Text

Proposed Tech Specs for Unit 1 & 2
	ML20154C545
Person / Time
Site:	Vogtle
Issue date:	09/02/1988
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML20154C537	List: ML20154C535; ML20154C545;
References
	NUDOCS 8809140403
	Download: ML20154C545 (184)
	v • d • e

.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...............................................

3/4 0-1 3/4.1 REACTIVITYdONTROLSYSTEMS 3/4.1.1 80 RATION CONTROL.

Shutdown Margin - M00ES 1 and 2..........................

3/4 1-1 Shutdown Margin - MODES 3, 4 and 5.......................

3/4 1-3 FIGURE 3.1-1 RE IRED SHUTDOWN MARGIN FOR MODES 3 AND 4 (WO WI A LgjPg 30R COOLANT, PUMP RUNNING). QUIT 1 /41-3a 3

O '

$E TH r

.1-R MAR NO N i-3/4 1-3p C.

WE 6. l'M S

.IOWb*[MoWfdifff.'

3/ 4 l. 3 ej Modera r emperatu cient........................

3/4 1-4 Mi nimus Temperature for Cri tical ity......................

3/4 1-6 3/4.1.2 BORATION SYSTEMS Flow Path - Shutdown.....................................

3/4 1-7 Fl ew P ath s - Op e rati ng...................................

3/4 1-8 Charging Pump - Shutdown.................................

3/41-9 Charging Pumps - Operating...............................

3/4 1-10 Borated Wate r Source - Shutdown..........................

3/4 1-11 Borated Water Sources - Operating........................

3/4 1-12 3/4.1.3 MOVABLE CONTROL ASSENBLIES G ro up He i g h t.............................................

3/4 1-14 TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE i

EVENT OF AN IN0PERA8LE CONTROL OR SHUTDOWN R00...........

3/4 1-16 Position Indication Systans - Operating,.................

3/4 1-17 l

l Position Indication Systas - Shutdown....................

3/4 1-18 R od D ro p T i me............................................

3/4 1-19 l

Shutdown Rod Insertion Limit.............................

3/4 1-20 control Rod Insertion Limits.............................

3/4 1-21 j

i FIGURE 3.1-3 R00 BANK INSERTION LIMITS VERSUS THERMAL POWER,......

3/4 1-22 8809140403 GC:0902 PDR ADOCK 0*_>000424 p

FDC V0GTLE UNITS-1 & 2 IV i

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIRENENTS SECTION PACE TA8LE 3.3-5 SEISMIC MONITORING INSTRUMENTATION....................

3/4 3-51 TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS.............................................

3/4 3-52 Meteorological Instrumentation...........................

3/4 3-53 TABLE 3,.3-6 METEOROLOGICAL MONITORING INSTRUMENTATION (CCMON SYSTEM).......................................

3/4 3-54 Remote Shutdown Systes...................................

3/4 3-55 TA8tE 3.3-7 REMOTE SHUTDOWN SYSTEM MONITORING INSTRUMENTATION.....

3/4 3-56 Accident Moni toring Instrumentation......................

3/4 3-58 TAILE 3.3-8 At DENT MONITORING INSTRUMENTATION...................

3/4 3-59 C

ine taction Systaas...............................

3/4 3-63 oosh Detection Systes (Deleted)..................

3/4 3-64 to e Liquid Effluent Monitoring Instrumentation...

3/4 3-65 TABLE 3.3-9 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION 3/4 3-66 TABLE 4.3-5 RADI0 ACTIVE LIQUID EFFLUENT MONITORING Ii4STRUMENTATION SURVEILLANCE REQUIRENENTS................

3/4 3-68 Radioactive Gaseous Effluent Monitoring Instrumentation..

3/4 3-71 TABLE 3.3-10 RADI0 ACTIVE GASEQUS EFFLUENT MONITORING INSTRUMENTATION..........................................

3/4 3 72 TABLE 4.3-6 RADI0 ACTIVE GASE0US EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS................

3/4 3-76 High-Energy Line Break Isolation Sensors.................

3/4 3-79 TABLE 3.3-11 HIGH-ENERGY LINE 8REAK INSTRUMENTATICN...............

3/4 3-80 i

3/4.3.4 TURBINE OVERSPEED PROTECTION..............................

3/4 3-81 4

(

l V0GTLE UNITS-1 & 2 VI

[

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION f.M.E.,

3/4.4.9 PRESSURE / TEMPERATURE LIMITS R e a c to r Co o l an t Sy s tem...................................

3/4 4-30 grin G -FIGURE 3.4-2.. RC'OTCP, 000L*F SYSTr." HEATOF LINITATIGH3 - C

,.,, r u ww u U F is in ir r r.................................

..,n, e

b0*

WIGURE3.4-3 REAciuk 6000mi 5hita svulD0wii LiniNuoMs -

?? P LI C? " L U? T0 16 EF F Y.................................

=3/4 4-3; q P re s s u r i z e r..............................................

3/4 4-33 Cold overpressure Protection Systems.....................

3/4 4-34 he, jmer.}-

e FI?JP.E 3.4 2 Mf2I.W".' LLC'.lABLE NGHINAL-PORY SETPolf F0" "iE 000 WEAPRt55 uni rxviisiiON 5YSTEM....%...................

(J/' ' 33,y 3/4.4.10 STRUCTURAL INTEGRITY.....................................

3/4 4-37 3/4.4.11 REACTOR COOLANT SYSTEM VENTS.............................

3/4 4-38 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACCUMULATORS.............................................

3/45-1 3/4.5.2 ECCS SUBSYSTEMS - T,yg GREATER THAN OR EQUAL TO 350*F....

3/45-3 3/4.5.3 ECCS SUBSYSTEMS - T LESS 1HAN 350*F avg ECCS Subsystaas..........................................

3/4 5-7 Safety Injection Pumps...................................

3/4 5-9 3/4.5.4 REFUELING WATER STO RAGE TANK.............................

3/4 5-10

(

e e

V0GTLE UNITS-1 & 2 VIII i

-~

INSERT AA FIGURE 3.4-2a UNIT 1 REACTOR COOLANT SYS1EM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY 3/4 4-31 i

FIGURE 3.4-2b UNIT 2 REACTOR _ COOLANT SYSTEM HEATUP v

LIMITATIONS - APPLICABLE UP TO 16 EFPY 3/4 4-31a l

FIGURE 3.4-3a UNIT 1 REACTOR COOLANT SYSTEM COOLDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY 3/4 4-32 FIGURE 3.3-3b UNIT 2 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY 3/4 4-32a e

I i

l' s

F I

i i

s

INSERT AB FIGURE 3,4-4a UNIT 1 MAXIMUM ALLOWA8LE NOMINAL PORY SETPOINT FOR THE COLD OVERPRESSURE PROTECTION SYSTEM 3/3 4-35 FIGURE 3.4-4b UNIT 2 MAXIMUM Al.LOWASLE NOMINAL PORY SETPOINT FOR THE COLO OVERPRESSURE PROTECTION SYSTEM 3/4 4-354 t

O e

4 t

I I

t i

l r

t I

I f

i I

i l

b i

I I

1N.,0E,1

.LIMITTNG CONDITIONS FOR__0PERATYON AND SURVE1LLANCE REQUIREMENTS SECTI0ft PAG 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES 0perating................................................

3/43-1 TA8LE 4.8-1 DIESEL GENERATOR TEST SCHEDULE........................

3/4 8-9 Shutdown........................'.........................

3/4 8-10 3/4.8.2

0. C3 SOURCES 0perating................................................

3/4 3-11 TA8LE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS.....................

3/4 8-13 Shutdown.................................................

3/4 8-14 3/4.8.3 ONSITE POWER DISTRIBUTION l

l 0perating................................................

3/4 8-15 Shutdown.................................................

3/4 6-18 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Containment Penetration Conductor Overcurrent Protective Devices and Feeder Breakers to Isolation Transferners 8etween 440 V Class if Susses and Non-Class 1E l

Equipment..............................................

3/4 6 19 L

Safety-Related Motor-operated Valves Therwal Overload l

P ro tecti on and Bypa s a Cev ices.........................,

3/4 8-21 f*

TA8t t 3.s-1 SAFETY-RELATEI) MOTOR-0 PEP} ED VALVES TH PROTECTION SYPAs5 M QQE4..............................

3/4 8 22,

l l

1 e

V0GTLE UNITS-1 & 2 XI l

.a.

INDEX 1

bases sECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS M0 COOLANT CIRCULATION.............

B 3/4 4-1 3/4.4.2 SAFETY VALVES.............................................

8 3/4 4-2 3/4.4.3 PRESSURIZER......'...'........~.............................

B 3/4 4-2

)

3/4.4.4 RELIEF VALVES............

B 3/4 4-3 l

3/4.4.5 iT EM G ENERATO RS......................................8 3/4 4-3 3/4.4.6 PEACTOR COO LANT SYGTEM LEA XdGE...........................

B 3/4 4-4 3/4.4.7 CHEMISTRY.................................................

B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITf.........................................

B 3/4 4-5

~

3/4.4.9 FRESSURE/TDPERATURE LIMITS...............................

B 3/4 4-7 hTMLIB,/4.?-1 "AC'0 " VE3 3 E L TOE i3 5..........................

s.v, C -

S - Q '

FIGURE 8 3/4.4-irfAST NEUTRON FLUENCE (D1MeV) AS A FUNCTION OF

^#%

FULL PCIER SERVICE LIFE..................................

8 3/4 4-10 RE 8 3/4.4-2 \\tFFECT OF FLUENCE M0 COPPEA CONTENT ON SHIFT OF RT FOR REACTOR VESSELS EXPOSED TO RM.:.ATION AT 550'F.

B 3/4 4-11 NOT 3/4.4.10 STRUCTURAL INTEGRITY.....................................

B 3/4 4-16 3/4.4.11 REACTOR C00 TANT SYSTEM VENTS.............................

B 3/4 4-16 3/4.5_ EERGENCY CD'tE C00 LING SYSTDes 3/4.5.1 ACCtMJLAT02$..............................................

8 3/4 6-1 3/4,5.2 and 1/4.5.3 ECCS SU85YST BS...............................

B 3/4 5-1 3/4.5.4 REFU(LING WATER STORAGE TANK............... '...............

B 3/4 5-2 V0GTLE UNITS-1 & 2 XVI h

1

e INDEX

$NINISTRAi!VECONTROLS SECTION PAGE 6.1 RESP 0NSIpILITY..............................................

6-1 t,. 2 ORGANIZATION................................................

bl 0FFSITE..Q.?6hMLEffT-KONS.........................

6-1 6.2.2 PLANT STAFF...............................................

6-1 FIGURE 6. 2-1 6GFf 3Iii G^GANIZATIOgf. (MW.W.9...............

6-3 FIGURE 6.2 2 n.. _.......-...-. v = - 4 u u a g.....( y..e.. p...

6-4 TA8LE 6.2-1 MINIMUM SHIFT CREW COMPOSITION SINGLI UNIT FACILITY.

6-5 6.2.3 IMOEPENDENT SAFETY ENGINEERING GROUP (ISEG)

Function..................................................

6-6 Compcaition...............................................

6-6 Responsibilities..........................................

6-6 Records...................................................

6-6 C.2.4 SHIFT TECHNICAL ADVIS0R...................................

6-6 6.3 TRAINING....................................................

6-6

6. 4 REVIEV AND AUDIT............................................

6-7 6.4.1 PLANT P.2 VIEW BOARD (PRS)

Function..................................................

6-7 Composition.,,............................................

6-7 A1tarr.ates................................................

6-7 Mee ti ng F reque ncy.........................................

6-7 Quorum....................................................

6-7 Responsibilities..........................................

6-7 Records...................................................

6-9 1

V0GTLE UNITS-1 & 2 XXII W

g 9

680 UNACCEPTA8LE OPERATION 660 2400 p.la N

N 640 A

,\\

w

[

2250 psia 2000 pela g

620 r,

x 3 22.,,i.

N

\\

600.,

i N

N W

580 f

ACCEPTABLE

~

OPERATION i

540 0

0.2 0.4 0.6 0.8 1.0 1.2 FRACTION OF RATED THERMAL POWER I

FIGURE 2.1-1 l

REACTOR CORE SAFETY LIMIT V0GTLE UNITS - 1 & 2 2-2

SAFETY LIMITS ANO LIMITING SAFETY SYSTEM SETTINGS

2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlock Setpoints shall be set consistent with the Trip Setpoint values.shown in Table 2.2-1.

l APP!ICABILITY: As shown for each channel in Table 3.3-1.

ACTION:

a.

With a Reactor Trip Systaa Instrumentation or Interlock Setpoint less conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 2.2-1, adjust the Setpoint consistent with the Trip Setpoint value.

b.

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

1.

Adjust the Setpoint consistant with the Trip Setpoint value of Table 2.2-1 and detemine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied ior the affected channel, or 2.

Declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to CPERA8LE status with its Setpoint adjusted consistant with the Trip Setpoint value.

Equation 2.2-1 I + R + 51 TA Where:

Z = The value from Column Z of Table 2.2-1 for the affected channel, R = The "as measured" value (in percent span) of rack errer for the affected channel, S = Either the "as measured" value (in percent span) of the sensor error, or the value from Column S (Sensor Error) of Tacle 2.2-1 for the affected channel, and TA = The value from Column TA (Total illowanci of Table 2.2-1 for the affected channel.

V0GTLE UNITS - 1 & 2 2-3

s'.

t k

i TABLE 2.2-1 (Conti.wsed)

REACTOR TRIP SYSTEM INSTRbHENTATION TRIP SETPOINTS 1

c>

-4M TOTAt.

SENSOR ALLOWAACE ERROR c$

FUNCTIONAL UNIT (TA)

~

(S)

TRIP SETPOINT ALLOWA8LE VALUE e

Z d

l S.

Pressurizer Pressure-Low 3.1 0.71 1.67

>1960 psig**

>1950 psig (PI-0455A,8fC PI-9456 &

~

PI-0456A, PI-0457 & PI-0457A, i

t e.

PI-0458 & PI-0458A) j u

10. Pressurizer F7 essure-High 3.1 0.71 1.67 12385 psig 12395 psig l'

I (PI-0455A,8&C, PI-0456 &

F1-0456A, PI-0457 s PI-0437A, PI-0458 & PI-0458A)

{

11. Pressurizer Water Level-High 8.0

2.18 1.67

<92% of instrument

<93.9% of instrument (LI-0459A,LI-3460A,LI-0461) spa span

12. Reactor Coolant Flow-Low 2.5 1.87 0.60

>90% of loop

>89E of loop i

(LOOP 1 LOOP 2 LOOP 3 iOOP4 Besign flow 3esign flow

i a

m FI-0414 FI-0424 FI-0434 FI-0444 J, -

FI-0415 FI-0425 FI-0435 FI-0445 FI-0416 FI-0426 FI-0436 FI-0446)

I

13. Steam Generator Water Level 18.5 17.18 1.67

>18.5%*** of narrow >17.8% of narrow j

Low-Low range instrument range instrument (8.00P1 LOOP 7 LOOP 3 LOOP 4 span span i

LI-0517 LI-0527 LI-0537 LI-0547 LI-0518 LI-0528 LI-0538 LI-0548 LI-0519 LI-0529 LI-6539 LI-0549 LI-0551 LI-0552 L1-0553 LI-0554)

14. Undervoltage - Reactor 6.0 0.58 0

>9600 volts

>9481 volts Coolant Pumps T70% bus voltage)

{69%busvoltage)

}

15. Underfrequency - Reactor 3.3 0.50 0

>57.3 Hz

>57.1 Hz Coolant Pumps i

Loop design flow = 95,700 gpa

- Time c ants utilized in the lead-lag controller for Pressurizer Pressure-Low are 10 seconds for lead and i

1 c

f lag. Channel Calibration shall ensure that these time constants are adjusted to these values.

- Unt sol tion of the Veritrak transmitter uncertainty issue this setpolat will be set at >22.5% (Unit 1) and 3.5% (

lt 2) of narrow range instrument span.

y

SECTIONS 3.0 AND 4.0 LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1

I a

i i

@ LE UNITS - 1 &

d i,a-4

^

K.

= - k.

- v :, 2.

....,s.,-

APPLICA8ILITY LIMITING CONDITION FOR OPERATION (Continued) 3.0.5 Unless specifically noted, all the information provided in the Limiting Condition for Operation including the associated ACTION requiraments shall apply to each unit individually.

In "

cases where a specification makes reference to sysi'ess or components lch a shared by both units, the affected systems or compor,Jnts will be cle 1 enti Nd in parentheses or footnotes declaring the reference to be "c Wh never the Limiting condition for Operation refers to systems or c t

ich are common, the ACTION nqeiraments will apply to both uni multaneously.

(This will be indicated in the ACTION section.) Whenever certain portions of a specification refer to systaas, components, operating parameters, setpoints, etc., which am different for each unit, this will be identified in parentheses or footnotes or in the APPLICA8ILITY section W A

as c.pmprih SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be set during the OPERATIONAL MODES or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Re.quirement.

l 4.0.2 Each Surveillance Requirement shall be performed within the specified time interval with:

a.

A maxieue allowable extension not to exceed 25% of the surveillance l

interval, but b.

The combined time interval for any three consecutive surveillance intervals shall not exceed 3.25 times the specified surveillance l

ir,terval.

1 4.0.3 Failum to perfore a Survalliance Requirement within the specified time interval shall cor.stitute a failun to meet the OPERA 8IL2TY requirements for a Limiting Condition for Operation.

Exceptions to these nquirements are stated in the individual specifications.

Surveillance Requirements do

~

not have to be performed on inoperable equipment.

4.0.4 Entry into an OPERATIONAL MODE or other specified condition shall not be i

sade unless the Surveillance Requirement (s) associated with the Limiting Condition for Operation has been performed within the stated surveillance interval or as otherwise specified.

V0GTLE UNITS - 1 & 2 3/4 0-2

ll l

Il i

O

)0 3

005

~

2

(

g l

00 1

42 T

I NU

)

GN I

NN 0

U 0

R 0

E 2

P L

C B

R A G E

T N N

PI O

ETN T

CA O S

C RI A

A EG 0

)

NPE 0 m E.

t UOR 6 p 1

p T

f A

(

N H

O T

I I

T W

A a

k R

l 4

T 1

E 1

N D

T 00 E 3

O I

N 2 C M

E

(

1 N

U R

O U

4 N

C G

I I

D G

0 N

F N

)

R 3

O A

E A

R LG M

_1 O

3 BN 0

B S

AE N

0 E

T TN W

9 0 S D

PA O O

0 C O

(

8 R M

ERI CEG D

CPE T

R AOR U

O F

H S

N D

I G

E R

R A

M IU 0

Q 0

W E

4 O

R D

T U

l iS D

ER I

U Q

0 E

0 0

R 0

0 0

n 0

0 0

S 4

3 e

1 0

gE4g iokg z5OQs u

E 2

2

7u*

<g#"' Enm. [ m

-a---

4 4

s

~N 4

y 8'

l E

\\

E

\\

~

\\

W o

J

' t) 5w 1,

<H 1

_h k

W

_T 3

m T

=

--a y5 f

5 e s,

,e 1

N

/

T e

2 a

2 f

(

2C e

b hM N

-j

- f g

~'

g smg e

s 8

k2 e

'3 w

2m I

k e no u

w T

_I.

"U

@H z

c zh i

.d

$g E!

S y

s,s!

