d)

E' 3 2

Indicator 0 to 1500 psig (c)

(d) l y

Reactor Pressure 5

2 Recorder -10 to +90 psig (c)

(d) 4 Drywell Pressure 2

Recorder 0 to 500 F (c)

(d)

U 5

Drywell Temperature 2

Recorder 0 to 500 F (c)

(d) 6 Suppression Chamber Air Temperature

=

2 Recorder 0 to 250 F (c)

(d) h7 Suppression Chamber Water Temperature 2

Indicator 0 to 300" (c)

(d) 8 Suppression Chamber Water Level 2

Recorder 0 to 30" (c)(e)

(d) 2 Recorder -10 to +90 psig (c)

(d) 9 Suppression Chamber Pressure 10 Rod Position Information System (RPIS) 1 28 Volt Indicating Lights (c)

(d) l 1

Recorder 0 to 52 (c)

(d)

.11 Hydrogen and Oxygen Analyzer 12 Post LOCA Radiation Monitoring System 1

Recorder (c)

(d) 6 R/hr (c)

(d)

Indicator 1 to 10 j

i 13 a) Safety / Relief Valve Position Primary 1/5EV Indicating Light at 85 psig (f)

Indicator

~

b) Safety / Relief Valve Position Secondary 1

Recorder 0 to 600 F (f)

Indicator

l m'

o i

Table 4.2-1 Check, Functional Test, and Calibration Minimum frequency for Instrumentation g,

Which Initiates Reactor Vessel and Primary Contalement isolation l

3

'5

$ Raf.

It'strument Check Instrtment functional Test Instrument Calibration No.

Instrument Minimum frequency Minista f requency Minista f requency (b)

(c)

(a) w 1

Reactor Vessel Water tevel Once/ shift once/ month Once/ operating cycle E

l M (Levels 1, 2, and 3) 2 Rear.*or Vessel Steam Dome Once/ shift Once/morth Once/ operating cycle l

l" Pressure (Shutdown Cooling l

Mode) 3 Drywell Pressure Once/ shift Once/ month once/ operating cycle l

4 Main Steam Line None Once/ week (e)

Every 3 months (f)

Radiation

."y 5

Main Steam Line None (d)

Every 3 months Pressure y

6 Main Steam Line Flow Once/ shift Once/ month Once/ operating cycle l

7 Main Steam Line Tunnel Once/ shift Once/ month Once/ operating cycle Temperature 8

Reactor Wr.ter Cleanup None (d)

Every 3 months l

System Differential Flow 9

Reactor Water Cleanup Once/ shift Once/ month Once/ operating cycle l

Area Temperature l

t

- - _ - _ -.-..-..-. _..-.. -. -.-..~. -

-._... ~. - _

_ _ ~.

A O

9' l

Table 4.2-1 (Cont'd) l l

Raf.

Instrument Check Instrument Functional Test Instrument Calibration No.

Instrument Minimum Frequency Minimum Frequency Minimum Frequency

- j (b)

(c) j (a) 10 Reactor Water Cleanup Once/ shift

.Once/ month Once/ operating cycle Area Ventilation Differential Temperature 1

11 Condenser Vacuum None (d)

Every 3 months Notes for Table 4.2-1 i

l The column entitled "Ref. No.".is only for convenience so that a one-to-one relationship can be o.

w established between items in Table 4.2-1 and items in Table 3.2-1.

• u I

Instrument functional tests are not required when the instruments are not required to be operable or b.

are tripped. However, if functional tests are missed, they shall be performed prior to returning the l~

instrument to an operable status.

c.. Calibrations are not required when the instruments are not required to be operable.

However, if calibrations are missed, they shall be performed prior to returning the instrument to an operable i

status, Initially once per month or according to Figure 4.1-1 with an interval of not less than one month nor d.

i more than three months. The compilation of instrument failure rate date may include data obtained

-=--.-

u -. --. -...-

n.

O P

E Table 4.2-2 g

'5 Check, Functional Test, and Calibration Minimum Frequency for Instrumentation Which Initiates or Controls HPCI

s" 5

Instrument Check Instrument Functional Test Instrument Calibration

% No.

Instrument Minimum Frequency Minimum Frequency Minimum Frequency R2f.

(b)

(c)

Q1

• 1 Reactor Vessel Water Level Once/ shift Once/ month Once/ operating cycle 5

(Level 2) 2 Drywell Pressure Once/ shift Once/ month Once/ operating cycle l

w R

-3 HPCI Turbine Overspeed None N/A Once/ operating cycle U

4 HPCI Turbine Exhaust Once/ shift Once/ month Once/ operating cycle l

w Pressure 1

s Once/ shift Once/ month Once/ operating cycle l

y l

5 HPCI Pump suction Pressure 6

Reactor Vessel Water Level Once/ shift Once/ month Once/ operating cycle (Level 8)

Once/ operating cycle 7

HPCI Pump Discharge Flow Once/ shift Once/ month 8

HPCI Emergency Area -

-- Once/shi f t Once/ month Once/ operating cycle Cooler Ambient Temperature 9

HPCI Steam Supply Pressure Once/ shift.

Once/ month Once/ operating cycle f

.9..

. ~. -. -

..~-

R-

?

D s.

a Table 4.2-2 (Cont'd) r*

E

[

Raf.

Instrument Check Instrument Functional Test Instrument Calibration

]t No.

Instrument Minimum Frequency Minimum Frequency Minimum Frequency (b)

(c)

{3}

G Once/ shift Once/ month Once/ operating cycle HPCI Steam Line AP (Flow) 11 HPCI Turbine Exhaust Once/ shift Once/ month Once/ operating cycle l

l Diaphragm Pressure l

12 Suppressicn Chamber Area Once/ shift Once/ month Once/ operating cycle Ambient Temperature u,

13 Suppression Chamber Area Once/ shift Once/ month Once/ operating cycle l

i Differential l

u l

a

. Air Temperature 14 Condensate Storage None (d)

Every 3 months m

l Tank Level 15 Suppression Chamber Once/ shift Once/ month Once/ operating cycle l

Water Level 16 HPCI Logic Power

None Once/ operating cycle None l

Failure Monitor 1

Notes for Table 4.2-2 I

The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be c.

