ML20012B698

Proposed Tech Specs,Reflecting Editorial &/Or Administrative Changes to Make Unit 1 Tech Specs Consistent W/Unit 2 Tech Specs & BWR/4 STS

ML20012B698
Person / Time
Site:	Hatch
Issue date:	03/02/1990
From:	GEORGIA POWER CO.
To:
Shared Package
ML20012B697	List: ML20012B696 ML20012B698
References
NUDOCS 9003160132
Download: ML20012B698 (39)

Text

'}

n J

~LIMlllNG COND1110NS FOR OPERATION SURVEllLANCE REOUIREMENTS 3.1.

REACTOR PROTECTION SYSTEM (RPS) 4.1.

REACTOR PROTECTION SYSTEM (RPS) 4

-Acolicability ADolicability i

The Limiting Conditions for Operation The Surveillance Requirements asso-associated with the Reactor Protection ciated with the Reactor Protection System apply to the instrumentation and System apply to the instrumentation associated devices which initiate a and associated devices which initiate reactor scram, a reactor scram.

Objective Objective

~The objective of the Limiting Condi-The objective of the Surveillance tions for Operation is to assure Requirements is to specify the type-the operability of the Reactor and frequency of surveillance to be Protection System, applied to the protection instrumen-tation to assure operability.

Specifications SDecifications l^

A.

Sources of a Trio Sional Which A.

Test and Calibration Reautrements Initiate a Reactor Scram for the RPS The instrumentation requirements RPS instrumentation systems and associated with each source of a associated systems shall be func-scram signal shall be as given in tionally tested and calibrated as Table 3.1-1.

indicated in Table 4.1-1.

I The~ action to be taken if the number The trip system containing the of operable channels is not met for unsafe failure may be placed in both trip systems is also given in the untripped condition during the Table 3.1-1.

period in which surveillance testing is being performed on the other RPS channels.

B.

Core Maximum Fraction of B.

Core Maximum Fraction of Limitina Power Density (CMFLPD)

Limitina Power Density (CMFLPD)

This section deleted.

This section deleted.

i HATCH - UNIT 1 3.1 -1 Proposed TS/03174/305-105 P

Table 85.1-1 (Cont.)

-4 O-instrument Check Instrument Functional Test instrument Calibration.;

Sc ram Number Source of Scram Trip Signal Group-Minimum Frequency Minimum Frequency Minimum Frequency a

fa)

(b)

(c) 9 Main Steam line High Radiation B

D Every 3 months (e)

Every 3 months (i) 10 Main Steam Line Isolation Valve A

NA Every 3 months

.(h)'

• -a

-Closure 11 Turbine Control Valve Fast A-NA Every 3 months (J)

Once/ Ope rat ing Closure

. Cycle (k) 12 Turbine Stop Valve Clusure A

NA Every 3 months th)

RPS Channel Switch A

NA Once/ Operating Cycle Not Applicable Turbine First Stage Pressure A

NA Every 3 months Every 6 months Permissive I

a.

The column entitled " Scram Number" is for convenience so that a one-to-one relationship can be established between items in Table 4.1-1 and items in Table 3.1-1.

b.

The definition for each of the four groups is as rollows:

w Group A.

On-orr sensors that prov8de a scram trip signal.

cc Group B.

Analog devices Coupled with bi-stable trips that provide a screm trip signal.

' Group C.

Devices which only serve a useful function during some restricted mode or operation, such as startup or shutdown, or for which the only practical test is one that can be perrormed at shutdown.

Group D.

Analog transmitters and trip units that provide a scram trip runction.

c.

Functional tests are not required when the systems are not required to be operable or are tripped.

~

However, if functional tests ars missed, they shall be performed prior to returning the systems to an operable status, s

d.

Ca l i b ra t i ons a re no t requi red when the systems a re no t requ i red to be ope rab l e o r a re t ri pped.

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

~o 7

e.

This instrumentation is exempted from the instrument functional test derinition. This instrument O

functional test will consist of injecting a simulated electrical signal into the measurement o

channels.

cL r.

Deleted d

g.

The water level in the reactor will be perturbed and the corresponding level indicator changes will N

be mon i to red.

This perturbation test will be performed every 3 months arter completion of the O

runctionaI test program,

s. s CD h.

Physical inspection and actuation or these position switches will be performed once per operating cycle.

oN W

i. Standard current source used which provides an instrument channel alignment. Calibration using a radiation source shall be made once per operating cycle.

e5 J.

Measure time interval rrom EHC pressure switch actuation to RPS relay K14 de-energization.

(.o Proposed TS/0318q/305-163 s.

.. ~ ~.

~~..

.=

Notes for Table 3.2-2 (Cent.)

>-e

.>c7z b'-

' When any CCCS subsystem is required to be operabis by Section 3.5, there shall be two operable trip systems. If the required number of operable channels cannet be set for one of the trip systees, place the inoperable channel in the tripped condition or declare the associGted CCCS inoperable

.C within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

If the required number of operable channels cannot'be met for both trip systems, 3,

declare the associated CCCS inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

-4<

w i

ro a

l N

l f

l l

T

,O O

vs fD C2.

l M

in NOw FO w

O NW C3 Ut tC L

a NOTES FOR TABLE 3.2-3 a.

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

b.

When any CCCS subsystem is required to be operable by Section 3.5, there shall be two operable trip systems.

If the required number of operable channels cannot be met for one of the trip systems, place the inoperable channel in the tripped condition or declare the associated CCCS inoperable within I hour.

If the required number of operable channels cannot be met for both trip systems, declare the associated CCCS t

inoperable within I hour.

I l

l l

1 l

L HATCH - UNIT 1 3.2-9a Proposed TS/0321q/305-101

aI a

Table 3.2-4 h

INSTRUMENTATION WHICH INITIATES OR CONTROLS ADS H$

Ref.

Instrument-Trip Required Trip Setting Rema rks No.

Condition OperabIe

-(a)

Nomenclatu re Channels c

per Trip

,2 System Ibl H

1.

Reactor vessel Water Level Low (Level 3) 1 210.0 inches

' Confirms low level, ADS' permissive Reactor Vessel Water Level Low Low Low 2

2-113 inches Permissive signal to ADS timer (Level.1) 2.

Drywell Pres su re High 2

51.92 psig Permissive signal to ADS timer 3.

RHR Pump Discharge High 2

2112 psig Permissive signal to ADS timer Pressure 4.

CS Pump D i scha rge High 2

2137 psig Permissive signal to ADS timer Pressure 5.

Auto Depressurization 2

513 minutes Bypasses high drywell pressure Low Water Level Timer permissive upon sustained Level' I 6.

Auto Depressurization 1

120 12 seconds With Level 3 and Level 1 and high w

Timer d rywe l l pressure and CS or RHR pump Nb at pressure, timing sequence begins.

If the ADS timer is not o

reset it will initiate ADS.

7.

Automatic Blowdown Control

'1 Mot applicable Moni tors ava i labi l i ty of" power to Power Fa ilure Monitor logic system The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be established between

.a.

items in Tsble 3.2-4 and items in Table 4.2-4.

y b.

When any CCCS subsystem is required to be operable by Section 3.5, there shall be two operable o

trip systems. If the required number of operable channels cannot be met for one of the trip systems, V

place the inoperable channel in the tripped condition or declare the associated:CCCS inoperable 3

witttin 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

If the required number of operable channels cannot be met for bcrth trip systems, g

declare the associated CCCS inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

YAM5g Proposed TS/0321g/305-122 -

• m l3

's-

[~

14:

llotes for Table 3.2-5.

l

.x 3=

-4 c7

r:

.a.

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

~

C b.

When any CCCS subsystem is required to be operable by Section 3.5, there shall be two operable 3

trip systems. If the required number of operable channels cannot be met for one of the trip systems,

-4 place the inoperable channel in the tripped condition or declare the associated CCCS Inoperable within l' hour. ' If the required number of operable channels cannot be met for both trip systems, g

declare the associated CCCS Inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.,

.ta N

3w (A3 7

,O O

(n tD O.

-4

(,n NO m

Proposed TS/0321g/305-0 w

O N

(4 O

tra 8

O

,;~

+f"_'l Table 3.2-6~

INSTRUMENTATION WHICH INITIATES OR CONTROLS CORE SPRAY

-4 Q Ref.

Instrument Trip Required Trip Setting Ramarks fio.

Condition Ope rable 8

(a)

NomencIature ChanneIs e-per Trip z

System fb1

--e M

1.

Reactor Vessel Water Level Low Low Low 2-113 inches Initiates CS.

( 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 AOS.

3.

Reactor Vessel Steam-Dome Low 2

2as22 psig*

Permissive to open CS Pressure injection valves.

Is.

Core Spray Spa rger 1***

$ 3.1 psid Monitors integrity of CS Di f fe rentia l Pressure greater (less piping inside vessel (between negative) than the nozzle and core shroud).

the normal g

indicated P at rated core power to and flow.

5 5.