i g

o o

=

V o

E {s g

O O

O w

g H

u e

E v

.=

E" 8

d I

Oi O

O r

~~~

g u

_ o F-S 8

8 8

8 s

e s

s a

s e

WW%) NIDWVW NMOG10HS 3

V0GTLE UNITS - 1 & 2 3/4 1-3b

)0 8

4 0052 (j

2 2

1 T

I N

U

[

00

)

E 0

G L

2 N

I B

N A N NU T O R

P I

ETN s

~

CA O P

C RI

)

C A EG m

R NPE 0

p UOR 0

p O

6

(

N 1

N l

i O

T I

I T

W A

4 RT a

2 E

N DO 1

E 1

M T

0 C

(

0 N 3

I N

2 O E

5 1

U C

RU E

E N

G D

LG N

O O

I BN R

F M

I AI G

O R

T TN R

B O

PA O A

F ERI S

CEG M

C N

CPE N

0 R I

0 G

AOR W

8 R

O A

M D

T N

U W

)0 O

H 0

D S

1 T

D U

E 0

I 0

S I

R 5

0 D

(

U 0

E O

4 R

I E

U H

Q ER 0

0 0

0 0_

0 0

5 A

3 2

1 0

22I6 z5K$az h 0>3I*

6 oM m c5 3 * " "

r-Nu h 2

i i'

8'l UNIT 2 E

N 3

cdu (2100,6A4) y 6.00

_.F ACCEPTA8 t g

OPE n A TING y

ntcson 2 5.00 I

y 3

f A

1 6

f 3 4.00 f

a 0

neouanto

=

MF k

SHU T DOWN uancsN p

E 2 3.00 g r 5

/

UNACCEPTABLE T

O J

ort nATING W

P nEcsom j

j m

2r

-d F

2 2F 1.00 (300, 1.0) i 0.00 j

0 0.2 0.4 0.6 0.8 1

2 14 1.6 1.8 2

(Thousarv.fs)

RCS BORON CONCENTRATION (ppm)

FIGURE 3.I-2b REQUIRED SHUTDOWN MARGIN FOR MODE 5 (MODE 4 WITil NO RCPs RUNNING) UNIT 2

1 REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

L ss siti th ' 0 k/* 'for heairod with awn / beg nin et, i 0504 cy e 11 e (

L), et z 1

RMAL EA ondi ion; hnd N-5.

ess nega va an 4.0 1

Ak/ /*F f the all ds ithd aw,

fcylefifep0L) and RAT THE L

EA c ndit, on.

APPLICA8ILITY:

Specification 3.1.1.3a. - MODES 1 and 2* only.'"

Specification 3.1.1.3b. - MODES 1, 2, and 3 only.**

ACTION:

a.

With the MTC acre positive than the limit of Specification 3.1.1.3a.

above, operation in MODES 1 and 2 may proceed provided:

1.

Control rod withdrawal limits are established ar.d maintained sinfficient to restore the MTC to less positive than 0 ak/k/*F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These withdrawal limits shall be in addition to the insertion limits of Specification 3.1.3.6; 2.

The control rods are aa'intained within the withdrawal ifaits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrswn condition; and 3.

A Special Report is prepared and subeitted to the Commission, pursuant to Specification 6.8.2 within 10 days, describing the value of the esasured MTC.' the Interia control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

i b.

With the MC acre negative than the limit of Specification 3.1.1.3b.

above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

\\

With K,ff greater than or equal to 1.

- See Special Test Exceptions Specification 3.10.3.

V0GTLE UNITS - 1 & 2 3/4 1-4 i

INSERT A0 a.

Unit 1:

l Less positive than +0.7 x 10-4 ak/k/'F for the all rods withdrawn, beginning of cycle life (COL) condition for power levels up to 70-percent RATED THERMAL POWER with a linear ramp to 0 ak/k/*F at 100-percent RATED THERMAL POWER.

Unit 2:

Less positive than 0 ak/k/*F for the all rods withdrawn, beginning of cycle life (BOL), hot zero THERMAL POWER condition.

i b.

Unit 1:

Less negative than -4.0 x 10'4 ak/k/*F for the all rods withdrawn, end of cycle (EOL), RATED THERMAL POWER condition.

Unit 2:.

i Less negative than -4.0 'x 10-4 ak/k/'F for the all rods withdrawn, end of cycle (E0L), RATED THERMAL POWER condition.

l l

l t

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i f

1 I

[

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l

_ REACTIVITY CONTROL SYSTEMS MINIMUM TEMPERATURE FOR CRITICALITY LIMITING CONDITION FOR OPERATION 3.1.1.4 The Reactor Coolant SystepHowest operating loop temperature (TI-0412, TI-0422, TI-0432 TI-0442) (T sha greater than or equal to 551'F.

g APPLICARILITY:

MODES 1 and.O' A

ACTION:

W With a Reactor Coolant Systee operating loop temperature (T,yg) less than 551*F, restore T,yg to within its Ifait within 15 minutes or be in HOT STAN08Y within the next 15 minutes.

SURVEILLANCE REQUIREMENTS 4.1.1.4 The Reactor Coolant Systaa temperature (Tavg) shall be datarmined to be greater than or equal to 551*F:

a.

Within 15 minutes prior to achieving reactor criticality, and b.

At least once per 30 minutes when the reactor is critical and the Reactor Coolant Systee T,yg (TI-W, TI-0422, U-0432 TI-042) is less than 561'F with the T,yg-Tref Deviation Alarn not reset.

L "V th K,ff greater than or equal to 1.

- e Special Test Exceptions Specification 3.10.3.

LE UNITS - 1 & 2 3/4 1-6

REACTIVITY CONTROL SYSTEMS 3/4.1.2 80RATICN SYSTEMS FLOW PATH - SHtJTDOWN l

LINITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following baron injection flow paths shall be OPERA 8LE:

i a.

A floc path from the beric acid storage tank via a boric acid transfer pump and a charging pusp to the Reactor Coolant System if the boric acid storage tank in Specification 3.1.2.5a. is OPERA 8LE, or b.

A flow path from the nfueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank in Specification 3.1.2.56. is OPERABLE.

APPLICA8!LITY:

M00E5 5 and 8.

ACTION:

l l

With none of the above flow paths OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS i

i 4.1.2.1 At least one of the above Nquired flow paths shall be demonstrated l

CPERA8LE:

a.

At least once per 7 days wheri the boric acid storage tatik is a re-1 quired water source, verify that the applicable portions of the auxiliary building (TI5L 12410 or TI5L 12411. TI5L 12412 or TI5L 12413 414 or TISL M412, TI5L 12416 or TISL 12417. TISL 20900 or TI5L 2

1. TISL 20902 or TI5L 20903, and TISL 20904 or TISL 20905) yand are nt* the portions of the flow path for which ambient toeperature j

indication are not provided are 1 45'F, and b.

At least once per 31 days by verifying that each valve (sanual,

)

power-operated., or automatic) in the flew path that is not locked, sealed, or otherwise secured in position, is in its correct position.

l 1

I V0GTLE UNITS - 1 & 2 3/4 1-7

_ REACTIVITY CONTROL SYSTEMS BORATED WATER $0VRCE - SHUTOOWN LIMITING CON 0! TION FOR OPERATION 3.1.2.5 As a sinisua, one of the following boratad water sources shall be OPERABLE:

a.

A Boric Acid Storage Tank with:

l 1)

A minimum contained borated water volume of 9504 gallons (19%

of instrument span) (L!-102A, LI-104A),

2)

A boron concentration between 7000 ppe and 7700 ppe, and 3)

A minimum solution temperature of 65'F (TI-0103).

b.

The refuelirq water sLeage tank (RWST) with:

li (ni '

d ol

/'of 00 g il a (

i r

w

/

of inst LI 0. 0 4

9 I-9

{ng g g L -0

)

2[/

n oc t a io b tyee ho p d 22 0 n

3)

A minimum solution temperature of 54*F (TI-10982).

APPLICA8ILITY:

MODES $ and 6.

ACTION:

With no borated water source OPERA 8LE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

SURVEILLANCE REQUIREMENTS 4.1.2.5 The above required borated water source shall be demonstrated OPERABLE:

a.

At least once per 7 days by:

1)

Varifying the boron concentration of the water, 2)

Verifying the contained borated water volume, and 3)

When the boric acid storage tank is the source of borated water and the ' ambient temperature of the horic acid storage tank reos l

(TISL-20902 TISL-20903) is 172*F;F. verify the boric acid storage tank solution temperature is > 65 b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST tescerature (TI-10982) when it is the source of borated water and the outside air temperature is less than SV F.'

V0GTLE UNITS - 1 & 2 3/4 1-11

i INSERT AE 1)

A minimim contained borated water volume of:

Unit 1 - 99.404 gallons (9 percent of instrument span)

Unit 2 - 70,832 gallons (5 percent of instrum nt span)

(LI-0990A & 8, LI-0991A & 8, LI-0992A, LI-0993A) 2)

A boron concentration between:

l Unit 1 - 2400 ppm and 2600 ppm Unit 2 - 2000 ppm and 2200 ppm G

1 i

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i e

REACTIVITY CONTROL SYSTEMS 80 RATED WATER SOURCES - OPERATING LINITING CON 0! TION FOR OPERATION 3.1,,2.6 As a minimum the following borated water source (s) shall be OPERABLE as required by Specification 3.1.2.2:

a.

A Boric Acid Storage Tank with:

1)

A minimum contained borated water vclume of 36674 gallons (81%

of ti strument span) (LI-102A, LI-104A),

2)

A baron concentration between 7000 ppe and 7700 ppe, and 3)

A sinimum solutien temperature of 65'F (TI-0103).

b.

The refueling water storage tank (RWST) with:

1)

A minimum contained borated water volume of 631478 gallons (85 of instrument span) (LI-0990A44, LI 0991A&8, LI-0992A, LI-0993A),

l-2) a _ sm n e n ne_._...+ e... ' w....-..a n.a.n. 7 7---...2.e.. n. - r -,

w r

g 3)

A minimum solution temperature of 54'F, and 4)

A maximum solution temperature of 116*F (TI-10982).

l APPLICABILITY:

M00ES 1, 2, 3, and 4.

1 ACTION:

i a.

With the Boric Acid Storage Tank inoperable and being used at one of the above required berattd water sources, restore the tank to CPERA8LE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STM08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and borated to a SHUTDOWN MARGIN as required by Figure 3.1 2 at 200*F restore the Boric Acid Storage Tank to 0PERA8LE status withI'n the next 7 days or be in l

COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With the RWST inoperable, restore the tank to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STAN0tY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

l l

i V0GTLE UNITS - 1 & 2 3/4 1-12 e'

INSERT AF

\\

2)

A boron concentration between:

Unit 1 - 2400 ppm and 2600 ppm Unit 2 - 2000 ppm and 2100 ppm

)

9

1 TA&i.E 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN IN0FERA8LE CONTROL OR SHUTDOWN R00 Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Decrease in Reactor Coolant Inventory d rtant Opening of a Pressurizer Safety or Relief Valve Leyk t Instrument Line or Other Lines free Reactor Coolant Press re Boundry That Penetrate Contatraent Lo f-Coolant-Accidents Increase in Heat Removal by the Secondary Systas (Steam Systes Piping Rupture)

Spectrum of Rod cluster Control Assembly Ejection Accidents.

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I V0GTLE UNITS - 1 & 2 3/4 1-16

i AEACTIVITY CONTROL SYSTEMS POSITION IN0! CATION $YSTEM - SHUTDOWN j.!MITINGCON0!TIONFOROPERATION l

3.1.3.3 One digital rod position indicator (excludin2 demand position indication) shall be OPEAA8LE and capable of determining the control rod position wtthin i 12 steps for each shutdown or control rod not fully

inserted, g

APPLICA8ILITY: MODES 3,*

4,* ** and 5,8 ACTION:

With less than the above required position indicator (s) OPERA 8LE, immediately open the Reactor Trip Systas breakers.

i l

SURVEILLANCE REQUIREMENTS 4.1.3.3 Each of the above nquired digital rod position indicator (s) shall be determined to be OPERA 8LE by verifying that the digital rod position indicators agree with the demand position indicators within 12 steps when exercised over the full-range of rod travel, at least once per 18 months.

l O

Wit the Reactor Trip Systas breakers in the closed position.

- es pecial Test Exceptions Specification 0.10.5.

l A

e V0GTLE UNITS - 1 & 2 3/4 1-18

REACTIVITY CONTROL SYSTEMS SHUTDOWN RCD INSERTION LIMIT LIMITING CONDITION FOR OPERATION 3.1.3.5 All shutdown rods shal u

ithdrawn.

,O" APPLICABILITY:

MODES 18 and 3

ACTION:

e With a maximum of one shutdown to not fully withdrawn, except for surveillance tasting pursuant to Specification 4.1.3.1.2, within I hour either:

a.

Fully withdraw the rod, or b

Declare the rod to be inoperable a.3d apply Specification 3.1.3.1.

i SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown rod shall be detarsined to be fully withdrawn:

Within 15 minutes prior to withdrawal of any rods in Control a.

Bank A, B, C, or 0 during an approach to reactor criticality, and b.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter.

l

_e.

'See'fpecialTestExceptionsSpecifications3.10.2and3.10.3.

4 it K,ff greater than or equal to 1.

d4

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V0GTLE UNITS - 1 & 2 3/4 1-20 i

REACT!v!TY CONTROL SYSTEMS CONTROL R00 INSERTION L.IMITS l,IMITING CONDITION FOR OPERATION 3.1.3.6 The : introl banks shall be limited in physical insertion as shown in Figure 3.1-3.

APPLICABILITY: MODES 18 and a

ACTION:

With the control banks inserted b'eyond the above insertion limits, except for surveillance testing pursuant to Specification 4.1.3.1.2:

Restore the control banks to within the limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or a.

b.

Reduce THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the bank post-tion using the above figure, or c.

Se in at least HOT STAN08Y within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each control bank shall be determined to be within i

the insertion limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months axcept during time intervals when the rod insertion limit monitor is inoperable, then verify the individual rod positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

"See pecial Test Exceptions Specifications 3.10.2 and 3.10.3.

i K,ff greater than or equal to 1.

A#

V0GTLE UNITS - 1 & 2 3/4 1-21 e'

I 7

7 220

/

l 200 I

/ BANK B

/

I 180 I/

/

I s

5

/

(100. al) E

$,ao (0, 161)

[

d

/ BANK C

/l s34,I

/

I E I

/

I E

I

/

I 5

I

/

_I 30o Bi I

/

I e*E l /

/

BANK D E

s I

/

if I

E I/

/

I g 60 g

3 (0,48)

/

I I

/

I

,0 l

I

/

I O

m-m 0

20 40 60 80 100 R ELATIVE POWER (percent)

FIGURE 3.1-3 R00 BANK INSPECTION LIMITS VERSUS THERMAL POWER V0GTLE UNITS - 1 & 2 3/4 1-22

3/4.2 POVER DI$TRIBUTION LIMITS 3/4.2.1 AXIALFLUXOIFFERENCl LIMITING CONDITION FOR OPERATION 3.2.1 The indica ed '(NI-00418, NI-00423, NI-00438 NI-0044B) AXIAL FUM DIFFERENCE (AFD) shall be saialained within the following traget band (flux difference units) about the target flux difference:

a.

2 5% for core avurage accumclated burnup of less tha

$r equal to 3000 MWD /KTU; and b.

+ 35, -15 for core average accumulated burnup of greater than 3000 MWD /MTU.

The indicated AFD may deviate outside the above required target bane. at greater than or equal to 505 but less than 90% of RATED THERMAL PCWER provided the indi-cated AFD is within the Acceptable Operation Limits of Figure 3.2-1 and the casa-lative penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

The indicated AFD say deviate outside the above required target band at greater than 15% but less than 50% of RATED THERMAL POWEP provided the cumulative penalty deviation time does not exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months during the Ieyfe 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

,APPLICA81LITW MODE 1, above 15% of RATED TNiiO4AL PCVER " **

ACTION:

With the indicated AFD outside of the above required t.arget band and 4.

with THERMAL PCWER gnatar than or equal to 9C% of RATED THERMAL POWER, within 15 minutas either:

1.

Restore the indicated AFD to within the target hand limits, or 2.

Reduce THERMAL PCWER to less than 90% of RATED THERMAL PCWER.

b.

With the indicated AFD outside of the above required target band for som than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time darin0 the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or outside the Acceptable Operation Limits of Figure 3.2-1 and with THERMAL PCWER less than 90% but equal to or gnatar than 505 of RATED THERMAL PCWER, nduce:

1.

THERMAL PCWER to less than 5C% of RATED THERNAL POWIR within 30 minutes, and 2.

The Power Range Neutron Flux" - High Setpoints to less than or equal, to 55% of RATED THERMAL PCWER within the naxt 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

" th5pecial Test Exceptions Specification 3.10.2.

uryeillance testing of the Power Range Neutron Flux Channel say be performed Celow 90% of RATED THERMAL PCVER) pursuant to $pecification 4.3.1.1 provind i the indicated AFD is maintained within the Acceptable Operation Limits of Figure 3.2-1.

A total of 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months operation say be accumuisted with the AFD outside of the above required target band during testing without penalty deviation.

V0GTLE UNITS - 1 & 2 3/4 2-1

e 140 I

120 l

UNACCEPTABLE 100 OPERATION

(.11,401 (11,60)

/

i i

E 80 i

/

\\

$zE l

W 60 ACCEPTABLE OPER ATION 8-(41.50)

(31,50) 40 l

i i

20

.I 0

-50

-40

-30

-20

-10 0

10 20 30 40 50 FLUX DIFFERENCE,.il (percent) 4 FIGURE 3.2-1 l

AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER YOGTLE UNITS - 1 & 2 3/4 2 3

1.50 r

1.25 1.00 8

0,75 CORE C.50 HEIGHT K(Z) 1 0.000 1.000 8.000 1,000 10.835 0.939 0.25 12.000 0.652 i

0 0

2.0 4.0 8.0 8.0 10.0 12.0 CORE HEIGHT (ft) l P

l l

FIGURE 3.2-2 K(Z) - NOR!tALIZE0 F (Z) AS A F1)NCTION OF CORE HEIGHT Q

4 YOGTLE UNITS - 1 12 3/4 2-6

TA8tE 3.3-1 (C wtinued)_

h AEACIM TRIP SYSTEM INSTERRENTATION MINIM M Q

14TAL NO.

CHAleIE!S OL'astELS APPLIC/3tt g

ILNCTIONAL utilT 0F CHAselELS TO TRE OFERASLE le0ES ACTION h

7.

Overtemperetwo AT 4

2 3

1, 2 6

OCI-M11C,181-9421C, 101-0431C,101-8441C) m s.

Overpower AT 4

2 3

1, 2 6

(161-04118,10I-04Z18, 101-M318,181-04414) 3 9.

fresswlier Pressure--Law 4

2 3

1 6"

~

(PI-M55A,84C, PI-BtM & PI-M56k, PI-0457 4 FI-M 57A, PI-0454 &

w PI-0453A) b w

10. Presswlor Presswe-Migh 4

2 3

1, 2 6

t (t!-04554,8sc, PI-M56 & ?I-M564, F1-0457 & PI-6457A, PI-C458 &

PI-M54A)

}

i y

11. Presswizer 'Ater Level-Migh" 3

2 2

l' 6

(LI-045M. LI-84604, it-9461A)

. _s

12. Reacter feelant flow--Les h

b StaO e Leap (Above P-4) 3/ loop 2/leep in esp in 1

g l

a.

any oper-each oper-ating leap eting leap (LOOP 1 LOOP 2 LOOP 3 LGV 4 FI-6414 FI-8424 FI-0434 FI-0444 fI-0415 fI-0425 F1-0435 F1-0445 fI-9416 FI-0426 fI-8436 fI-M45)

"See Specification 3.3.3.E

f TABLE 3.3-1 (Continued)

TABLE NOTATIONS When the Reactor Tri9 System breakers are in the closed position and the a

Control Rod Drive System is capable of rod withdrawal.

b The provisions of Specification 3.0.4 are not applicable.

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

~

d Below the P-10 (Low Setpoint Power Range Neutron Flux Inbrlock) $ctpoint.

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

e f Above the Power Reacto'r Trip Block) Setpoint, g The applicab.

Modes and Action Statement for these channels noted in Table 3.3-3 are more restrictive and, therefore, applicable.

h Above the P-8 (Single Loc es o low) int.

i Trip logic consists o undeholtage/ unde]

ency for Reactor Coolint Pumps I or 2 and 3 or 4 V

j The Source Range High Flux at u' own ars say bc blocked during reactor startup in accordance with approvea procedures.

ACTION STATEMENTS ACTION 1 - With the number of OPERAdLE channels one less than the Minimus Channels OPERABLE requirement,.(store the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels,ing condition;s are satisfied:STARTUP and/

provided the follow a.

The inoperable channel is placed in the tripped condition within 6 houn, b.

The Minimum Channels OPERABLE requirement is met; howeer, the inoperable channel say be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for surveillance tasting of other channels per Sp(cification 4.3.1.1, and c.

Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERW's POWER and the Power Range Neutron Flux Trip Setpoint is nduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the QUACRAN1 P7 DER TILT RATIO is monitored at icist once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Spe.ification 4.2.4.2.

V0GTLE UNITS 1 & 2 3/4 3-6 i

TABLE 4.3-1 (Continued) l-REACTOR TRIP SYSTEM INSTkUMENTATIOM SURVEILLANCE REQUIRENENTS

~

TRIP g

ANALOG.

ACTUATING MODES FOR 3

CHANNEL DEVICE WICH v

C gANNE ERATI OPERATI TUATI

SURVEILLANCK FUNCTIONAL UNIT CHECK LIBk.. ION EST

/

TEST OGIC TE S REQUIRED 8.

Overpower AT S

R Q(17)

N.A.

1, 2 (I01-04118. T01-04218, TDI-04312,101-04418) 9.

Pressurizer Pressure-Low 5

R Q(17)

N.A.

M.A.

1*

l (PI-0455A,8&C, PI-8456 &

l PI-0456A, PI-0457 &

i PI-0457A, PI-0458 &

)

q PI-0458A)

+

w

10. Pressurizer Pressure--High 5

R Q(17)

M.A.

1, 2 (PI-0455A,8&C, PI-0456 I

& PI-0456A, PI-0457

& PI-0457A, PI-0458

& PI-0458A) i,,

11. Pressurizer Water level--

S R

Q(17)

N.A.

l' High" (LI-0459A,LI-0460A, LI-0461A)

12. Reactor Coolaat F1:nt--Low S

1 l'

y (LOOP 1 LOOP 2 LOOP 3 LDOP4 FI 9414 FI-0424 FI-0434 FI-0444 FI-0415 FI-0425 FI-0435 FI-0445 FI-0416 FI-0426 FI-0436 FI-0446) l 1

See Specification 4.3.3.6

I 4

TABLE 4.3-1 (Continued)

REACTOR TRIP SYSTEM INSTRtMENTATION SURVEILLANCE EEQUIREMENTS E

i TRIP E

ANALOG ACTUATING MODES FOR q

CHANNEL DEVICE WICH a

CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLAK7E l

FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED e

e-

13. Steam Generator Yater Level--

S R

Q(17,18)

N.A.

1, 2 a

Low-Low" (LOOP 1 LOOP 2 LOOP 3 LOOP 4 LI-0517 LI-0527 LI-0537 LI-0547 LI-0518 Li-0526 LI-0538 LI-0548 LI-0519 LI-0529 LI-0539 LI-0549 i

LI-0551 LI-0552 LI-0553 LI-0554)

14. Undervoltage - Reactor Coolant N..t.

R N.A.

Q(17)

N.A.

l' Pumps

15. Underfrequency - Reactor N.A.

R N.A.

Q(17)

N.A.

1*

Coola..c Pumps

16. Turbine Trip l

b a.

tow Fluid Oil Pressure N.A.

R S/U (1, 10)

N.A.

I (PT-6161,P1-6162,PT-6163) b b.

Turbine Stop Valve Closure N.A.

R N.A.

S/U(1,10)

N.A.

I

17. Safety Injection Input from N.A.

N.A.

R N.A.

1, 2 ESF

18. Reactor Trip System Interlocks a.

Intermediate Range C

Neutron Flux, P-6 N. I..

R(4)

R N.A.

N.A.

2 Ml W R D&E, NI-00368,0&(&

(ee,Specificatios4.3..

TA8Li 4.3-1 (Continued)

P4LENOTATIOM a When the Reactor Trip Systes breskers are closed and the Control Rod Drive Systes is capable of rod withdrawal.

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

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

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

e Above P-7 (Low Power Reactor Trip Block) Setpoint.

(1) If not performed in p.avious 31 days.

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

Adjust cxcore channel gains consistent with calorimetric power if absoluta difference is greater than 25.

The provisions of Specification 4.0.4 aN not applicable to entry into MODE 2 or 1.

(3) Sirgle point comparison of incore to excore AXIAL FLUX DIFFERENCE chove 15% of RATED THERMAL POWER.

Recalibrata if the abselute difference is greater than or equal to 3%.

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

For the purpose of this surveillance requirement, sontnly shall mean at least once per 31 EFPO.

l (4) Neutron detectors say be excluded from CHANNEL CALIBRATION.

(5) Detector plateau curves shall be obtained, and evaluated.

For the Intermediata Range and Power Range Neutron Flux char.nels the provisions of Specification 4.9.4 are not applicable for entry into MODE 2 or 1.

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

This is the determinatica of the response of the axcore power range detectors to the incore measured axial power distribution to generata setpoints for the

'HANNEL CM 18 RATION.

The provisions of Specification 4.0.4 are not appli-cable for antry into MODE 2 or 1.

For the purpose of this surveille.nce requirement, quarterly shall mean at least once per 92 EFPO.

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

(8) Not used.

8 8

8 (9) Quarterly surveillance in M00ES 3, 4, and 5 shall also include verifi-cation that peruf ssives P-6 and P-10 are in their required state for existing plant conditicas by observation of the permissive window.

Quarterly surveillance shall include verifiestion of the Source Range High' Flux at Shutdown Alars Satpoint of less than or equal to % M times background.

A A3 OlALU 01 V0GTLE UNITS - 1 & 2 3/4 3-13 L

TABLE 4.3-1 (Continued)

TABLE NOTATIONS (Continued)

(10) Setpoint verification is not applicable.

(11) The TRIP ACTUATING DEVICS OPERATIONAL TEST shall include independent verification of the OPERABILITY of the Undervoltage and Shunt. trip of the Reactor Trip Breaker.

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

(13) Not used.

(14) The TRIP ACTUATING DEVICE OPERATIONAL TEST sh. il independently verify the OPERABILITY of tne undervoltage and shunt trip circuits for the Manual Reactor Trip function, ne test shall also oarify the OPERABILITY of the Bypass Breaker trip circuit (s).

(15) Local senual shunt trip prior to placing breaker in service.

16) Automatic undervoltage trip.

N F

Each channel shall be tastad at least every 92 days on a STAGCERED TEST gv sAS:5.

(18) The surveillance frtquency and/or H0 DES specified for these channels in Table 4.3-2 are more restrictive and, therefor ~a, applicable.

P i

V0GTLE UNITS - 1 & 2 3/4 3-14 l

l 1

1 INSTRUMENTATION 3/4.3.2 ENGINEERED _ SAFETY FEATURES ACTUNfION SYSTEM :95TRUMENTATION LIMIfING CONDITION FOR OPERATION i

3.3.2 The Engineered Safety Features Actuation Systere (ESFAS) instrumentation channels and interlocks shown in Table 3.3-2 shall be OPERABLE with their Trip Setpoints set consistent with the values shown in the Trip Set Table 3.3-3 and with response times within their limit value. point column of APPLICABILITY:

As shown in Tale 3.3-2.

ACTION:

With an ESFAS Instrumentation or Interlock Trip Setpoint trip less a.

conservative than the value shown in the Trip Setpoint column but more conservative than the value shown in the Allowable Value column of Table 3.3-3, adjust the Setpoint consistent with the Trip Setnoint value.

b.

With an ESFAS Instrumertation or Interlock Trip Setpoint less conserva-tive than the value shown in the Allowable Value column of Table 3.3-3, either:

1.

Adjust the Setpoint consistent with the Trip Setpoint value of Table 3.3-3, and detarsine within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months that Equation 2.2-1 was satisfied for the affected channel, or 2.

Declare the cha.or.el inoperable and apply the applicable ACTION statement requirments of Table 3.3-2 until the channel is restored to OPERA 8LE status with its Setpoint adjusted consistant with the Trip Setpoint value.

Equation 2.2-1 Z + R + 5 < TA Where:

lZ = The value from Column I of Table 3.3-3 for the affected channel, R = The "as measured" value (in percent span) of rack error for the affected char.ael, S = Either the "as seasured" valu (in per:ent span) of the sensor error, or the value from Column S (Sensor Error) of Table 3.3-3 for the affected channel, ar.d TA = The value from Column TA (Total Allowance) of Table 3.3-3 for th affected channel, c.

With an ESFAS instrumentation channel or interlock inoperable, take the ACTION shown in Table 3.3-2.

V0GTLE ' NITS - 1 & 2 3/4 3-15 J

_ _. =,

i l

TABLF 3.3-2 (Continued) l 8

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRLSENTATION i

-4 G

C MINIMUM

)

Q TOTAL NO.

CHANNELS CHAl#IELS APPLICABLE

^

FUNCTIONAL UNIT OF CHANNELS TO TRIP

OPERABLE,

_ M00ES ACTION l

~

4.

Steam Line Iselation (Continue

)

d c.

Cratainment Pressne-Hf-2 2

2 1, 2, 3 15 (PI-0934. PI-0935. PI-69 d.

Steam Line Pressure-Low

3/ steam 2/ steam 2/ steam d

1, 2, 3*'#

15 line line any line steam line j

(LOOP 1 LOOP 2 LOOP 3 LOOP 4 PI-0524A,8&C PI-0524A&R PI-uS34A&8 PI-0544A,8&C, y

PI-0515A PI-0525A PI-0535A PI-0545A, g

PI-0516A PI-0526A PI-0536A PI-0546A) b d

3.f 35 e.

Steam Line Pressur 3/ steam 2/ steam 2/ steam

^

Negative Rate--Hi M

line line any line steam line (LOOP 1 LOOP 2 LOOP 3 L0dP4 l

PI-0514A,8&C PI-0524A&8 PI-0534A&8 PI-0544A,8&C PI-0515A PI-0525A PI-0535A VI-0545A

{

PI-0516A PI-0526A PI-0536A PI-0546A) 4 5.

Turbine Trip and Feodwater Isolation 2

1 a.

Automatic Actuation Logic 2

1 2

1, 2 25

)

and Actuation Relays b.

Low RCS Tavg Coincident I

with Reactor Trip **

d 1.

Low RCS T,,g 4

2 3

1, 2 20 1

2.

Reactor Trip, P-4 See Functional Unit 9b for P-4 requirements.

e.

e

..sst.

ns

~

~s e

i TABLE 3.3-2 (Continued)

<8 ENGINEERED SAFETY Frf.TURES ACTUATION SYSTEM INSTRilNENTATION j

4 WININUM i

i E

TOTAL NO.

CHANNELS CHMNELS 1.PPLICABLE Q

FUNCTIONAL UNIY OF CHANNELS TO TRIP,

OPERABLE, M00ES ACTION us 4

5.

Turbine. Trip and Feeduater Isolation (Continued) e.

,i e

c.

.Steae Generator Water 4/sta. gen.

2/sta. gen.

3/sta. gen.

1, 2 20d l

Level-HigerHigh (P- ),

in any oper-in each j

ating sta.

operating i

gen.

sta. gen.

(LOOP 1 LOOP 2 LOOP 3 LOOP.1 LI-e517 LI-0527 LI-0537 LI-0547 LI-0518 LI-0528 LI-0536 LI-0548 m1 LI-0519 LI-0529 LI-0539 LI-0549 1