. established between items in Table 4.2-2 and items in Table 3.2-2.

l

A.

A A

5

$a 2:s" Table 4.2-3 5

Check, Functional Test, and Calibration Minimum Frequency for Instrumentation Which Initiates or Controls RCIC I.

M R2f.

Instrument Check Instrument Functional Test Instrument Calibration

" No.

Instrument Minimum Frequency Minimum Frequency Minimum Frequency (b)

(c)

(a)

D'.

1 Reactor Vessel Water Level Once/ shift Once/ month Once/ operating cycle (Level 2) 2 RCIC Turbine Overspeed Electrical /

None N/A Once/ operating cycle Mechanical None N/A Once/ operating cycle 3

RCIC Turbine Exhaust Once/ shift Once/ month Once/ operating cycle l

Pressure g

4 RCIC Pump Suction Once/ shift Once/ month Once/ operating cycle l

Y Pressure 5

Reactor Vessel Water Level Once/ shift Once/ month Once/ operating cycle (Level 8) 6 RCIC Pump Discharge Flow Once/ shift Once month Once/ operating cycle 7

RCIC Emergency Area Once/ shift Once/ month Once/ operating cycle Cooler Ambient Temperature 8

RCIC Steam Supply Pressure Once/ shift Once/ month Once/ operating cycle

1 A-l A

A E

Pe5

$e 95 Table 4.2-3 (Cont'd) 5

~

Instrument Check Instrument functional Test Instrument Calibration R:f.

D3 No.

Instrument Minimum Frequency Minimum Frequency Minimum Frequency (b)

(c)

(n) 9 RCIC Steam Line Once/ shift Once/ month Once/ operating cycle AP (Flow) 10 RCIC Turbine Exhaust once/ shift Once/ month Once/ operating cycle l

Diaphragm Pressure 11 Suppression Chamber Area Once/ shift Once/ month Once/ operating cycle Ambient Temperature u

Y 12 Suppression Chamber Area Once/ shift Once/ month Once/ operating cycle 71 Differential Air Temperature 13 RCIC Logic Power Hone Once/ operating cycle None l

Failure Monitor 14 Condensate Storage None Monthly Every 3 months l

Tank Level 15 Suppression Pool Hone Monthly Every 3 months l

Water Level Notes for Table 4.2-3 c.

The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be established between items in Table 4.2-3 and items in Table 3.2-3.

b.

Instrument functional tests are not required when the instruments are not required to be operable or are tripped. However, if functional tests are missed, they shall be performed prior to returning the instrument to an operable status.

____m._

-_m.__..

_.__.________m-i

.O' O

O t

Notes for Table 4.2-3 (Cont'd)

However, if Calibrations are not required when the instruments are not required to be operable.

l c.

calibrations are missed, they shall be performed prior to returning the instrument to an operable status.

i i

t 2'

I uL u

Logic system functional test and simulated automatic actuation shall be performed once each operating cycle for the following:

1.

RCIC Subsystem Auto Isolation The logic system functional tests shall include a calibration of time relays and timers necessary for proper functioning of the trip systems.

4 4

v e

w.

-.,_.3_

O.

A O

i k

Table 4.2-4 a

2 Check, Functional Test, and Calibration Minimum Frequency for Instrumentation Which Initiates or Controls ADS

[

Instrument Check Instrument Functional Test Instrument Calibration z,o Ref.

No.

Instrument Minimus Frequency Minimum Frequency Minimus Frequency (b)

(c) y g (a) l g1 Eeactor Vessel Water Level Once/ shift Once/ month Once/ operating cycle t

I (Level 3) l Reactor Vessel Water Level Once/ shift Once/ month Once/ operating cycle (Level 1) 2 Drywell Pressure Once/ shift Once/ month Once/ operating cycle l

Once/ shift Once/ month Once/ operating cycle l

3 RHR Pump Discharge 7

Pressure 4

CS Pump Discharge Once/ shift Once/ month Once/ operating cycle l

l Pressure 5

Auto Deprassurization None N/A Once/ operating cycle Timer 6

Automatic Blowdown None Once/ operating cycle None Centrol Power Failure Monitor l

Notes for Table 4.2-4 The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be c.

established between items in Table 4.2-4 and items in Table 3.2-4.

i R

R l

^

1 Notes for Table 4.2-4 (Cont'd)

Instrument functional tests are not required when the instruments are not required to be operable or b.

However, if functional tests are missed, they shall be performed prior to returning the are tripped.

instrument to an operable status.

Calibrations are not required when the instruments are not required to be operable. However, if c.

calibrations are missed, they shall be performed prior to returning the instrument to an operable status.

i 1

Gu

-8w l

Sm l

l l

Logic system functional tests and simulated automatic actuation shall be performed once each operating 1

cycle for the following:

1.

ADS Subsystem The logic system functional tests shall include a calibration of time relays and timers necessary for proper functioning of the trip systems.

l

A-A A

N S

Table 4.2-5 I

$a 5

Check, Functional Test, and Calibration Minimum Frequency for Instrumentation Which Initiates or Controls the LPCI Mode of RHR Instrument Check Instrument Functional Test Instrument Calibration w

R;f.

E No.

Instrument Minimum Frequency Minimum Frequency Minimum Frequency (b)

(c)

• M i b1 Reactor Vessel Water Level Once/ shift Once/ month Onc'e/ operating cycle (Level 1) 2 Drywell Pressure Once/ shift Once/ month Once/ operating cycle l

3 Reactor Vessel Steam Once/ shift Once/ month Once/ operating cycle Dome Pressure Once/ operating cycle 4

Reactor Shroud Water Level Once/ shift Once/ month (Level 0)

'f 5 LPCI Cross Connect Valve None Once/ Operating cycle None g

Open Annunciator 6

RHR (LPCI) Pump Flow

- Once/ shift Once/ month Once/ operating cycle 7

RHR (LPCI) Pump None N/A Once/ operating cycle Start Timers 8

Valve Selection Timers None N/A Once/ operating cycle l

l l

9 RHR Relay Logic Power None Once/ operating cycle None Failure Monitor

R.