CS Pump Discharge Flow Low 1

-2610 gpm Minimum flow bypass line is (2 as.13 inches) closed when low flow signal

'is not present.

6.

Core Spray Logic Power 1

Not AppIIcable Monitors availability of Failure Monitor power to logic system.

o

• This trip function shall be 5500 psig.

(D a.

The column entitled "Ref. No." is only for convenience so that a one-to-one relationship can be estabilshed between items in Table 3.2-6 and items in-Table 88.2-6.

-4 Q

b.

When any CCCS subsystem is required to be operable by Section 3.5, there shall be two operable o

trip systems. If the required number of operable channels cannot be met for one of the trip systems, W

place the inoperable channel in the tripped comiition or declare the associated CCCS inoperable U

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, if the required number of operable channels cannot be met for both trip systems, Q

declare the associated CCCS inoperable within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />._,_

c.

Ala rm only. When inoperable, verify.that the core spray differential pressure is within limits at least once per y

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or, declare the associated core spray loop inope rab le.

5 Proposed TS/0321g/305-1as6

.,,, - mm

-y d'

Table 3.2-8'

.m-D.

RADI ATI001 MOIIITORIIIG. SYSTEMS WH'ICH LIMIT. RADIOACTIVITY. RELEASE n

Z Ref.

Instrument

' Trip'..

Operable there are not two operable.

Remarks Required Trip Setting

- Action to be ta ken i f '

e No.

Condition e (a)

Nomencla-Channels or tripped trip systees -

z ture per Trip q

System ib) e*

1.

Of f-os s '

Upscale /-

1~

- At a value not (c)

(d)

- 2 upscales,' or 1 Post Treatment Downscale to exceed the downscale and 1

+

Radiation equivalent of.

Upsca le, or. 2 down.

Monitors the stack re-scales will' isolate lease limit the SJAE off-gas indicated in EnvironmentaI Tech Specs 2.

Rsfueling Floor Upscale 2

. At a va lue not '

Cease refueling opera-2 upscale vi18' Exhaust vent to exceed the tions, if in progress.

Isolate the secondary Rad ia tion equivalent of isolate the secondary containment and Moni to rs the stack re-

- containment and start initiate the standby.

Ioase limit the standby gas treet-gas treatment system indicated in ment system.

EnvironmentaI Tech Specs 3.

Reactor Bldg.

Upscale 2

520 ar/hr Isolate the secondary

' 2 upscale vill isolate y

Exhaust vent containment, sta rt stand-the secondary con -

to Radiation by gas treatment system, tainment and initiate b

Mon i to rs close primary conta in-the standby gas cp ment and vent valves.

treatment system.

4.

Control Room Downscale.

1.

to.015 mr/hr Refer to Specifications I upscale or 2 down-Intake 3.12.C. and 3.12.D.

scales will actuate Radia t ion HI 51.0 ar/hr the MCRECS in the-Mon i to rs control room pres-surization mode.

t3

,o O

to O.

M Proposed TS/0321g/305-51 O

ta f9Y N

C4 O

ts1 9

(31 I

J.--

u.

~,,

we n.,

.x a

.,n c

c e.,

~

^

'. ~

~

flotes' for Table 4.2-8 (Cont'd)

T3=

2.

Standby Gas Treatment System Actuation -

-A

'S

3. _ Stese Jet-Air' Gector off-gas Actuation-4.

Primary Containment Purge and Vent Valve Closure e

5.

MCRECS Control. Room Pressurization Mode Actuation

--4 6.

(Deleted) 7.

Mechanical Vacuum Pump Isolation' The logic system functional test shall include a calibration of-time delay relays and timers necessary for proper functioning of* the trip systems.

e f-r0 I

b A

r T

,O O

us CL H

t.n NOus n

NW O-o

~

m

..,s.

4-g err y

g M.

y

- wg g

t',' iip-t

..g gW7g-i.-ggr*

.g

.gg e

a,

' - y

+

}

W je._

t.

LIMlilNG COND,111pNS FOR OPERAllDN SURVEILLANCE REOUIREMENTS l

4.4.A.3.

Each Doeratino Cvele (Continued) c.

vessel? This test checks the explosive charge, proper opera-tion of the associated valves-and selected pump operability.

The replacement charge to be installed will be selected from a manufactured batch which has been tested.

d.-

Both loops including both explo-sive valves should be tested in the course of two operating cycles.

e..

Prior to startup, verify (by i

analysis) that the sodium pentaborate enrichment is within prescribed limits.

l 3.4.B.

Doeratina with Incoerable B.

Surveillance with Inocerable Components Components If one Standby Liquid Control (Deleted) redundant component is inoperable the reactor may remain in operation for a period not to exceed seven (7) days provided the redundant component is operable.

l l

C.

Sodium Pentaborate Solution C.

Sodium Pentaborate Solution At all times when the Standby The following tests shall be Liquid; Control System is re-performed to verify the avail-quired to be operable the bility of the liquid control following conditions shall be solution:

not:

'1.

Volume l '. Volume The volume of the liquid Check the standby liquid i

control solution in the control tank volume at least liquid control tank shall once per day, be maintained as required

'in Figure 3.4-1.

2.

Concentration 2.

Concentration The concentration of the Check the concentration of the liquid control tank shall liquid in the standby liquid be maintained as required control tank by chemical in Figure 3.4-1.

analysis:

HATCH - UNIT 1 3.4-2 Proposed TS/0319q/305-142

p=

n j E, a

LINillNG CONDI110NS FOR OPERATION

$URVEILLANCE REOUIREMENTS 3.5.

CpRE AND CONTAINMENT COOLING 4.5.

CORE AND CONTAINMENT COOLING SYSTEMS SYSitMS

'ADDliCabilitV

- ADDlicabilitV The Limiting Conditions for The Surveillante Requirements Operation apply to the apply to the core and containment operational status of the core cooling systems when the corres-and containment cooling systems, ponding limiting conditions for operation are in ef fect.

Ob.iective QLiective The objective of the Limiting The objective of the Surveillance Conditions for Operation is to Requirements is to verify the assure the operability of the operability of the core and con-core and containment cooling tainment cooling systems under all systems under all conditions conditions for which this cooling for which this cooling capa-capability is an essential response bility is an essential to plant abnormalities, response to plant abnor-malities.

Specifications Specifications A.

Core Sprav (CS) System A.

Core SDraV (CS) SVstem 1.

Normal Svstem Availability 1.

Normal Operational Tests a.

The CS System shall be operable CS system testing shall be performed as follows:

'(1) Prior to reactor startup from a cold condition, or 11q!g -

Freauency (2) When irradiated fuel is in the a.

Simulated Once/ Operating reactor vessel and the reactor Automatic

Cycle, pressure is greater than Actuation atmospheric pressure, except as Test stated in Specification 3.5.A.2.

b.

System flow Once/3 months.

rate: Each CS pump can develop at least 4250 gpm against a system head-corresponding to a reactor vessel pres-sure of at least 113 psig.

c.

Valve lineups: Once/31 days.

Verify that each valve in the flow path that is not locked, sealed.

l or otherwise secured in posi-tion is in its correct position.

d.

(Deleted) l 1

l.

HATCH - UNIT 1 3.5-1 Proposed TS/0330g/305-150 i

~

.. 5 F

I".,

l LIMillNG CONDil10NS FOR OPERAlION SURVilLLANCE REQUIREMENI$

~

'3.5.A.2.- onaration with Inonarable 4.5 A.2.

Surveillance with Inonarable Comoonents Components if one CS system loop is inoper-(Deleted) able, the reactor may remain in operation for a period not to exceed 7 days providing all l

active components in the other CS system loop, the RHR system LPCI mode and the diesel generators (per Specif hei'....

4.g.A.2.a) are operable.

When performing an inservice hydrostatic or leakage test with the reactor coolant temperature above or below 212'F.the CS system is not required to be operable.

3.

Shutdown Requirements if Specification 3.5.A.l.a. or 3.5.A.2. cannot be met the reactor shall be placed in 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 />.

B.

Residual Hett Removal (RHR)

B.

Rgi@al Heat Removal (RHR)

System (LPCI and Containment Sysiem (LPCI and Containment Coo lino Model Coolino Modf1

+

1.

Normal System Availability 1.

Normal Doerational Tests RHR system testing shall be performed as follows:

11.ent Frecuency e

a.. The RHR System shall be operable:

a.

Air test on Once/10 years.

-l~

drywell head-

' (1) Prior to reactor startup ers and nozzles from a cold condition, or and air or water test on (2) When irradiated fuel is in torus headers the reactor vessel and the and nozzles reactor pressure is greater than atmospheric except as stated in Specification 3.5.B.2.

l l

l HATCH - UNil 1 3.5-2 Proposed TS/0330q/305-160

a

. +

j;

~ *:

LIMlilNG CONDlilDNS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.5,8.1.

Normal System Availability (Cont.)

4.5.B.I.

Normal Coerational Tests

b. ' One RHR loop with two pumps or two

,[13!!