LI-0551 LI-0552 LI-0553 LI-0554) 4 j

m k

d.

Safety Injeciton See Furxtional Unit I above fer all Safety l

Injection initiat.ing functions and requi:eme ts..

.i 6.

Aex111ary feesdater i

t.t.tomatic Actuation Logic 2

1 2

1,2,3 22 4.

~

ud Actuation Relays l

Se SpedG ccaion 3. 3. 3 4

i TABLE 3.3-2 (Centinued)

TABLE NOTATIONS a Trip function say be blocked in this H00E below the P-11 (Pressurizer Pressure Interlock) Setpoint.

b Trip function automatically blocked above P-11 and may be blocked below P-11 when Safety Injection on low steam line pressure is not blocked.

c During movement of irradiated fuel or movement of loads over irradiated fuel within containment.

d The provisions of Specifiestion 3.0.4 are r.ot appl. :able.

e Ouring movement of irradiated feel or movement of loads over irradiated fue'.

f Not applicable if one main steam isolation valve and associated bypass iso-laiion valve per steaalN is closed, g Contairueent Ver.tilaticn Radiation (RE-2565) is treated as one channel and ts considered OPERABLE if the particulate (FJ-2565A) and iodiae monitors (RE-25658) are OPERA 8LE or the nobic gas monitor (RE-2565C) is GPERA8LE.

h Manual initiation of Auxilia ptar is accomplished via the pump handswitches.

5 GI W f i Whenever irradiated fuel i in a stora pool.

j For actions associated with ino e instrumentation, follow actions specified in Specification 3.9.12.

ACTION STATENENTS, ACTION 14 - With the number of 0PERABLE channels one less than the Minista Channels OPERA 8Li requirement, be in at least HOT STAN0BY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the N11 ewing 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ; however, one channel say be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.2.1, provided the other channel is 0FERA8LE.

ACTION 15 - With the number' of OPERA 8LE channels one less than the Total Number of Channels, operation say proceed until performance of the next rcquired ANALOG CHANNEL OPERATIONAL TEST rarovided the inoperable channel is placed in the tripped conditicn within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

ACTION 16 - With the number of OPERA 8LE channels less than the Minfeus Channels OPERA 8kE requirement, comply with the ACTION requirements of Specification 3.9.3 (Mode 6).

l l

l V0GTLE UNITS - 1 & 2 3/4 3-25 l

t I

- ~ -

~

TABLE 3.3.1 (Continued) 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEi4 INSTRUMENTATION TAIP SETPOINTS TOTAL SEN504 f

i ALLOWANCE ERROR I g (TA)

I (5)

TAIP SETPOINT A1LOWA8tf VALUE

g FUNCTIONAL UNI 7 i

yj 4.

Steam Llae Isolation

~

a.

Manual In1ttatIon N.A.

N.A.

M.A.

N.A.

k 1

b.

Auteatic Act::stion Logic M.A.

N.A.

M.A.

N.A.

p and Actuation Aalays c.

Cwntainment Pressure--Nfgh-2 3.1 0.71 1.67

<14.5 psig

<15.4 psig (PI-0934 PI-0935. PI-0936) d.

Ste.w Line Pressure--Low 13.0 10.71 1.67 1585 pstg" 1570 p>fg (LOOP 1 LOOP 2 LOOP 3 LOOP 4 o

PI-0514A,8&C PI-0524A&8 fI-0534A&6 PI-0544A,8&C 5;*

PI-0515A PI-0525A PI-0535A PI-0545A g

PI-0516A PI-0526A PI-0536A PI-0546A)

Steam Line Frassure - Negativa 3.0 0.50 0

e.

Rate--Migh

-<100 ps:""

<125 pst (LOOP 1 LOOP 2 LOOP 3 LOOP 4

?I-0514A,0&C PI-0524A&8 PI-0534A&8 PI-0544A,8&C PI-0515A PI-0525A PI-0535A PI-0545A j

PI-0516A PI-8526A PJ-0536A PI-0546A) 5.

Turbine irty end feedwater Isolation i

a.

Automatic Actuation Logic and M.A.

N.A.

M.A.

Actuation Relays b.

Loe ACS Yavg Celacident with Reactor Tripa#

1.

Low ACS T,,g 4.3 0.82 0.87 1564*r 1561.5*r

.?.

Reactor Trip, P-4 N.A.

M.A.

N.A.

M.A.

l l

p i

TABLE 3.3-3 (Continued)

$'l ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM IN$1AUMENTATION TR W SETPOINTS r-TOTAL SENS08 ALLOWANCE ERROR E

FUNCTIONAL UNIT (TA)

Z (S)

TRIP SETPOIN)

A_LLOWABLE VALUE d

5-Turbine Trip and feedwater Isolation (Continued) a c.

Steam Generatur Water 5.1 2.18 1.67

<78.0% of

<79.9% of narrow 4

[

Level--High-High (F-14) narrow range range instrument instrunent span. '

span.

(LOOP 1 LOOP 2 LOOP 3 LOOP 4 LI-0517 LI-0527 LI-0537 LI-0547 i

LI-0518 LI-0528 LI-0538 LI-0548 LI-0519 LI-0529 LI-05?9 LI-0549

.g LI-0551 LI-G552 LI-0553 LI-0554) l

[

d.

Safety Injection

, See functional Unit I above for all Safety Injection Trip j

Setpoints and Allowable Values.-

1 6.

Auxiliary feessater-

)

a.

Automatic Actuation Logic N.A.

N.A.

]~

and Actuation Relkys b.

Steam Generator Water Level--

l Lc,w-Low (LOOP 1 LOOF 2 LOOP 3 LOOP 4 LI-0517 LI-0527 LI-0537 LI-0547 l

LI-0518 LI-0523 LI-0538 LI-0548 l

LI-0519 LI-0529 LI-0539 LI-0549 i

LI-G551 LI-0552 LI-0553 LI-0554)

I 1.

Start Motor-0 riven Pumps 18.5 17.18 1.57 118.5%

117.8% of narrow range narrow range instrument instrument span.

2.

Start Turbine-Oriven Pump

- 18.5 17.18 1.67

>17.8K of narrow range narrow range -

fnstrument instrument

span, span.

TABLE 3.3-3 (Continued)

TABLE NOTATIONS kTime constants utilized in th'a lead-lag controller for Steam Line Pressure-Lc are tg 1 50 seconds and 12 < 5 seconds.

CHANNEL CALIBRATION shall ensure that these time constants are adjusted to these values.

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

CHANNEL CAllBRATI shall ensure that this time constant is adjusted to this value.

- - Untti resolution of the Veritrak transmitter uncertainty issue this setpoint will be set at 11885 psig.

Until reso'lution of the Veritrak transmitter uncertainty issue the cutootnt will be set at 122.5% (Unit 1) and 123.EE (Unit 2) of narrow range instrument span.

Edwaterisolationonly.

Turbine trip occurs on reacter trip.

NFe 80 aring refueling operations, bD Jring power operation.

This is an initial setpoint only.

The trip setpoint wi ll be set at 50 times background level.

Backgrovad level shculd be detar-ned at or near the end of the first fuel cycle.

a C

Se tpoints will not exceed the lint 4 of Specification 3.11.2.1.

1 4

a V0GTLE UNITS - 1 & 2 3/4 3-35 i

s TABLE 4.3-2 (CenLinued) g FEINEEREf, SAFETY FEATURES ACTUATION SYSTEN INSTRUMENTATION p

SURVEILLANCE REQUIREN NTS E19 g

TRIP q

ANALOG ACTUATING MODES CHANNEL DEVICC MASTER SLAVE FOR WICN 6

CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY. RELAY SURVEILLANCE FUNCTIONAL UNIT CHECs(

CALIBRATION TEST TEST

_ LOGIC TEST TEST TEST IS REQUIRED e

s

5. Turbine Trip and Teedater.

Isolation

a. Autcoatic Actuation N.A.

N.A.

M.A.

M(1)

N(1)

Q 1, 2 Logic and Actuation Relays

b. Low RCS Tavg y

Colacident with Reactor Trip

y 1.

Low RCS T,,,

5 R

H N.A.

N.A.

1, 2 5

2.

Reactor Trip, P-4 See functional Unit 9b for P-4 Survel11ance requirements.

c. St as Generator S

R M

N.A.

1, 2 er Level-N yh P-14 *

(L 00P2 LOOP 3 LOOP 4 LI-0517 LI-0527 LI-0537 LI-0547 LI-0518 LI-0528 LI-0.538 LI-9548 LI-0519 LI-0529 LI-0539 LI-0549 LI-0551 LI-0552 L1-0553 LI-0554) d.

Safety injection See functional Unit I above for all Safety Injection Surveillance Requirements.

feedwater isolation only. Turbine trip occurs on reactor trip.

"See Specification 4.3.3.6

TABLE 4.3-2 (Continued)

ENGINEE2ED SAFETY FEATURES ACTUATION SYSTEM INSTRUNENTATION p

SURVEILLANCE REQUIREMNTS m

g TRIP q

ANALOG ACTUATIts M00ES m

CHANNEL DEVICE MASTER SLAVE FOR WHICN CHANNEL CHANNEL CHAPNEL OPERATIONAL OPERATIONAL ACTUATION RE!L' RELAY SURVEILLANCE FUNCTIONAL UNIf CHECK CALIBRATIGN TEST TEST LOGIC TEST TEST TEST IS REQUIRED H

e-10.

Cantrol Room Ventilation Emergency Mode Actuation (Continued) c.

Safety Injection 56e Functional Unit I above for all Safety Injection Surveillance Requirements.

d.

Intake Radiogas Monitor 5

2 M

N.A.

N. A..

1, 2, 3, 4 (RE-12116 RE-12117)

SG, 60 11.

Fuel Handling Building Post Accident Ventilation Actuation g

(Comacn System) a.

Manual Initiation N.A.

N.A.

R N.A.

N.A.

.(2) b.

Fuel Handling S'

R M

N.A.

(2)

Building Exhaust Doct Radiation Signal (ARE-2532 A&B ARE-2533 A&8) c.

Automatic N.A.

N.A.

N.A, N.A.

M(1)

N.A.

(2)

Actuation Logic

~

and Actuation Relays C

TABLE NOTATION (1) Each train shall bz tested ar. I ast ave y 62 days on a STAGGERED TEST BASIS.

(2) Whenever irradiated fuel is in ste age pool.

0 During movement of irradiat d ue movement of loads over irridated fuel.

INSTRUMENTATION SEISMIC INSTRUMENTATION mmon ushm) g

/

LIMITING CONDITION FOR OPERATION 3.3.3.3 The seismic sonitoring instrumentation shown in Table 3.3-5 shall be OPERA 8LE.

APPLICABILITY: At all times.

ACTION:

With one or more of the above required safsnic monitoried instruments a.

inoperable for more than 30 days, propan and submit a Special Report to the Coesission pursuant to Specification 6.8.2 within the next 10 days outlining the cause of the salfunction and the plans for restoring the instrument (s) to OPERA 8LE status.

b.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable, s

SURVEILLANr.E REQUIREMENTS 4.3.3.3.1 Each of the above requind safsaic monitoring instruments shall be demonstrated OPERA 8LE by the performance of the CHANNEL CHECK, CHANNEL CALI-8 RATION, and ANALOG CHANNEL OPERATIONAL TEST at the frequencies shown in Table 4.3-4.

4.3.3.3.2 Each of the above requind seisaic monitoring instrtments which is accessible during power operations and which is actuated during a seismic event greater than or equal to 0.01 g shall be nstond ta OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and a CHANNEL CALIBRATION performed within 15 days following the seismic event. Data shall be retrieved free actuated instruments and analyzed to determine the magnitude of the vibratory ground action. A Special Report shali be pnpared and submitted to the Commission pursuant to Specifica-tion 6.8.2 within 14 days describing the magnitude, frequency spectrum, and nsultant effect upon facilit/ featuns important to safety.

Each of the above seisaic monitoring instruments which is actuated du*ing a seismic event gnater than or equal to 0.01 g but is not accessible during power operation shall be nstored to OPERA 8LE status and a CHANNEL CALIBRATION per-fonsed the next time Unit 1 enters MODE 5 or below.

A supplemental report shall then be prepared and submitted W the Coamission with 14 days pursuant to Specification 6.8.2 describing the additional data from these instruments.

V0GTLE ONITS 1 & 2 3/4 3-50

R TABLE 3.3-5 SEISMIC MONITORING INSTRUMENTATION MINIMUM INSTRUME!

MEASUREMENT INSTRUMENTS TAG INSTRUMENTS AND SENSOR LOCATIONS RANGE

_ OPERABLE NUMBER

1. Triaxial Time-History Accelerographs

4. Fne Field (500 ft from containment)

-1 to 1 g 1

AXT-1990C

~. Unit 1 Containment Gallery (basemat)

-1 to 1 g 1

AXT-1990*.

o

c. Unit 1 Containment Operating Floor

-1 to 1 g 1

AXT-19907.

d. Auxiliary Building Basemat

-1 to 1 g 1

AXT-19906

e. Unit 1 Containment Pressurizer Support -1 to 1 g 1

AXT-19903

f. Auxiliary Building Level 1

-1 to 1 g 1

AXT-19902

2. Triaxial Peak Accelerographs

a. Unit 1 Reactor Coolant Pug Motor (210 ft) 110gHoriz/15gVert 1 AXR-19910

b. Unit 1 Staae Generator (185 ft) 22gHoriz/t5gVert 1

AXR-1991

c. Unit 1 NSCW Piping Outside Aux Bldg (220 ft)

-10g to +10g 1

AXR-19913

3. Triaxial Seismic Switch

a. Unit 1 Containment Tendon Gallery (basemat) 1*

AXSH-1992

4. Triaxial Reponse-Spectrus Analyzer

a. Control Roos Input: -1 'to Ig 1*

AXA-19930 Output: 0.03g to 9.99g iax *1 Setssic Triggers

_a Jlnig 1 Containment Tendon Gallery (ban

)

18 AXSH-1992.'

s t 1 Containment Operating Floor la AXSH-1992 Vith reactor control room indication.

- Triaxial seismic switch is set at the CBE acceleration level of 0.17g horizontal and 0.23g vertical.

"* Triaxial seismic trigger is set at 0 Olg all axes.

V0GTLE UNITS - 1 & ?

3/4 3-51 i

TABLE 4.3-4 SEISMIC MONITORING INSTRUMENTATION _ SURVEILLANCE REQUIREMENTS.

ANALOG CHANNEL ACCESS CHANNEL CHANNEL OPERATIONAL DURI INSTRUMENTS AND SENSOR LOCATIONS CNECK CALIBRATION

___ TEST MODE 1.

Triaxial Time-History Accelerographs a.

Free Field (500 ft from containment)

M R

SA All.

b.

Unit 1 Containment Gallery (basamat)

M R

SA All c.

Unit 1 Containment Operating Floor M

R SA All d.

Auxiliary Buf1 ding Basemat M

R SA All e.

Unit 1 Containment Pressurizar Support M R

SA 5

f.

Auxiliary Building Level 1 M

R SA All 2.

Triaxial Peak Accelerographs a.

Unit 1 Reactor Coolant Fusp Motor (210 ft)

N.A.

R N.A.

5, b.

Unit 1 Staan Generator (185 ft)

N. A.

R N. A.

5, c.

Unit 1 KSCW Piping Outside Aux. Bldg.

(220 ft)

N.A.

R N. A.

All 3.

Triaxial Seismic Switches a.

Unit 1 Containment Tendon Gallery (basemat)

M R

SA All 4.

Triaxial Responst-Spectrus Analyzer a.

Control Rooe*

M R

N. A. -

All 5.

Triaxial Seismic Triggers

~(din a

1 Containment Tendon Gallery basomat M

R SA All Jf )

hit 1 Containment Operating Floor M

R SA All

'With reactor control room indications.

V0GTLI UNITS - 1 & 2 3/4 3-52 h

TABLE 3.3-6 METEOROLOGICAL MONITORING INSTRUMENTATION

MINIMUM INSTRUMENT LOCATION OPERABLE 1.

Wind Speed a.

Lower, Primary Tower Nominal Elev. 10 a 1

b.

Upper. Primary Tower Nominal Elev. 60 m 1

2.

Wind Direction a.

Lower, Primary Tower Nominal Elev.10 m 1

b.

Upper, Primary Tower Nominal Elev. 60 m 1

3.

Air Temperature - AT a.

AT, Primary Tower Nominal Elev.10e-60s 1

E t

l I

i

(

"Tnis instrumentation is connon to Units 1 and 2.

I V0GTLE UNITS - 1 & 2 3/4 3-54 i

l

,_e---

1781E 3.3-7 h

REMOTE SHUTDOWN SYSTEM MONITORING INSTRUMENTATION m

NINIMM g

READOUT 8 CHANNELS CHANNELS INSTRUMENT Function g

LOCATION AVAILABLE OPERABLE

[

1.

Source Range Net. tron Flux A

1 (hl-31E) 1

[

2.

Ex ange Neutron Flux 8

1 (:41-13135 Caa) 1 3.

R P Cold Les Tesperal.wre A, 8 1/ Loop 1/ Loop (Loop 1 TI-04130, Panel A)

(Loop 2 TI-04230, Panel 8)

(Loop 3 TI-04330, Panel t)

(Loop 4 TI-04430, Panel A) 4.

RCS Hot Leg Temperature A

2 2

}

(Loop 1 TI-0413C Loop 4 TI-0443C)

E 5.

Core Exit Thermocouples 8

2 2

(Loop 2 Cere quandrant TI-10055 Loop 3 Core Qaandrant TI-10056) 6.

RCS 'Jide Range Pressure A, B

'2 2

,~

(PI-405A, Panel A)

- (PI-403A, Paael 8) 7.

Steana k.:;rr.= Level Wide Range A, 8 1/ Loop 1/ Loop (Loop 1 LI-5018, Panel A)

(Loop 2 LI-5028, Panel 8)

(Loop 3 LI-5038, Panel 8) l (Loop 4 LI-5048, Panel A) 8.

Pressurizer Level AG 2

2 1

(LI-459C, Pene! A)

(LI-460C, Panel E) i 9.

RWS,T Level L

1 (LI-099^C) 18 i

s TABLE 3.3-7 (Continued) 8 REMOTE illUIDOWH SYSTEM HONITORING INSTRt#tENTATION MsNIM48 E

READOUT 8 CHANNELS CHANNELS Q

INSTRUMENT Function LOCATION AVAILABLE OPERABLE

. h

10. 8AST Level L

1 (PI-10115a}

is

31. CST Level

~

L 2

28 n

(Tank 1 LI-3100)

(Tank 2 LI-5115)

12. Auxiliary feedsater Flow A, 8 1/ LOOP 1/ LOOP (LOOP 1 FI-51528, Panci A)

(LGOP2 FI-51518, Panel 3)

(LOOP 3 FI-51538, Panel 8)

(LOGP4 FI-51508, Panel A) m1

13. Steam Generator Pressure A, 6 1/ LOOP 1/ LOOP m

(LOOP 1 PI-9514C, Panel A)

(LOOP 2 PI-05258 Panel 8)

~

(LOOP 3 PI-05358, Panel 8)

(LOOP 4 PI-0544C, Panel A)

A - Remote Shutdown Panel P50A 8 - Remste Shutdown Panel P508 L - Local Indicat.lon 2

Graph will be provided to determine level from pressure reading 3 Alternate local level indication may be established to fulfill the minlaus channels OPERA 8tE.

O

TABLE 3.3-8 h

ACCIDENT HONITORING INSTRUMENTATION t

-4E TOTAL MINIMlM c

NO. OF CHANNELS INSTALMENT CHAfstELS OPERABLE ACTION us l

1.

Reactor Coolant Pressure (Wide Range) 4 1

35 (Loop 408, 418, 428, & 438) 2.

Reactor Coolant System That (Wide E4898) 1 388P

8P 32 Y

(Loop 413A, 423A, 433A 1 443A) 3.

Reactor Coolant System Tcold (WId* E8808) 1/I**P l/ loop 32

)

(Loop 4138, 4238, 43M & 4438) 4.

SG Water Level (Wide Range) 1/SG 1/SG 32 l

(Loop 501, 502, 503 & 504) i j

5.

SG Water Level (Narrow Range)

N 4/SG 1/SG 35 j

g (Loop 517, 518, 519, 527, 528, 529, 537, 538, 539, 547, 548, 549, 551, 552, 553, 554 i

6.

Pressurizer Level 3

1 30 l

(Loop 459, 460, 461) 7.

Contalament Pressure 4

1 35 (Loop 934, 935, 936, 937) 8.

Steamline Pressure 3/<. line 1/sta. line 30 (Loop 514, 515, 516, 524, 525, 526, 534, 535, 536, 544 'L545 & 546 9.

RWST Level 4

1 35 (Loop 990, 9S1, 992 & 993) 1

10. Containment Normal Sumps Level (Narrow Ran0e) 2 1

31 (Loop 7777 & 7789) 11.

Containment Water Level (Wide Range) 2 1

31 (toop 0764,& 0765)

i TABLE 3.3-8 (Continued)

ACCIDENT MONITORING INSTRUMENTATION

[

TOTAL MINIM M z

NO. OF CHANNELS

}

INSTRtmENT CHANNELS OPERABLE ACTION

12. Condensate storage Tank Level 2/ tank 1/ tank 31 (Leop 5101, 5111, 5104 & 5116) e-

13. Auxiliary feedater Flow 2/ feed line 1/ feed ilne 31 (Loop 5152, 15152, 5153, 15153, 5151, 15151, 5150 & 15150)

\\

14. Containment Radiation Level (High Range) 2 A

33 l

(Loop 0005 & 0006)

15. Steamline Radiation Monitor 1/sta. line 1/sta. line 33 1

(Loop 13119, 13120, 13121 & 13122) 2:'

16. Core Exit Thermocouples 4/ quad / train 2/ quad / train 30

'2'

17. Reactor Coolant System Subcoollag 2

1 31

18. Neutron Flux (Extended Range) 2 1

31 1

(Loop 13135A & 131358) l 19.

RW.IS 2

1 34 l

20.

Containment Hydro 9en Concentration 2

1 31 l

(Loop 12979 & 12980) 21.

Containment Pressure (Extended Rance) 1 31 (Loop 19942 & 10943) l

22. Containment Isolation Valve Position. Indication

1/ valve 1/ valve 36 I

"Appitcable for containment isolation valve position indication de i t~d as post-accident monitaring instru-e mentation (containment isolation valves Alch receive containment isolation Phase A or containment ventilation j

isolation signals).

\\

TABLE 3.3-8 (Continued)

ACTION STATEMENTS ACTION 30 - a. With the number of 0PERA8LE haone s one less than the Total Number of Channels requir

ntg, store the inoperable channel to OPERA 8LE status within 31

, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. With the number of OPERA 8LE channels two less then the Total Number of Channels requirement, restore at least one inoperable

~

channel to OPERA 8LE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

'c. Wi of OPERABLE channels less than the Minimum chan-nel utrement, restore at least one inoperable chan-nel status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in H07 SHUTDOWN with the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. The provisions of Specification 3.0.4 are not applicable.

ACTION 31 - a. With the number of CPERA3LE channels one less than the Total Number of Channels requirements, restore one inoperable channel to CPERA8LE status within 7 days, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

T

b. With th OPEAA8LE 1s less than the Ninimum Channel le uirene

' store at least one inoperable channel status n 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or be in HOT SHUTDOWN within the nex 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. The provisions of Specification 3.0.4 are not applicable.

4 ACTION 32 -

Wi of OPE annels less than the Minteum Chan-nel le ui store at least one inoperable chan-l nel status 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The provisions of Specification 3.0.4 are not applicabis, i

ACTION 33 -

the neer of OPERA 8LI channels less than th i aus 1s OPERA 8LE requirement, initiate the alter a method of i

ring the parneetar within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and either restore the e

inoperable channel (s) to OPERA 8LE status within 7 days or prepare and submit a. Special Report to the Commission, pursuant to Speci-j fication 6.8.2, within 14 days that provides actions taken, cause of the inoperability, and the plans and schedule for restoring l

the channels to OPERA 6LE status.

The provisions of Specifica-j tion 3.0.4 are not applicable.

[

I V0GTLE UNITS - 1 & 2 3/4 3-61 l

j

TABLE 3.3-8 (Continued)

ACTION STATEMENTS ACTION 34 a With the number of OP c'

s less than the required num-ber of channels or e

a els OPERABLE requirement, restore the inoperabl n

OPERABLE status as per Action 31a or b as as able repair is feasible during plant operation.

If npair is not feasible, prepara and submit a Special Report to the Caunission pursuant to Specification 6.8.2 within 14 days that provides actions taken, cause of the in-operability, and the plans and schedule for restoring the chan-nels to OPERABLE status. The provisions of Specification 3.0.4 are not applicable.*

ACTION 35 - a. With the number of OPERABLE

)1stwolessthantheTotal Number of Channels mquirene poston the inoperable channel to CPERABLE status within 31

, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ,

b. With the number of OPERABLE channels thne less than the Total Number of Channels requineant, restors at least one inoperable channel to OPERABLE status within 7 days, or he in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. With a umQr of OPERABLE channels less than the Minimus Chan-nel Operable fequirement, reston at least one inoperable chan-nel MST.E status with 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT SHUTDOW within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

d. The provisions of Specification 3.0.4 are not applicable.

ACTION 36 -

iTD the number of OPERABLE Aels less than th.31 aus

.ch4hnels OPERABLE requirene caply with the p

~ ions of deification 3.e.3. for an rable containment isolation valre.

"Action Statement 34 applies to the first fuel cycle only.

Action Statement 31.

is applicable thenafter.

V0GTLE UNITS - 1 & 2 3/4 3 62 i

I

TABLE 3.3-9 RADI0 ACTIVE LIQUID EFFLLTNT MONITORING INSTRtBENTATION E

53 MINIMM CHAletELS INSTRtBENT OPERA 8LE ACTICII 1.

R diogctivity Monitors Providing Alarm and 8*

Aut tic Terminetten of Release Liquid Radueste Etfluent Line (RE-0018) 1 37 b.

Steam Generator Blewdown Effluent Line (RE-0021) 1 38 ine Building (Floor Drains) Sunps Effluent Li

-0848) 38 c.

2.

Rad tivity Monitors Providing Alara But Not Providing 2:a Automatic Terminetten of Release y

a.

Nuclear Service Coeling Water System Fffluent Line j

g (RE-0020 A & 2) 1 39 i

3.

Flow Rate Measurement Devices 1

Liquid Rad este Effluent Line (FT-0018) 1 40 a.

b.

Steam-Generator Blowdeun Etfluent Line (FT-0021) 1 40 "1

i c.

Flow to 21 Suep (AFQI-7620, FR-7620, pen 1) 1 40 (bunoq)

~

l 1

e

-,____m--

.m

I l

I TABLE 4.3-5 (Continued) 1 h

RADIDACTIVE LIQUID EFFLUENT MONITORING INSTRUMEN!Allitt SURVElttANCE RE@tREMENTS m

E "4

ANALOG CHANNEL CHANNEL maarF CHAledEL OPERATIONAL INSTRLSENT OICK CIECK CALIBRATIGIl TEST c.

3.

Flow Rate Measurement Devices u

Liquid Raduaste Effluent Line (FT-0018)

D(4)

N.A.

R N.A.

a.

1 b.

Steam Generator Slowdown Effluent Line (FT-0021)

D(4)

N.A.

R N.A.

c.

Elpw Blowdown Susp 0(4)

N.A.

R q

(AFq 7620,FR-7620 pen 1) t:"

i Y

i O

4 G

O

J TABLE 3.3-10 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTmatE'dTAVIGN MINIMUM CHANNELS E

INSTRUMENT OPERABLE APPLICARILITY ACTION 1.

GASEOUSWASTEPROCES51NGSYSTEM a.

Noble Gas Activity Moni r-Providi Alarm Terminat on of u014) 45 b.

Effluent Syst F

\\

g/[M Measuring Dev e AFT-0014 1

46 unc A) 2 2.

GASEOUS WASTE PROCES "L

losiveG]a

{MonitoringSystem s*

a.

Hydrogen Monitor 1/recombiner 50 i

b.

Oxygen Monitor 2/recombiner 49 3.

Condenser Air Ejector and Steam Packig Exhauster System a.

Noble Gas Activity Monitor 1

47 (RE-12839C) b.

Iodine Sampler 1

maa 51 1

(RE-128358) c.

Particulate Supler 1

51 (RE-12839A) d.

riow Rate Monitor 1

46 (FT-12839),

(FIS-12862) e.-

Sampler Flow Rate Monitor 1

m**

46 (FI-13211)

TABLE 3.3-10 (Continue,q),

TABLE NOTATIONS l

At all times.

- During dASE005 WASTE PROCES$1NG SYSTEM operation.

l

- - During radioactive releas this pathway.

Ouring Emergency Filtrati n, ACTION STATEMENTS ACTION 9-44 (Not Used)

ACTI0ti 45 -

With the num6ar of channels OPERABLE less than required by the Minimum Channels OPERA 8LE requiresetii!, the cor. tents of the tank (s) say be released to the environment provided that price to initiating the release:

s.

At least two independant samphs of the tank's contents are analyzed, and b.

At least two technically qualified membe-t of the facility staff independently verify the release rate calculations and discharge valve lineup, g

Otherwise, susptnd release of radioactive effluents via this pathwaf.

ACTION 46 -

With the nu-ter of channels OPERABLE less tun nquired by the Minimus Chawls OPERA 8LE requirement, affluent n1 eases via this pathwn say cantinue provided the flow rata is estimated at least once per 4 houn.

ACTION 47 -

With the nuabar of channels 0FTRA8LE less.than required by the Minimum Channels OPERA 8LE mquirement, affluent releases vias this p thway say continue provided grab sr.nples are talien at least w e per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these seaples are analyzed for radio &ctivity witnin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION 48 -

With the number of channels OPERABLE less taan nquind b*r the Minious Channels OPERA 8LE nquirer ". ' -ediately suspen'd containment PURGING of radioactiv off N 's via this pathway.

ACTION 49 -

a.

With the outlet oxygen moniter e %. inoperable, opera-tion of the system say continue provided grab samples are taken and analyzed at least once per 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months and the oxygen concentration ressins less than 1 percent.

b.

With the inlet oxygen monitor inoperable, operaticn say con-OC:

tinue if inlet hydrogen monitor is OPERA 8LE.

4 c.

With both cxygen channels or both of the inlet. oxygen and inlet hydrogen monitors inoperable, suspend oxygen supply to the recombiner.

Addition of waste gas to the system say continue provided grab samples are takes and analyzed at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations or at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during other operations and the oxygen concentration ressins less than 1 percent.

V0GTLE UNITS - 1 & 2 3/4 3-74

IA8tE 4.3-G RADIDACINE GASEOUS EFFLUENT NONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS E

5 ANALOG l

CHANNEL M00ES FOR WICh CHAMhEL 500RCa.

PWL OPERATIONAL SURVEILLANCE INSTRUMENT CHECK CHECK ZALIBRATIOn TEST 15 REQUIRED _

1.

M

""5 U TE **0CE551NG SYSTEM

u. 0 r Activity Moniter -

' <,it' - Alers and Automatic N

mWm.%n of Release (ARE-0014) P P

R(3)

Q(1) c J-F N

.luent System Flow Rate P

N. A.

A u.a.

c

~ is-ik, Device (AFT-0014) r-7 y

. > V '.3TE PROCESSING SYSTEM Exolosive 1 f44 -.erT4ering Sys h

a. Hydrogen Monitor.

O N.A.

Q(4)

M b

b. Oxygen Monitors 0

N.A.

Q(5)

M b

3.

Condenser Air Ejector and Steam Pxking Exhauster System I ('2)

a. Nobis Gas Activity r witor 0

M R(3)

Q c

(RE-12839C) 1

. >. Iodine Sampler W(6)

N.A.

N.c N.A.

c (PE-128398)

c. Particulate Sampler W(6)

N.A.

M.A.

N.A.

c (RE-12'3%)

d. Flow Rate Monitor D

N.A.

R N..'..

c (FT-12839)

e. -Swler F. low Rate Mon! Lor D

N.A.

R Q

c (FI-1321;)

l

9 TABLE 3.3-11 HIGH-ENERGY LINE BREAK INSTRUMENTATION Isolation Instrument Ninteum Applicable Function Channel k nels operable Modes i

1.

Electric Steam ATE 19722A (RD52) 1 Soller Isolation ATE 19723A (RD52)

(Common Instrumentation)

ATE 197228 (RC41) 1 ATE 197238 (RC41)

ATE 19722C (RC65) 1 ATE 19723C (RC64)

ATI 197220 (RC47) 1 4

ATE 197230 (RC64)

AFT 19722 1

AFT 19723 I

l ATE 19722E (RC95) 1 ATE 197231 (RCSS) l L

Minine Iselation Instraent Instrument Chr.nnels Applicable Function Channel (bnft 1}

Channel (Unit 2) Operable _ Modes i

2.

Staan Generator TE 15212A (R804)

TE 15212A(RS UI) 1 1,2,3,4 Slowdown Lins Tt 15216A (R300)

TE 15216A(RSUI) 1 I'*I*ti'"

TI 152123 (RC104) TI 152128(RC103) 1 1,2,3,4

,~

TE 152168 (RC106) M 152188(RC103) l Ti 15212C (RC107) TI 15212C(RC101) 1 1,2,3,4 TI 15216C (RC107) TI 15n5C(RC101)

TT 1.52120 (RC108) TE 15212D(RC102) 1 1, ', 3, 4 t

i TL 152160 (RC104) TE 152180(RC102) i FT 15212A (Loop 1) FT 15212A(Loop 1) 1 1,2,3,4 l

FT 15n6A FT 15215A L

j FT 182123 (Loop 2) FT 15212S(Loop 2) 1 1,2,3,4 FT 152148 FT 152168 FT 15212C (Loop 3) FT 15212C(Loop 3) 1 1,2,3,4 i

M unE R unE 1

FT 152120 (Leep 4) FT 15212O(Loop 4'. 1 1,2,3,4 FT 15n50 FT 151160 t

3.

Letdown Line TE 15214A (A07)

TE 15214A(A100) 1 1,2,3,4 i

l Isolation TI 152UA (A07)

TI 15n5A(A100)

TI 152148 (A04)

TE 152184(A101) 1 1,, 2, 3, 4 4

TI 152158 (A04)

TI 152158(A101)

TE 15214C (A09)

TE 15214C(A103) 1 1,2,3.4 TE 15215C (A09)

TE 1522.5C(A103)

)

~

Req aired during all iCOES when electric steam boiler is 'a operation.

4

)

V0GTLE UNITS - 1 & 2 34 3-80

REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPCRATION 3.4.2.3 At least two of the hops / trains listed below shall be OPERABLE and at least one of these loops / trains shall be in operation:"

Reactor Coolant Loop 1 and its associated steam generator and a.

reactor coolant pump,**

b.

Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,**

c.

Reactor Coolant Loop 3 and its associated stew generator and reactor coolant pus.p,**

J.

Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump,**

e.

RHR train A, and f.

RHR train 8.

APPLICABILITY:

MODE 4.

ACTION:

a.

With less than the above required loops / trains OPERA 8LE, inmediately initiate corrective action to return the required loops / trains to.

OPERA 8LE status as soon as possible; if the remaining OPERA 8LE loop / train is an RPR train, be in COL.D SHUTDOWN within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

l b.

'dith no loop / train in operation, suspend all operations involving a i

reduction in baron concentration of the Reactor t.olent System and f amediately initiate corrective action to return the required loop / train to operation.

"All reactor coolant pumps and RHR ?uses say be deenergized for up to 1 nour provided:

(1) no opsrations are perst tted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation taepeettttre.

- A reactor coolant pump shall not be started un1 ss qe bcondary water temperature of each stead gener.ator is less th n 5(F a)ove each of the t

l Reactor Cculant Systes cold leg temperatures.

l V0GTLE UNITS - 1 & 2 3/4 4-3 i

o REACTOR COOLANT SYSTEM COLD SHUT 00WN - LOOPS FILLED LIMITING CON 0! TION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) train shall be OPERA 8LE and in operaticn*, and either:

a.

One additional RHR train shall be OPERABLE **, or b.

The secondary side water level of at least two steam generators shall be greater than 17% of wide range (LI-0501, LI-0502, LI-0503, LI 0504).

APPLICA8ILITY: MODE 5 with reactor coolant loops filled ***.

ACf!0N:

4.

With one of the RHR trains inoperable or with less than the required steam generator water level, immediately initiate corrective action to siturn the inoperable RHR train to OPERA 8LE status or restere the t

required steam generator water level as soon as possible.

b.

With no RHR train in operation, suspend all operations involving a reduction in baron concentration of the Reactor Coolant System and immediately initiate ccrrective action to return the required RHR train to operation.

I SURVEILLANCE REQUIRENENTS j

4.4.1.4.1.1 The secondary side water level of at least two steam generators when required sna11 be determined to be within limits at least once per t

I 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l 4.4.1.4.1.2 At least one RHR train shall be determined to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i l

"The RHR pump may be deenergized for up to I hour provided:

(1) no operations are persitted thet would cause dilution of the Rea: tor Coolant System boror; concentratica, and (2) core outlet temperature is natntained at laast 10*F below saturation temperature.

- 0ne RHR train say be inoperable for up to 2 he.urs for surveillance testing l

provided tha other RhR train is OPERA 8LE and in operation.

""*A reactor coolant pump shall not be started unigss 4he s condary water temperature of each steam generator is less th n 5q,/F apovo each of the Reactor Coolant System cold leg temperatures.

1 I

I V0GTLE UNITS - 1 & 2 3/4 4-5 i

i

REACTOR COOLANT SYSTEM CCl.0 SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.2 Two residual heat removal (RHR) trains shall be.0PERABLE" and at least one *HR train shall be in operation.** Reactor Makeup water Storage Tant (RMWST) discharge valves (1208-LM-175,1208-U4-176,1208-U4-177 and 1208-U4-183) shall be closed and secured in position.

APPLICABILITY:

MODE 5 with reactor coolant loops not filled.

ACTION:

With less than the above required RHR trains OPERA 8LE, innedtately a.

initiate corrective action to return the required RHR train. to OPERABLE status as soon as possible.

b.

With no RHR train in operation, suspend all operations involving a reduction in boron concentrati n of the Reactor Coolant System and f amediately initiate corrective action to return the required RHR train to operation.

With the Reactor Makaue Water Storage Tank (RNWST) discharge valves c.

(1208-04-175, 1208-04-176, 1204-U4-177, and 1208-U4-183) not closed and secured in position, immediately close and secure in position the RMWST discharge valves.

SURVEILLANCE REQUIREMENTS 4.4.1.4.2.1 M circula1,1 4 reac(toleast ene RHR train shall be detetuined to be in operation and coolant at least once per 12 nours.

4.4.1.4.2 2 Valv 1204-04-175, 1208-04-176, 1204-04-177, and 1208-U4-183 shall be verifi

'cTosed and secured in position by mechanical stops at least once per 31 dr/s.

"One RHR train may be inoperable for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance, ssting provided the other RHR train is OPERA 8LE and in operation.

- The RHR puso say be deenergized for up to I hour provided:

(1) no operations are permitted that would cause dilution of the Reactor Ceoltnt System boron concentration, and (2) core outlet tamperature is saintained at least 10*F below saturation temperature.

VDGTLE UN7f5 1 & 2 3/4 4 6 i

REACTOR CtCLANT SYSTEM 3/4.4.4 RLt!EF VALVES LIMITING CON 0! TION FOR, OPERATION P.

3.4.4 All power-operated ' relief valves (PORVs) and their associated block valves shall be OPERABLE.

APPLICABI.8.!TY:

M00ES 1, 2, and 38 ACTION:

a.

kita one or more PORV(s) inoperable, because of excessive seat leakage, within I hour either restore the PORV(s) to CPERA8LE status or close the associated block valve (s); othe'rwise, he in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN wipin the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b.

With one or more PORV(s) inoperable due to causes other than exces-sive seat leakage, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either restore the PORV(s) to CPERABLE status or close the associated block valve (s) and remove power from the block valve, and 1.

With only one PORY OPERA 8LE, restore at least a total of two PORVs to OPERA 8LE status within the following 72 tours or be in HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDCVN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or 2.

With no PORVs OPERA 8LE, restors at least one PORV to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in NOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and H0T SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c.

With one or more block valve (s) inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (1) restore the block valve (s) to 0PERA8LE status or t1ose the biock valve (s) and remove power from the block valve (s) or close the PORY and remove power free its associated solenoid valve; aod (2) apply ACTION b above, as appropriate, for the isolated PORV(s),

d.

The provisions of Specification 3.0.4 are not applicable.

SURVEft.LAMCE REQUIREMENTS 4.4.4 1 Ea ch' RV shall be demonstrated CPERA8LE at least once per 18 months by:

a.

Operating the va0 e through one complete cycle of fu'

svel, and b.

Performing a CHANNEL CALIBRATION.

"The provisions of tais specification are not applicable to Unit 2 until initial entry of Unit 2 into M00E 2.

V0GTLE UNITS - 1 & 2 3/4 4-10 i

REACTOS COOLANT SYSTEM STEAM GENERATOR SURVEILLANCE REQUIREMENTS (Continued) 9)

Preyervice inspection seans an inspection of the full length of each tube in each steam generator performed by eddy current techniques prior to service to establish a baseline condition of the tubing.

This inspection shall be performed prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent inservice inspections, b.

The steam generator snall be determined OPERAaLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table 4.4-2.

4.4.5.5 Repo rt7 a.

Within 15 days following the cocolation of each inservice inspection of steam generator tubes, the number of tubes plugged in each steam generator shall be reported to the Consission in a Special Report pursuant to Specification 6.8.2; b.

The complete results of the steam generator tube inservi.ce inspection shall be submitted to the Commission in a Special Repe-t pursuant to Specification 6.3.2 within 12 manths following the cozpletion of the inspection.

This Special Report shall include:

1)

Number and extent of tubes inspicted, 2)

Location and percent of well-thickness penetration for each indication of an imperfection, and 3)

Identification of tubes plugged.

Resulks of steam generatar tube inspections which fall into Category c.

1 be reported in a special report to the Cosmission pursuant spec fication 6.8.2 within 30 days and prior te resumption of plant ration.

This report sliall provide a description of investigations co '

ed to determine cause of the tube degradation and corrective sensures taken to prevent recurrence.

V0GTLE UNITS - 1 & 2 3/4 4-16 i

REACTOR COOLANT SYSTEM OPERATIONALLEAXAGJ SURVEILLANCE REQUIREMENTS 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within each of the above limits by:

a.

Monitoring the containment atmosp'iere (gasecus or particulate) radioactivity monitor at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; b.

Monitoring the containment no mal sumps and reactor cavity sump inventory at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; c.

Measurcaent of the CONTROLLED LEAXAGE to the reactor coolant pump seals when the Reactor Coolant System pressure is 223$ 2 20 psig at least once per 31 days.

The provisions of Specification 4.0.4 are not applicable for entry into PCOE 3 or 4; d.

Performance of a Rene colant System water inventory balance at g

least once.ner 72 Nur and iOSCN AG.

Monitoring the Reactor Htad Flange Leakoff Syites at least once per e.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

4.4.6.2.2 Each Reactor Coolant Systes Pressure Isolation Valve specified in fable 3.4-1 shall be demonstrated OPERA 8LE by verifying leakage to be within its limit:

a.

At least once during each refueling avuge (leak te;, ting should be perfonned after all disturbances to the valves are completa, such as before reachir.g powe.' operation during a refueling outege);

b.

Prior to returning the valve to service following maintenance, repair, or replacement work on the valva (leak testing should be perfonped after all disturbances to the valves are complete);

c.

For systems rated at less than 50% of RCS design pressure, each time the valve has moved free its fully closed position except for valves MV-8701 A/B and HV-8702 A/t.

d.

Prior to entering M00E 2 whenever the plant has been in COLD SHUTDOWN for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or more and if leakage testing.ias not been perforsed in the previous 9 months except for valves HV-8701 A!8 and HV-8702 A/8.

e.

As outlined in the ASME Code,Section XI, caragra;

.W-3427(b).

The provisions of Specification 4.0.4 are not applicable for entry into MODE 3 or 4.

V0GTLE UNITS - 1 & 2 3/4 4-21 i

INSERT AG T

p ovis ns of S'ecification 4.0.4 are not applicable for entry into J

e 9

TA8tf 3.4-1 REACTCR COOLANT SYSTEM PRESSURE ISOLATION VALVES MAXIHUH ALLOWA8LE VALVE NUMBER VALVE SIZE (in.)

FUNCTION l.EAKAGE(com) 1.

HV-8701A 12 RHR Suction (gate valve) 5.0 2.

HV-87013 12 RHR Suction (gate valve) 5.0 3.

HV-8702A 12 RHR Suction (gate valve) 5.0 4.

HV-87028 12 RHR Suction (gate valve) 5.0 5.

1204-04-120 2

SI-Hot Leg 2nd Isolation Valve 1.0 6.

1204-04-121 2

SI-Hot Leg 2nd Isolation Valve

1. 0 7.

1204-U4 122 2

51-Hot Leg 2nd Isolation Valve

1. 0 8.

1204-04-123 2

SI-Hot Leg 2nd Isolation Valve

1. 0 9.

1204-U6-079 10 Accumulator 2nd Isolation Valve 5.0

10. 1204-06-080 10 Accumulator 2nd Isolation Valve 5.0

11. 1204-U6-081 10 Accumulator 2nd Isolation Valve 5.0

12. 1204-06-082 10 Accumulator 2nd Isolation Valve 5.0

13. 1204-U6-063 10 Injection Line 1st Isolation Valve 5.0

14. 1204-U6-084 10 Injection Line 1st Isolation Valve 5.0

15. 1204-06-085 10 Injection Line 1st Isolation Valve 5.0

16. 1204-06-086 10 Injection Line 1st Isolation Valve 5.0

17. 1204-U6-124 6

51-Hot Leg ist Isolation Valve

3.,0 1A. 1204-U6-125 6

SI-Hot Leg ist Isolation Valve 3.0 19.1204 U6-126 6

SI-Hot Leg ist Isolation Valve 3.0 f

20. 1204-06-127 6

SI-Hot Leg ist Isolation Valve 3.0 l

21. 1204-06-128 8

RHR-Hot leg 2nd Isolation Valve 4.0 22, 1204-06-129 8

RHR-Hot Leg 2nd Isolation Valve 4.0

23. 1204-04-143 2

SI-Cold Leg 2nd Isolation V:1ve

1. 0

24. 1204 04-1A4 2

SI-Cold i.eg 2nd Isolation Valve

1. 0 25, 1204-04-145 2

SI-Cold Leg 2nd Isolation Valva

1. 0 26, 1204-04-144 2

SI-Cold Leg 2nd Isolation Valve

1. 0 27, 1204-06-147 6

RHR Cold Leg 2nd Isolation Valve 3.0

28. 1204-06-148 6

RHR Cold Leg 2nd Isolatisn Valve 3.0

29. 1204-v6-149 6

RHR Cold Leg 2nd Isolation Valve 3.0

30. 1204-U6-150 6

RHR Cold Leg 2nd Isolation Valve 3.0 V0GTLE UNITS - 1 & 2 3/4 4-22 m

250 k

j O

.S b

Ea D

200 2

UNACCEPTABLE.

OPERATION W

\\

5 S

A 150 h

b a

8u 100 m

a.

5 2

9 ACCEPTABLE OPERATION Y

50 2

8w W

\\

O

=

20 30 40 50 60 7C 80 90 100 PERCENT OF RATED THERMAL POWER l

l l

FIGURE 3.4-1 l

DOSE E0VIVALENT I-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS l

PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY >,1 uCl/ gram DOSE EQUIVALENT I-131 V0GTLE UNITS - 1 & 2 3/4 4-26

e REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE tIMITS l0 St4@ttst0 TO REfttti ann dj REACTOR COOLANT SYSTEM LIMITING CON 0! TION FOR OPERATION 3.4.9.1 The Reactor Coolant System (ucept the pressurizer) taeperature and pressure shall be limited in accortlance with the likit linas shown on W.igunt gv-2 =d 3.

durin heatup, cooldown, criticality, and intsrvice leak and mser4-M. hydrostatic tes ing wgt3:

A saximum hectup of 100*F in any 1-hour period, a.

b.

A maxismo cooldown of 100*F in any 1"hour period, and A maximum temperature chsnge of less than or equal to 10*F in any c.

1-hour period during 2nservice hydrostatic and leak tasting operations above the heatup and cooldown limit curves.

APPLICA8ILITY: At all times.

ACTION:

With any of the above limits axceeded, restore the tasperature and/cr pressure to within the limit within 30 sinutes; perform an engineering evaluation to determine the effects of the out of-limit condition on '.he structural integrity of the Reactor Coolant Systas; datermine that the Reactor Coolant System remains acceptable for c;ntinued operation or be in at Ivest HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the ACS T,yg and pressure to lass than 200*F and 500 psig, respectively, within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILL/_ ACE REQtIREMENTS

=

4.4.9.1 The Reactor Coolant Systas tasp dature and p ssura shall be determined to be within the limits at least once per 30 minutes during systaa heetup, cooldown, and inservice leak and hydrostatic testinc operations.

RE 3.4 V0GTLE UNITS - 1 li 2 3/4 4-30 i

INSkRT AH Figures 3.4-2a (Unit 1) and 3.4-3a (Unit 1), Figures 3.4-2b (Unit 2) and 3.4-3b (Unit 2) i r

i 1

i l

l 2

l I

I

]

i l

l l

j

3000 CURVE APPLICABLE FOR THE SERVICE.

PERIOD UP TO 16 EFPY RTNDT After 16 EFPY 0

LEAK TCST a.1/4 T = 110 F LIMIT 0

87 F w

b. 3/4T =

e CRITICALITY

}

LIMIT G

0 FOR 60 F/hr ----

g 2000 UNACCEPTABLE r

HEATUP OPERATION 2w J

CRITICALITY h

/ A LIMIT 0

FOR 100 F/hr 3

HEATUP

@u 8

h 0

60 "/hr H^

BASED ON INSERVICE -

y 1000 CURVE /

O r

HYDROSTATIC TEST 0

TEMPERATURE (255 F) p C

FOR THE SERVICE d

PERIOD UP TO 16 EFPY 5

2 0

~

100 F/hr HEATUP ACCEPTABLE J

CURVE OPERATION i

l l

I 0.0 0.0 100 200 300 400 500 TEMPER ATURE (OF)

LOYtEST INDICATED RCS TCOLD MATERIAL BASIS Ceesee Coateat An w - 10 wt %

I ActwW f4 Wt %

j RTggy in,t.c A. w. af f i Astw - 30' f l e 1908F afggy Ahee 16 ( F PY 9 t/47

~

FIGURE 3.'4-2 a UNIT 1 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY V0GTLE UNIT'; - 1 & 2 3/4 4-31

e e

2500 l

CURVE APPLICABLE FOR THE SE RVICE PERIOD UP TO it EFFY LEAK TEST f

LIMIT RT

'* II IE NOT e.1/4 7 = 123*P

a. 3/4 Y = 97*P 2000 CRITICALITY LIMIT FOR g

D UNACCEPTABLE WF/hr HE ATUP OPE R ATION e

3 g 1500 60*F/hr HEATUP CURVE CRITICALITY LIMIT FOR M

100*F/hr HEATUP

>=

AMEPTABLE OPERATION

. 000 h

1000F/hr HEATUP CURVE %

O<w EASEU ON INSERVICE HYDROSTATIC TEST TEMPERATURE 1264*F) FOR w

T;lE SERVICE FERIOD UP TO IS EFPY g

MA.f RI AL 5AS35

-.n10 m %

Aammed..

Cooper Coment:

g3,,,,, d. N A *F Asume RT Indaad:

NOT o

,gn pg RT Ah it Em 9 tief = 123*F NOT 9 3/4T = 97'F

+

I 0.00.0 100 200 300 400 500 l

k.OWEST INDICATED RCS T TEMPERATURE (OF)

OOLD l

f i

(

FIGURE 3.4-2b i

UNIT 2 REACTOR COOLANT SYSTEM HEATUP LIMITATIONS - APPLICABLE UP TO 16 EFPY

(

V0dTLE UNITS - 1 & 2 3/4 4-31a

3000 i

i CURVE APPLICABLE FOR THE SERVICE PERIOD UP TO 16 EFPY__.

l 1

W PRESSURE. TEMPERATURE LIMITS 2000 E

3 UNACCEPTABLE h

OPERATION ACCEPTABLE M

OPERATION 3

8o m

2

/

MATERIAL BASIS 1000 o

Copper Content: Assumed

.10 Wt %

(Actual

.06 Wt %)

l h

I 0

RTNDT nitial:

Ar.umed - 40 F O p" (Actual

- 300 Q

- 20 F) 77 COOLDOWN 40 f

RATE 60 /

0 110 F.

100 RTNDT After 14 EFPY 0 1/4T

=

(CF/ht) 0 87 F

3/4T

=

0.0

~

0.0 100 200 300 400 500 TEMPERATURE (OF)

LOWEST INDICATED ".CS TCOLD FIGURE 3.4-3a UNIT 1 REACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 FFPY j

V0GTLE UNITS - 1 & 2 3/4 4-32

2500 i

CURVk APP'.lCABLE FOR THE SERVICO PERIOD UP TO 18 EPPY uJ z

2000 4.

Et; g

1s00 UNACCEPTABLE OP E R ATION i

O C

ACCEPTA3LE OPE R ATION U

COOLDOWN RATE PPhel bi

%2 MATERIAL SASIS 0

20 -

Copose Centesit:

Aamumed.10 m %

40 -

( Actwel. 05 m %

$4

/

RT ist6el:

Assumee. nae 7 100 NOT (Aetual 508Pl RT After 18 EPPY # 1/47

123eP NOT 0.0 O0 100 200 303 400 500 TEMPERATURE (OF)

LOWEST IN0lCATED RCS TCOLD f

4 i

e FIGURE 3.4-3b UNIT 2 RF. ACTOR COOLANT SYSTEM C00LDOWN LIMITATIONS - APPLICABLE UP TO 16 EFPY V0GTLE UNITS - 1 & 2 3/4 4-32a

REACTOR C00 TANT SYSTEM COLD ovFRPRESSURE PROTECTION SYSTEMS I

LIMITING CONDITION FOR OPERATION 3.4.9.3 At laast one of the following Cold Overpressurs Protectiot. Systems shall tm 0PERA8LE:

Two power operated relief valves (PORVs) 4th lift settir.gs which a.

vary with RCS temperature and which do not exceid the limits estab-g

-q 11shed inghur; :.4-4, er g

{M b.

Two residual heat removal (RHR) suction nlief valves each with a setpoint of 450 psig

3%, or c.

The Reactor Coolant Systes (RCS) pr,e s rized with an RC3 vaat capable of relieving at least 67 g water flow st 470 psig.

N MODES 4, 5, and 6 with the re.M APPt !CA81LITY:

actor vessel head on.

ACTI0ft-With one PORV and one RHR suction relief valva ineptr21e, either a.

restore two PORVs or two RHA suction relief valves to CPERA8L.E status within 7 days or depressurixe and vent the RCS as specified in b

QCMCO 3.4.9.3.c, above, within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

b.

With both PORVs and both RHR suction relief valves inoper.able, depressurize and vent the RCS as specified in 3.4.9.3.c, above, x

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

Siccdicar\\

In the event either the PORVs, th9 RHR suction n11ef valvet, or tr.e c.

RCS vent (s) are used to mitigate an RCS pnssure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 8.8.2 within 30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs. the RHR suction relief valves er RC3 vent (s) on the transient, and any cornctivo action necessary to prevent recurrence.

d.

The provisions of Specification 3.0.4 are not applicable.

t V0GTI.E UNITS - 1 & 2 3/4 4*34 i

INSERT Al Figure 3.4-44 (U9it 1). Figure 3.4-4b (Unit 2), or O

e f

C 5

t

)

l l

l i

l I

l l

I I

0 300

~

-f

-[

g z

{

8 700 s

/

. _ ~

~

8 I

I u

.5 E

-x y

l 2 00 ~1

/

i 2

,/

z x(

y

~

/- ' ' '

500 50 100 150 200 250 300 350 l

TRTD AUCTIONEERED LOW MEASURED RCS TEMPERATURE (*F) i I

I FIGURE 3.4-4a UNIT 1 MAXIMUM ALLOWABLE _ NOMINAL PORY SETPOINT FOR THE COLD OVERPRESSURE PROTECTION SYSTEM V0GTLE UNITS - 1 & 2 3/4 4-35

800

/

i 7%

\\y 8

a>

$5 **

3<

=

3 E

500 450 50 100 150 200 250 300 350 T RTD - AUCTIONEERED LOW MEASURED RCS TEMPERATURE (*F)

FIGURE 3.4-4b UNIT 2 MAXIMUM ALLOWABLE NOMINAL PORV SETPOINT FOR THE COLD OVERPRESSURE PROTECTION SYSTEM V0GTLE UNITS - 1 & 2 3/4 4-35a

3/4.*

DERGENCY CORE COOLING SYSTEMS 3'4.5.1 ACCUmjLATORS LIMITING CON 01 TION FOR OPERATION 3.5.1 Each Reactor Coolant System (RCS) accumulator shall be OPERA 8LE with:

a.

The isolation valve open, b.

A contained b ed water volume keen 6616 (36% of instrument span) and LI-0952, LI-0953 LI-0954 LI-0955, LI-Osso, u)-0957),64% of instr g g g

+ c.

A boron concentration of between 1900 and hM-pend y-

[h7.

d.

A nitrogen cover-pressure'of between 617 and 678 psig.

(PI-0960A&B, PI-0941A&B, PI-0962A&8[ -0963A&8, PI-0964A&8, PI-0965A&8, PI-0964A&B, PI-0947 APPLICAstt.!TY:

MODES 1, 2, and 3*,

9 ACTION:

a.

With one accumulator inoperable, except as a Msult of a closed

_}

isolation valve, restore the inoperable acetsulator to OPERA 8LE

^

status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STAN08Y within the next i

6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , b.

With one accumulator inoperable due to the isolation valve being closed, either tunediately open the isolation valve or be in at least HOT STANCRY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce pressurizer pressure to less than 1000 psig within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIRD8ENTS 4.5.1.1 Each accumulator shall be demonstrated CPERA8LE:

i a.

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by:

1)

Verifying the contained borated water voltmo and nitrogen i

cover-pressure in the tanks, and i

2)

Verifying that each accumulator isolation valve is ope g. "

~

l (HV-8804A, B, C, f.

"Pressurizer pressure above 1000 psig.

V0GTLE UNITS - 1 & 2 3/4 5-1 l

INSERT AJ c.

A boron concentration of between:

Unit 1 - 1900 ppm and 2600 ppm Unit 2 - 1900 ppm and 2100 ppm, and r

e e

EMERGENCY CORE COOLING SYSTEMS 9'

SURVEILLANCE REQUIRENENTS 4.5.2 Each ECCS subsystem shall be demonstrated OPERA 8LI:

At least on'ce per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the following valves a.

are in the indicated positions with power lockout switches in the lockout position:

Valve Number Velve' Function Valve Position W 8835 51 Pump Cold Leg. Inj.

OPEN W-8840 RHR Pump Hot Leg. Inj.

CLOSED W-8813 51 Ptap Mini. Flow Isol.

OPEN W-8806 SI Pump suction from RWST QPEN W 8802A, B SIPimpHotLegInj.

CL0 W 8809A, R RHR Pump Cold Leg Inj.

OP F

b.

At least once per 31 days by:

1)

Verifying that the ECC5 piping is full of water by venting the ECCS pump casings and accessible discharge piping high points, and 2)

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

9 By a visual inspection which verifies that no loose debris (rags, c.

'~ /

trash, clothing, etc.) is present in the containment which could be I

transported to the Containment Emergency Sump and cause restriction of the pimp suctions during LOCA conditions.

This visual inspection shall be performed:

1)

For all accessible areas of the containnut prior to establish-ing CONTAIMENT IXTIGAITY, and 2) of the areas affected within containment at the completion of j

aach containment entry when CONTAINMENT INTEGRITY is established.

i d.

At least once per la months by:

i 1)

Verifying automatic isolation and interlock action of the RHR system from the Reactor Coolant System by ensuring that:

I a)

With a simulated or actual Reactor Coolant System pressure

{

signal gnater than er equal to 377 psig the interlocks prevent the valves from being opened, and b)

With a simulated or actual Reactor Coolant Syntas pressure l

signal less than or equal to 750 psig the interlocks will cause the valves to automatically close.

2)

A visual inspection of the containment Emergency Sump s.nd verify-ing that the subsystem suction inlets are not restricted by debris and that t.ie sump cosconents (trash escks, screens, etc.) show no,

evidence of structural distress or abnormal corrosion.

o, W VCLIVd. Mcty Ec rec.li ned in MCD6 3 Er -fesbnJ puthnt-C t

.YDGTLE UNITS Y2

'4b' i

3/4 5 a o

i

_ ~ _ _. _ _ _ _ _ _ _ _ _ _ _ _

BORON INJECTION SYSTEM 3/4.5.4 REFUELING WATER STORAGE TANK LIMITING CON 0! TION FOR OPERATION

'3.5.4 The refueling water storage tank (RWST;i. hall be OPERABLE with:

a.

A sintaus contained borated water volume of 631,478 gallons (86% of instrument span) (LI-0990A&B, LI-0991A&8, LI-0992A, LI-0993A).

$ b.

hree cerentretien ef ht= 2000 pp= mad M00 ~;= ef bere, 7

{RSTM b K.

A m' inimum solution tamperature of 54*F, and c.

d.

A saximum soletion temperature of 116*F (TI-10982).

RWST Sludge Mixing Pump Isolation valves capable of closing on RWST e.

low-level.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

a.

With the RWST inoperable except for the Sludge Mixing Pump Isolation Valves, restore the tank to OPERA 8LE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in at least HOT STANOBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHtITDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b.

With a Sludge Mixing Pumo !sclation Valve (s) inoperable, restore the valve (s) to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or isolata the sludge sixing systes by either closing the manual isolation valves or deenergizing the OPERA 8LE solenoid pilot valve within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and maintain closed..

SURVE!LLANCE REQUIREMENTS 4.5.4 The RWST shall tk demonstrated CPERA8LE:

a.

At least once per 7 days by:

1)

Verifying the contained borated water volume in the tank, and 2)

Verifying the boron concentration of the water.

b.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the RWST temperature when the wtsHe air temperature is its: than 50'F.

c.

At 14ast onsa per 18 months by verifying that the sludge mixing pump isolation valves automatically close upon'an RWST low-level tast signal.

V0GTLE UNITS - 1 & 2 3/4 5-10 i

i

.i INSERT AK t

b.

A boron concentration of between:

Unit 1 - 2400 ppe and 2600 ppe Unit 2 - 2000 ppm and 2100 ppm l

t L

i t

i i

l i

(

[

(

t 9

i I

l, l

i I

f s

CONTAINNENT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) b.

If any periodic Type A test fails to meet 0.75 L, the test schedule for subsequent Type A tests shall be reviewed and approved by the Commission.

If two consecutive Type A tests fail to meet 0.75 L,,

a Type A test shall be performed at least every 18 months until two consecutive, Type A tests meet 0.75 L, at which time the above test schedule may be resumed; c.

The accuracy of each* Type A test shall be verifled by a supplemental test which:

1)

Confirms the accuracy of the test by verifying that th's absolute value of the supplemental test nsult, L,, minus the 3,

sum of the Type A and the superimposed leak L,, is equal to or less than 0.25 L,;

2)

Has a duration sufficient to establish accurately the change in leakage rata between the Type A test and the supplemental tast; and 3)

Requires that the rata at which gas is injected into the centain-ment or bled free the containu.it during the supplemental test is between 0.75 L, and 1.25 L,.

j d.

Type B and C tests shall be conducted with gas at a pressure not less than P,, 45 psig, at intervals no greater than 24 months except for tests involving:

1)

Air lock.:

2)

Purge ly and axhaust isolation valves with nsilient satarial seals.,,

Air locks shall be tested and demonstrated OPERA 8LE by the require-e.

ments of Specification 4.6.1.3; f,

Purge supply and axhaust isolation valves with nsilient material seals shall be tasted aM demonstrated OPERA 8LE by the nquirements of Specification 4.4.1.7.2; g.

The provisions of Specification 4.0.2 are not applicable.

'a V0GTLE UNITS a 1&2 3/4 6-3

0

\\

CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATION 3.6.1.5 Primary containment average air temperature (TE-2563. TE-2612 TE-2613) shall not exceed 120'F.

APPLICABILITY: M00E5 1, 2, 3, and 4.

ACTION:

With the containment average air temperature greatar than 120'F, reduce the average air temperature to within the limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.5 T)'e primary containment average air temperature shall be the arith-setical average of the temperatures at the following locations and shall be detaruined at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

Location Tao Numbers

l Level 2 TE-2543 b.

Level B TE-2613 c.

Level C TE-2612 "Or local sample at corresponding location V0GTLE UNITS - 1 & 2 3/4 6-7 i

CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGRITY LIMITING CONDITION FOR OPERATION

3. 6.1. 6 The structural integrity of the containment shall be saintained at a level consistent with the acceptance criteria in specification 4.6.1.6.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

(

a.

With more than one Unit 1 tendon with an cbserved lift-off force between the predicted lower limit and 90% of the predicted lower

limit or with one tendon below 90% of the predicted lower limit, nstore the tendon (s) to the required level of integrity within 15 days and perform an engineering evalua ton of the Unit I contain-dont and provide a special Repo:'t to the ission within 30 days in accordance with Specification 6.8.2 o in at least H0T STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD 5 ithin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

In addition, within 90 days of completion of the Unit 1 evaluation, perfore an engineering evaluation of Unit 2 containment and provide a Special R rt to the Cossaission in accordance with Specification 6.8.2 or bgFt6 at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD w

e following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , o

b.

With abnormal degrada on o 1 structural integrity as defined in 4. 6.1. 6.1.1 items b or e, naton the containment to the required level of integrity within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and perfors an engineer-ing evaluation of Unit 1 containment and provide a special Report to ge Commission within 15 days in accordance with Specification 6.8.2 gig 0F65"'ilf'at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOCWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

In addition, within 90 days of cospletion of the Unit 1 evaluation, perfore an engineering evaluation of the Unit 2 containment and provide a special Re rt to the Commission in accordance with Specification 6.8.2 o n at least HOT STAND 8Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD $

within the fo lowing 30 grsg yng},L_

fdegradationofthestructuralintegrityotherthan c.

Vi abne

-kat defined M ACT MLa 'nd b at a level below the acceptance criter14 of Specifica-tion.

1.6, nstore the applicable containment to the nquired level of integrity within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and perfore an engineering evalua-tion of the applicable containment and provide a Special Report to LCL h ~.

camission within 15 days in accordance with Specification 6.8.2 f(gliCde UJhh

' trit least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD or within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIRENENTS 4.6.1.6.1.1 Containment Tendons (Unit 1).

The structural integrity of the i

Unit 1 contafnment tencons snall be oemonstrated at the end of 1, 3, and 5 years following the initial containment vessel structural integrity test a-d at 5 year intervals thereafter.

The structural integrity of the tendons shall be demon-strated by:

4.

Detersining that a random but representative sample of at least 13 tendons (4 inverted U and 9 hoep) each have an observed lift-off force within predicted limits for each.

For each subsequent V0GTLE U!4ITS - 1 & 2 3/4 6-8 i

CONTAINMENT SYSTENS CONTAINMENT VENTILATION SYSTEM LIMITING CONDITION FOR OPERATION

3. 6.1. 7 Each containment purge surply and exhaust isolation valve (W-2626A&B, W-2627A&B, W-2628A&B, W-2629A&B) shall be OPERABLE and:

~

a.

24-inch containment purge supply and exhaust isola tion valve shall be closed and sealed closed, and b.

The 14-inch containment purge supply and phaust isolation valve (s) shall be closed to the maximuu extent practicable but may be open for purge systes operation for pressure control, for ALARA and respirable air quality considerations for personnel entry and for surveillance tests that require the valve (s) to be open.

APPLICA8tLITY:

H00ES 1, 2, 3, and 4.

ACTION:

'a.

With a 24-inch containment purge supply and/or exhaust isolation valve open or not sealed cl;, sed, close and seal that valve or isolata the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least lMT STM08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SWTDCW within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

m.'

8pec.hccdico b.

With the 14-inch containrent purge supply or exhaust isolation

' valve (s) open for reasons other than given i 3.6.1.7b above, close the open 14-inch valve (s) or isolate the penetration (s) within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , otherwise be in at least HOT STAN06Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHLITDOW within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

With a containment purge supply and/or exhaust isolation valve (s) c.

having a measurcd leakage rate in excess if the limits of Specif t-cation 4.6.1.7.2, restore the inoperable valve (s) to OPERA 8LE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , otherwise be in at least HOT STAN08Y within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , and in COLD SHIJTDOW within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.6.1.7.1 Each 24-inch containment purge supply and exhaust isolation valve (W-2626A, W-2627A, W-2628A, W-262SA) shall be verified to be sealed closed at least once per 31 days.

k e

V0GTLE UNITS - 1 & 2 3/4 6-11 i

CONTAINMENT SYSTEMS A

..- )

'd SURVEILLANCE REQUIREMENTS (Continued)

4. 6.1. 7. 2 At least once per 3 months the containment ptge valves with resilient material seals in each sealed closed co en purge supply an exhaust penetration shall be demonstrated OPERA 8 by vert ying that the measured penetration leakage rate is less than 0. 6 g wh pressurized o 4.6.1.7.3 Each 14-inch containment purge supply st isolation va v (W-26268, W-26278, W-26288, W 26298) shall be ver ied to be closed or open in accordance with Specification 3.6.1.7b at least once per 31 days.

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V V0GTLE UNITS - 1 & 2 3/4 6-12 i

.n 3 4.7 PLANT SYSTL%

3/4.7.1 TUR81NE CYCLE SAFETY VALVES LIMITING CONDITION FOR OPERATION 3.7.1.1 All sain steam line Code safety valves associated with e'ach steam generator shall be OPERA 8LE with Itft settings as specified in Table 3.7-2.

APPLICA8!t.!TY: MODES 1, 2, and 3.

ACTION:

a.

With four mactor coolant loops and associated steam generators in operation and with one or more sain steaa line Code safety valves inoperable, operation in MODES 1, 2, and 3 may proceed, provided that within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , either the inoperable valve is restored to 0PERA8LE status or the Power Range Neutron Flux iligh Trip Setpoint (NI-00415&C, MI-00428&C, NI-00438&C, NI-00448&C) is n duced per Table 3.7-1; othenvise, be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

The provisions of Specification.1.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.7.1.1 No additional requirements other than those required by Specif cation 4.0.5.

V0GTLE UNITS - 1 & 2 3/4 7-1 i

TA8LE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP LIFT SETTING VALVE NUMBER

(* 10" QRIFICE SIZE SG-1

$G-2 sc 3 SG-4 l.

PSV 3001 PSV 3011 PSV 3021 PSV 3031 1185 psig 16.0 nl 2.

PSV 3002 PSV 3012 PSV 3022 PSV 3032 1200 psig 16.0 n!

3.

PSV 3003 PSV 3013 PSV 3023' PSV 3033 1210 psig 16.0 n!'

4.

PSV 3004 PSV 3014 PSV 3024 PSV 3034 1220 psig 14.0 nl 5.

PSV 3005 PSV 3015 PSV 3025 PSV 3015 1235 psig 16.0 n

. _)

"The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating toegerature and pressure.

J V0GTLE UNITS - 1 & 2 3/4 7-3 l

e

TA8LE a.7-1

_ SECONDARY C00LANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM

......[OF MEASUREMENT TYPE SAMPLE AND ANALYSIS AMD ANALYSIS FREQUENCY y

a 1.

Gross Radioactivity At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , etermination*

2.

Isotopic Analysis for DOSE

. a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity determination indicates concentrations greater than 105 of the allowable lieit for radiciodines, b) Once per 8 monthr., when-ever the gross radio-activity determination indi.a.atas concentrations less than or equal to 105 of the allowable limit

(.

for radiciodines.

\\ ': l "A gross radioactivity analysis shall consist of the quantitative measurement of the total specific activity of the secondary coolant except for radio-nuclides with half-lives less than 14 minutes.

Detaruination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level.

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V0GTLE UNITS - 1 & 2 3/4 7-8 1

l 5

(Te te Aavhed to Re#het tMt. 2]

PLANT SYSTD45 2,4,.7.6_CONTnJLROCMEMERGENCYFILTRATIONSYSTEM LIMITING CON 0! TION FOR OPERATION J ".:..

3.7.6 Two independent Control Room Emergency Filtration Systems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

MODES 5 and 6 during novament of Trractated fuel or movement of loads over irradiated fuel.

ACTION:

MODES 1, 2, 3 or 4:

With one Control Roon Emergency Filtration Systas inoperable, restore the inoperable system to 0PERA8LE status within 7 days or be in at least H0T STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD $HUTDC%N within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

MODES 5, and 6 during movement of irraatated fuel or movement of loads over irradiated fuel:

With one Control Roos Emergency Filtration System inoperable, a.

restore the inoperable systas to CPERA8LE status within 7 days or initiate and maintain operation of the remaining CPERA8LE Control Roos Emergency Filtration Systas in the neergency mode, b.

With both Control Roos Emerger.cy Filtration Systans inoperable, or with the OPERA 8LE Control Roce Emergency Filtration System, required to be in the emergency mode by ACTION a.

not capable of being powered by an CPERA8LE emergency power so,urce, suspend all operations involving movement of irradiated fuel or movement of loads over irradiated fuel.

)

$URVEILLANCE REQt'IRDeOf75 l

4.7.6 Each Lontrol Roce Emergency Filtration Syntaa shall be demonstrated OPERA 8LE:

At least once p r 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that ta[econtrol room a.

air tamperature is less than or equal to (DOS b.

At least once per 31 days on a STAGGERED TEST BASIS by initiatir1, from the control room, flow (FI-12191, FI-12132) through the HEPA filters and charcoal adsorters and verifying that the systee ope 'ates for at least 10 continuous hours with the heater control circuit i

energized.

I V0GTLE UNITS - 1 & 2 3/4 7-14

Pt, ANT SYSTEMS SURVE!LLANCE REOUIRENENTS (Continued) 3)

Verifying that the systes maintains the control room positive pressure of greater than or equal to 1/8 i Water Gauge at less than or equal to a pressurtzstion fl of cfm relative to adjacent areas during system operation; g

4)

Verifying that the heaters dissipate 118 2 6 kW when t(dja.

accordance with Section 14 of ANSI M510-1980; and 5)

Verifying that on a Control Room / Toxic Gas !selation test signal, the control room isolation dampers close within 6 seconds and the systes automatically switches into an isolation mode of operation with flow through the HEPA filters and charcoal adsorbers, f.

After each complete or partial replacement of a HEPA filter bank, by verifying that the HEPA' filter banks ren:ve greater than or equal to 99.953 of the MP when they are tasted in place in accordance with Section 10 cf AN5! M510-1980 while operating the systes at a flow rata of 19,000 cfm 2105; and g.

After each completa or partial replacement of a charcoal adsomer bank, by verifying that the charcoal absorters remove greater than C,

or equal to 93.95% of a halogenated hydrocarton refrigerant test gas

\\-

when tested in place in accordance with $ action 12 of ANSI M510-1980 V

while operating the systee at a flow rata of 19,000 cfm 2103.

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E V0GTLE UNITS - 1 & 2 3/4 7-16 h

PLANT SYSTEMS 3/4.7.7 PIPING DENETRATION AREA FILTRATION AND EXHAUST SYSTEM LIMITING CON 0! TION FOR OPERATION 3.7.7 Two independent Piping Penetration Area Filtration and Exhaust Systems shall be OPERAR E.

APPLICABILITY: MODES 1, 2, 3,* and 4' ACTION:

With one Piping Penetration Area Filtration and Exhaust Systa inoperable, nston the inoperable syntes to OPERA 8LE status with!n 7 days or be in at least HOT STAN08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN vithin the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

SURVEILLANCE REQUIREMENTS 4.7.7 Each Piping Penetration Area Filtration and Exhaust Systes shall be demonstrated OPERA 8LE:

At least once per 31 days on a STAGGERED TEST BASIS by initiating, a.

from the control roce, flow (FI-12629, FI-12542) through the HEPA filters and charcoal adsorters and verifying that the system operatas for at least 10 certinuous hours with the heater control circuit energized, b.

At least once p<c 18 months or (1) after any structural maintenance on the HEPA filter or charcoal adsorter housings, or (2) following painting, tire, or chemical n1 ease in any ventilation zone communi-cating with the systas by:

1.

Verifying that the c1 Ny tee satisfies the in-place testing acceptance criterp gnate an or equal to 99.95% filter retention while perating the sy es at a flow rata of 15,500 cfm

105 (Unit 1) cfm

ICT (Un

2) performing the following tests:

I (a) A visual spec $1on of e iping penetration area filtration and exhaust (tee,s be sade before each 00P test or activated carton adsorter section leak test in accordance with Section E of ANSI N530-1980.

(b) An in-place 00P test for the HEPA f11 tars shall be performed in accordance with Section 10 of ANSI M510-1960.

(c) A charcoal adsorter section leak test with a gaseous halogenated hydPocarbon refrigerint shall be perfonsed in accordance with Section 12 of ANSI M510-1980.

'The provisions of this specification are not applicable to Unit 2 until its initial entry into MODE 2.

V0GTLE UNITS - 1 & 2 3/4 7-17 1

PLANT SYSTEMS 3/4.7.8 SNUBBERS LIMITING CON 0! TION FOR OPERATION 3.7.8 All snubbers shall be OPERA 8LZ.

Theonlysnubbersexcludedfrobthe requirements are those installed on nonsafety-related systems and then anly if their failure or failure of the system on which they are installed would have to adverse effect on any safety-related system.

APPLICABILITY:

MODES 1, 2 3, and 4.

MODES 5 and 6 for snubbers located on systems required OPERABLE In those MODES.

ACTION:

With one or more snubbers inoperable an any systes, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or re-store the inoperable snubber (s) to OPERA 8LE status and perfone an engineering eval-uation per Specification 4.7.49. on the attached component or declan the attached system inoperable and follow the appropriate ACTION statement for that system.

SURVEILt#0E_RFQUIRENENTS Each snubber shall be demonotrete[0PERAALE by performance of the following

4. 7. 4 augmented inservice inspection program in addition to the Nguirements of specification 4.0.5.

a.

Inspection Ty m As used in this specification, type of snubber shall mean snubbers q

of the same design and sanufacturer, irrespective of capacity.

q,i b.

Visual Inseections

' Snubbers are categorized as inaccessible or accessible during reactor operation.

Each of these groups (inaccessible and accessible) rsay be inspected independently according to the schedule below.

The first in-service visual inspection of each type of snubber shall be performed after 4 nonths but within 10 months of commencing POWER OPERATION and shall include all inubbers.

If all snubbers of each type are found OPERA 8LE during the first inservice visual inspection, the second.

inservice visual inspection shall be performed at the first refueling outage.

Otherwise, subsequent visual inspections shall be performed in accordance with the following schedule:

No. of Inoperable Snubbers of Each subsequent Visual Type per inspection Period Inspection Period" i

0 M mnus 2 2M i

1 12 months a 25%

2 6 months 2 25%

3,4 124 days a 25%

5,6,7 62 days 2 25%

4 or more 31 days 2 25%

4 "The inspection interval for each type of snubber shall not b4 lengthened more than one step at a time unless a generic probles has been identified and corrected; in that event the inspection interval may be lengthened one 7 top the first time and two steps thereafter if no inoperable snubbers of i

that, type are found.

i he provisions of Specification 4.0.2 are not applicable.

%As. m 2 m 1 1, i

PLM T SYST us SURVEILL.ANCE REQUIRDENTS (Continued) c..

e.

Functional Tests (Continued) 2)

A mpresentativa sample of each type of snubber 'shall be func-tionally tested in accordance with Figure 4.7 1.

"C" is the total ntaber of snubbers of a type found not s,eeting t!ie accept-ance requirements of Specification 4.7.4f.

The cumulative number of snubbers of a ty At the r.ad of each day's testing,pe testad is denoted by "N".

the new values of "N" and "C" (pre-vious day's total plus curnnt day's increments) shall be -

plottedonFigure4.7-1.

If at any time the point plotted falls in the Reject" region, all snubbers of that type shall be functionally tested.

If at any time the point plotted falls in the "Accept" region, testing of snubbers of that type may be taminated.

When the point plotted lies in the "Continue Testing" region, additional snubbers of that type shall be tested untti the point falls in the "Accept" region or the "Reject" region, or all the snubbers of that type have been t

tested; or 3)

An initial np nsentative sample of 55 snubbers shall be func-tionally tasted.

For each snubber t functional test acceptance critaria,ype which does not meet the another sample of at least one-half the size of the initial sample shall be tested until

/ _..

the total number tasted is equal to the initial sample size j

multiplied by the factor,1 + C/2, dare "C" is the nuncer of l

Anyhters found which do not meet the functional test acespta r

' ceiteria.

The results from this sample plan shall be MIGd using an "Accept" line dich follows the equation M =

(1 +

l C/f).

Each snubber point should be plotted as sm n a the 1

er is tasted.

If the point plotted falls on or beloe>

j "Accept" line, testity of that type of snubber may be terminated.

If the point plotted falls above the "Accept" line, tasting auet continue until the point falls in the "Accept" region or all the snubbers of that type have been tasted.

l Testing equipment failure during functional tasting may invalidata that day's testing and allow that day's testing to restme anew at a later time provfded all snubbers tasted with the failed equipment during l

the day of equipment failure are retested.

The meresentative sasole i

selected for the functional test sample plans shall be randomly selected free the snubbers of each type and reviewed before begir.ning the testing.

The nyiew shall ensure, as far as practicable, that they an repnsen-tative of the varinus configurations, operating environment!,, range of i

size, and capacity of snubbers of each type.

Snubbers placed in the I

same location as snubbers which failed the previous functional test i

shall be ntasted at the time of the next functional test but shall i

not be included in the sample plan.

If during the functional testing, additional sampling is required due to failun of only one type of l

snubber, the functional test results shall be nyiewed at that time I

to determine if additional samples should be limitad to the type of snutber dich has failed the functional testing.

V0GTLE UNITS - 1 & 2 3/4 7*21

)

1 10 9

8 REJECT 6

C 5

e' 4

CONTINUE

/

TEsTiNo 3

1 Y

~,$

e

,9 '

9 2

ACCEPT f

/

0 10 20 30 40 60 60 70 80 90 100 N

FIGURE 4.7-1 SAMPLE PLAN 2 FOR SNUBBER FUNCTIONAL TEST V0GTL2 UNITS - 1 & 2 3/4 7-24

_TA8LE 3.7-3 AREA TEMPERATURE MONITORING l

1 MAXIMUM MAXIMUM BUILDING Room N) f ggyAL TEMP M No w L m p FUEL 80 104 120 AUXILIARY 110 100 104 AUXILIARY 202 100 lo' AUXILIARY 203 100 1cy AUXILIARY A017 100 U2 AUXILIARY Ao47 100 107 AUXILIARY 8017 100 100 AUXILIARY C113 100 gog AUXILIARY C120 100 106 1

AUXILIARY 0053 100 uo AUXILIARY 0064 100 104 AUXILIARY 0072 100 105 AUXILIARY 0075 100 gog AUXILIARY 0113 100 105 AUXILIARY 0121 100-103' CCNTROL 147 30 37 CONTROL 149 30 37 CONTROL A054 100 u4 CONTROL A062 100 103 CONTROL 8065 100

' 106 CONTROL 8074 100 107 CCNTROL 5078 100 104 i

(Rooms associated with Unit 2 will be added latar]

i 4

t 4

VoGTLT UNITS - 1 & 2 3/4 7 28 i

3/4.7.13 O!ESEL GENERATOR BUILDING AND AUXIL!ARY FEE 0 VATER PtHPHOUSE SYSTEMS LIMITING CON 0! TION FOR OPERATION 3.7.13 The diesel generator building and auxiliary feedwater pumphouse ESF HVAC systems shall be OPERABLE with:

At least two ESF supply fans and associated dampers per diesel a.

generator train, and b.

At least three ESF auxiliary feedwater pumphouse HVAC systems.

APPLICA8!LITY:

Whenever the associated diesel generator or auxiliary fesowater pumps are required to be operable.

ACTION:

With 1 or more supply fans to a given diesel generator train inoper-a.

able, restore the inoperable fan (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following urs.

b.

With 1 or acM ESF auxiliary feedwater pumphouse HVAC. systee incper-able, follow the ACTION specified % specif cation 3.7.1.2 for an inoperable auxiliary feedwater p

.E SURVEILLANCE REQUIREFENTS 4.7.13 The diesel generator building and auxiliary feedwater pumphouse ESF HVAC systes shall be demonstrated QPERA8LE:

a.

At least once per la months by:

l 1)

Verifying that the diesel generator ESF supply fans U2-1566 B7-001 1

(train A) and U2-1564-87 002 (train B) start automatically are the associated intake and discharge dampers actuata to their j

cornet position on their train associated diesel generator running signal.

2)

Verifying that diesel generator ESF supply fans V21566 87-003 and V2-1566-87-004 start automatically and the associated intake and discharge dampers actuate to their correct position on a f

high diesel generator building toeparature signal coincident with diesel generator running signal.

b.

At least once per la months by:

1)

Verifying that the auxiliary feedwater pumphouse E5F supply fans, V2-1593 57 001 and U2-1593 87-002 and associated shutoff dampers actuate to their correct position on a high reca tencerature signal.

VCGTLE UNITS - 1 & 2 3/4 7-31 i

9LANT 615TEMS o

b 3/t 7.17 (Continued

$URVE!LLANCE RfCUfREMENT$

j 4.7.13 (Continued) 4 l

2)

Verifying that the t$F eutside air intake and exhaust dampers i

for the turbine driven auiliary feedwater pump actuate to the correct position en a turbine driven auiliary feedwater pump automatic start signal.

a 1

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YOGTLE UNIT 1 - 1 & 2 3/4 7-32 l

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3/4.8 ELECTRICAL POWER SYSTEMS h

3/4.8.1 A.C. SOURCES d

OPERATING LIMITING CON 0! TION FOR OPERAT!0N 3.8.1.1 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

Two physically independent circuits between the offsite transatssion a.

network and the onsite Class 1E Distribution Systen, and

\\

b.

Two separate.?nd independent diesel generators, each with:

1)

A day tank contat Log a i

instrument span)'A inisue volume of 450 gallons (52% of of fue LI-9018, LI-9015),

2)

/. separate Fuel Storage Systes containing a sintaus volume of 68,000 gallons of fuel (76% of instrument span) (LI-9024, l

LI-9025), and 3)

A separata fuel transfer pump, APPt.!CA8!LITY:

MODES 1, 2, 3, and 4.

ACTION:

/"'g With one offsite circuit of the above-required A.C. electrical power a.

1, sources incperable, demonstrate the OPERA 8!LITY of the remaining A.C.

sources by performing Surveillance Requirement 4.4.1.1.1.a within I hour and at least, once per a hours thereafter.

If either diesel generator has not been successfully tastW within the 24st 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , coeonstrata its CPEAA8ILITY by performing Surveillance Requirements 4.4.1.1.2.a.4 and 4.4.1.1.2.a.5 for each such diesel generator, separately, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless the diesel generatar is already operating.

Restart the offsita circuit to CPEAA8LI status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months er be in at least HOT STAN04Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and

)

in COLD SHUTDOW within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months , b.

With either diesel generator inoperable, demonstrate the OPERA 4!LITY of the above required A.C. effsite sources by performing Surveillance Requirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter.

If the diesel generatar became inoperable due to any cause 1

other than preplanned preventive natntenance or testing, demonstrate

{

the OPEAASILITY of the remaining CPERA8LI diesel generator by perfare-ing Surveillance Requirements 4.4.1.1.2.4.4 and 4.8.1.1.2.a.5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 4.

Restore the inoperable diesel generator to CPERA8LE status i

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANC4Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months l

and in COLD SHUTDOW within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

i "This test is required to be completed regardless of den the inoperable diesel generator is restored to CPEAA8ILITY.

The diesel shall not be rendered inoperable by activities perforSed to y

support testing pursuant to the Action Statement (e.g., an air roll).

I VCGTLE UNITS 1 & 2 3/4 8-1 t

ELECTR! CAL POWER SYSTEMS LIMITING CONDITION FOR OPCRATION ACTION (Continued)

With one offsite cir< utt and one diesel generator of the above nquired c.

i A.C. electrical power sources inoperable, demonstrate the OPERASILITY of the remaining A.C. offsite source by performing Surveillance Re-quirement 4.8.1.1.1.a within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereaf ter, and, if the diesel generator became inoperable due to any cause other than pnplanned preventative maintenance or testing, demon-strate the OPERASILITY of the remaining CPERA8LE diesel generator by m

performing Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 kC&[ $

within 8 bours*, unless the OPEAA8LE diesel generator is already s

operating)in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or be in at least HOT STANOBY within the nex status with 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months and in COLD SHLJTDOWN within Jw following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

Reston the other A.C. power source (offsite.ircuit or diesel generator) to CPEAA8LE status in accordance with the provisions of 3.8.1.1, Action Statement a or b, as appropriata, with the time requirement of that Action Statement based on the time of initial loss of the remaining troperable A.C. power source.

A successful test of diesel generator CPEAASIUTY per Surveillance Requirements 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 performed under the Action Statement for an CPERABLE diesel generator or a nstored to OPEAA8LE diesel generator satisfies the diesel generator test nauirement of Action Statement a or b.

I d.

With one diesel generator inoperable in addition to ACTICN b. or c.

i above, verify that:

l All required systees, subsystems, trains, ceaconents, and devices 1.

that depend on the remaining OPERA 8L1 diesel generator as a i

source of emergency power are also CPERA8LE, and 2.

When in MODE 1, 2, or 3, the staae &iven auxiliary feedwater ptmo is CPERA8LI.

l If these conditions are not satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be in at least HOT STAM08Y within the next 6 houn and in COLD SHLJTDOWN within the l

following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I e.