D l

5" 3

l l

Se 5

Notes for Table 4.2-5 h c.

The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be established between items in Table 4.2-5 and items in Table 3.2-5.

b.

Instrument functional tests are not required when the instruments are not required to be operable or are tripped. However, if functional tests are missed, they shall be performed prior to returning the instrument to an operable status.

Calibrations are not required when the instruments are not required to be operable. However, if c.

calibrations are missed, they shall be performed prior to returning the instrument to an operable y

status.

ww Logic system functional tests and simulated automatic actuation shall be performed once each operating cycle for the following:

1.

LPCI Subsystem 2.

Containment Spray subsystem

__.m l

R.

M O

o 5

Table 4.2-6 se if Check, Functional Test, and Calibration Minimum Frequency for Instrumentation Which Initiates or Controls Core Spray EE 2* Ref.

Instrument Check Instrument Functional Test Instrument Calibration

_. No.

Instrument Minimum Frequency Minimum Frequency Minimum frequency (b)

(c)

D! (a) 1 Reactor Vessel Water Level Once/ shift Once/ month Once/ operating cycle (Level 1) 2 Drywell Pressure Once/ shift Once/ month Once/ operating cycle 3

Reactor Vessel Steam Dome Once/ shift Once/ month Once/ operating cycle Pressure 4

Core Spray Sparger Once/ day N/A Once/ operating cycle w

Differential Pressure i

5 CS Pump Discharge Flcw Once/ shift Once/ month Once/ operating cycle l

None Once/ operating cycle None 6

Core Spray Logic Power Failure Mcaitor Notes for Table 4.2-6 The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be c.

established between items in Table 4.2-6 and items in Table 3.2-6.

6

- - ~ _ - _ _ _

_ ~ ~ _

... ~ _ _

n-n

-i Notes for Table 4.2-6 (Cont'd)

Instrument functional tests are not required when the instruments are not required to be operable or b.

are tripped. However, if functional tests are missed, they shall be performed prior to returning the instrument to an operable status.

Calibrations are not required when the instruments are not required to be operable. However, if c.

I calibrations are missed, they shall be performed prior to returning the instrument to an operable status.

t

[

u u

suw Logic system fuxtional tests and simulated automatic actuation shall be performed once each operating cycle for the following:

l 1.

Core Spray Subsystem The logic system functional tests shall include a calibration of time delay relays and timers necessary for proper functioning of the trip systems.

n

~ _ _ _ _ _... _ _ _,.. _,.. _

~... _ _,,,

m

_...a_

.R

• A, D

t Table 4.2-11 i

Check and Calibration Minimum Frequency for Instrumentation S

Which Provides Surveillance Information R.

Instrument Check Instrument Calibration II Ref.

Instrument Minimum Frequency Minimum frequency 5

No.

(b)

(c) g (a)

=

w 1

Reactor Vessel Water Level Each shift Once/ operating cycle (f)

=

2 Shroud Water level Each shift Once/ operating cycle (f) 3 Reactor Pressure Each shift Once/ operating cycle (f)

'S Each shift Every 6 months 4

Drywell Pressure l

Each shift Every 6 months l

5 Drywell Temperature 6

Suppression Chamber Air Each shift Every 6 months Teaperature 7

Suppression Chamber Water Each shift Every 6 months t

Temperature l

w L

8 Suppression Chamber Water Each shift Every 6 months i

Level

=

Each shift Every 6 months 9

Suppression Chamber Pressure 10 Rod Position Information Each shift N/A System (RPIS)

Each shift Every 6 months 11 Hydrogen and Oxygen i

Analyzer 12 Post LOCA Radiation Each shift Every 6 months 13 a) Safety / Relief Valve Position Pri-Monthly Every 18 months mary Indicator b) Safety / Relief Valve Position Monthly Every 18 months i

Secondary Indicator

I l

N 8

.A k

Notes for Table 4.2-11 ic The colume entitled "Ref No." is only for convenience so that a one-to-one relationship can be 3 4 a.

established between items in Table 4.2-11 and items in Table 3.2-11.

l

==

o 2-l j

bh b.

Instrument checks are not required when the instruments are not required to be operable or are 3

tripped. However, if instrument checks are missed, they shall be performed prior to returning the c-2 instrument to an operable status.

f U

Calibrations are not required when the instruments are not required to be operable or are

~~

c.

tripped. However, if calibrations are missed, they shall be performed prior to returning the instrument to an operable status.

l d.

Functional tests are not required when the instruments are not required to be operable or are tripped. However, if functional tests are missed, they shall be performed prior to returning the instrument to an operable status.

Calibration of a drywell high range monitor shall consist of an electronic calibration of the

.'d e.

channel, not including the detector, for range decades above 10 R/hr and one point calibration check of the detector below IG R/hr with an installed or portable gasuna source.

t f.

The entire loop shall be calibrated once per 18 months; however, the recorder itself must be calibrated at least once per 12 months.

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2 PROTECTIVE INSTRUMENTATION In addition to the Reactor Protection System (RPS) instrumentation which in-itiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of accidents which are beyond the operators ability to control, or terminates operator errors before they result in serious consequences. This set of Specifications provides the lim-iting conditions for operation of the instrumentation:

(a) which initiates reactor vessel and primary containment isolation, (b) which initiates or controls the core and containment cooling systems, (c) which initiates control rod blocks. (d) which initiates protective action, (e) which monitors leakage into the drywell and (f) which provides surveti-lance information. The objectives of these specifications are (i) to assure the effectiveness of the protective instrumentation when required by preserv-ing its capability to tolerate a single failure of any component of such sys-tems even during periods when portions of such systems are out of service for maintenance, and (ii) to prescribe the trip settings required to assure ade-quate performance. When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.

r l

Instrumentation Which Initiates Reactor Vessel and Primary Containment A.