Frecuency loops with one pump per loop shall be operable in the shutdown cool-b.

Simulated Once/ Operating

= ing mode when irradiated fuel is Automatic Cycle.

in the reactor vessel and the Actuation reactor pressure is atmospheric Test except prior to a reacter startup as stated in Specification 3.5.B.l.a.

During an inservice hydrostatic or leakage test, one RHR loop with two pumps or two loops with one pump per loop shall also be operable in the LPCI mode.

c.

The reactor shall not be started up c.

System flow Once/3 months, with the RHR system supplying ratet Each cooling to the fuel pool.

RHR pump shall deliver d.

During reactor power operation, the at least 1700 LPCI system discharge cross-tie gpm against a valve, Ell-F010, shall be in the system head closed position and the associated corresponding valve motor starter circuit to a reactor breaker shall be locked in the vessel pressure

off position. In addition, an of at least 20 annunciator which indicates that psig.

the cross-tie valve is not in the fully closed position shall be d.

Valve lineups: Once/31 days.

available in the control room.

Verify that each valve in e.

Both recirculation pump discharge the flow path valves shall be operable prior to that is not reactor startup (or closed if per-locked, sealed, mitted elsewhere in these speci-or otherwise fications),

secured in posi7 tion is in its correct position.

2.

Doeration with InoDerable e.

(Deleted)

Components f.

Both recirculation pump discharge a.1 One LPCI Pumo Inoperable valves shall be tested for oper-ability during any outage exceeding.

If one LPCI pump is inoperable, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, if operability tests have the reactor may remain in opera-not been performed during the tion for a period not to exceed preceding month.

7daysprovidedthattheremainingl LPCI pumps, both LPCI subsystem 2.

Surveillance with Inocerable flowpaths,theCSsystem,andthel Components associated diesel generators are operable (per Specification a.

(Deleted) 4.9.A.2.a).

b.

One LPCI Subsystem Inocerable b.

(Deleted)

A LPCI subsystem is considered to be inoperable if (1) both of the LPCI pumps within that system are inoperable or (2) the active valves in the subsystem flow path are inoperable.

HATCH - UNIT 1 3.5-3 Proposed TS/0330q/305-160

5 LIMITING CDNDITIONS FOR OPERAllDN SURVEILLANCE RE0VIREhEN15 3.5.B.2. Doeration with lnoperable 4.5.B.2. (Deleted)

{omponents (Con",inued) b, if one LPCI subsystem is inoper-able, the reactor may remain in operation for a period not to exceed 7 days provided that l

a11' active components of the remaining LPCI subsystem, the CS system, and the assoc)eted diesel generators are operable (per Specification 4.9. A.2.a).

c.

When performing an inservice hydrostatic or leakage test with the reactor coolant temperature above or below 212'F comply with Specification 3.5.B.l.b.

P w

i 1

l L

L F

HATCH - UNIT-1 3.5-4 Proposed TS/03304/305-160

r-q

3,~

e-F LIMiilNG COND1110NS FOR CPERATION SURVillLANCE REOUIREMENTS 3.5.B.3.

Shutdown Reauirements if Specification 3.5.B.1.a. or 3.5.8.2. cannot be met, the reactor shall be placed in 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 />.

C.

RHR Service Water System 4.5.C.

RHR Service Water System 1.

Normal System Availability 1.

Normal Operational Tests The RHR service water system RHR service water system testing shall be operable:

shall be performed as follows:

111m Freauency a.

Prior to reactor startup a.

Valve lineups: Once/31 days.

from a Cold Shutdown Verfiy that Condition,-or each valve in the flow path that is not locked, sealed, or otherwise secured in posi-tion is in its correct position, b.

When irradiated fuel is in b.

Pump Capacity Once/3 the reactor vessel and-the Testa months.

reactor vessel pressure is Each RHR ser-greater than atmospheric vice water pressure except as stated in pump shall Specification 3.5.C.2, or l

deliver at least 4000 gpm at a system head of at least 847 feet.

c.- When irradiated fuel is in the reactor vessel and the' reactor is depressurized at least one RHR service water loop shall be operable.

2.

One Pump Inoperable 2.

One Pumo Inoperable if one RHR service water (Deleted) pump is inoperable the reactor may remain in operation for a period not to exceed 30 days provided all other active components of both subsystems are operable.

When performing an inservice hydrostatic or leakage test, comply with Specification 3.5.C.I.c.

HATCH - UNIT 1 3.5-5 Proposed TS/03304/305-160

t L-. -

JIMlilNG CON 01110NS FOR OPERA 110N SURVEILLANCE REQUIREMENTS 3.5.C.3.. Two Pumps inopfrikit 4.5.C.3.

Two Pumps Inoperable If two RHR service water pumps are (Deleted) inoperable, the reactor may remain 3

i in operation for a period not to exceed 7 days provided all redun-l dent active components in both of the RHR service water subsystems are operable.

4.

Shutdown Reouirements If Specifications 3.5.C cannot be i

met, the reactor shall be placed

-in the Cold Shutdown Cendition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D.

)figh Pressure Coolant Injection D.

Hinh Pressure Coolant injection (HPCI) System (HPCI) System 1.

Normal System Availability 1.

Normal Doerational Tests HPCI system testing shall be performed as follows:

ligg Freauency a.

The HPCI System shall be a.

Simulated Once/ Operating operable:

automatic

Cycle, actuation (1) Prior to reactor startup test from a cold condition, or b.(1) Flow rate Once/3 (2) When irradiated fuel is in for a system

months, the reactor vessel and the head corre-greaterthan150psig,exceptl reactor vessel pressure is sponding to a reactor as stated in Specification vessel pres-3.5.0.2.*

sure of 11000 psig when steam is being sup-plied to the turbine at 5 1000 psig, and

-(2) Flow rate for Once/ Operating a system head Cycle.

corresponding to a reactor vessel pres-sure of 1 165 psig when

. steam is being supplied to the turbine at 165 i 15 psig.

• HPCI is not required to be operable for performance of inservice hydrostatic or leak testing with reactor pressure greater than 150 psig and all control rods inserted.

HATCH.- UNIT 1 3.5-6 Proposed TS/03300/305-147

I LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 4.5.0.1.b.

Normal Operational Tests The HPCI pumps shall deliver at least 4250 gpm during each flow rate test.

3.5.D.2.

Doeration with inocerable c.

Valve lineups: Once/31 days.

Comoonents Verify that each valve in If the HPCI system is it,9perable, the flow path the reactor may remain it, opera-that is not tion for a period not to exceed locked, sealed.

' fourteen (14) days provided the or otherwise ADS, CS system, RHR system LPCI secured in posi-mode, and RCIC system are operable.

tion is in its correct position.

With the surveillance requirements of Specification 4.5.0.1. not per-2.

Surveillance with Inocerable formed at the required frequencies

[omoonents due to low reactor steam pressure, reactor startup is permitted and (Deleted) the appropriate survelliance will be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> af ter reactor steam pressure is adequate (i.e., reactor pressure is such that the required steam pressure is maintained at the turbine for the duration of the test) to perform the tests.

3.

Shutdown Reauirements If Specification 3.5.D.1. or 3.5.0.2. cannot be met, an orderly shutdown shall be initiated and the reactor vessel pressure shell be reduced to 150 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

E.

$tg.ctorCoreIsolationCoolina E.

Peactor Core Isolation Coolino (RCIC) System (RCICI System 1.

Normal System Availability 1.

Normal Ooerational Tests a.

The RCIC system shall be RCIC system testing shall be per-operable with an operable formed as follows:

flow path capable of (auto-l-

matica11y) taking suction its Freauency i

from the suppression pool i

and transferring the water a.

Simulated Once/ Operating 1

l to the reactor pressure Automatic Cycle.

I vessel:

Actuation i

(and restart *)

l (1), Prior to reactor startup Test.

E from a cold condition, or I

a.(2)- When there is irradiated l.

fuel in the reactor vessel and the reactor pressure I

is above 150 psig, except as stated in Specification 3.5.E.2.*

i

• Automatic Restart on a low Water Level which is subsequent to a High Level Trip.

HATCH - UNIT 1 3.5-7 Proposed TS/0330g/305-116

.~

g

)

i

('

l

~!

1.

t e

p ri 8

c..

LIN111NG CONDITIONS FOR OPERAll0N SURV[lLLANCE REQUIREMENTS

^

f[

(3.5.E.1.

Nonaal System Availability (cont.)

4.5.E.1. Normal Doerational Tests (Cont.)

j L,

b.

Verifying that suc-Once/-

tion for the RCIC Operating system is automati-Cycle.

cally transferred from the CST to the suppression pool on:

a simulated-low CST 1evel or high sup-pression pool level signal.

.i

.E c.(1)- Flow rate when Once/3 steam is being

months, supplied to the turbine at nor-mal reactor ves-sel operating pressure,1000 +.

l 20,-80 psig, and (2) Flow rate when Once/

steam is being operating i

supplied to the Cycle.

turbine at a pres-sure of 150 + 15

-0 psig.