With two of the above required offette A.C. circuita inoperable, demon-l strata the OPERA 8!UTY of two diesel generators separately by perfore-l ing_the requirements of Specification 4.8.1.1.2.a.4 and 4.8.1.1.2.a.5 u_

Qb within 8 hourid, unless the diesel generators are alnady operating; notare et least one of the inoperaale offsita sources to CPERA8LE I.

status within 24 houn or be in at least HOT STAA08Y within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

Fellowing nsteration of one offsite sourts, follow Action Statement a with the time requirement of that Action statement based "This test is nquind to be completed regardless'of when the incoerable EDG is restored to CPEAA8!WTY.

The diesel shall not be nndend inoperable by activities perforsed to support testing pursuant to the Action Statement (e.g., an air roll).

V0GTLE UNITS 1 & 2 3/4 8 2

ELECTRICAL PCvtR SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) v b)

A kinematic viscosity at 40*C of greater than or equal to 1.9 contistokes, but less aan or equal to 4.1 centistokes, if gravit supplier'y was not deterni..ed by comparison with s certification; c)

A flash point equal to or greater than 125'F; and d)

A clear and bright appearance with proper color when tasted in accordance with ASTM-04176 42.

2)

Sy verifying within 30 days of obtaining the sagle that the other properties specified in Table 1 of ASTM 0975-81 an set l

when tested in accordance with ASTM 0975-41 except that the analysis for sulfur may be performed in accordance with ASTM 01552-79 er ASTM-02622-82.

At least once every 31 days by obtaining a sample of fuel oil in e.

accordance with A$TM D22/6-78, and verifying that total particulate l

contamination is less than 10 mg/litar when checked in accordance with ASTM 02276 78, Method A; f.

At least once per 92 days and from new fuel prior to addition to the storage tank obtain a sample and verify that the neutralization number is less than 0.2 and the mercaptan content is less than 0 bY15Cfif

\\

g g.

At least once per 144, days by:

l j

...)

1)

Verifying the diesel starts' from ambient conditions and the generator voltage and frequency a n 4160 + 170, -410 volts and 60 21.2 Hz within 11.4 seconds after the start signal.

The diesel generatar shall be started for this test by using one of the signals listed in Surveillance Requineent 4.8.1.1.2.a.4 l

{

This tast, if it is performed so. it concides with the tasting Mquired by Surveillance Requirement 4.4.1.1.2.a.4, may also l

serve to concurnntly meet these, requirements as well.

I

All engine starta for the punene of surveillance testing as nquired by

,$ctrh.CO 4.8.1.1.2 may be proceded by an e ine prelube. period as ncommended by the manufacturer to stataize mechanica stnes en the diesel engine.

i

Mercaptan content shall not be reqJired to be vartfied within specification for new fuel prior to its additten, for up to 15,000 gallons of fuel added to the tank, if the last tank samle had a mercaptan cont 4nt of less than 0.007hnA11 subsequent new fuel addition will require mercaptan content veri-ficafton trier to its addition until the tank cantants an verified to be less t

fed {nG0075.

th tFsYatipnnus[er have[to be verified its' add tion, unti) 60 d4ys after Itcenst[issuanie.ssthd0 l

11 n l

1 Until th t' tfine f

L n of newj uel pecif ations 11beicepletpwith430 s/f I

VCGTLE tl NIT 5 1 & 7 3/4 8 5 l

l i

l I

l r

g(CTRICAt. PCVER SYSTD45

$URVEtt.t>NCE REQUIRD4ENTS (Continued)

/

2)

Verifying the Tadedtoanindicated value of $100 generator (s tyncleMaed.#or equal t 7000 W" umluuthan and operates with a load of f400-7000 W*** for at least 60 minutes lhts test, if it is performed so it coincides with the testing nguired by Surveillance Requirement 4.8.1.1.2.e.5, may also serve to concurreatly meet those requirements as well.

h.

At least once per la months,** during shutdown, by:

i 1)

Subjecting the diesel to an inspection in accordance with proce-duns prepared in conjunct,fon with ita manufacturers' recosesenda-tions for this class of standby service; 2)

Verifying the diesel generator capability to nject a load of greater than or equal to 671 W (notor-driven auxiliary feedwater pump) while maintaining voltage at 4160 + 240. -410 volts and speed of less than 444 rpa (less than nominal speed plus 75%

l of the differenca between nominal speed and the overspeed Trip Setpoint); and ncovering voltage to within 4160 + 170, -410 volts within 3 seconds.

3)

Verifying the diesel generator capability to reject a load of 7000 W without tripping.

The generator voltage shall not exceed 5000 volta during and following the load njection;

{

4)

Simulating a loss-of offsite power by itself, and:

(

Verifying deenergitatien of the emergency busses and load shedding free the emergency bussas, and la Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 11.5 seconds

energites the auto-connected shutdown loads through the Iced sequencer and operates for gnater than er equal to 5 minutes while ita generator is loaded with the shutdown leads. After energization, the steady-stata voltage and frequency of the emergency busses i

shall be asintained at 4160 +170, -410 volta and 60

1.2 Hz l

during this test.

~

5)

Verifying that on an 13F Actuation test signal, without loss-of-l offsita power, the diesel generatar starta en the auto-start signal and operates en standby for gnater than er equal to 5 minutes.

The generater voltage and frequency shall be 4160

+170. -410 volta and 60 s 1.2 Ma within 11.4 seconds af ter the l

"All engine starts for the purpose of surveillance testing as required by Specification 4.8.1.1.2 may be preceded by an engine preluce period as recce-asMed by the manufacturer to minialze mechanical stnes and wear on the j

diesel engine, i

- For any start of a diesel, the diesel sust be operated with a load in accor-t dance with the manufacturer's recommendations.

- - This ttand is meant as guidance to avoid routine overloading of the engine.

t Loads in excess of this band or somentary variations due to changing bus loads shall not invalidate this test.

l l

VCGTLE LNITS 1 & 2 3/4 8 6 i

I f

SURVE!LLANCE REQUIRENENTS auto-start signal; the steady-state generator voltage and frequency shall be saintained within these limits during this test; 6)

Simulating a loss-of offsite power in conjunction with an ESF Actuation test sigmal, and a)

Verifying deenergization of the emergency busses and load shedding from the energency busses; b)

Verifying the diesel starts on the auto-start signal, energizes the emergency busses with permanently connected loads within 11.5 seconds," energizes the auto-connected l

emergency (accident) loads through the load sequencer and operatas for gnator than or equal to 5 minutes while its generator is losJed with the emergency loads.

After energi2ation, the steady-state voltage e-d frequency of the emergency busses shall be saintained at 4160 +170

-410 volts and 60 + 1.2 Mz during this tast; and c)

Verifying that all automatic diesel generator trips, except engine overspeed, low lube oil pressure, high jacket watar i

temperatuns and generator differential, are automatically bvpassed upon loss of voltage on the emergency bus concurnnt with a Safety Injection Actuation signal.

7)

Verifying the diesel generator operatas for at least 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Durirl the first 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of this test, the diesel generator shall 3

be lo'ded ta an indicated 7600 to 7700 W,** and during the I

resei'ing 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months of this test, the diesel generator shall be leadeo to an indicated 6800-7000 W. ** The generator voltage and frequency shall be 4160 + 170 - 410 volts and 60 21.2 Hz I

within d.4 seconds after the start signal; the steady-state l

generator voltage and f uency shall oe maintained within these 1

limita during this test # Within 5 minutes after completip.

l this 24-hour test, perform ecification 4.8.1.1.2h.6)b) N 8)

Verifying that the auto-connec loads to each diese i

generater de not exceed the contin a rating of 70 W;

j 9)

Verifying the diesel generatar's apab gh J 5 C ri

'All engines starts for the purpose of survet lance tes(h hdk/chCn 4.8.1.1.2 say be prwceded by an engine preline period as recommended by the as requireo by i

l nanufacturer to sintatte nochanical st nes and wear on the diesel engine.

- This band is meant as guidance to avoid routine overloading of the engine.

l Loads in excess of this band or momentary variations due to changing bus loads i

j shall not invalidata the test.

Failure to natntain voltage and frequency requirements due to grid disturtances does not render a 24-hour test as a failure.

N!f Specification 4.8.1.1.2h.6)b) is not satisfactorily ccepleted, it is not necessary to Npeat the preceding 24-hour test.

Instead, the diesel generator wy be operated at the load requind by Surveillance Requinsent 4.8.1.1.2.a5 W for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or until operating temperature has stabilizad.

t 4

. V0GTLE UNITS - 1 & 2 3/4 4-7 i

j

TA8LE 4.0 1 O!!$tL GENERATOR TEST SCHEDULL Number of Failures Heer of Failures in in Last 100 Valid t,ast 20 Valid Tests

Tests

_ Test Frequency 11 14 One.e per 31 days

, 1 2'*

>5 Once per 7 days i

f

-I I

l

Critaria for detaruining number of failures and numeer of valid tests shall be in accordance with Regulatary Position C 2.e of Aegulatory Guide 1.108, but detamined en a per diesel generatar basis.

For the r,urposes of detamining the requind test frequency, the previous test failure count may be reduced te zere if a completa diesel overhaul to lika-newconditifL(

empleted, provided that the everhaul, including appro-priata post as nance ration and tasting, is specifically app nved by I

the manufactu r 3g1 if ecceptaale n11 ability has been descastrated.

The reliability cr tamn st)sil be the successful completion of 14 consecutive tests in a si I q fst.

Ten of these tests shall be in accordance with the ruutine Survet11ance Requiraments 4.4.1.1.2.a.4 and 4.4.1.1.2.a.5 and four tests in accordance with the 144 day tasting requirement of Surveillance I

Requirement 4.4.1.1.2.f.

If this critarion is net satisfied during the first i

series of tests, any altamate critarion to be used to transvalue the failure count te zero requins NRC aoproval.

- The associated test frequency shall be ea M n(d until seven consecutive i

failure fne comands have been perfo g the!nuncer of failures in the l

last 20 valid desands has been reduced one.

l 1

YOGTLE UNITS - 1 & 1 3/4 8 9 l

t,

O.C. SOURCES gg LIMITING CON 0!T!0N F0.1 OPERAT 3.8.2.2 As a minimus one trtl related pair of 125 V ttery ks (either 125 V battery banks 13 and Q3018 or 125-V battery banh 13012 a one associated full-capacity. charger per required battery { bank shall 10018) and e

OPEAA8LI.

APPLICA4!L.f2: MCOES 5 and 8.

ACTION:

With the required battery bank and/or both chargers inoperable, immediately suspend all operations involving Co?E ALTEAATIONS, positive reactivity changes, or movement of irradiated fuel; initiate cornctive action ta restore the required battery bank and at least one charger to CPP'~ 'tatus as soon as possible, and wiain 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , provide relief capabil' the Reactor Coolant Systes in accordance with Specification 3.4.9.3.

SURVE!LLANCt REQUIRDt2NT5

(...

(-

4.8.2.2 The above required 12.5 volt battery bank and full cacacity char shall be demonstrated CPERA8LE in accordance with Specification 4.4.2.1.ger I

t e

YCGTLE UNIT 5 - 1 & 2 3/4 S*14

,tLtcTRICAL (CutPw!NT PROTECTIVE OEvtCES SAFI'TY RELATED MOTOR CPERATro VALVES THERMAt. OVERL0A0 PROTECTION AND L

DEV CIS LIN!T!NG CON 0! TION FOR OPERATION 3.4.4.2 The thermal overload protection bypass devices of each safety related actor-operated valve given in Table 3.81 shall be CPERA:LE.

APPLICA8!LITY: Whenever the motor-operated valve is required to be CPERABLE.

ACTION:

With the thersal everload protection bWass devices for any one or more safety-related notor-operated valves inoperable, declare the affected valve (s) inoper-able and apply the appropriate ACTION $tatement(s) of the affected valve (v).

SURVE!LLANCE REQUIR@ENT$

4.1.4.2 The above required thermal overload protection bpass devices shall be verified to to CPERABLI.

a.

Following natntenance on the valve actor starter, and b.

Following any periodic testing during which the thersal overload device was tassorarily placed in force.

c.

At least once per la month uringshutdod 4

U VCGTLE UNITS 1 & 2 3/4 8-21 i

TAALE 3.4-1

$AFE"'Y R(LATED MOTOR OPERATED Vat.VE5 THER. MAL OVEr.t.0A0 PROTECTION 6YPA55 CEVICES VALVE NW8ER FUNCTION V2W-1668A Nuclear Service C1g Twr, A Return 1/2W-16688 Muclear Service Cig. Twr, A Fans typass 1/2W-1669A Muclear Service C1g. Twr, B Return U2W-16698 Nuclear Service C1g. Tvr. I Fans I pass

/

V2W-1806 CTNT Air Cool A700VA7002 CW Inlet V2W 1807 CTNT Air Cool A7003/A7004 CW Inlet V2W-1404 C M tir Cool A7005/A7004 CW Inlet V2W-1809

?:f Air Cool A7007/A7004 CW Inlet V2W-1422 lM iir Cool A700VA7002 CV Cutiet 1/2W 1823 7Id' Air Cool A7003/A7004 CV Cutlet V2W 1430 CfMT Air Clr A7005/A7006 CW Outlet V2W-1831 CTNT Air Clr A7007/A7004 CV Cutlet I

V2W-1974 Aux Comp CW Trn B Return Iso U2W-1975 Aux camp CW Trn A Return Iso i

U2W 1978 Aux Coop CW Trn 8 Supply V2W-1379 D CW Trn A Supply '4 V2W 2041 Reteter, Coolant Pumps Tharu.arrier XCWS (utlet Heeder 1/2W n34 4eactae cavity C1g coil D001 Inlet Iso V2W-2135 R4 attar Cavity C1g coil D002 Inlet Iso 1/2W 2138 Reactor Cavity C1g coil D001 Outlet Iso V2W 2139 Reactor Cavity C1g coil D002 Outlet Iso U2W-uu4 CR Outside Air Intake Iso V2W un5 CR Outside Air Intake Ise 1/2W-12n3 C8 CR Filter linits N7001 Inlets V2W-u113 C3 CR Filter Ur.its N7002 Inlets V2W-Mut CS CR Filter Units N7001 Outlets V2W-M129 CS CR Filter Units N7002 Outlets 1/2W-u130 CS CR Return Air Fans 57005 Inlets U2W-u131 CS CA Return Air Fans B7004 Inlets i

U2W-u727 CS 3R Sattery As hh 87002 Damper V2W 12742 CS SF Battery An E.xh B7001 Desper V2W u744 CS 5/ Battery Re Exh 57003 Damper V2W-u749 C8 SF Battery Re Exh 57004 Desper U2W-19051 Thermal Barrier Coeling Wtr RCP 001 V2W-19013 Thermal Barrier Coeling Wtr RCP 002 1/2W-19055 Thersel Barrier Coeling Wtr RCP 003 l

V2W 19057 Thersal Barrier Cooling Wtr RCP 004 i

U2W 4920 Safety-Injection Puse Mintflow ! solation V2W-2624A CT1 Post LOCA Purge E.xhaust Iso U2W 26244 CT3 Post LOCA Purge Exhaust Iso i

U2W 2626A CTNT Bldg Norg Purge Supply !so

{

l i

i i

VCGTLE UNITS

1 & 2 J/4 8 22

TA8LE 3.8-1 h n45NC SAFETY-RELATED MOTOR OPERATED vat.VES THERMAL OVERLOAD PROTECTION 8YPASS DEVICES VALVE NUMBER FUNCTION V2HV-2627A CTB Nom Purge Supply Iso 1/2HV-2628A CTMT Bldg Nors Purge Exhaust Iso U2HV-2629A CTB Nom Purge Exhaust Iso U2HV-5106 Aux FDW Pump Turt ine t

V2HV-5113 Conds Stor TX V4002 to Pump P4001 V2W-5118 Conds Stor TX V4002 to Pump P4002 1/khV-5119 Conds Stor TX V4002 to Pump P4003 1/2HV-5120 Aux FDW Pump P4001 Discharge Trn C 1/2HV-5122 Aux FDW Pump P4001 Discharge Trn C V2HV-5125 Aux FDW Pump P4001 Discharge Trn C 1/2HV-5127 Aux FDW Pump P4001 Discharge Trn C 1/2HV-5132 Aux FDW Pump P4002 Discharge Trn 8 1/2HV-5134 Aux FDW Pump P4002 Discharge Trn 8 1/2HV-5137 Aux FDW Pump P4003 Discharge Trn A 1/2W-5139 Aux FDW Pump P4003 Discharge Tro \\

1/2FV-5154 Aux FDW Pump P4002 Miniflow l

1/2FV-5155 Aux FUW Pump P4003 Miniflow 1

1/2HV-8428 Charging Pump 8 Discharge 1

1/2HV-8471A Alt Charging Pump A Suction 1/2HV-84718 Alt Charging Pump 8 Suction 1/2HV-8445A Charging Pump A Otscharge

, 1/2HV-84858 Charging Pump 8 Discharge U2HV-8508A, 8 Charging Pump Miniflow Iso to RWST 1/2HV-8509A, 8 Charging Pump Miniflow Iso to RWST 1/2HV-930A CTNT ATM Unit 1 SVCE Air i

U2HV-93803 CTMT ATM Unit 1 SVCE Air 1/2W-12005 Trn 8 Aux FW Pump Ra Inlet Damper 1/2HV-12006 Trn A Aux FDW Pump Ra Inlet Damper U2HV-12050 DG8 Exh Fan 87001.015ch Damper (Trn A)

U2HV-12051 DG8 Exh Fan 87003 Dicch Dampar (Trn A) 1/2HV-12052 DG8 Exh Fgn 87005 Disch Damper (Trn A)

V2HV-12053 DG8 Exh Fan 87002 Disch Damper (Trn 8) 1/2HV-12054 DG8 Exh Fan 87004 Disch Damper (Trn 8)

U2HV-12055 DG8 Esh Fan 87006 Otsch Osaper (Trn 8)

V2HV-8000A, 8 PORY 81ockline V2HV-8147 Reg..Hx Tube Outlet to RCS Alternate Chg V2HV-8146 Reg. Mx Tube Outlet to RCS Normal Chg V2W-8100, 8112 No 1 Seal Leakoff I

U2HV-8103A, 8, C, O RCP No 1 Seal froe Chg V2W-8111A, 8, 8110 Chg Pump Mintflow V2LV-0112C, 8 VCT Discharge Header U2LV-0112E, 0 SIS RW5T Discharge to Chg/S! Pump Suction U2HV-8104 CVC5 Boric Acid Filler to Charging Pump suction i

i V0GTLE UNITS - 1 & 2 3/4 8-23 i

((onhued)

TABLE 3.8-1 f 7 SAFETY-RELATED

~IDR-CPERATED VALVES THERMAL QVERLOAD MO Q ROTECTION SYPASS DEVICES VALVE NUMBER FUNCTION 1/2HV-8105; 8106 Chg Pump to RCS Isolation 1/2HV-8807A, 8; 8924 HHSI Suction to Chg/SI Suction V2W-8801A, 8 BIT Discharge 1/2W-8808A, 8, C, D Accumulator Otscharge 1/2HV-8811A, 8 Containment Emergency Sump Isolation 1/2HV-8812A, 8 RHR Suction from RWST 1/2HV-8809A, 8 RHR Discharge Header 1/2HV-8804A RHR Nx No. 1 Outlet to Charge Pump 1/2HV-88048 RHR Nx No. 2 Outlet to 51 Pumps 1/2HV-8806 RWST Discharge Header to SI pumps 1/2HV-8'J23A, 8 SI Pump Suction Isolation 1/2HV-8813; 8814 SI Pump Mintflow 1/2HV-8821A, 8 SI Pump Crosschannel 1/2HV-8835 51 Pump Discharge to Cold Legs 1/2HV-8840 RHA Pump Discharge to Hot Legs 1/2HV-8802A, 8 SI Pump Discharge Header 1/2HV-8701A, 8; 8702A, 8 RHR Suction free RCS Hot Legt,1, 4 1/2FV-0610; 06 u RHR Miniflow 1/2HV-8716A, 8 RHR Cross Connect 1/2HV 9002A, 8 Spray Pump Containment Emergency Sump isolation V2HV-9003A, 8 Spray Pump Containment Emergency Susp Isolation 1/2HV-9017A, 8 Spray Pumo Suction free RWST V2HV-9001A, 8 Spray,' ump Discharge Header 1/2HV-8994A, 8 Spray Additive Tank Discharge V2HV-u600 MSCW Pump Discharge Isolation U2HV-u605 NSCW Pump Discharge Isolation 1/2HV-u606 MSCW Ptap Discharge Isolation V2HV-u607 NSCW Pump Discharge Isolation U2W-u612 M5CW Pump Discharge Isolation l

V2HV-u6U NSCW Pump Discharge Isolation V2PV-2550A Piping Penetration Room to Atmosphere t

V2PV-2551A '

Piping Penetration Room to Atmosphere L

V2HV-3009 TDAFP Staan supply Isolation i

U2HV-3019 TDAFP Steam Supply Isolation U2HV-8u6 Charging Pump Discharge Baron Injection t

V2PV-15121 TDAFP Trip and Throttle Valve V2HV-2582A CT5 Cooling Unit A7001 V2HV-25828 CT5 Cooling Unit A7002 V2HV-2543A CTS Cooling Unit A7003 V2HV-25838 CTS Cooling Unit A7004 V2HV-2584A CT3 Cooling Unit A7005 U2HV-25848 CTI Cooling Unit A7006 V2HV-2545A CTB Cooling Unit A7007 V2HV-25458 CTB Cooling Unit A7004 V0GTLE UNITS - 1 & 2 3/4 8-24

REFUELING OPERATIONS li,.9.12 FUEL HANDLING BUILDING POST ACCIDENT VENTILATION SYSTEM (COMMO 1.IMITING CONDITION FOR OPERATION 3.9.12 Two independent Fuel Handling BuildingJom ident Vent"ation Systems shall be OPERABLE.

Sker APplICA8ILITY: Whenever irradiated fuel is i M orage pool.

ACTION:..

ygg(,

gjfygf a.

With one Fuel H dling Building Pos Accident Ventilation System inoperable, fuel movement within hstorage pool or crane operation with loads over storage pool say proceed provided the OPERA 8LE Fuel Handling Building Post Accident Ventilation System is capable of being powered from an OPERA 8LE emergency power source and is in operation and discharging through at least one train of HEPA filtars and charcoal adsorbers.

With no Fuel Handling Building Post Accident Ventilation System PERA8LE, suspend all operations involving movement of fuel within 14h8 torage pool or crane operation with loads over % storage pool until at least one Fuel Handling Building Post Accidon apti?ation System is restored to OPERA 8LE status.

'gg c.

The provisions of Specifications 3.0.3 and 3.0.4 are not app icable.

SURVEILLANCE REQUIREMENTS 4.9.12 The above required Fuel Handling Building Post Accident Ventilation Systees shall be demonstrated OPERA 8LE:

a.

At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HLPA filters and charcoal adsorbers and verifying that the system operates for at least 10 continuous hours with the heater control circuit energized; b.

At least once per 18 sonths or (1) after any structural maintenance on the HEPA filter or charcoal adsorter housings, or (2) following painting, fire, or chemical release in any ventilation zone communicating with the system by:

V0GTLE UNITS - 1 & 2 3/4 9-14

3/4.11 RA0!0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS CONCENTRATION

~

I,IMITIMG CONDITION FOR OPERATION 3.11.1.1 The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 5.1-1) shall be Itatted to the concentrations specified in 10 CFR Part 20, A oendix B, Table II, Column 2 for radionuclides.

other than dissolved or entrained noble gases.

For dissolved or entrained noble gases, the concentration shall be 11af ted to 2 x 10 4 aicrocurie/m1 t.ntal activity.

APPt.ICA81LITY: At all times.

ACTION:

Wkth the concentration of radioactive saterial released in liquid effluents a.

tb UNRESTRICTED AREAS exceeding the above limits, immediately restore the concentration to within.the above liatts.

t,.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMMTS-(,4

4. n.1.1.1 Radioactive ifquid wastad shall be sampled and analyzed according to the sampling and analysis program of Table 4.11-1.

4. u.1.1. 2 The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the 00CM to assure that the concentrations at the point of release are saintained within the limits of Specification

3. u.1.1.

i l

l I

f I

l l

i l

\\

.a V0GTLE UNITS 1 & 2 3/4 u-1 9

TABLE 4.11-1 A)

RADI0 ACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM V

LOWER LIMIT MINIMUM 0F DETECTION LIQUID RELEASE SAMPLING AXALYSIS

' TYPE OF ACTIVITY (LLD)(1)

TYPE FREQUENCY FREQUENCY ANALYSIS (WCi/al) 1.

Batch Waste P

P Release Each Batch Each Batch Principal Gamma 5x10 7 Tanks (2)

EsittersI3) a.

Waste-Monitor I-131 1x10.s Tank 1901-76-009 P

M Dissolved and 1x10.s One Batch /M Entrained Gases b.

Waste-Monitor Tank 1901-T6-010 P

M H-3 1x10 8 Each Batch Composite (4) c.

Drainage of Systems P

Q Sr-89, Sr-90 5x10.s Each Batch Composita (4)

Fe-55 1x10-6 2.

Continuous W

Principal Gamma 5x10 7 Releases (5)

ContinuousIO) C g osi u(6)

Emitun(3) l I-131 1x1..e a.

W Wat r i

ret tio bas d

9 M

M Dissolved and 1x10.s.

l*

rab sample Entrained Gases (Ganna Esitters)

M H-3 1x10.s Continuous (6) C g osi u(6) l l

Gross Alpha 1x10 7 9

$r-89, Sr-90 1x10.s Continous(6)

Composite (6)

Fe-55 1x10 8 V0GTLE UNITS - 1 & 2 3/4 11-2 i

TA8LE 4.11-1 (Continued)

TA8LE NOTATIONS (1)The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive saterial in a sample that will yield a not count above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation npresents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

4.66 sh LLD =

. Y. exp (-AAt)

E Y

2.22 x 108 Where:

LLD = She "a priori" lower limit of detection (microcurie per unit mass or volume),

h

the standard deviation of the background counting rate or of the s

count' ng rate of a blank' sample as appropriate (counts per minute),

E = the counting efficiency (counts per disintegration),

V = the sample size (units of mass or volume),

~~'

2.22 x 10s a the number of disintegrations per minuta per microcurie, i

Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec-1); and

' at = the elapsed time between the ai@oint of sample collection and the u se < f counting (sec).

Typical values of E, V, Y, and at should be used in the calculation.

It should be recognized that the LLD it defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a, posteriori (after the fact) limit for a particular seasureeint.

(2)A batch release is the discharge of liquid westes of a discrete volume.

l Prior to sampling for analyses, each batch shall be isolated, and then thorougnly mixed by a method described in the 00CM to assure representative

sampling,

.v VCGTt.E UNITS - 1 & 2 3/4 11-3 i

l

RADIOACT!vEEFFLUEN3 3/4.11.2 GASEOUS EFFLUENTS J

DOSE RATE LIMITING CONDITION FOR OPERATION

3. u. 2.1 The dose rate due to radioactive materials released in gaseous effluents from the site ;,o areas at and beyond the SITE BOUNDARY (sea Figure 5.1-1) shall be Ifmited to the following:

s.

For noble gases:

Less than or equal to 500 areas /yr to the whole body and less than or equal to 3000 areas /yr to the skin, and b.

For Iodine-131, for Iodine-133, for tritium, and for all radio-nuclides in particulate form with half-lives greater than 8 days:

Less than or equal to 1500 aress/yr to any organ.

APPLICABILITY:

At all times.

ACTION:

With the dose rata (s) exceeding the above limits, immediately restore a.

. the release rate to within a limit (s).

r b.

The provisions of Specific ton $3.

3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

4. u.2.1.1 The dose rata due to noble gases in gaseous affluents shall be determined to be within the above limits in accordance with the methodology and parameters.in the 00CM.

4. u.2.1.2 The dose rata due to Iodine-131, Iodine-133, tritium, and all radionuclides in particulata form with half-lives greater than 8 days in gaseous effluents shal.1 be detarsined to be within the above limits in accordance with the methodology and parameters in the 00CM by obtaining representative saeples and perforsing analyses in accordance with the sampling and anslysis program specified in Table 4.u-2.

V0GTLE UNITS - 1 & 2 3/4 11-8 i

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

_TA8LE 4.11-2'(Continued)

TABLE NOTATIONS

^

3b say b al ed provided the absence of a primary to secondary leak has been des n r ed; that is, the gamma activity in the secondary water does nc '

exceed be und by more than 20%.

(1)The LLD is defined, for purposes of these specifications, as the smallest concentration of radioactive material in a sample that will yield a not count above system background, that will be detected with 95% probability with only 57, probability of falsely concluding that a blank observation represents a "real" signal.

For a pa'rticular seasurement system, which may include radiochemical separation:

4. H s g,

b E

V 2.22 x 108 Y

exp (-Aat)

Where:

_ t.LD = the "a priori" lower limit of detection (microcurie per unit mass or volume),

b = the standard deviation of the background counting reta or of the s

/'

counti ig rate of a blank sample as appropriate (counts per minute),

E = the counting efficiency (counts per disintegration),

V = the sample size (units of mass or valuee),

2.22 x los = the number of disintegrations per minuta per :sicrocurie, Y = the fractional radiochemical yield, when applicable, A =J,he radioactive decay constant for the particular radionuclide (sec-p),and at a he elapsed time between the midpoint of sample collection and time co nting (sec).

~ Typical values of E, V, Y, and at should be used in the calculation.

It'should be recognized that the LLD is defined as an a Driori (before the fact) limit representing the capability of a measureceiit systee and not as an a posterioe_t (4f ter the fact) limit for a particular seasurement.

I I

V0GTLE UNITS - 1 & 2 3/4 11-10

TABLE 3.17-1 RADIOLOGICAL ENVIR000 ENTAL MONITORING PROGRAM-h{

NUMBER Of REPRESENTATIVE g

EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY gg)

AND/OR SAMPLE SAfrLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS 1.

Direct Radiation (2) Thirty-six routine monitoring quarterly.

Gamune dose guarterly.

P stations either with two or more y

dostosters or with one instrument for measuring and recording dose i

rate continuously, placed as follows:

+

P An inner ring of stations, one in each meteorological sector in the general area of the SITE BOUNDARY; An outer ring of stattens, one in g

each meteorological sector in the 6 alle range free-the site;

~S and

{

The balance of the stattens i

to be placed in special Interest g

areas such as population centers.

nearby residences, schools, and In one er two areas to serve as control stations.

g e

'O_-

J

~

TABLE 3.12-1 (Continued)

RADIOLOGICAL ENVIR0tetENTAL MONITORING Pit 0 GRAM Oh

'NUPSER OF l

d REPRESENTATIVE h