Isolation (Table 3.2-1)

Isolation valves are installed in those lines which penetrate the primary con-tainment and must be isolated during a loss of coolant accident so that the radiation dose limits are not exceeded during an accident condition. Actua-tion of these valves is initiated by protective instrumentation shown in Table

(,

3.2-1 which senses the conditions for which isolation is required. Such in-strumentation must be available whenever primary containment integrity is re-quired. The objective is to isolate the primary containment so that the guidelines of 10 CFR 100 are ret exceeded during an accident. The events when isolation is required are ciscussed in Appendix G of the FSAR. The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.

1.

Reactor Vessel Water Level Reactor Vessel Water Level' Low (Level 3) (Narrow Range) a.

The reactor water level instrumentation is set to trip when reactor water level is approximately 14 feet above the top of the active fuel. This level is referred to as Level 3 in the Technical Speci-fications and corresponds to a reading of 10.0 inches on the Narrow Range Scale. This trip initiates Group 2 and 6 isolation but does not trip the recirculation pumps, b.

Reactor Vessel Water level Low Low (Level 2)

The reactor water level instrumentation is set to trip when reactor l water level is approximately 9 feet above the top of the active fuel.

This level is referred to as Level 2 in the Technical Spect-fications and corresponds to a reading of -47 inches.

l This trip initiates Group 5 isolation, starts the standby gas

(,

treatment system, and initiates secondary containment isolation.

Amendment No. JM.121 3.2-50

BASES FOR LIMITING CONDITIONS FOR OPERATION

(

3.2.A.1.c.

Reactor Vessel Water Level Low Low Low (Level 1)

The reactor water level instrumentation is set to trip when the reactor water level is approximately 51 inches above the top of l

the active fuel. This level is referred to as Level 1 in the Technical Specifications and corresponds to a reading of of -113 inches. This trip initiates Group 1 isolation.

l 1

l l

l l

l l

l l

l l

l l

i i

I I,

3.2-50a Anendment flo. 103, 121 i

a 1,4 I

BASES FOR LIMITING CONDITIONS FOR OPERATION

~

3.2.A.2.

Reactor Vessel Steam Dome Pressure (Shutdown Cooling Mode) Low Permissive l j

(_

This setpoint is chosen to preserve the pressure integrity of the RHR

~

system under conditions of increasing reactor pressure (startup). The RHR suction valves from the reactor (shutdown cooling mode) would be closed when the 145 psig setpoint is reached. This function protects l

against RHR system pipe breaks during the shutdown cooling mode of op-4, eration. Additionally, at reactor pressures'below this setpoint the primary containment isolation signals are permttted to close the in-i board motor operated injection valve (LPCI mode).

3.

Drywell pressure High i

The Bases for Drywell Pressure High are discussed in the Bases for Specif-j ication 3.1.A.5.

Pressure above the trip setting starts the SGTS and in-1 intiates primary and secondary containment isolation.

4 Main Steam Line Radiation High I

Radiation monitors in the main steam line tunnel have been provided to l

detect gross fuel failure as in the control rod drop accident.

This in-strumentation causes a Group 1 isolation. With the established setting j

of reproximately three times normal full power background, fission product re-j lease is limited so that 10 CFR 100 guidelines are not exceeded for this 7

i ac:ident.

Ref. Section 14.4.4 FSAR.

5.

Main Steam Line Pressure Low j

{N The Bases for Main Steam Line Pressure Low are discussed in the Bases for 1

l Specification 2.1.A.6.

J 6.

Main Steam Line Flow High Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.

In addition to monitoring steam flow, instru-mentation is provided which initiates Group 1 isolation. 'The primary func-1 tion of the instrumentation is to detect a break in the main steam line, For the worst case accident, a main steam lieg break outside the drywell.

the trip setting of 115 psid, corresponding to 138". of-rated stean flow, l

i

.I in conjunction with the flow limiters and main steam isolation valve clo-sure, limits the mass inventory loss such that fuel is not uncovered.

Fuel j

temperatures remain approximately 1000$F and release of radioactivity to the environs is well below 10 CFR 100 guidelines.

Ref. Section 14.6.5 of the FSAR.

c

.e i

7.

Main Steam Line Tunnel Temperature High i

l Temperature monitoring instrumentation is provided in the main steam line tunnel to detect leaks in this area. Trips are provided on this instrumen-i tation and when exceeded cause a Group 1 isolation.

Its setting is low enough to detect liais of the order of five to 10 gpm; thus, it-is capable l

of covering the entire spectrum of breaks.

For large breaks, it is a back-l j (

up to high steam flow instru.entation discussed above, and for small breaks s

i i

[

3.2-51 i

l 1

BASES FOR LIMITING CONDITIONS FOR OPERATION

'3.2.A.7 Main Steam Line Tunnel Temperature High (Continued) with the resultant small release of radioactivity, gives isolation before the guidelines of 10 CFR 100 are exceeded.

l 8.

Reactor Water Cleanup System Differential Flow High

~

Gross leakage (pipe break) from the reactor water cleanup system is detected by measuring the difference of flow entering and leaving the system. The set point is low enough to ensure prompt isolation of the l

cleanup system in the event of such a break but, not so low that spurious isolation can occur due to normal system flow fluctuations and instrument noise. Time delay relays are used to prevent the isola-tion signal which might be generated from the initial flow surge when the l

cleanup system is started or when operational system adjustments are made which produce short term transients.

9.

Reactor Water Cleanup Area Temperature High and 10.

Reactor Water Cleanup Area Ventilation Differential Temperature High Leakage in the high temperature process flow of the reactor water cleanup system external to the primary containment will be detected by temperature sensing elements. Temperature sensors are located in the inlet and outlet ventilation ducts to measure the temperature difference.

Local ambient temperature sensors are located in the compartment containing equipment and

(,-

piping for this system. An alarm in the main control room will be set to annunciate a temperature rise corresponding to a leakage within the identi-fied limit.

In addition to annunciation, a high cleanup room temperature will actuate automatic isolation of the cleanup system.

11. Condenser Vacuum Low The Bases for Condenser Vacuum Low are discussed in The Bases for Specifica-tion 2.1.A.7.

B.

Instrumentation Which Initiates or Controls HPCI (Table _3.2-2) 1.