The RCIC pump shall deliver at least 400 gpm during each flow test.

d.

Valve lineups:

Once/31 days.

Verify that each valve in the flow path 2.

Doeration with Inocerab11 that is not

'i Gamponents locked, sealed.

or otherwise If the RCIC system is inoperablo,

' secured in posi-the reactor may remain in oper-tion is in its t

ation for a per.iod not to exceed correct position.

7 days if the HPCI system is l

operable during such time.

~

.e.

(Deleted)

With the surveillance requirements of Specification 4.5.E.1 not 2.

Surveillance with inocerable

' performed at the required fre-Components quencies due to' low reactor steam pressure, reactor startup (Deleted)

(

is permitted and the appropriate surveillance will be performed i

within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate (i.e., reactor pressure is such that the required steam pressure is raintained at the turbine for the duration of the test) to perform the test.

3.

If Specification 3.5 E.1. or 3.5.E.2. is not met, an orderly shutdown shall be initiated and the reactor shall be depressurized to less than 150 s

psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

• RCIC is not required to be operable for performance of inservice hydrostatic or leak testing with reactor pressure greater than 150 psig and all control rods inserted.

HATCH - UNIT 1 3.5-8 Proposed TS/0330q/305-137 i

m

y LIMlilNG COND1110NS FOR OPERA 110N SURVEILLANCE RE0VIREMEN15 3.5.F.

Automatic Depressurization System 4.5.F.

Automatic Depressurization System (ADS)

(ADS) 1.

Normal System Availability 1.

Normal Doerational Tests The seven valves of the Automatic Depressurization System shall be operable:

a.

Prior to reactor startup f rom a a.

A simulated automatic cold' shutdown, or actuation test shall be performed on the ADS prior to startup after each refueling outage. Surveil-lance of all relief valves is covered in Specification 4.6.H.

b.

When there is irradiated fuel in b.

A leak rate test of each the reactor vessel and the ADS valve accumulator, check reactor is above 113 psig except valve, and actuator assembly as stated in Specification shall be performed during 3.5 F.2.*

each refueling outage at a pressure of 90118 psig.

The leakage rate shall be verified to be $ 4.5 SCFH.

2.

Doeration with Inoperable 2.

Surveillance with inoDerable CoaJonents-Components j

If one of the seven ADS valves is (Deleted) known to be incapable of automatic operation, the reactor may remain in operation for a period not to exceed 7 days, provided the HPCI system l

1s operable. (Note that the pres-sure relief function of these valves L

is assured by Specification 3.6.H.;

Specification 3.5.F. only applies to the' ADS function).

.3.

Shutdown Reouirements If Specification 3.5.F.1. or 3.5.F.2.

l cannot be met, an orderly shutdown will be initiated and the reactor pressure shall be reduced to 113 psig or less within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

t r

-r l

• The ADS valves are not required to be operable for performance of inservice hydrostatic or leak testing with reactor pressure greater than 113 psig and all control rods inserted.

HATCH - UNIT't 3.5-9 Proposed TS/0330q/305-137

7 x

i; LIMillNG CONDill0NS FOR OPERA 110N SURVllLLANCE RIOUlREMENIS

'3.5.6.

Mindmum Core and ConitiD9tD1 4.5.G.

Surveillance of Core and Contain-ho'ine Systems Availability ment Coolino Systems During any period when one of (Deleted) the standby diesel generators is inoperable, continued reactor operation is limited to 7 l

t.

days unless operability of the t

l.

within this period. During such diesel generator is restored 7 days all of the components l

in the RHR system LPCI mode and containment cooling mode mode shall be operable. If this requirement cannot be met, an orderly shutdown shall be initi-ated and the reactor shall be in 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 />. Specification 3.9. Provides further guidance on electrical system availability.

Any combination of inoperable components in the core and con-tainment cooling systems shall not defeat the capability of the remaining operable components to fulf111 the core and contain-ment cooling f unctions.

I l

When irradiated fuel is in the reactor vessel and the reactor i

is in the Cold Shutdown Condi-tion, both CS systems and the l

LPCI and containment cooling subsystems of the RHR system may be inoperable provljed that the shutdown cooling subsystem of the RHR system is operable in accordance with Specification 3.5.B.l.b and that no work is being done which has the potential for draining the i

l-reactor vessel.

HATCH - UNIT 1 3.5-10 Proposed TS/0330g/305-147

_y

.4 LIMITING CONDITIONS FOR OPERA 110N SURVEILLANCE REOUIREMENIS 3.5.H.= Beintenance of Filled Discharce 4.5.H.

Maintenance of Filled Discharoe Ultti -

Unti Whenever the CS system LPCI, The following surveillance re-HPCI, or RCIC are required quirements shall be performed to be operable, the discharge to assure that the discharge piping f rom the pump discharge piping of the CS system, LPCI, l

of these systems to the last HPCI and RCIC are block valve shall be filled.

filled when required:

The suction of the HPCI pumps shall be aligned to the conden-1.

Every month, the discharge sate storage tank.

piping of the LPCI and CS systems shall be vented from the high point and water flow observed.

2.

Following any period where the LPCI or CS systems l

have not been required to be operable, or have been inoperable, the discharge

_l piping of the system or sys-tems being returned to ser-vice shall be vented from the high point prior to re-turn of the system to service.

3.

Whenever the HPCI or RCIC system is lined up to take suction from the condensate storage tank, the discharge piping of the HPCI and RCIC shall be vented from the high point of the system.

and water flow observed on a monthly basis.

(.

4.

The level switches which I

moni;or the discharge lines shall be functionally tested every month and calibrated-every 3 months.

l_

(

l.

Minimum River Level 1.

Binimum River Level-I 1.

If the water level, as The water level as, measured D

measured in the pump well, in the pump well, and the is less than 61.2 ft MSL, level in the river

• shall the discharge f rom each plant be verified with the follow-l service water (PSW) pump will ing fruquencies:

be throttled such that each pump does not exceed 7000 gpm.

Level (MS1).

Frecuency 2.

If the water. level, as measured l 1.

> 61.7 ft Biweekly.

In the pump well, decreases to less than 60.7 ft MSL, or if 2.

5 61.7 ft Every 12 hrs.

-the level in the river

• drops to a level equivalent to less i

1

• 0nly pump well monitoring is required if a temporary weir is not in place.

HATCH - UNIT 1 3.5-11 Proposed TS/0330g/305-154

r b

b LIMITING COND1110NS FOR OPERA 110N SURVLILLANCE RE0VIREMEN15

~

than 60.7 f t in the pump well of the intake structure, an 4'

~ orderly shutdown of the reactor shall be initiated, and the reactor shall be in 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 /> until the level in the river is greater than or equal to 60.7 ft MSL equivalent in the pump well.

3.5.J.

Plant Service Water System 4.$.J.

Plant Service Water System 1.

Normal Availability 1.

The automatic pump start functions and automatic The reactor shall not be isolation functions shall made critical from the be tested once per operating Cold Shutdown Condition

cycle, unless the PSW System (including four PSW i

pumps and the standby service water pump) is operable.

l 2.

Inoperable Components 2.

Inocerable Components a.

The standby service water a.

With the standby service pump may be inoperable for water subsystem inoperable i

a period not to exceed 60 for up to 60 days, provide l

days provided that an alter-Unit I service water cooling' 1

nate Unit 1 PSW water cool-l to the IB diesel generator ing source to the 18 diesel by verifying OPERABILITY of an generator is OPERABLE.

alternate Unit 1 service water cooling source within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Otherwise, declare the 18 diesel generator inoperable and take the action required by Specification 3.9.B.2.

b.

One PSW pump may be in-b.

(Deleted) operable for a period not j

to-exceed 30 days provided j

all other PSW pumps and the standby service water pump are operable.

c.

One PSW pump and the stand-c.

(Deleted) by service water pump may be inoperable for a period not to exceed 30 days pro-vided all other PSW pumps are operable.

d.

Two PSW pumps or one PSW.

d.

(Deleted) division may be inoperable for a period not to exceed 7 days provided all other PSW pumps and the standby service water pump are operable.

HATCH - UNIT 1 3.5-12 Proposed 15/0330q/305-147

9 1

9:

i LIMITING CDND111DNS FOR OPERAT!DN SURVi!LLANCE RE0VIREMEN15 f

1 3.5.J.

Plant Service Water System 4.5.J.

Plant Service hhter System 2.- Inonerable Components (Cont'd) 2.

Inocerable Components (Cont'd) e.

Two PSW pumps or one PSW e.

When cooling water to division, and the standby diesel generator IB is service water pump may be intertied with the PSW i

inoperable for a period divisional piping supply.

'not to exceed 7 days operability of the div-provided all other PSW.

1sional interlock valves pumps are operable, shall be demonstrated.

For each condition above in which the standby service water pump is inoperable, cooling water to diesel generator 1B shall be intertied with the PSW divisional piping supply.