EXPOSURE PATHWAY SAMPLES AND Sm!M m M M FRWU AND/OR SAMPLE SAMLE LOCATIONS (g)

COLLECTION FREQUENCY CF ANALYSIS P

2.

Airborne 1

p

, Rad _f oi ne and Samples from five locations Continuous sampler oper-Radiolodine Cannister:

l Partic lates atton with sample collec-I-131 analysis weekly.

I l

tion weekly, or more J

Three samples from close to frequently 1f required by i

i the ti.ree SITE BOUNDARY dust loading.

Particulate Sampler:

h locations, in different l

sectors; P Gross beta radioactivity analysis following filter change;I and gamma Isotopft. analysis (

I of composite (by location) l M

One sagle from the vicinity quarterly.

g of a community having the l

highest calculated annual g

average groundlevel D/Q; and l

One sample from a control M

location, as for example a

.I population center 10 to 20 miles distant and in the i

least prevalent wind direction.

l 3.

Waterborne Surfac=I I One sample upstream Composite sample over Gamma isotopic analysis a.

One sample downstream 1-month period.I I a erly.

l t

s l

l

TABLE 3.12-1 (Continued)

O RADIOLOGICAL ENVIR0f9 ENTAL MONITORING PROGRAM 7d MUMDER OF hi REPRESENTATIVE EXPOSURE PATlWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY II)

F AND/OR SAMPLE SAMPLE LOCATIONS COLLECTIOM FRFQUENCY OF ANALYSIS 3.

Waterborne (Continued)

M b.

Drfnking Two samples at each of one to Composite sample of I-131 analysis on each i

three of the nearest water river water near intake sample when the dose treatment plants that could be at each water treatment calculated for the con-

~

affected by its discharge.

plant over 2-week sumption of the water 4

rid (G) whien I-131 8"* 'I

mP es at a contml

,g analysis is perforleed, per year Composite monthly composite other-for gross beta g amma wise; and grab sample of isotopic analyses finished water at each monthly. Composite for water treatment plant tritium analysis quarterly.

g

~

every 2 weeks or monthly, as approprlate.

-P-

. Sedine One sample from downstream area Sealannually.

Gamma isotopic analysis Trom with existing or potential sentennually.

s R2 Sho Ine recreational value.

6 4.

Inges(1 a.

Milk Samples from allking animals Semimonthly.

Gamusa Isotopic **'}

I In three locations within 3 analysts semimonthly.

miles distance having the highest dose potential.

If there are none, then one sample from allking animals (8) in each of three areas between 3 and 5 miles distance where doses are calculated to be greater than 1 arem per yr.N}

l l

J f

n v

1

g TA9tE 3.12-1 (Continued) j RADIOLOGICAL ENVIRoletENTAL MONITORING PROGRAM j

e

'It REPRESENTATIVE EXPOSURE PATNWAY SAMPLES AND SAMPLING AND TYPE MID FREQUENCY II}

9 AlW/0R SAPPLE SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS i

GE 4.

Ingestion (Continued) d a.

Milk One sample from milking

~

'{

animals (s) at a control loca-q tien about 10 miles distent b

or beyond and preferably in j

a wind direction of lower j

prevalence.

1 b.

Fish At lesst one sample of any Sealannually.

Gamma isetopic analysis (4) l M

commercially and recreationally on edible portions.

4::

Important species la vicinity of plant discharge area.

~D At least one sample of any species in areas not inflhenced by plant discharge.

i N

At least one sample of any During spring spawning C==== isotopic analyses anadronous species in vicinity season.

on edible portion.

j of plant discharge.

I Crass or One sample from two onsite Monthly during Gamme isotopic **'I j

Leafy locations near the SITE growing season.

,Veget ten BOUIGnRY In different sectors.

4

}

One sample from a control Monthly during r-1st, topic location at about 15 miles growing season.

distance.

i

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rep 0,

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TABLE 4.12-1 OETECTION CAPABILITIES FOR ENVIR000 ENTAL SAffLE ANALYSI5(1) (2)

LOWER LIMIT OF DETECTION (LLO)I '

s g{

GRASS OR LEAFY MTER AIR 80RNE PARTICULATE FISH MILK E iTATION SEDINENT J

ANALYSIS (pCl/1)

OR GASES (pC1/m )

(pCl/kg, wet) (pCl/I)

(pCl/kg, wet) (pC1/kg, dry) s di I

Gross seta 4

0.01 4

N-3 2000*

D Mn-54 15 130 Fe-59 30 260 Co-58 15 130 g

Co-60 15 130 Zn-65 30 260 Zr-95 30 m-95 15 O

I-131 1**

0.07 1

60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Be-140 60 60 La-140 15 15 "If no drinking water pathway exists, a value of 3000 pCf/1 may be used.

- If no drinking water pathway exists, a value of 15 pCl/l may be used.

t.)

l

BASES FOR SECTIONS 3.0 AND 4.0 LIMITING CON 0!TIONS FOR OPEAATION AND

$URVEILLANCE REQll!REMENTS t

4 4

1 J

a 4

)

@TLEUNITS-1&2 Q 3/4 0-4 h

E3 The BASES contained in succeeding pages summarize

.the reasons for the Specifications in Sections 3.0 and 4.0, but in accordance with 10 CFR 50.36 are net part of these Technical Specifications.

i

)

9 i

1 s

G

[Lf, UNITS-1&

i

3/4.0 APPLICABILITY BASES The specifications of this section provide the general requirements app' icable to each of the Limitting Conditions for Operation and Surveillance 1equiremen ur.

within Section 3/4.

In the event of a disagreement between the requirements statec in these Technical Specifications and those stated in an applicable Federal I agula-tion or Act, the requirtuents stated in the applicable Federal Regulation or Act i

shall take precedence and shall be set.

3.0.1 This specification defines the applicability of each specificati< n in terms of defined OPEMTIONAL MODES or other specified conditions and is prov' ded te delineate specifically when each specification is applicable.

3.0.2 This specification defines those conditions necessary to constitute coe ance with the terms of an individual Limiting Condition por Operation and as: ociate ACTION requinment.

3.0.3 The specification delineates the neasures to be taken for those ci rcum-stances not directly provided for in the ACTION statements and whose occurrence wou violate ;he intent of a specification.

For example, Specification 3.5.2 requires t-independent ECCS subsystems to be OPERA 8LE and provides explicit ACTION nqu ineent one ECCS subsystem is inoperable.

Under the requirements of Specification 3 0.3, i both the Mquired ECCS subsystems an inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months measures must be in ated to place the unit in at least HOT STANOBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in at 1 HOT SHUTD0%N within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

As a further example, Specificat ion 3.6 requires two Containment Spray Systems to be OPERABLE and provides explicit ACTION requirements if one Spray System is inoperable.

Under the requirements of Specifi' tion 3.0.3, if both the requind Containment Spray Systems are inoperable, within seasures must be initiated to place the unit in at least HOT STAN08Y within 2te nex 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , in at least HOT SHUTDC%N within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD iHUTCCV within the subsecuent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

It is acceptable to initiate and complete a reduct in OPERATIONAL MODES in a shorter time interval than requind in the ACTION statese-and to add the unused portion of this allowable out of-service time to that srovice-for operation in subsequent lower OPERATION M00E(5).

Stated allowable out-o F servic times are applicable ngardless of the OPERATIONAL M00E(S) in whi:n the inope nbili-is discovered but the times provided for achieving 4 mode reduction are not applica: l if the inoperability is discovered in a mode lower than the applicable mode.

For a ple if the Containment Spray System was discovered to be inoperable while in START 1b the ACTION Statement would allow up to 156 houn to achieve COLD SHUTOOWN.

If HOT STAN08Y is attained in 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months rather than the allowed 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months ,140 hours would s-be available befon the plant would be required to be in COLD SHUTOC%N However, if this systas was discovered to be inoperable while in HOT STAN08Y, the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provi<

to achieve HOT STANOBY would not be additive to the time available to achieve COLD SHUTDOVH so that the total allowable time is reduced free 156 hours0.00181 days 0.0433 hours 2.579365e-4 weeks 5.9358e-5 months to 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months .

3.0.4 This specification provides that entry tre.a an CPERATIONAL N00E or othe specified applicability condition must be sade with:

(1) the full complement of nquired systems, equipment, or components OPERA 8LE and (2) all other paramters as specified in the Limiting Conditions for Operation being met without regard for allowable deviations and out-of-service provisions contained in the ACTION t tatamen' The intent of this provision is to ensun that facility operation is nc t initi<

with either required equipment or systees inoperable or other specified lirai ts beint exceeded.

1

,d V0GTLE UNITS - 1 & 2 B 3/4 0-1 i

REACTI ITY COMTROL SYSTEMS i:%

D0; l00 QCdLCCS (lbSY lhj 07;(N0 don Q

8ASES (ini g).

803ATION SYSTEMS (Continu #)

d MARGIN from expected ope ting conditions as defined by Specification 3/4.1.1.1 (MODES 1 and 2) and Speci teation 3/4.1.1.2 (MODES 3 and 4) ef ter xenon decsy and cooldown to 200*F.

he maximum expected boration capability requirement occurs at E0L from full ower equilibrium xenon cenditions and requins 31740 gallons usable v 1.

of 7000 ppe borated water from the boric acid storage tanks or-0700 v.11 m usable volume of M 0 m borated water from 7

the refueling water storage tank (RWST).

dM ppm Gind,6CCO ppm With the RCS temperature below 200'F, one Boron Injection System isbN acceptable without single failu n consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive nactivity changes in the event the single Baron Injection Systas becomes inoperable, g

g g

The beration capability requind 200*F is sufficient to idI SHtITDOWN MAAGIN as defined by Specification

.1.2 (M00E 5) after xenon decay and cocidown from 200*F to 140*F. This co tion requires either 4570 gallons usable volume of 7000 ppe berated wat - free the boric acid storage tanks or 30 gelhM usable volme of 20.0 99 berated water from the CAbabgcdlem OJ.ni+1), lA 6303c4 ens (_ltd+a)

RWsT.

9 The contained water volume limits provided in Specifications 3/4.1.2.5 l'

and 3/4.1.2.6 include allowance for water not available because of discharge s

line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensun a pH value of between 8.5 and 10.5 for the solution recirculated.

within containment after a LOCA.

This pH band minimizes the ivolution of iodine and einimizes the effect of chloride and caustic stress corrosion on mechanical systees and components.

The OPERASILITY of one Boeon Injection System during REF1)f. LING ensures

, that this systas is available for reactivity control while in MODE 6.

3/4.1.3 MOVASLI CONTROL AssDel!!5 The specifications of this sect. ion are necessary to ensure that the follow-ing requirements an met at all times during normal operation.

By observing that the RCCAs are positioned above their respective insertion limits during normal operation, 1.

At any time in life for MODE 1 and 2 operation, the einimum SHUTDCWN l

MARGIN will be maintained.

For operational MODES 3, 4, 5, and 6, the nectivity condition consistent with other specifications will be sain-tained with all RCCAs fully inserted by observing that the boron concentra '

tion is always gnator than an appropriate minisua value.

2.

During norsal operation the enthalpy rise hot enannel facto, F w 11 be saintained within acceptable limits,

,j V0GTLE lJN!TS

1 & 2 5 3/4 1 3 l

/

POWER O!STA!BUTION LIMITS

~

BASES'"

AXIAL FLUX OIFFERENCE (Continued)

Although it is intended that the plant will be operated with the AFD within the target band required by Specification 3.2.1 about the target flux difference, during rapid plant THERMAL POWER reductions, control rod motion will cause the AFD to deviate outside of the target band at reduced THERMAL POWER levels. This deviation will not affect the xenon redistribution suffi-ciently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POWER (with the AFD within the target band) provided the time duracion of the deviation is limited.

Accordingly, a 1, hour penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limita of Figure 3.2-1 while at THERMAL POWER levels between 50% and 90% of RATED THERMAL POWER.

For THERMAL POWER levels between 13% and 50% of RATED THERMAL POWER, deviations of the AFD outside of the target band are less significant.

The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alars.

The computar deter-eines the 1-minuta average of each of the OPERA 8LE axcore detector outputs and provides an alan message immediately if the AFD for two or non 0# ERA 8LE excore channels an outside the target band and the THERNAL POWER is greater than 90% of RATED THERMAL POWER.

During operation at THERMAL POWER levels between 50% and,9C% and between 13% and 50% RATED THERMAL POWER, the computer outputs an alars message when the penalty deviation accumulates beyond the liatts of I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

3/4.2.2 and 3/4.2.3 MUX HOT CHANNEL FACTOR and NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR F

The limits on by f

hot channel factor and nuclear.enthalpy rise hot channel factor ensun '

t:

(1) the design limits on peak local power density and minimus OMBR are not exceeded and (2) in the event of a LOCA the peak fuel clad temperatun will not exceed the 2200*F ECCS acceptance criteria limit.

Each of these is seasurable but will norw11y only be detenined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic l

surveillance is sufficient to ensure that the limits are maintained provided:

1 a.

Control rods in a single group move together with no individual rod insertion differing by more than 212 steps, indicatsd, from the group demand position; b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; V0GTLE UNITS 1 & 2 8 3/4 2 2

1.00 I

1 I

q. J 0.90 g

I f

1 I

- ? =,(

4 0.80 --

3 ~;

I I

r

)

J l

I o.70 l

g

=

g I

TARGET FLUX l

DIFFERENCE

.J j

0.60 l

i 5

I E

0.50 l

a:

1 u.

I I

2 0.40 9

\\

l 0

1 a:

l 0.30 I

I r

1 0.20 l

l 0.10 l

I I

o

-30%

-20%

-104 0

+10%

+20%

+30%

INDICATED AXlAL FLUX DIFFERENCE FIGURE B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER V0GTLE UNITS - 1 & 2 B 3/4 2-3

INSTRLHENTATION 8ASES REACTOR TRIP SYSTEM and ENGINEEREO SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)

(7) steam line isolation, (8) turbine trip (9) auxiliary feedwater pumps start and automatic valves position, (10) containment fan coolers start and automatic valves position, (u) Nuclear Service Cooling and Component Cooling water pumps start.nd autoestic valves position, and (12) Control Roos Ventilation Energency Actuation Systems start.

The Engineered safety Features Actuation Systes interlocks perform the following functions:

P-4 Reactor tripped - Actuatas Turbine trip, closes main feedwater-valves on T,yg below setpoint, prevents the opening of the main feedwater valves which won closed by a safety Injection or High Staan Generator Water Level signal, allows Safety Injection block so that components can be nset or tripped.

Reactor not tripped - prevents manual block of Safety Injection.

P-u With pressurizer pnssure below the P-U setpoint, allows manual block of safety injection actuation on low pressurizer pressure signal.

Allows manual block of safety injection actuation and steam

.. "j line isolation on low coeoensated steam line pressun signal and allows staas line isolation on high steam line negative pressure rata.

With pressurizer pressure abocy the P-u setpoint, defeats sanual block of safety injection actuation on low pressurizer pres-sure and safety injection and staas line isolation on low stsaa line pressure and defeats steam line isolation on high staas line negative pressure rata.

P-14 On inenasing staan generator water level, P-14 automatically trips all feedwater isolation valves, initiatas a turbine trip, and g

inhibits feedwater control valve modulation.

(($:(4EL.'3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADTATION MONITORING FOR PLANT OPERATIONS The OPEAASILITY of the radiation acnitoring instrumentation for plant i

operations ensu ns that:

(1) the associated action will be initiated when the radiation level monitored by each channel or comeination thereof naches its setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out-of service for testing or maintenance.

The radiation monitors for plant operations senses I

radiation levels in selected plant systans and locatione and determines whether or not predetenined limits are being exceeded.

If they are, the signals are comeined into logic matrices sensitive to coscinations indicative of various

^

accidents and abnormal conditions.

Once the required logic r.ooeination is V

completed. the systes sends actuation sigr.als to initiate alarus or automatic isolation action and actuation cf Leergency Exhaust or Ventilation Systems.

V0GTLE UNITS 1 & 2.

5 3/4 3-3

l t

INSERT AL I

t The Source Range High Flu'x at Shutdown Alarm Setpoint is an analysis assumption for mitigation of a Boron Dilution Event in MODES 3, 4, and 5.

4 i-I

{

1 l

l I

l t

]

i a

i l

t 1r 1

I l

1 4

t 1

b i

i i

1 t

i

INSTRLHENTATION BASE $

=

3/4.3.3.2 MOVA8LE INCORE DETECTORS The OPERA 8ILITY of the novable incere detectors with the specified minimum complement of equipment ensuns that the seasurements obtained from use of this system accurately npresent the spatial neutron flux distribution of the core. The OPERA 8ILITY of this system is demonstrated by irradiating each detector used and detemining the acceptability of its voltage curve.

For the purpose of measuring F (Z) or Fh a full incore flux sap is used.

q Quarter-con flux 44ps, as defined in WCAP-8648, June 1976, may be used in ncalibration of the Excon Neutron Flux Detection System, and full incere flux naps or synastric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable.

3/4.3.3.3 $EISMIC INSTRUMENTATION The OPERASILITY of the seismic instrumentation ensuns that sufficient capability is available to promptly detemine the magnitude of a seismic event and evaluata the response of those featuns important to safety.

This capa-bility is requind to permit comparison of the sensured response to that used in the design basis for the facility to datamine if ant shutdown is required pursuant to Appendix A of 10-UJ_ic bttumentatig acerresponding Technica lart 100.

The i station on Unit 1 is shand with Unit 2 anWseisa Specifications meet teq W esenta of.Re. latoW. Guide

'.12, Revision 1, April 1974.

mngg 3/4.3.3.4 NETEOROLOGICAL INSTRLW A

The OPERASILITY of the meteorological instrumentation snsures that sufficient meteorological data are available for estinating potential radiation doses to the public as a result of routine or accidental release of radioactive saterials to the atmosphere.

This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistant with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs,' February 1972.

3/4.3.3.5 RD40TE $HUTDOWN SY$ TEM The OPERASILITY of the Remota $hutdown Systes ensuns that sufficient capability is available to pruit safe shutdown of the facility free locations outside of the control roce.

This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of 10 CFR Part 50.

The OPERA 8!LITY of the Reeute Shutdown System ensures that a fire will not preclude achieving safe shutdown.

The Remote Shutdown System instrumentation, V0GTLE UNITS - 1 & 2 3 3/4 3 4

/

e REACTOR COOLANT SYSTEM BASES SPECIFIC ACTIVITY (Continued)

Based upon the above considerations for excluding certain radionuclides from the' sample coepleting, analysis, the allowable time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months between sample taking and the initial analysis is based upon a typical time necessary to per-a fore the sampling, transport the sample, and perfore the analysis of about 90 minutes. After 90 minutes, the specific count should be made for gases (i.e., xenons and kryptons) and particulates (i.e., cobalt and casiums) in a nproducible geometry of sample and counter having reproducible beta or gama self-shielding properties.

The counter should be reset to a reproducible efficiency versus energy.

It is not necessary to identify specific nuclides.

The radiochemical determination of nuclides should be based on multiple count-ing of the sample within typical counting basis following sampling of less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , about 2 houn, about 1 day, about I week, and about 1 month.

The identification of 95% of the grgss specific activity by definition does not obligata VEGP into calculating E every time gross activity is determined.

Reduciqq 1ess than 500*F prevents the release of activity should a stana gene rupture since the saturation pressure of the reactor coolant is bel lift pnssun of the atmospheric steam relief valves.

The Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to take corrective action.

A nduction in frequency of isotopic analyses following power changes may be persissible if justified by the data obtained.

PRES $URE/TD4PERATtJRE LIMITS kTe % hvised-Te-Aef4eet-UMt

[

3/4,4.9 The temperature and pressun changes during heatup and cooldown are limited to be consistant with the nquineents given in the ASNE Boiler and Pressure Yusel Code, Section !!!, Appendix G:

I 1.

The nactor coolant temperature and pressure ud systes heatup and cooldown ratop (with the exception of the pressurizer) shall be limited in accordance withtSW:;-3 2-2 d ' 4-3 for the service period specified therson:

^\\

a.

Allowable combinations of pnssun and temperature for specific temperature change rates are below and to the right of the limit Q

lines shown.

Limit lines for cooldown rates between those presented Qce,lMgft M\\.

say be obtained by interpolation; and Figuresk.'-2 ed-3r+4 define lietts to assure prevention of I

non-ductile failure onl For nonsal aperation, other inherent plant characteristics, e.g., y. pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be I

achieved over ceruin pressure-temperature ranges, i

i VCGTLE UNITS - 1 & 2 8 3/4 4-7

(

O INSERT AM Figures 3.4 2a and 3.4-3a (l' nit 1)

Figures 3.4-2b and 3.4-3b (Unit 2) l 9

f P

e I

I i

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

REACTOR COOLANT SYSTEM to,.

si.

n,.....a...,

.a.

..... e.. n..... u..... y PRESSURE / TEMPERATURE LIMITS (Continued) 2.

These limit lines shall be calculated periodically using methods provided

below, 3.

The secondary side of the steam generator sust not be pressurized above 200 psig if the temperature of the steam generator is below 70*F, 4.

The pressurizer heatup and cooltiown rates shall not exceed 100*F/h and 200*F/h, respectively.

The auxiliary spray shall not be used if the temperature difference between the pressurizer and the auxiliary spray fluid is greater than 625'F, and 5.

System preservice hydrotests and inservice leak and hydrotests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

The fracture toughness properties of thJ ferritic materials in the reactor vessel are detemined in accordance with the NRC Standard Review Plan, ASTM E185-82, and in accordance with additional nactor vessel nquirements.

These properties are then evaluated in accordance with Appendix G of the 1972 Summer Addenda to Section III of the A!ME Boiler and Pnssure vessel Code and the calculation methods described in WCAP-7924 A, "Basis for Heatup and Cooldown Limit Curves," April 1975.

nFqurds3/-2Md.4-hre co Ido M li t c rves peat a

ChE app i Je Vo le-nit 1 fo up o1 EFP.

e st 1si ng teri it h

an ini al of 3 F nd co er ente of.06 ople-g s'r g

Ugg-ere s t)ie h, an coo cury si Fi re

.4 a

3.4 ar ba d.n t}41{g of K'F coppercpnte i

o 0.

WT

/

Heatup and cooldown limit curves are calculated using the most limiting val WWif.ygfgrence temperature, ATMT, at the end of 1 gfective ull gewer years fEFPY) of service life.

The 14 EFPY service I tagefod t~ chos Qst W U a limiting AT at the 1/4T location in MT the core ngion s greater than the ATM T 'I th' II*III"i ""I'#8dI*t'.d ****'I*I*

The selection of such a limiting RT assures that all components in the ET Reactor Coolant System will be operated conservatively in accordance with applicable Code requirements.

.g g The reactor vessel materials have been tested to determine their initial RTNOT; the results of these tests are shownVin Table B 3/4.4-Reactor opera-tion and neultant fast neutron (E gnator than 1 MeV) teradi tion can cause an incnese in the RT Thenfore, an adjusted reference esperature, based MT.

upon the fluence, copper content, and phosphorus content of the saterial in question, can be predicted using Figure B 3/4.41 and the argest value of aATNOT computed by either Regulatory Guide 1.99, Revisio 1, "Effects of O C fy] h V0GTLE UNITS - 1 & 2 5 3/4 4-4 4.

respechvel 3

l INSERT AN The heatup and cooldown limit curves shown in Figures 3.4-2a and 3.4-3a are applicable to Unit 1 for up to 16 EFPY and are based on Westinghouse-developed generic curves which were developed assuming a 40'F initial RTNDT and a copper content of 0.10 WT5 for the most limiting material.

These curves are applicable to Unit 1 since its most limiting material (Table 8 3/4.4-la) has both a lower initial RTNDT (30'F) and a lower copper content (0.05 WTX). These curves, however, are not applicable to Unit 2, since its most limiting material (Table B 3/4.4-1b) has a higher initial RTNDT (50 compared to 10'F).

Separate heatup and cooldown 4

limit curves were developed based on the actual material properties of the :nost limiting material for Unit 2 up to 16 EFPY.

The Unit 2 curves are shown in Figures 3.4-2b and 3.4-3b.

4 I

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4 1

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I 4

TABLE 8 3/4.4-ILL

<8

-f GOLT1 REACTOR VESSEL TOUGNNESS T

g ASME ypg-50 FT-LS

,T

.mmy p f, q

Caer MATERIAL CU Ni P

W 35 MIL ggy sa CGMP 00s[MI C00E TYPE M

M M

M TEIF #4)

(4)

- )

FT-TE)

- if ll e

Closure Need Dome See87-1 A5330CL1

.16

.67

.006 -50 75 15 48 e-Closure Need Terus 80808-1 A5330CL1

.14

.56

. Ole -30 68 8

85 Closure Mead Flames 30001-1 A500CL2

.70

. ell 20

<40 20 132

=

Vesset flange 30002-1 A58KL2

.71

.07.4 0

<60 0

119 Inlet Nozzle 80009-1 A500CL2

.46

.011 -20

<10

-20 107 Inlet Mezzle 30009-2 A500CL2

.84

.014 -10

<50

-10 95 Inlet Mezzle BM09-3 A508CL2

.82

.013 -18

<10

-10 117 Inlet Mozzle 88009-4 A50GCL2

.87

.014 -20

<10

-20 105 Outlet Mezzle Seele-1 asaart 2

.82

.006 -10

<50

-10

>124 Outlet Mezzle Seele-2 asaart 2

.79

.006 -10

<50

-10

>100 Outlet Nozzle 84814-3 A500CL2

.77

.006 -10

<50

-10

>102 a=

y Outlet Dzzle 38016.4 A500CL2

.80

.006 -10

<10

-10

>75 Nozzle h il Bee 64-1 A5330CL1

.14

.62

.611 -le 88 28 94 Nozzle h 11 30004-2 A5330CL1

.10

.58

.006 -40 75 15 104 i

am Mezzle h il 88004-3 A5330CL1

.14

.69

.013 -30 100 40 92 Inter. h 11 30005-1 A5330CL1

.08

.59

. 004 8

60 0

90 g

p Inter. h 11 84405-2 A5330CL1

.08

.59

.004 -10 80 20 100 e

Inter. h 11 80005-3 A5330CL1

.06

.60

.003 -20 90 30 107 o

g tower hil 34606-1 A5330CLI

.05

.59

.005 -50 80 20 116 tower h il wa606-2 A5330CL1

.05

.58

.009 -14 80 20 113 a

tower h 11 30606-3

. A5330CL1

.06

.64

.007 -20 70 10 118 Betten Need Terus 88813-1 A531ers.1

.13

.50

.009 -40 50

-10 88 g

Betten Need Some 34412-1 as11erL1

.10

.53

.009 -48 32

-28 122 Inter & teuer h 11 G1.43 SAW

.03

.10

.007 -80

<-20

-80

>129

_c Vertical Wald Seams and Girth y

km 3

e jh1 to major working directions

[

"deM J1sg structiser Limi4inc Mcdeirin1.

[,

-it(pec skelE eneg

IA8tE 8 3/4.4-lb l

imIT 2 REACTOR VESSEL TOUGHNESS A5ME l

1

<:8 Comp.

Material Cu Ni P

T RT 888 8

agy NOT i

l ra-panent Code Type 111 111 111 1*F1

[*F)

(ft-Ib)

E Closure head dame R9-1 A5338 Cl. 1 0.07 0.61 0.000