Reactor Vessel Water level Low Low (Level 2) t The reactor vessel water level instrumentation setpoint which initiates HPCI i

is 2 -47 inches. This level is approximately 9 feet above l

the top of the active fuel and in the Technical Specifications is refer-red to as Level 2.

The reactor vessel low water level setting for HPCI system l initiation is selected high enough above the active fuel to start the HPCI j

system in time both to prevent excessive fuel clad temperatures and to pre-vent more than a small fraction of the core from reaching the temperature i

at which gross fuel failure occurs. The water level setting is far enough below normal levels that spurious HPCI system startups are avoided, 2.

Drywell Pressure High e

!g The drywell pressure which initiates HPCI is s2

(

psig. High drywell pressure could indicate a -failure of the nuclear j

system process barrier. This pressure is selected to be as low as possible j

without inducing spurious HPCI system startups. This instrumentation ser-i ves as a backup to the water level instrumentation described above.

3.2-52' Amendment No. 193, 121

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.B.3 HPCI Turbine Overspeed The HPCI turbine is automatically shut down by tripping the HPCI turbine stop valve closed when the 5000 rpm setpoint on the mechanical governor is reached. A turbine overspeed trip is required to protect the physi-cal integrity of the turbine.

4.

HPCI Turbine Exhaust Pressure High When HPCI turbine exhaust pressure reaches the setpoint (s 146 osig) the HPCI l turbine is automatically shut down by tripping the HPCI stop valve closed.

HPCI turbine exhaust high pressure is indicative of a condition which threat-ens the physical integrity of the exhaust line.

5.

HPCI Pump Suction Pressure Low A pressure switch is used to detect low HPCI system pump suction pressure and is set to trip the HPCI turbine at s 12.6 inches of mercury vacuum.

l This setpoint is chosen to prevent pump damage by cavitation.

6.

Reactor Vessel Water Level High (Level 8)

A reactor water level of +56.5 inches is indicative that the HPCI system has performed satisfactorily in providing makeup water to the reactor vessel. The reactor vessel high water level setting which trips the HPCI turbine is near the top of the steam separators and is sufficient to

(

prevent gross moisture carryover to the HPCI turbine. Two analog dif-ferential pressure transmitters trip to initiate a HPCI turbine shutdown.

I 7.

HPCI Pump Discharge Flow High l

J To prevent damage by overheating at reduced HPCI system pump flow, a pump discharge minimum flow bypass is provided. The bypass is contro led by an automatic, D. C. motor-operated valve. A high flow signal from a flow meter downstream of the pump on the main HPCI line will cause the bypass valve to close. Two signals are required to open the valve: A HPCI pump discharge pressure transmitter high differential pressure signal must be j,

received to act as a permissive to open the bypass valve in the presence of a low flow signal from the differential pressure transmitter.

NOTE:

Because the steam supply line to the HPCI turbine is part of the nuclear system process barrier, the following con-ditions (8-13) automatically isolate this line, causing shutdown of the HPCI system turbine.

8.

HPCI Emergency Area Cooler Ambient -Temperature High l

f High ambient temperature in the HPCI equipment room near the emergency area cooler could indicate a break in the HPCI system turbine steam line.

The automatic closure of the HPCI steam line valves prevents the ex-cessive loss of reactor coolant and the release of significant amounts of

(-

radioactive material from the nuclear system process barrier. The high Amendment No. N3,121 3.2-53 I

i

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.B.8 HPCI Emergency Area Cooler Ambient Temperature High (Continued) temperature setting of s 169 F was selected to be far enough above anti-cipated normal HPCI system operational levels to avoid spurious isolation but low enough to provide timely detection of HPCI turbine steam line break.

9.

HPCI Steam Supply Pressure Low Low pressure in the HPCI steam line could indicate a break in the HPCI steam line. Therefore, the HPCI steam line isolation valves are auto-matically closed. The steam line low pressure function is provided so in the event that a gross rupture of the HPCI steam line occurred up-stream from the high flow sensing location, thus negating the high flow indicating function, isolation would be effected on low pressure. The allowable value of 2 100 psig is selected at a pressure sufficiently high enough to prevent turbine stall.

10. HPCI Steam Line AP (Flow) High HPCI steam line high flow could indicate a break in the HPCI turbine steam line. The automatic closure of the HPCI steam line isolation valves l

prevents the excessive loss of reactor coolant and the release of signi-l ficant amount of radioactive materials from the nuclear system process barrier. Upon detection of HPCI steam line high flow the HPCI turbine steam line is isolated. The high steam flow trip setting of 303% flow

('

was selected high enough to avoid spurious isolation, i.e., above the high steam flow rate encountered during turbine starts.

The setting was selected low enough to provide timely detection of an HPCI turbine steam line break.

11. HPCI Turbine Exhaust Diaphragm pressure High High pressure in the HPCI turbine exhaust could indicate that the turbine rotor is not turning, thus allowing reactor pressure to act on the turbine exhaust line. The HPCI steam line isolation valves are automatically closed to prevent overpressurization of the turbine exhaust line. The turbine ex-haust diaphragm pressure trip setting of s 20 psig is selected high enough l

l to avoid isolation of the HPCI if the turbine is operating, yet low enough e

to effect isolation before the turbine exhaust line is unduly pressurized.

l

12. Suppression Chamber Area Ambient Temperature High i

A temperature of 169 F will initiate a timer to isolate the HPCI turbine steam line.

i l

l k

Amendment No. 9, 70it,121 3.2-54 i

t BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.B.13 Suppression Chamber Area Differential Air Temperature High A differential air temperature greater than the trip setting of s 42*F between the inlet and outlet ducts which ventilate the suppression chamber i

area will initiate a timer to isolate the HPCI turbine steam line.

i

14. Condensate Storage Tank Level Low e

The CST is the preferred source of suction for HPCI.

In order to provide an adequate water supply, an indication of low level in'the CST automat-ically switches the suction to the suppression chamber. A trip setting of 0 inches corresponds to 10,000 gallons of water remaining in the tank.

15. Suppression Chamber Water Level High A high water level in the suppression chamber automatically switches HPCI suction to the suppression chamber from the CST.