3.

Shutdown Recuirements If the requirements of Specifi-cations 3.5.J.1. and 3.5.J.2.

cannot be met the reactor shall be placed in 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 />.

3.5.K.

Eauinevnt Area Coolers 4.5 K.

[AuiDeent Area Coolers 1.

The equipment area coolers 1.

Each equipment area cooler serving the Reactor Core Iso-is operated in conjunction lation Cooling (RCIC), High with the equipment served by Pressure Coolant injection that particular cooler; (HPCI), Core Spray or Residual therefore, the equipment area Heat Removal (RHR) pumps must coolers are tested at the-be operable at all times when same frequency as the pumpt the pump or pumps served by which they serve.

that specific cooler is con-sidered to be operable.

2.

Wnen an equipment area cooler is not operable, the pump (s) served by that cooler must be considered inoperable for Technical Specification purposes.

HATCH - UNIT 1 3.5-13 Proposed TS/03304/305-147 y

J BASES FOR LIM 111NG COND1110NS FOR OPERA 110N AND SURVEILLANCE R[0UIREMIN15 13.5.

CORE AND CONTAINMENT COOLING SYSTEMS A.

Core Sorav (CS) System 1.

Normal System Availability Analyses presented in Reference 1 demonstrated that the CS systeni provides adequate cooling to the core to dissipate the energy associated with the loss-of-coolant accident and to limit fuel clad temperature to below 2200'F which assures that core geometry remains intact and to limit any clad metal-water reaction to less than one percent. CS distribution has been shown in tests of systems similar l

[

in design to HNP-1 to exceed the minimum requirements. In addition,

~

cooling effectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiat-d fuel.

The intent of the CS system specifications is to prevent operation above atmospheric pressure without all associated equipment being operable. However, during operation, certain components may be out of service for the specified allowable repair times. The allowable repair times have been selected using engineering judgment based on i

experiences and supported by availability analysis. Assurance of the availability of the remaining systems is increased by demonstra-ting operability immediately and by requiring selected testing during the outage period.

When the reactor vessel pressure is atmospheric, the limiting conditions for operation are less restrictive. At atmospheric pressure, the minimum requirement is for one supply of makeup water to the core._ Requiring two operable RHR pumps and one CS pump provides redundancy to ensure makeup water availability.

2.

Doeration with inoperable Components Should one CS loop become inoperable, the remaining CS loop and the RHR system are required to be operable to ensure their availability should the need for core cooling arise. The-surveillance testing required by Specification 4.5.A. 4.5.H. and 4.6 K ensures the availability of the remaining CS loop. The surveillance testing required by Specifications 4.5.B.

4.5.H. and 4.6.K ensures the availability of the RHR system.

These provide extensive margin over the operable equipment needed for adequate core cooling. With due regard for this margin, the allowable repair time of 7 days was chosen.

B.

Residual Heat Removal (RHR) System (LPCI and Containment Coolina Mode) 1.

Normal System Availability I

The RHR system LPCI mode is designed to provide emergency cooling to the core by flooding in the event of a loss-of-coolant accident.

I f-This system-is completely independent of the CS system; however, it does function in combination with the CS system to prevent exce551Ve fuel clad temperature. The LPCI mode of the RHR system and the CS system provide adequate cooling for break areas of approximately 0.2 square feet up to and including the double-ended recirculation line break without assistance f rom the high-pressure emergency core cooling systems.

1 HATCH - UNIT 1 3.5-14 Proposed TS/0330q/305-150

L 6.i -

t BASIS FOR LIMITING CDNDITIDNS FOR OPERATION AND SURVEILLANCE REQUIREMENTS

3.5.8.1.

Nornal System Availability (Continued)=

Observation of the stated requirements for the containment cooling mode assures that the suppression pool and the drywell will be sufficiently cooled, following a loss-of-coolant accident, to prevent primary contain-ment over pressuritation. The containment cooling function of the RHR system is permitted only after the core has reflooded to the two-thirds core height level. This prevents inadvertently diverting water needed for core flooding to the less urgent task of containment cooling. - The two-thirds core height level interlock may be manually bypassed by a keylock switch.

-?

l:

The intent of the RHR system specifications is to prevent operation-above atmospheric pressure without all associated equipment being oper-

+

able. However, during operation, certain components may be out of service for the specified allowable repair times. The allowable repair times have been selected using engineering judgment based on experiences and supported by availability analysis. Assurance of the availability of the remaining systems is increased by demonstrating operability immediately and by requiring selected testing during the outage period.

When the reactor vesset pressure is atmospheric, the limiting conditions t

for operation are less restrictive. At atmospheric pressure, the minimum i

requirement is for one supply _of makeup water to the core.

2. -Operation with Inoperable Components

' With one LPCI pump inoperable or one LPCI subsystem inoperable, adequate core flooding is assured by the required operability of the' redundant LPCI pumps and LPCI subsystem and the CS system. The surveillance testing required by Specifications 4.5.8, 4.5.H. and 4.6 K ensures the availability of the redundant LPCI pump and LPCI subsystem. The surveillance testing required by Specifications 4.5.A. 4.5.H, and 4.6.K ensures the availability of the CS system. The reduced redundancy justifies the specified 7 day cut-of-service period.

4 i

i t

HATCH - UNIT 1 3.5-15 Proposed 15/03304/305-147

E 3

.c 2

l RA$[5 FOR LINITING CONI fl0N$ FOR OPLkATION AND SURVilLLANCE kt00lREMEWi$

3.$.D.I.' Goeration with Inocerable Censonents The HPCI system serves as a backup to the RCIC systeis as a source of feedwater makeup during primary system isolation conditions. The AD$ serves as a backup to the HPCI system for reactor depressuritation for postulated transients and accidents. The ADS must he operable if the HPCI system is determined to be inoperable. In addition, the survet11ance testing required by the specified Specifications ensures the availability of the following: C$ (4.6.A. 4.$.H.

and 4.6.K), LPCI (4.5.0, 4.$.H. and 4.6.K), RCIC (4.5.C. 4.$.H, and 4.6.K), and AD$ (4.5.F and 4.6.K).

Considering the redundant systems, an allowable repair time of 1 days was selected.

C.

Reactor Core 1 solation Coolina (RCIC) System 1.

Normal $vstem Availability The various conditions under which the RCIC system plays an essential role in providing makeup water to the reactor vessel have been identified by evaluating the various plant events over the full range of planned operations. The specifications ensure that the function for which the RCIC system was designed will be available when Seeded.

Because the low-pressure cooling systems (LPCI and C$) are capable of provid-l ing all the coolinD required for any plant event when nuclear system pres-sure is below 150 psig, the RCIC system is not required below this pres-sure. RCit system design flow (400 gpm) is sufficient to maintain water level above the top of the active fuel for a complete lors of feedwnter flou at the design Iower.

Tw t,ources Of iatn ere availt.blJ to the RCIC systee. ' $vetin is uit'selly taken f ree the cadetsate storage tank and is avtomatita,1y tiensfer sd to the buppress %n pool epor low LST level or high suporessian pool level.

,f.

WtttDftlyto,htparJble C9522"1DM l

Cons 19eration of the Dvallnbility of the RCIC system reveals that the avtrage rssk attociated with failure of the RCIC system to cool the core when reabired it not 4 creased if the RCIC system is inoperable fo" r.o longer then 7 t'ay.,

h provided that the HPCI system is operable durtt:d this period. The survell-lance testing required by Specifications 4.5.0, 4.$,H. and 4.6.K ensures the availability of the HPCI system.

F.

Automatic Depressurization system (ADS) 1.

Normal System Availability This specification ensures the operability of the ADS under all conditions for which the depressuritation of the nuclear system is an essential response to Unit abnormalities.

The nuclear system pressure relief system provides automatic nuclear system depressurization for small breaks in the nuclear system 50 that the LPCI and the C$ systems can operate to protect the fission product barrier.

l Note that this Specification applies only to the automatic feature of the pressure relief system.

I 1

HATCH - UNii 1 3.5-17 Proposed TS/03300/305-107

h i

ball 5 f0R LIM 111NG COND1110NS FOR OPERA 110N AND $URVillLANCE REQUIREMEN15 3.5.f.1.

Moriaal system Availability (continued)

Specification 3.6. states the requirements for the pressure relief function of the valves, it is possible for any number of the valves assigned to the AD$ to be incapable of performing their AD5 functions because of instrumentation failures yet be fully capable of perforwing their pressure relief function.

Because the automatic depressuritation system does not provide mekeup to the reactor primary vessel, no credit is taken for the steam cooling of the core caused by the system actuation to provide further conservatism to the Core Standby Cooling Systems.

b 1he ADS valve accumulators are stred such that, following loss of the pr.eumatic supply, at least two valve actuations will be possible with the drywell at 705 of its design pressore. This drywell pressure results f rom the largest break which could lead to the need for rapid depressuritation through the ADS valves. The allowable accumulator leakage criterion ensures the above capability for 30 minutes following loss of the pneumatic supply.