16. HpCI Logic Power Failure Monitor 4

e-The HPCI Logic Power Failure Monitor mcnitors the availability of power i

to the logic system. In the event of loss of availability of power to the logic system, an alarm is annunciated in the control room.

C.

Instrumentation Which Initiates or Controls RCIC (Table 3.2-3) 1.

Reactor Vessel Water Level Low Low (Level 2)

The reactor vessel water level instrumentation setpoint which initiates RCIC is t-47 inches. This level is approximately 9 feet above the top of the active fuel and is referred to as Level 2.

This setpoint insures that RCIC is started in time to preclude cont'itions which lead to inade-quate core cooling.

2.

RCIC Turbine Overspeed i

The RCIC turbine is automatically shutdown by tripping the RCIC turbine stop valve closed when the 125% speed at rated flow setpoint on the mech-anical governor is reached. Turbine overspeed is indicative of a condi-tion which threatens the physical integrity of the system. An electrical tachometer trip setpoint of 110*4 also will trip the RCIC turbine stop valve l

closed.

i 3.

RCIC Turbine Exhaust Pressure High When RCIC turbine exhaust pressure reaches the setpoint (s 45 psig), the l

l RCIC turbine is automatically shut down by tripping the RCIC turbine stop i

valve closed.

RCIC. turbine exhaust high pressure is indicative of a con-i dition which threatens the physical integrity of the exhaust line.

l 4.

RCIC Pump Suction Pressure Low 1

i One differential pressure transmitter is used to detect low RCIC system pump e

l(

suction pressure and is set to trip the RCIC turbine at s 12.6 inches of mer-j cury vacuum.

i

_ l Amendment No. JM, JM,121 3.2-55

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.C.5 Reactor Vessel Water Level High (Level 8)

A high reactor water level trip is indicative that the RCIC system has performed satisfactorily in providing makeup water to the reactor vessel.

The reactor vessel high water level setting which trips the RCIC turbine is near the top of the steam separators and sufficiently low to prevent gross moisture carryover to the RCIC turbine. Two differential pressure trans-mitters trip to initiate a RCIC turbine shutdown. Once tripped the system is capable of automatic reset af ter the water level drops below Level 8.

This automatic reset eliminates the need for manual reset of the system before the operator can take manual control to avoid fluctuating water levels.

6.

RCIC pump Discharge Flow l

To prevent damage by overheating at reduced RCIC system pump flow, a pump discharge minimum flow bypass is provided. The bypass is controlled by an automatic, D. C. motor-operated valve. A high flow signal from a flow meter downstream of the pump on the main RCIC line will cause the bypass valve to close. Two signals are required to open the valve: A RCIC pump discharge pressure transmitter high differential pressure signal must be l

received to act as a permissive to open the bypass valve in the presence of a low flow signal from the differential pressure transmitter.

l Note:

Because the steam supply line to the RCIC turbine is part of C,

the nuclear system process barrier, the following conditions (7 - 13) automatically isolate this line, causing shutdown of the RCIC system turbine.

7.

RCIC Emergency Area Cooler Ambient Temperature High l

High ambient temperature in the RCIC equipment room near the emergency area cooler could indicate a break in the RCIC system turbine steam line.

The automatic closure of the RCIC steam line valves prevents the exces-sive loss of reactor coolant and the release of significant amounts of radioactive material from the nuclear system process barrier. The high l

temperature setting of s 169 F was selected to be far enough above anti-cipated normal RCIC system operational levels to avoid spurious isolation but low enough to provide timely detection of a RCIC turbine steam line break.

8.

RCIC Steam Supply pressure Low Low pressure in the RCIC.

tm supply could indicate a break in the RCIC steam line. Therefore, the KCIC steam supply isolation valves are auto-matically closed. The steam line low pressure function is provided so that in the event a gross rupture of the RCIC steam line occurred up-stream from the high flow sensing location, thus negating the high flow The iso-indicating function, isolation would be ef fected on low pressure.

lation setpoint of 2: 60 psig is chosen at a pressure below that at which the RCIC turbine can effectively operate.

Amendment No. $$, /p3,121 3.2-56 l

s

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.C.9 RCIC Steam Line (AP) Flow High l

RCIC turbine high steam flow could indicate a break in the RCIC turbine steam line. The automatic closure of the RCIC steam line isolation valves prevents the excessive loss of reactor coolant and the release of sigaffi-cant amounts of radioactive materials from the nuclear system process barrier.

Upon detection of RCIC turbine high steam flow the RCIC turbine steam line is isolated. The high steam flow trip setting of 306% flow l

was selected high enough to avoid spurious isolation, i.e., above the high steam flow rate encountered during turbine starts. The setting was selected low enough to provide timely detection of a RCIC turbine steam line break.

10.

RCIC Turbine Exhaust Diaphragm Pressure High High pressure in the RCIC turbine exhaust could indicate that the turbine rotor is not turning, thus allowing reactor pressure to act on the turbine exhaust line.

The RCIC steam line isolation valves are automatically closed to prevent overpressurization of the turbine exhaust line.

The tur-bine exhaust diaphragm pressure trip setting of s 20 psig is selected high l enough to avoid isolation of the RCIC if the turbine is operating, yet low enough to effect isolation before the turbine exhaust line is unduly pres-surized.

11.

Scopression Chamber Area Ambient Temperature High l

As in the RCIC equipmer.t room, and for the same reason, a temperature of

(,

s 169 F will initiate a timer to isolate the RCIC turbine steam line.

l 12.

Suopression Chamber Area Differential Air Temperature High A high differential air temperature between the inlet and outlet ducts which ventilate the suppression chamber area will initiate a timer to l

isolate the RCIC turbine steam line.

13.

RCIC Logic Power Failure Monitor The RCIC Logic Power Failure Monitor monitors the availability of power to the logic system.

In the event of loss of availability of power to the logic system, an alarm is annunciated in the control room.

14.

Condensate Storage Tank Level Low The low CST level signal transfers RCIC suction from the CST to the 6

suppression pool. The setpoint was chosen to ensure an uninterrupted supply i

of water during suction transfer.

i

15. Suppression Pool Water Level High A high water level in the suppression chamber automatically switches RCIC suction from the CST to the suppression pool.

Amendment flo. 107, 103, 121 3.2-57

BASES FOR LIMITING CONDITIONS FOR OPERATION D.

Instrumentation Which Initiates or Controls ADS (Table 3.2-4)

The ADS is a backup system to HPCI.

In the event of failure by HPCI l

to maintain reactor water level, ADS will initiate depressurization of the reactor in time for LPCI and CS to adequately cool the core. Four signals are required to initiate ADS: Low water level, confirmed low water level, high drywell pressure, and either a RHR or Core Spray pump available. The simultaneous presence of these four signals will initiate a 120 second timer which will depressurize the reactor if not reset.

1.

Reactor Vessel Water Level Reactor Vessel Water Level Low (Level 3) a.

The second reactor vessel low water level initiation setting

(+10.0 inches) is selected to confirm that water level in the vessel is in fact low, thus providing protection against inadvertent depressurization in the event of an instrument line (water level) failure.

Such a failure could produce a simultaneous high drywell pressure. A confirmed low level is one of four signals required to initiate ADS.

b.

Reactor Vessel Water Level Low Low Low (Level 1)

The reactor vessel low water level setting of -113 inches is l

selected to provide a permissive signal to open the relief valve C

and depressurize the reactor vessel in time to allow adequate cooling of the fuel by the core spray and LPCI systems following a LOCA in which the other make up systems (RCIC and HPCI) fail to maintain vessel water level. This signal is one of four required to initiate ADS.

2.

Drywell pressure High A primary containment high pressure of 2 2 psig indicates that a breach of the nuclear system process barrier has occurred inside the dr'ywell. The signal is one of four required to initiate the ADS.

l t

k i

i Amendment No. 183, 121 3.2-58 i

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.0.

3.

RHR Pump Discharge Pressure High An RHR pump discharge pressure of 2 112 psig indicates that LPCI l

flow is available when the reactor is depressurized. The presence of this signal means low pressure core standby cooling is available.

Low pressure core standby cooling available is one of the four signals required to initiate ADS.

4.

Core Spray Pump Discharge Pressure High A core spray pump discharge pressure of 2137 psig indicates that' l

Core Soray flow is available when the reactor is depressurized.

The presence of this signal means low pressure core standby cooling is available.

Low pressure core standby cooling available is one-of the four signals required to initiate ADS.

5.

Auto Depressurization Timer The 120-second delay time setting-is chosen to be long enough so that the HPCI system has time to start, yet not so long that the core spray system and LPCI are unable to adequately cool the core if HPCI fails to start. An alarm in the main control room is annunciated each time either.of the timers is timing. Resetting the automatic depressurization system logic in the presence of tripped initiating signals recycles the timers.

6.

Automatic Blowdown Control Power Failure Monitor The Automatic Blowdown Control Power Failure Monitor monitors the availability of power to the logic system.

In the event of loss of availability of power to the logic system, an alarm is annunciated in the control room.

E.

Instrumentation Which Initiates or Controls the LPCI Mode of RHR 4

(Table 3.2-5) 1.

Reactor Vessel Water Level Low Low Low (Level 1)

Reactor vessel low water level (Level 1) initiates LPCI and indi-cates that the core is in danger of being overheated because of an insufficient coolant inventory. This level is sufficient to allow the timed initiation of the various valve closure and i

loop selection routines to go to completion and still successfully perform its design function.

i i

k Amendment No. 193, 121 3.2-59

f BASES FOR LIMITING CONDITIONS FOR OPERATION

(

2.

Drywell Pressure High l

Primary containment high pressure could indicate a break in the nuclear system process barrier inside the drywell. The high drywell l

pressure setpoint is selected to be high enough to avoid spurious starts but low enough to allow timely system initiation.

3.

Reactor Vessel Steam Dome Pressure Low l

The Bases for Reactor Pressure (Shutdown Cooling Mode) are discussed in the Bases for Specification 3.2.A.2.

With an analytical limit of 2 300 psig and a nominal trip setpoint of 370 psig, the recirculation discharge valve will close successfully during l

a LOCA condition.

Once the LPCI system is initiated, a reactor low pressure setpoint of 460 l

psig produces a signal which.is used as a permissive to open the LPCI in-jection valves. The valves do not open, however, until reactor pressure falls below the discharge head of LPCI.

4 l

I l

3.2-60 Amendment No. 793, 121

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.E.4 Reactor Shroud Water Level Low (Level 0)

A reactor water level 2-202 inches below instrument zero is indicative that LPCI has made progress in reflooding the core. A simultaneous high drywell pressure trip indicates the need for containment cooling. The 2-202 inch l

setpoint acts as a permissive for manual diversion for some of the LPCI flow to containment spray.

5. LDCI Cross Connect Valve Open Annunciator With the modified LPCI arrangement, the cross connect valve status was changed from normally open to normally closed.

Inadvertent opening of this valve could negate the LPCI system injection when needed. The annunciator will alarm when the LPCI cross connect valve is not fully closed.

6.

RHR (LPCI) pump Flow Low A flow element and differential pressure transmitter are provided down-stream of each pair " RHR pumps in their common line. To protect the pumps from overheatu.g at low flow rates, a minimum flow bypass with a Cl restricting orific is provided for each pump which routes water through the common motor,perated valve to the suppression chamber. This mini-mum flow byr:. valve automatically opens upon sensing low flow in the common discharge piping. The valve automatically closes whenever the flow (whether from both pumps or a single pump) is above,the low flow setting.

7.

RHR (LPCI) Pamp Start Timers l

If normal AC power is available, four pumps automatically start without delay.

If normal AC power is not available one pump starts without de-lay as soon as power becomes available from the standby sources. The i

i other three pumps start after a 10-second delay. The timer provides cor-i rect sequencing of the loads to the diesel generator.

i-i I

I k

Amendment No. 92,121 3.2-61

BASES FOR LIMITING CONDITIONS FOR OPERATION 3.2.E.8 Valve Sele en Timers l

After 10 minutes, a timer cancels the LPCI signals to the injection valves.