2.

Opnation with Inocerable _Cotspgnali With one f.D$ valve known to t,e incapabia of aatomatic operstion six v?1ves remain operable to perform their AD$ f unction. However, Since the LCCS Loss of Costant Acsident aralysis for small litie breaks assun.ed that all seven ADS valves were t,perable, reactor operit %n with ont ADS i

volve inoperable is only 611n nd to coittiwe for 7 Ccys provided that the HPCI sntem 1s (perable ard inat the (rometring) sin ACS vt.1vis are operable. In sedition, surve1Nrice testD.g tvt.Jired by the l

specified $rM ificatic<ns tr.sures the evallad itty of the io11cwin3:

HpC) (4.6.D. 4.5.H. srid 4.6.ir.) and ACS (t 5.F s nd 4.6.L).

l i

6.

Ejnimyo Core Ag(igptaimactf0glj.cg Systems Availability j

The f.urpose of this $be;.ification is to asture that ader,ut% ; ore l

cooling envifornt is available et all straes. If, for tsample, one C5 l

loop were out of service and the utesel which powered the opposite CS were out of service, only 2 RHR pumps hoL1d be availabIr.

Specification 3.g. must also be consulted to determine other i

j requirements for the diesel generators, This specification establishes conditions for the performance of major maintenance, such as draining of the suppression pool. The availability of the shutdown cooling subsystem of the RHR system and the RHR service water system ensure adequate supplies of reactor cooling and emergency makeup water when the reactor is in the Cold Shutdown Condition. In addition this specification provides that, should Irajor maintenance be l

performed, no work will be performed which could lead to draining the i

water f rom the reactor vessel.

HAICH - UNIT 1 3.5-1B proposed 15/0330g/30$-123 4

BA$t$ f 0R LIMillNG COND1110NS FOR OPlRA110N AND SURVEILLANCE RIOUIREMlNI$

3.$.H.

Maintenance of Filled Discharte Pipn if the discharge piping of the C$ LPCI, HPCI, and RCIC systems l

are not filled, a water hanumer can develop in this piping when the pump and/or pumps are started. To minimite damage to the discharge piping and to ensure added margin in the operation of these systems, this Technical Specification requires the discharge lines to be filled whenever the system is in an operable condition. If a discharge pipe is not filled, the pumps that supply that line must be assumed to be inoperable for

$pecification purposes.

The CS and LPCI distharge piping high point vents are visually checked for water flow once a month to ensure that the lines are filled.

Assurance that the HPCI and RCIC discharge piping remains filled is pro-vided by observing water flow from these systems high points monthly.

1.

Minimum River Flow A very low-flow river stage discharge relationship was developed at the Plant Hatch intake structure loc 6 tion. USG3 rating data w?rtr available for flows above 1740 cis 64 the BaalPY lauge f at U.S. Highway No. I bridge, on the plant site). This detr, which includes bathymetric surseys or the rating cross-section, were used to extenti the USGS rating curve t.y compu-tation. Since the l'5CS tlata use(/ in these comentaticas result in the highest flow for a given low flow stat,e c'rtr rJcordtd at the location, t

the cansputed rating curve s14.u13 giv1 a ronterrative low stag? for a givcm f lw. Vhe rive;* ratthe curve M the Plant. Mattit in'ake structure wet deseloprj by scibitccting 0.1 f t from st.o t!sE sjusge uvaluettun fer a gtwn dischJrge. 1!.e 0.1-f t adjustmen.'4 was e5 term.oed by Itvel surver shen the river level at th# USGS Quset was appres.<mately 62 f t M5L. At the Plent Hikh site, the rher level would he 61.3 f t 81$L for 1200 r.fs which is the low flow of record at Charlotte: and 60,9 f t MSL for tae hypothetitut minimam low flow of g50 cfs.

1he mlhimri 10w flow is irpnttans because of its effect on thc oDeratirn of P5W ard R;ll: service water pumps. 7f o RHR servire water pumps at ra(ed-flow conditions require for net positive suct1Jn head (NP5H) a river stage of only

$g.0 ft. Thus, no further considcration is required on river stage with regard to submergence of these pumps.

At the rated flow of 8500 gpm each for the P5W pumps, 4 ft of submergence will satisfy the NPSH and vortexing requirement. This corresponds to a stage in the pump well of 61.2 ft. Normal operation requires about 7840 gpm for each of three pumps. Shutdown or emergency conditions require only one pump with a discharge flow of 4428 gpm. This corresponds to a pump well level of 59.9 ft for safe shutdown. For a 0.1-f t-head loss through the trash rack and traveling screen, the corresponding river level would be 60.0 f t MSL, which corresponds to a flow of 660 cf s.

Similarly, l

l HAICH - UNii 1 3.5-19 Proposed 15/0330g/305-154

I I

i t

Basts IDR tlN111NG COND1110NS FOR OPERA 110N AND $URYllllANCE Rl0VIRintN1$

3.6.J/4.6.J flant Service Water Svitte the Plant service Water (PSW) system consists of two subsystems (divisions) of two pumps each and a separate standby service water pump system for diesel generator 18. During norinal f ull power operation the two subsystems f unction as a 3 out of 4 pune cross connected system supplying cooling water to the turbine and reactor building cooling systems. In the event of an accident signal, nonsaf ety-related cooling loads are isolated and the PSW pumps in the two subsystems supply cooling water to diesel generators 1A and 1C, the reactor building cooling syster., and the control room air conditioners, while the standby service water pump is available to automatically supply cooling water to diesel generator 1B should it be needed. Additionally, diesel 18 has a manual backup water supply available f rom the Unit 1 Division 1 or Division 2 P5W subsystems to that during maintenance on the standby diesel service water pump, either division of the PSW system can manually be aligned to supply cooling water to the 18 diesel. The tuo subsystems and the standby service water pump system are split in the accident mode for greater reliability with one pump in each of the two subsystems automatically starting while a Start signal f rom diesel penerator 1B initiates standby service water pump operation. Only one of the Division 1 PSW pumps and one of the Livision 2 PSW pumps are required f or cooling diesel generators I A and IC. respect 1=ely, while the standby service water p6mp provides adequete cooling water in sit tel generator IP.

In the event that the stau6Q te.'vice wcter pump is (noperable. *.he HNP-1 Liviston I-Divisiw 2 interta fitply piping tan be ellened to cool the Ib diesel. 1H thts condition, ore PsW pump is espatile of supplyiaC tho coolleg requirements idr the teactor tsu11 ding coolleg systsm, the contrc1 room air conJittorers, and the 1 A,18, and 1C diesel generatars, lhe P5W systeer can scpply all power ger.eration systenis at full load and l

the diesti generators with reJ9ndancy if one PSW pump and/or the standby l

service water pump att innperable. Hence, a 60 4ay 9uttge tima is i

l justified if the standby screlce water pura is inoperable since all f ohr l

PW pumps are evalle%e idivisicoal inten tie to 19 diesel rebutted). In addition, a 30-dal ou$. age is justified if one PSW pu"ap is it opersble, or 18 one PSW pumi and the standby service water pump are inoperable (divisional intertie to Ib diesel required). $hould two P5W pumps (3r one subsysteit) become inocerable, or should two PSW pumps (or one subsystve) and the st&ndby l

service water pump oecome inoperable (division intertie to IB diesel required) i plant operation will probably only continue at less than full power. However, safety-related loads are sill) adequately powered for these conditions.

Therefore, a 7-day outage time is justified for such events. The surveillance

(

testing required by Specifications 4.5.J and 4.6.K ensures availability of the redundant pumps and subsystem.

I i

K, bgjgrino 5tfety Features touloment Area Coolers lhe equipment area cooler in each pump compartment is capable of providing adequate ventilation flow and cooling. Engineering analyses indicate that the temperature rise in safeguard compartments without adequate ventilation flow or cooling is such that continued operation of the safeguard equipment or associated auxiliary equipment cannot be assured, i

The surveillance and testing of the eculpment area coolers in each of their various modes is accomplished during the testing of the equipment served by these coolers. The testing is adequate to assure the operabiltiv of the equipment area coolers.

L.

References 1.

• ldwin 1. Hatch Nuclear Plant Units 1 and 2 SAFER /GESIR-LOCA Loss-of-Coolant i

Accident Analysis,' NEDC-31316-P, December 1g86.

HATCH - UNil 1 3.5-21 Proposed 15/0330q/305-150

r-LIMlllN6. CONDITIONS FOR OPERAll.QN SURVEILLANCE REQUIREMENTS 3.6.J.

Recirculation System 4.6.J.

Recirculation System 1.

Core therwel power shall not exceed 1.

Recirculation pump speeds shall be 15 of rated thermal power without recorded at least once per day.

forced recirculation.

2.

With only one recirculation loop 2.

Whenever the reactor is in the in operation, verify that the START & HOT STANDBY or RUN reactor operating conditions are modes, at least one outside the Operation Not Allowed recirculation loop shall be in Region in Figure 3.6-5:

operation.