The cancellation of the signals allows the operator to divert the water for other post-accident purposes. Cancellation of the signals does not cause the injection valves to move.

9.

RHR Relay Logic Power Failure Monitor The RHR Relay Logic Power Failure Monitor monitors the availability of power to the logic system.

In the event of loss of availability of power to the logic system, an alarm is annunciated in the control room.

F.

Instrumentation Which Initiates or Controls Core Spray (Table 3.2-6)

1. Reactor Vessel Water Level Low Low Low (Level 1)

I A reactor low water level of -113 inches (level 1) initiates Core Spray l

This level is indicative that the core is in danger of being overheated because of an insufficient coolant inventory.

2.

Drywell Pressure High Primary containment high pressure is indicative of a break in the nuclear system process barrier inside the drywell. The high drywell pressure set-point of s 2 psig is selected to be high enough to avoid spurious system initiation but low enough to allow timely system initiation.

3.

Reactor Vessel Steam Dome Pressure Low Once the core spray system is initiated, a reactor low pressure setpoint of 460 psig produces a signal which is used as a permissive-to open the core l

spray injection valves. The valves do not open, however, until reactor pres-sure falls below the discharge head of the core spray system.

4.

Core Spray Sparger Differential Pressure A detection system is provided to continuously confirm the integrity of the core spray piping between the inside of the reactor vessel and the core shroud. A differential pressure switch measures the pressure difference be-tween the top of the core support plate and the inside of the core spray sparger pipe just outside the reactor vessel.

If the core spray sparger pip-ing is sound, this pressure difference will be the pressure drop across the core resulting from inter-channel leakage.

If integrity is lost, this pres-sure drop will include the steam separator pressure drop. An increase in the normal pressure drop initiates an alarm in the main control room.

4 k.

Amendment No. 183, 121 3.2-62 l

1

E BASES FOR LIMITING CONDITIONS FOR OPERATION i

3.2.F.5.

Core Spray pump Discharge Flow A differential pressure transmitter is provided downstream of each core spray l

l pump to indicate the condition of each pump. To protect the pumps from over-heating at low flow rates a minimum flow bypass line, which routes water from the pump discharge to the suppression chamber, is provided. A single motor-operated valve controls the condition of each bypass line. The minimum flow bypass valve automatically opens upon sensing low flow in the discharge line.

The valve nutomatically closes whenever the flow is above the low flow setting.

4 6.

Core Spray Logic Power Failure Monitor s

The Core Spray logic Power Failure Monitor monitors the availability of I

power to the logic system.

In the event of loss of availability of power to the logic system, an alarm is annunciated in the control room.

I G.

Neutron Monitoring Instrumentation Which Initiates Control Rod Blocks (Table 3.2-7)

These control rod block functions are provided to prevent excessive control

~

rod withdrawal so that MCPR does not decrease to 1.07.

The trip logic for

{-

this function is 1 out of n:

e.g., any trip on one of six APRM's~, eight IRM's or four SRM's will result in a rod block.

i The minimum instrument channel requirements assure sufficient instrumentation to assure that the single failure criteria is met.

1.

SRM 4

a.

Inoperative i

This rod block assures that no control rod is withdrawn during low ~

i neutron flux level operations unless proper neutron monitoring capa-bility is available, in that all SRM channels are in service or properly bypassed, I

f b.

Not Fully Inserted i

Any source range monitor not fully inserted into the core when the SRM count rate level is below the retract permit level will cause a rod i

i block. This assures that no control-rod is withdrawn unless all SRM detectors are properly inserted when they must be relied upon to pro-vide the operator with a knowledge of the neutron flux level.

c.

Downscale This rod block assures that no control rod is withdrawn unless the k.

SRM count rate is above the minimum prescribed for low neutron flux level monitoring.

i Amendment No. 52,.121 3.2-63

Table 3.7-1 (Concluded)

Primary Containment Isolation Valves These notes refer to the lower case letters in parentheses'on the previous page.

1 i

NOTES:

a.

Key:

0 = Open SC = Stays closed C = Closed GC = Goes closed b.

Isolation Groupings are as follows:

GROUP 1: The valves in Group 1 are actuated by any og of the following conditions:

1.

Reactor vessel water level Low Low Low (Level 1) 2.

Main steam line radiation high 3.

Main steam line flow high 4.

Main steam line tunnel temperature high 5.

Main steam line pressure low 6.

Condenser vacuum low GROUP 2: The valves in Group 2 are actuated by one om of. the following conditions:

1.

Reactor vessel water level low (Level 3) 2.

Drywell pressure high GROUP 3:

Isolation valves in the high pressure coolant injection (HPCI) system are actuated by any one of the following conditions:

1.

HPCI steam line flow high 2.

High temperature in the vicinity of the HPCI steam line 3.

HPCI steam supply pressure low 4.

HPCI turbine exhaust diaphragm pressure l

GROUP 4: Primary Containment Isolation valves in the reactor core isolatian cooling (RCIC) system are actuated by any one of the following conditions:

RCIC steam line flow high

'igh temperature in the vicinity of the RCIC steam lin s kc'" steam supply pressure low l

GROUP 5: The a:;as in Group 5 are actuated by any one of the following conditions:

1.

Reactor vessel water level Low Low (Level 2) l 2.

Reactor water cleanup area temperature high 4

3.

Reactor water cleanup area ventilation differential temperature high 4.

Reactor water cleanup system differential flow high i

5.

Actuation of Standby Liquid Control System - closes outside valve only

)

6.

High temperature following non-reg'enerative heat exchanger - closes outside valve only GROUP 6: The valves in Group 6 are actuated by the following conditions:

I 1.

Reactor vessel water level low (Level 3) 2.

Reactor vessel steam dome pressure low permissive l

?

Requires a Group 2 signal or a Reactor Building ventilation high radiation c.

i isolation signal.

d.

For redundant lines, only one set of valves is listed. Other sets are identical i

except for valve numbers, which are included. Valve numbers are listed in order j

f from within primary containment outward for each line.

e.

Not applicable to check valves.

I I

Amendment No. 703, 121 3.7_19

-