(a) At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 3.

The requirements applicable to single-loop operation as identified (b) Whenever thermal power has in Sections 1.1. A. 2.1. A. 3.1. A, been changed by at least 5% of 3.2.G. 3.11. A. and 3;11.C shall be rated thermal power and steady-in effect within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following l state conditions have been the removal of one recirculation reached.

loop from service, or the unit shall be placed in the Hot Shutdown Cenditien within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUT 00WN within the following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.

l(Ith only one recircelatien Icop in operation and the unit in the Oper6 tion Not Allowed Region, specified in Figure 3.6-5, initiate action within 15 minutes to place the unit in the Operation Allowed l

Region, identified in Figure 3.6-5, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Otherwise, place the reactor in the Hot Shutdown Condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

5.

Following one pump operation the q

discharge valve of the low speed i

pump may not be opened unless the speed of the faster pump is less than 50% of its rated speed.

HATCH - UNIT 1 3.6-9c Proposed TS/0320g/305-141

]

h e

ADMlWISTRATIVE CONTROLS 6.3 UNIT STAFF OUALIFICATIONj 6.3.1 Each member of the unit staf f shall neet or exceed the minimum qualifications of AN$1 N18.1-1971 for comparable positions, except for the Health Physics Superintendent who shall meet or exceed the cualifications of Regulatory Guide 1.8, September 1975, and the Shif t lechnical Advisor who shall have a bachelor's degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING A retraining and replacement training program for the unit staff 6.4.1 shall be maintained ender the direction of'the Manager of Training and shall neet or exceed the requirements and recommendations of Section 5.5 of ANSI N18.1-1911 and Appendix 'A' of 10 CFR Part 55.

6,4.2 The Fire Protection Program, except trainteg, is maintained under the dirtttlen el the Manager-Engin';ering Support. The fire Protection J

Program meets c.r etcceds the guidelines of NFPA Code 27, It;5.

Fire Protection Training is pointcined unt.tr the direction of the Training and Emergency Preparedness M4n49er. Fire Prtte(tion Training reets or l

exceeds the guidelines of NFPA tode 27.1g75, except retraining f recuency.

Fire Brigade and Fire Emergency support Group (TB/FISG) members are recluired j

to attend retraining once per calendar guarter.

6.5 REVJEW.AND AUOff 6.5.1 PLANT R[ VIEW BOARD fPggi l

FUNCil0N 6.5.1.1 The PRB shall f unction to advise the Plant Manager on all t

matters related to nuclear safety.

{0MP051110N The PRB shall be composed of, as a minimum, a supervisor or 6.5.1.2 higher level individual f rom each of the departments listed below Operatiots Maintenance Quality Control (OC) i Health Physics Nuclear Safety and Compliance Engineering Support The Chairman, his alternate, and other members of the PRB shall be designated by the *1 ant Manager. The Chairman and his designated alternate shall both be managers of one of the six above listed departments or a higher level onsite manager.

ALTERNATCS All alternate members shall be appointed in writing by the PRB 6.5.1.3 Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in PRB activities at any one time.

HATCH - UNIT 1 6-6 Proposed 15/0322q/305-145

( w i

5.3.2 Audit Responsibility 5.3.2.1 The General Manager-Quality Assurance is responsible for an audit, i

conducted annually, of the activities of the Plant Manager and the Manager-Environmental Affairs, related to compliance with ETS.

'5.3.2.2 Audits of facility activities shall be performed annually under the cognizance of the SRB to ensure conformance of facility operation to provisions of the ETS.

5.4 State and Federal Permit and Certificates s

Section 401 of PL 92-500, the Federal Water Pollution Control Act Amendments of 1972 (FWPCA), requires any applicant for a L

Federal license or permit to conduct any activity that may P

result in any dl5 charge into provisions of Sections 301, 302, 106, and 307 of the FWPCA.

S2rtion 401 of PL 92-500 further requires that any certification provided under this section shh11. set any of fioant limitationt 2nd other limitations and monitortng requirements necessary te assure that any applicant far a Federal license or permit vill comply with the anplicable 11mitat19ms. Certifications provided in oce-ardance with Section 401 ret forth conditions on the Federal license or permit for which the certivicttien is rrovided. Accordingly, the licensee stc11 comply with tne reaairemants set forth in the currently

' applicable 401 certification and amenaments thereto issued to tha licensee by the Georgia Environmental Protection Division.

In acccrdance with the provisions of the Georgia Water Quality Control Act, the FWPCA and the rules and regulations promulgated l

pursuant to each of these acts, the Georgia Environmental Protection Division, under authority delegated by the U.S. EPA, issued NPDES permit No. GA 0004120 to the licensee. The NPDES permit authorizes the-licensee to discharge from HNP Units 1 and 2 to the Altamaha River in accordance with effluent limitations, monitoring requirements, and other conditions stipulated in the permit.

. Subsequent revisions to the certifications will be accommodated in accordance with the provisions of section 5.6.3.

l 5.5 Procedures Detailed written procedures, including applicable checklists and instructions, shall be prepared and followed for all activities involved in implementing the ETS. All procedures shall be maintained in a manner convenient for review and inspection.

l Procedures that are the responsibility of the Plant Manager shall be kept at the plant.

Procedures that are the responsibility L

of the Manager-Environmental Affairs shall be kept at the Georgia Power Company General Office.

l l

HATCH - UNIT 1 5-3 Proposed TS/0323q/305-145 l

ll

w 3

o o

l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM 1

LIMITING CONDITION FOR OPERATION j

3.5.1 The High Pressure Coolant Injection ( rCI) system shall be OPERABLE with:

a.

One OPERABLE HPCI pump, and b.

An OPERABLE flow path capable of taking suction from the

[

suppression chamber and transferring the water to the reactor

{

pressure vessel.

APPLICABILITY: CONDITIONS 1*, 2* and 3* with reactor vessel steam dome pressure > 150 psig.

MTION:

a.

With the HPCI system inoperable, POWER OPERATION may con.tinue and the provisions of 3.0.4 de not apply *. previded the RCIC system, ADS, CSS, and LPCI system are OPERABLE; restore the inoperable HPCI system to OPERABLE status within 14 days or be in at least HOT SHilTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor sterm dome pressure. to t. 150 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b.

With the surveillance requirements of Specification 4.5.1 not performed at the required frequencies due to low reactor steam pressure, the provisions of Specification 4.0.4 are not applic-able provided the appropriate surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate (i.e.,

reactor pressure is such that the required steam pressure is maintained at the turbine for the duration of the test) to perform the tests.

SURVEILLANCEREQUIREMENTS l

4.5.1 The HPCI shall be demonstrated OPERABLE:

l a.

At least once per 31 days by:

1.

Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water, and

• See Special Test Exception 3.10.5 HATCH - UNIT 2 3/4 5-1 Proposed TS/0324q/305-86

T CONTAINMENT SYSTEMS PRIMARY CONTAINMENT PURGE SYSTEM LIMITING CONDITION FOR OPERATION 3.6.6.5.1 The drywell and suppression chamber 18-inch purge supply and exhaust isolation valves shall be OPERABLE with:

a.

Each valve closed except for purge system operation for inerting, deinerting, and pressure control, b.

A leakage rate such that the provisions of Specification 3.6.1.2-are met.

APPLICABILITY: OTERATIONAL CONDITIONS 1, 2, and 3.

ACTION:

a.

With an 18-inen drywell and suppression cht.mber purge supply and/or exhaust isolation valve (s) inoperable or open for other than jrerting, deinerting or p* essure control, close the open 18-inch valve (s) or otherwise isolate the penetrations (s) within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.6.6.5.1 The primary containment purge system shall be demonstrated OPERABLE:

a.

In addition to the requirements of Specification 3.6.3, at least once per 31 days, when not PURGING and VENTING, by verifying that each 18-inch drywell and suppression chamber isolation valve is l

closed.

b.

At least once per 18 months by replacing the valve seat of each j

18-inch drywell and suppression chamber purge supply and exhaust isolation valve having a resilient material seat and verifying that the leakage rate is within its limit.

HATCH - UNIT 2 3/4 6-46 Proposed TS/0327q/305-58

w PLANT SYSTEM 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 The Reactor Core Isolation Cooling (RCIC) System shall be OPERABLE with an OPERABLE flow path capable of (AUTOMATICALLY) taking suction from the suppression pool and transferring the water to the reactor pressure vessel.

APPLICABILITY:

CONDITIONS 1, 2, and 3 with reactor steam dome pressure

> 150 psig.

ACTION:

a.

With the 1CIC system inoperable, operation may centinue and the provisions of Soecification 3.f).4 are net cpplict.ble provided the HPCI 7ystem is ODERABLF; re', tere the RCIC e stem to OPERABLE status vithin 14 days or be in at inst HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor Strem dome prersure to < 150 psig withiri the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

With the surveillance requirements of Specift:ation 4.7.3 not performed at the required intervals due to low reactor steam pressure, the provisions of Specification 4.0.4 are not applicable provided the appropriate surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate (i.e., reactor pressure is such that the required steam pressure is maintained at the turbine for the duration of the test) to perform the tests.

SURVEILLANCE REQUIREMtiNTS 4.7.3 The RCIC system shall be demonstrated OPERABLE:

a.

At least once per 31 days by:

1.

Verifying that the system piping from the pump discharge valve to the system isolation valve is filled with water, and l

2.

Verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

b.

At least once per 92 days by verifying that the RCIC pump develops a flow of 400 gpm on recirculation flow when steam is being supplied to the turbine at normal reactor vessel operating pressure, 2000 + 20,

- 80 psig.

HATCH - UNIT 2 3/4 7-9 Proposed TS/0328q/30S-38

I o

ELECTRICAL p0WER SYSTEMS A.C. CIRCUITS INSIDE PRIMARY CONTAINMENT LIMITING CONDITIONS FOR OPERATION I

i 3.8.2.5 The following A.C. circuits inside primary containment shall be l

l' de-energized *:

a.

Breaker Numbers 2, 4, 6, 8, 10, 12, 14, 40 and 42 in panel 2T51-5003, b.

Dret,ker Numbers 2, 4, 6, 8,10,12, 40 and 42 in panel 2T51-l f

S004, e.

Bre ner Numbers 28 and 34 in panel 2R25-5105, and d.

' ompartment ILL on NCC 2R24-5014.

l APPLICABILITY:

CONDITIONS 1, 2 and 3.

ACTION:

With any of the above required circuits energized, trip the associated circuit breaker (s) in the specified panel within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

SURVEILLANCE REQUIREMENTS 4.8.2.5 Each of the above required A.C. circuits shall be determined to -

be de-energized at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that the associated circuit breakers in the specified panels are in the tripped condition.

1

• Except during entry into the drywell.

HATCH - UNIT 2 3/4 8-17 Proposed TS/0329q/305-0 i-

l l

o..

I l

TABLE 3.8.2.6-1 (Continued)

PRIMARY CONTAINMENT PJNETRATION CONDUCTOR I.

OVERCURRENT PROTliCTIVE DEVICES i

DEVICE NUMBER

$YSTEN/ COMPONENT AND LOCATION

• POWERED c.

Type 3:

1.

600 VAC, MCB, T.M.

RECIRC. PUMP MOTOR HEATER 2R24-$014, COMPT. SE 2831-C0018 l

2.

600 VAC, MCB, T.M.

REACTOR RECIRC. PUMP MOTOR 2R24-$013, COMPT. SB HEATER 2831-C001A MCB, T.M.

DRYWELL COOLING UNIT 600 VAC,3, COMPT. 3B 3.

2R24-$01 2T47-8010A 4.

600 VAC, MCB, T.M.

DRYWELL COOLING UNIT 2R24-$014, COMPT. 8A 2T47-80108 d.

Type 4:

1.

120 VAC, MCB, T.M.

CABLE BHXT8C05 2R25-$102 CKT 10 2.

120 VAC, MCB, T.M.

CABLE dGX708C05 2R25-$101, CKT, 10 e.

Type 5:

1.

600 VAC, MCB, M.0.

ORYWELL EQUIP. DR. $ UMP 2R24-$014 COMPT. 2A 015CH. MOV 2Gil-F018 2.

600 VAC, MCB, M.O.

ORYWELL EQUIP DRAIN $ UMP 2R24-$014, COMPT. 6C RECIRC. MOV 2G11-F015

^

3.

600 VAC, MCB, M.0.

RCIC STEAMLINE INBOARD 2R24-$0128, COMPT. 4A 150. MOV. 2E51-F007 4.

600 VAC, MCB, M.0.

RHR HEAD $ PRAY !$0LAT10N 2R24-$0ll, COMPT, 9A MOV. 2E11-F022 5.

600 VAC, MCB, M.0.

HPCI STEAM LINE INBOARO 2R24-$011A, COMPT. 4A ISOLATION MOV. 2E41-F002 6.

600 VAC, MCB, M.0.

RWCU INBOARD !$0LAT10N 2R24-$011 COMPT. 14C MOV. 2G31-F001 7.

600 VAC, MCB, M 0, MAIN STEAM LINE DRAIN 2R24-$011 COMPT, ISB MOV. 2821-F016 SM.C.B. - molded case circuit breaker M.0. - magnette only T.M. - thermal magnetic HATCH - UNIT 2 3/4 8-21 Proposed TS/0329q/305-64

ADMINISTRATIVE CONTROLS 6,3 UNIT STAFF OVALIFICATION$

Each member of the unit staff shall meet or exceed the minimum 6.3.1 qualifications of AN$1 hlb.1-1971 f or comparable positions, except for the Health Physics Superintendent who shall meet or exceed the qualifications of Regulatory Guide 1.8 September 1975, and the $htf t Technical Advisor who shall have a bachelor's degree or equivalent in a scienttite or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents.

6.4 TRAINING A retraining and replacement training program for the unit staff 6.4.1 shall be maintained under the direction of the Manager of Training and shall meet or exceed the requirements and recommendations of section 5.5 of AN$1 N18.1-1971 and Appendtx A of 10 CFR part 55.

6.4.2 The Fire Protection Program, exr.ept t~e_ining, il maintained under the direction of the Mansp r Lnginearing Support. The Fire Protection Program meets or excee:Is the guidelines n.f NFPA Code 27,197F.

Fire Prote.tton Training is N intair.ed under the direction of the Training and Emergency Preparedness Mansger. Fire Protection Train.ng meets or 27, 1975, exceps re; raining frequency.

exceeds the guidelinas of NFPA Code Fire Brigade and Fire Emergency Support Group (FB/FEiG) members are required to attend retraining once per calendar quarter.

6.5 REVIEW AND AUD11 6.5.1 PLANT REVIEW BOARD (PRB)

FUNCTION 6.5.1.1 The PRB shall function to advise the Plant Manager on all matters related to nuclear safety.

COMPOSI1 ION The PRB shall be composed of, as a minimum, a supervisor or 6.5.1.2 higher level individual from each of the departments Itsted below:

Operations Maintenance Quality Control (QC)

Health Physics Nuclear Safety and Compliance Engineering Support The Chairman, his alternate, and other members of the PRB shall be designated by the Plant Manager. The Chairman and his designated alternate shall both be managers of one of the six above listed departments or a higher level onsi.e. manager.

t ALTERNATE $

6.5.1.3 All alternate members shall be appointed in writing by the PRB Chairman to serve on a temporary basts; however, no more than two alternates shall participate as voting members in PRB activities at any one time.

HATCH - UNIT 2 6-5 Proposed T$/0326q/305-94

e s o

c;,

5.3.2 Audit Responsibility 5.3.2.1 The General Manager-Quality Assurance is responsible for an audit, conducted annually, of the activities of the Plant Manager and j

the Manager-Environmental Affairs, related to compliance with ETS, 5.3.2.2 Audits of facility activities shall be performed

{

annually under the cognizance of the SRB to ensure conformance of facility operation.to provisions of the ETS.

5.4 State and Federal Permit and Certificates

.T Section 401 of PL 92-500, the Federal Water Pollution Control Act Amendments of 1972 (FWPCA), requires any applicant for a Federal license or permit to conduct any activity that may result in any dischargt into provisions of Sections 301, 302, t

305, and 307 of the FVPCA.

Section 401 of PL 92-500 further ree.uires that any certification provided under this section shall set any effluent limitations and other limitations and monitoring requirements necessary to assure that any applicant for a Federal license or permit will comply with the applicable

-limitations. Certifications provided in accordance with Section 401 set forth conditions on the Federal license or permit for which the certification is provided. Accordingly, the licensee

-shall comply with the requirements set forth in the currently applicable 401 certification and amendmonts thereto issued to the licensee by the Georgia Environmental Protection Division.

l In accordance with the provisions of the Georgia Water Quality Control Act, the FWPCA and the rules and regulations promulgated pursuant.to each of these acts, the Georgia Environmental Protection Division, under authority delegated by the U.S. EPA, issued NPDES permit No GA 0004120 to the licensee.

The NPDES permit authorizes the licensee to discharge from HNP Units 1 and 2 to the Altamaha River in accordance with effluent limitations, monitoring requirements, and other conditions stipulated in the permit.

1 Subsequent revisions to the certifications will be accommodated in accordance with the provisions of section 5.6.3.

l 5.5 Procedures L

Detailed written procedures, including applicable checklists and instructions, shall be prepared and followed for all activities involved in implementing the ETS. All procedures shall be l

maintained in.a manner convenient for review and inspection.

Procedures that are the responsibility of the Plant Manager shall be kept at the plant.

Procedures that are the responsibility of the Manager-Environmental Affairs shall be kept at the Georgia Power Company General Office.

l L

HATCH-UNIT 2 5-3 Proposed TS/0325q/305-80 L