-40

-30 123 y

Closure head tores R10-1 A533s C1. 1 0.07 0.64 0.010

-30 0

84 Closure head flange R7-1 A508 C1. 2 0.72 0.011 10 10 130 l

Vessel flange Al-1 A500 C1. 2 0.s7 0.011

-60

-60 115 Inlet nozzle 99006-1 A506 Cl. 2 0.07 0.84 0.010

-50

!!9 N

Inlet nozzle

. 39006-2 A508 C1. 2 0.M 0.83 0.009

-40

-40 128 Inlet nozzle RS-1 A508 Cl. 2 0.09 0.87 0.000

-20

-20 147 I

Inlet nozzle RS-2 A500 Cl. 2 0.08 0.85 0.009

-20

-20 134 Outlet nozzle A6-3 A508 C1. 4 0.69 0.011

-10

-10 122 Outlet nozzle R6-4 A508 C1. 2 0.66 0.010

-10

-10 140 Outlet nozzle 99007-3 A508 C1. 2 0.66 0.005

-30

-30 114 5

Outlet nozzle 99007-4 A508 C1. 2 0.64 0.010 10 10 132 R

Nozzle shell R3-1 A5338 C1. 1 0.20 0.67 0.015 0

20 79 Nozzle shell R3-2 A5338 C1. 1 0.20 0.67 0.015 0

40 79 Nozzle shell R3-3 A5338 C1. 1 4.15 0.62 0.010

-10 60 84 y

Intenmediate shell 24-1 A5338 C1. 1 0.06 0.64 0.009

-20 10 95 Intenmediate shell R4-2 A5338 C1. 1 0.05

'O.62 0.009

-10 10 104 latenmediate stsli R4-3 A5338 Cl. 1 0.05 0.59 0.009 0

30 84 Lauer shell 80825-1 A5338 Cl. 1 0.05 0.59 0.006

-20 40 83 Lower shs11 RS-1 A5338 C1. 1 0.06 0.62 0.007

-20 40 SF Lower shell 80628-1 A5338 Cl. 1 0.05 0.59 0.007

-20 50 85 Bottom !sead torus R12-1 A533s Cl. 1 0.17 0.64 0.012

-20

-20 89 Sottom head dame Ril-1 A5338 Cl. 1 0.10 0.62 0.008

-30

-30 115 latermediate and fouer Gl.60 SAW 0.07 0.13 0.007

-10

-19 147 shell vertical weld seams l

Intermediate to louer E3.'43 5AW 0.06 0.12 0.007

-50

-30 90 shell girth weld seam 1

4

Upper shelf Energy; sete - numal to major working direction.

- Lielting Material.

i i

REACTOR COOLANT SYSTEM SASES i3:

c -

PRES $URE/ TEMPERATURE LIMITS (Continued) b 'msv+ AO.

Residual Elements on Predicted Radiation Jamage to Reactor Vessel Materials,"

or the Westinghouse Copper Trend Curves snown in Figure B 3/4.4-2.

The heatup

)

and cooldown limit curves of 9-i' : M :.

7RETude predicted adjust-ments for this shift in RT at the end of 16 EFPY as well as adjustments ET j

for possible errors in the pressure and terparature sensing instruments.

t J

(

Values of ART @T detaruined in this manner may be used until the results from the saterial surveillance program, evaluated according to ASTM E185, are available.

Capsules will be removed in accordance with the nquinments of ASTM E185-82 and 10 CFR Part 50,' Appendix H.

The surveillance specimen with-drawal schedule is shown in Table 16.3-3 of the VEGP FSAR.

The lead factor npresents the relationship between the fast neutron flux density at the loca-l tion of the capsule and the inner wall of the nector vessel.

Therefore, the nsults obtained from the surveillance specimens can be used to predict future i

radiation damage to the reactor vessel material by using the lead factor and the withdrawal time of the capsule.

The heatte and cooldown curves must be recalculated when the ART deterni.ned from the surveillance capsule exceeds ET the calculated tRT for the equivalent capsule radiation exposure.

NOT J-Allowahle pressu n-temperature relationships for various heatup and i

cooldown rates are calculated using methods derived fN J Appendix G in Sec-tion III of the ASME Boiler and Fressure Vessel Code as required by Appendix G l

to 10 CFR Part 50, and these methods are discussed in detail in the following paragraphs.

j 4

The general method for calculating heatup and cooldown limit curves is i

l based upon the principles of the linear elastic fractun mechantes (LIFM) tech-l nology.

In the calculation procedures a sesielliptical surface defect with a

[

j depth of one quarter of the wall thickness T, and a length of 3/2T is assumed t

i to exist at the inside of the vessel wall as well as at the outside of the,

(

vessel wall.

The dimensions of this postulated crack, refernd to in Appendix G i

of ASME Section !!! as the reference flaw, amply exceed the current capabilities of inservice inspection techniques.

Therefore, the reactor operation limit cur-i ves developed for this reference crack are conservative and provide sufficient i

safety margins for protection against nonductile failure.

To assun that the l

radiation embrittlement effects an accounted for in the calculation of the i

limit curves, the most limiting value of the nil-ductility refennce tempera-l j

ture RTET, is used and this includes the radiation-induced shift,.iRTNOT' l

corresponding to the end of the period for wnich heatup and cooldown curves l

an genented.

f The A$ME approach for calculating the allowable liait curves for various l

l heatup and cooldown rates specifies that the total stnss intensity factor,

.+

K for the combined thermal and pressure stmsses at any time during heatup g

\\

V0GTLE UNITS - 1 & 2 B 3/4 10 i

i r

INSERT A0 Figures 3.4-2a and 3.4-3a (Unit 1), Figures 3.4-2b and 3.4-3b (Unit 2) e e

6

l 8

6 4

,,,, a SU R F 3.17 x 10

i 2

,,,, = 1/4T 1.78 x 10

/

10

l-f 2

g 6

3

/ /

d W

,,, *3/4T 3.49 x 1018 w

3:

/

o 2

6

[

Sz 10

/

I

/

r r

/

4 2

1 10" 5

10 15 20 25 30 35 EFFECTIVE FULL POWER (years)

FIGURE B 3/4.4-1 UNITS 1 AND 2 FA3T NEUTRON FLUENC.E (E>1MeV) AS A FUNCTION OF FULL POWER SERVICE LIFE

\\

VLGTLE UNITS'- 1 & 2 B 3/4 4-11

[

S T;

3 a*

4#

]

i i

i i I I I

i i i:

i i

i I i a

i i i Uppor tiniit j

em comu nau. ems uto

.oo 1

~

1 j

200 l

i

\\

l Louw Lisnit

~

100

-o e

/

n m o em coma nase.e.m mte M

e.sen cama. ass. emes moo e

)

I i

1 1 11 1

1 I 1111 1

I I

I I I e

aaaI 1

l 10" 2

4 6

8 10 "

2 4

6 8

1038 l

)

FAST NEUTRON FLUENCE (N/CM, E 78 heeV) 2 FIGLT.E B 3/4.4-2 i

UNITS 1 & 2 EFFECT OF FLUENCE AND COPPER CONTENT ON SHIFT OF RTNOT FOR RFACTOR VESSELS EXPOSED TO RADIATION AT 5500F 4

l

(

\\

REACTOR C00LMT SYSTEM 8ASES PRESSURE / TEMPERATURE LIMITS (Continued) the rate of heatup and the time (or coolant temperature) along the heatup ramp.

Furthermore since the thersal stresses at the outside are tens'11e and increase with incre,asing heatup rate, a lower bound curve cannot be defined.

Rather, each heatup rate of interest must be analyzed on an individual basis.

Following-the generation of pressure-temperature curves for both the steady-state and finita heatup rate situations, the final limit curves an produced as follows.

A composite curve is constructed based on a point-by-point comparison of the steady-state and finite heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the thne values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to *.he outside and the pnssure limit must at all times be based on analysis j

of the most critical criterion.

Next, the composita curves for the heatup rate data and the cooldown rata data are adjusted for possible errors in the pressure and temperature sensing instruments by the value indicated on the respective curves, Finally,theneM1 F 0 Appendix G Rule which addresses the metal temperature l

-^

.f :

of the closun head fla go and vessel flange ngions is considered.

This rule states that the minimus metal temperatun of the closun flange regions should be i

at least 120*F higher than the limiti RT for these regions when the pnssure ET exceed: 20 parcant of the reservice hydro t.atic test pressure (621 psig for Westinghouse Plants).

Fo Unit le

' minista temperature of the closun flange # wessel *13 pe

, since the limiting RTMT isp*F(see Spas Tabie a v4-4.

Thnvocieunitih.at, curve shown on Ftgun 3-4.#rs not impacted by new ICCF rule.

Howey Vogtle Unit 1 cooldown curve shown in Figure 3-4 is impac by the new 1 0 rule. A Although e pressurizer operates i erature ranges above those for which then is reason for concern of nonductile failure, operating limits an provided i

to assun compatibility of operation with the fatigue analysis perfomed in accordance with the ASME Code requirements.

COLD OVERPRESSURE PROTICTION SYSTEMS

~

The OPERASILITY of the PORYs, two RHR suction relief valves or an RCS vent capable of n11eving at least 670 gpa water flow at 470 psig ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the ACS cold legs an less than or equal to 350*F.

Either PORY or either RHR suction relief valve has adequata n11eving capability ta protect the RCS free overpressurization when the transient i's limited to either:

(1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F (6

above the RCS cold leg temper? tuns, or (2) the start of all thne charging pumps and subsequent injection into a water-solid RCS.

V0GTLE UNITS - 1 & 2 8 3/4 4-15

INSERT AP For Unit 2, the minimum temperature of the closure flange and vessel flange regions is 130'F. since the limiting RTNOT is 10 F (Table B 3/4.4-1b). The Unit 2 heatup curve shown in Figure 3.4-2b and the cooldown curve shown in Figure 3.4-3b are not impacted by the new 10 CFR 50 rule.

t h

3/4.5 EMERGENCY CORE C0OLING SYSTEMS l

t 8ASES 1

3/4.5.1 ACCUMJLATORS The OPERASILITY of each Reactor Coolant System (RCS) accumulator ensures that a sufficient volume of borated water will be immediately forced into the reactor core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators.

This initial surge of water into the core provides the initial cooling mechanism during large RCS pipe ruptures.

The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulater injection in the safety analysis are net.g in5CN h.

The accumulator power operated isolation valves are considered to be "operating bypasses" in the contaxt of IEEE Std. 279-1971, which nquires that bypasses of a protective function be removed automatically whenever permissive conditions are not net.

In addition, as these accumulator isolation valves fail to meet single failure critaria, removal of power to the valves is requind.

The Ifnits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator

'S which may result in unacceptable peak cladding temperatures.

If a closed

)

isolation valve cannot be immediately opened, the full capability of one accumulator is nct available and prorgt action is requind to place the reactor in a mode where this capitility is not required.

i 3/4.5.2 and 3/4.5.3 ECCS 51185YSTEMS The OPERA 8!LITf of two independent ECCS subsystans ensuns that sufficient emergency core cooling capability will be available in the event of a LOCA

'j assuming the loss of one subsystem through any single failure consideration; Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures i

within acceptable limita for all postulated'bnak sizes ranging from the double ended bMak of the largest RCS cold leg pipe dowriward.

In addition, each ECCS subsystem provides long-tern core cooling capability in the recirculation soie during the accic%nt recovery period.

With the ACS temperature below 350'F, one OPERA 8LE ECCS

esystee is 1

i acceptable without single failure conaideration on the basis o? the stable reactivity condition of the reactor and the Itaited core cooling requirements.

I L

e

,..o V0GTLE UNITS - 1 & 2 B 3/4 5-1

1 i

INSERT AQ The minimum boron concentration must ensure the reactor core will remain I

suberitical during the accumulator injection period of a small-break LOCA.

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EMERGENCY CORE COOLING SYSTEMS SASES - --

ECCS SUBSYSTEMS (Contin 1ed)

The limitatien for all safety injection pumps to be inoperable belcw 350*F provides assurance that a mass addition pressure transient ccn be nifeved by the operation of a single PORV.

The Surveillance Requirements provided to ensure OPERA 81LIN of each component ensures that at a sinfaum, the assumptions used in the sti3ty analyses are met and that subsystes CPERASILITY is maintained.

Surveillance Requirements for thrqtle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA.

Maintenance of proper flow resistance and pressure drop in the piping systee to each injection point is necessary to:

(1) prevent total pump flow from exceeding renout enditions when tt.e system is in its minimum resistance configuration, (2) provi(.a the proper flow split betwen injection points in accordance with the assumptions used in the ECCS-LOCA analyse:, (3) provide' an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses and (4) to ensure that centrifugal i

I charging pump injection flow which is directed through the seal injection path is less than or equal to the amount assumed'in the safety anal O

3/4.5.4 REFUELING WATER STORAGE TANK b l1Ecf4 bU-The OPERMILITY of the Refueling Water Starsge Tank (RWST) as part of the ECCS ensuns that sufficient negative nactivity is injected into the core to counteract any positive increase in reactivity caused by ACS cooldown.

RCS cooldown can be caused by inadvertant depressurization, a loss of-coolant accident. or a steas line ruptu n.

1 The 11af ts on NWST sintaus voltme and boroa concentration ensun that

1) sufficient water is available within containment to permit ncirculation cooling flow to the core, 2) the nacter will remain subcritical in the cold condition following a small LOCA or staaeline b nak, assuming completa sixing I

of the

, RCS, and ECCS water volumes with all control rods inserted except the reactive control assembly (ARI-1), and 3) the reacter will naain stec itical hn the cold condition fellwing a large bnak LOCA (bnak flow 1

0

) afstming completa mixing of the RWST, RCS, ECCS water and other so es of y ter that may eventually Mside in the sump, post-LOCA with all

]

con est reds asstaed to be out.

The contained water volume Itait includes an allowence for water not usable

)

because of tank discharge line location or other physical characteristics.

c 8 0(uAtO 8.s awwo The limits on contained water voltde and boro,n concentration 6f the RWST l

also ensure a pH value of betweCh*r'and 10.5 for the solution recirculated withiri containment after a LOCA.

This pH band minimizes the evolution of iodine and sinimizes the effect of chloride and caustic stnss corrosion on mechanical systees and coeconents.

v l

(

V0GTLE UNITS - 1 & 2 8 3/4 5-2

INSERT AU The surveillance requirements for leakage testing of ECCS check valves ensure a failure of one valve will not cause an intersystem LOCA.

In MODE 3, with either HV-8809A or 8 closed for ECCS check valve leak testing, adequate ECCS flow for core cooling in the event of a LOCA is assured.

t

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FR

~

$E iN3/4.5 CONTA!WENT SY EMS SPECIFICATIONS A

$TI SE atmos EAIC TYPE AINMENT l

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I CCNTAf*ENT SYSTEMS

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3/4.6.1.5 AIR TEMPERATURE i

The 11attations on containment average air temperature ensure that the over-all containment average air temperature does not exceed the initial temperature l

condition assumed in the safety analysis for a steam line bnak accident.

i Measurements shall be made at all listed locations, whether by fixed or port-able instruments, prior to determining the average air tamperature.

3/4.G.I.6 CONTA!*ENT STRUCTURAL INTEGRITY 1

This limitation ensures that the structural integrity o ntainment 1

will be maintained comparable to the original design standard 1tfe of l

the facility.

Structural integrity is mquired to ensure that the containment l

4 will withstand the maximum pnssure of 41.9 psig in the event of a steam line i

h m ak accident.

The seasurement of containment tendon If ft-off force and the l

tensile tests of the tendon stra,nds for Unit 1, and the visual examination of tendons, anchorages and exposed interior ans exterior surfaces of the contain-ment and the Type A leakage test for both units are sufficient to demonstrate this capaht11ty.

(The tendon strand sr.mples will clso be subjected to stros cycling tests and to accelerated corrosion tests to simulate the tarpior.'s 1

operating conditions and environment.)

Unit 1 and Unit 2 containe..ss's isfy the recommendations of Regulatory Guide 1.35 Revision 2, PositP C.1. 3.

j Therefon, Unit 2 containment is subject to visual inspectica, on I

I The Surveillance Requirements for demonstrating the structural integrity of each containment is in compliance with the rosamendations of Revision 2 of Regulatory Guide 1.35. "Inservice Surveillance of Ungrouted Tendons in Pn-

{

stressed Concreta Containment Structures," er.d proposed Regulatory Guide 1.35.1, i

j "Determining Prestmssing Forces for Inspection of Prestressed Concrete Con-tainments," April 1979.

The mquired Special Reporta free any engineering evaluation of contain-ment abnomalities shall inc1'Ae a description of the tandon condition, the condition of the concreta (repecially at tendon anchorages), the inspectior I

procedures, the tolerances on cracking, the results of the engineering evalua-tion, and the corrective actions taken.

t 3/4.6.1.7 CONTA!WENT VENTILATION SYSTEM l

)

The 24-inch containment purge supply and exhaust isolation valves an required to be sealed closed during plarc. operations since these valves have not t

been demonstrated capable of closing during a LOCA or steam line break accident.

[

Maintaining these valves sealed closed during plant operation ensures that exces-sive quantities of radioactive materials will not be released via the Containment Purge Systas. To provide assurance that these containment valves cannot be inad-i vertently opened, the valves are sealed closed in accordance with Standard Review Pl an 6. 2. 4.

Sealed closed isolation valves are. isolation valves under admini-strative control to assun that they cannot be inadvertently opened.

Admini-strative control includes mechanical devices to seal or lock the valve closed.

l the use of blind flanges, or removal of power to the valve operator.

l V0GTLE UNITS - 1 & 2 8 3/4 6-2 t

I

CONTAINMENT SYSTEMS 8ASES CO,NTAI*ENT VENTILATION SYSTEM (Continued)

The use of the containment

purge lines is restricted to the 14-inch purge supply and exhaust isolation valves since, unlike the 24-inch valves, the 14-inch valves are capable of closing during a LOCA or steam line break accident. There-fon, the SITE BOUNDARY dose guideline of 10 CFR Part 100 would not be exceeded in the event of an accident during containment PURGING operation.

Only safety-related masons; e.g., containment pressure control or the reduction of air-borne radioactivity to facilitate personnel access for surveillance and main-tenance activities, should be used to justify the opening of these isolation valves.

Leakage integrity tasts with a maximum allowable leakage rate for containment purge supply and exhaust supply va*ves will provide early indication of resilient material seal degradation and will allow opportunity for npair before gross leak-age failures could develop.

The 0.60 L leakage limit of Specification 3.6.1.2b.

shall not be exceeded when the leakage fates detamined by the leakage integrity

'a ts of these valves are added to the previously determined total for all valves and permtrations subject to Type B and C tests.

3/4.6.2 OEPRES$URIZATION AND COOLING SYSTDtS 3/4.6.2.1 CONTAINNENT SPRAY SYSTEM The OPERA 8ILITY of the Containment Spray Systes ensuns that containment depressurization and cooling capas111ty will be available in the event of a LOCA or steaa line break.

The pressure reduction and resultant lower containment leakage rata are consistant with the assumptions used in the safety analyses.

The Contairment Spray Systas and the Containment Cooling System both provide post-accident cooling of the containment atmosphere.

However, the Containment $ prey Systas also provides a mechanisa for removing iodine free the containment atmosphere and therefore the time requirements for restoring an inoperable Spray System to CPERABLE stat 9s have been natntained consistent with that assigned other inoperable ESF eqt.ipment.

3/4.6.2.2 SPRAY ADOITIVf $YSTEM 8.0 Mi+ A 8.MUmti The OPERA 4!LITY of the Spray Additive Systen sures that sufficient Na0H is added to the containment spray in the event of a L The limits on Na0H volume and concentration ensure a pH value of between drand 10.5 for the solution recirculated within cuntainment after a LOCA.

This pH band ainimizes the evolution of todine and minimizes the effect of chloride and caustic stress corrosion on mechanical systema and components.

The solution volume limits (3700-4000 gallons) represent the nquired solution to be delivered (i.e., the delivered solution volume is that volume above the tank discharge)

These assumptions are consistent with the iodine removal efficiency assumed ir the safety analyses.

V0GTLE UNITS - 1 & 2 8 3/4 6-3 i

)]4. 7 PLANT SYSTEMS BASES 3/4.7.1 TUR9fNE CYCLE 3/4.7.1.1 SAFETY VALVES The OPERA 8!LITY of the main steam line Code safety valves ensures that the secondary System pnssun will be Itaited to within 1105 (1304 psig) of its design pressure of*1145 psis during the most severe anticipated systes operational transient.

The maximum n11eving capacity is associated with a Turbine trip from 1005 AATED THERMAL POWER coincident with an assumed loss of condenser heat sink (i.e., no staan bypass to the condenser).

The specified valve lift settings and n11eving capacities an in accordance with the requirements of Section !!! of the ASME Boiler and Pressure Code, 1974 Edition.

The total relieving capa:ity for all valves on all of the staan lines is 18,607,220 lbs/h which is 1235 of the total secondary staan flow r

of 15,135,453 lbs/h at 1005 AATED THERMAL POWEt, A mintaus of two CPERA4LE safety valves per steen generatar ensures that sufficient relieving capacity is available for the allowable THERMAL POWEA restriction in Table 3.7-1.

i STARTUP and/or POWEA CPERATICN is allowable with safety valves inoperable within the limitations of the ACTION Mquirements on the basis af the reduction in Secondary Coolant Systaa staae flow and THERMAL POWER required by the i

reduced Reactor trip settings of the Power Range Neutron Flux channels.

The

{

Reactor Trip Setpoint redt.ctions an derived on the following basis:

For four loop operation

$P=0){

} x (109) i l

Where:

l P=

educed Reactor Trip setpoint in percent of RATED THERMAL

f0WEA, V =Paxista nLabor of inoperable safety valves per steam line, i

l 109 =

Power Range Neutron Flux-High Trip $etpoint for four loop operation, i

X =

Tot,a1 relieving capacity of all safety valves per staas line in 1bs/ hour, and MaximusCPebeing capacity of any one safety valve in Y =

i Ibs/ 'ur, V0GTLE UNITS - 1 & 2 5 3/4 7-1 i

L

PLANT SYSTC45 O

s.

O!!st, Gl!NhTOR BUILG!NG AND AUXILIARY Ftt0 WATER PUN 3/4.7.13

(

MVAC nV57EM5 hoarabof the diesel generator building and auxiliary feedwater pumphouse ESF HVAC systees ensures that the ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment served by these systems.

l l

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i V0GTLE UNIT 5 1 & 2 8 3/4 7-7 i

$3 3/4.8 ELECTRfCAL PCVER SYSTEMS O

SASES 3/4.4.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES. 0.c. SOURCES and ON3fTE PCVER DISTRIBUTION The OPERA 81 LIT'r of the A.C. and 0.C power sources and associated distribu-tion systems during operation ensuns that sufficient power will be available to supply the safety-related equipeent required fort (1) the safe shutdown of the facility, and (2) the sitigation and control of accident conditions within 3

the facility. The mininua specified independent and redundant A.C. and O.C.

power sources and distribution systans satisfy the requirements of General Design Criterion 17 of Appendix A to 10 CFR Part 50.

The ACTION requirements spccified for the levels of degradation of the 4

power sources provide restriction upon continued facility operation ccamensurate with the level of degradation.

The OPERA 81LITY of the power sources are con-sistant with the initial condition assumptions of the safety arialyses and are l

based upon saintaining at least one redundant set of onsite A.C. and 0.C. power sourtes and associated distribution systems CPERAALE during accident conditions coincident with an assumed loss of-offsita power and single failure of the other onsite A.C. source.

The A.C. and 0.C. source allowable out-of service times are based on Regulatory Guide 1.93 "Availability of Electrical Power Sources," Decent. r 1974 and Appendix A to Generic Letter 84-15. "Proposed Staf f

[

N, Position t41.eprove and Maintain Diesel Generater Reliability." When one Q..

diesel generator is inoperable, then is an additional ACTION requirement to i

verify that all required systems, subsystans, trains, components and devices, that depend on the maaining CPERMLI diesel generatar as a source of emergency power, are also CPERA8LE, and that the staardriven auxiliary feedwatar pump is i

CPERA8LE.

This requirement is intended to provide assurance that a loss-of-offsite power event will not nault in a completa loss of safety function of critical systans during the period one of the diesel generators is inoperable, i

The ten, verify, as used in this contaxt means ta administrative 1y check by examining logi or other information ta detarsine if certain components an out of service for maintenance er other nasons.

It does not mean to perfore l

the Surveillance Requirements needed te demonstrata the OPERA 8!LITY of the component, j

The opt.RASILITY of the sintaus specified A.C. and 0.C. power sources and associated distribution systema during shutdoom. and refueling. ensures that:

4 i

(1) the facility can be maintained in the shutdown er refueling condition for i

extended time periods, and (2) sufficient instrumentation and control capa-bility is available for monitoring and maintaining the unit status.

The Surveillance Requirements for demonstrating the OPERA 81L!rY of the I

diesel generatort are based on the recommendations of Regulatory Guides 1.9 I

Revision 2 '5 election of Diesel Generator Set Capacity for Stan@y Power i

Supplies,' Decenter, 1379; 1.108, "Periodic Tasting of Diesel Generator Units Used as Onsite Electric Power Systans 61 Nuclear Power Plants," Revision 1, August 1977; and 1.137, "Fuel 011 Systems for Staney Diesel Generston,"

Revision 1, October 1979, Appendix A to Generic Letter 84-15 and Gen-ic Letter 83-26, "Clarification of Surveillance Requirements for Of esel Fuel bourity s

Laval Tests."

i WCGTkE-a-WIT 1 8 3/4 8-1 i

ycone usTs - h-S

(LECTRICAL POWER SYSTEMS 8,ASES A.C. 500RCE5m 0.C. SOURCES, and ONSITE POVER O!$TRIBUT!0N (Continued)

The Surveillance Requirement for demonstrating the OPERA 8!LITY of the station batteries are based on the recommendations of Regulatory Guide 1.129, "Maintenance Testing and Replacement of Large Lead storage 8atteries for l

Nuclear Power Plants,' February 1978, and IEEE Std 450-1975, "!EEE Reconsnanded Practice for Maintenance, Testing, and Replacement of Large Lead Storage i

latteries for Generating Stations and Substations,' and 444-1975 "Recomended Practice fer Installation Design and Installation of Lead Storage Batteries for Generating Stations and Substations.'

Verifying average slutrolyta temperature above the ainfaum for which the battery was sized, total battery tarsinal voltage on float charge, connection resistance values, and the perforsance of battery service and discharge tasts ensures the effectiveness of the charging syntaa, the ability to handle high discharge rates, and compans the battery capacity at that time with the rated capacity.

Table 4.8 2 specifies the normal limits for each designated pilot cell and each connected cell for electrolyte level, float voltage, and specific g,.

gravity.

The limits for the designated pilot cells float voltage and specific l

gravity, greater than 2.13 volts and 0.015 below the manufacturer's full charge specific gravity or a battery charger current that had stabilized at a low value, is characteristic of a charged cell with adequate capacity.

The normal limits for each connected cell for float voltage and scocific gravity, greater l

than 2.13 volts and not more than 0.020 below the manufacturer's full charge specificgravitywithanaveragespecificprivityofalltheconnectedcells not more than 0.010 below the manufacturer : full charge specific gravity,

{

ensures the OPEAA8!LITY and capability of the battery.

)

Operation with a battery cell's parameter ouWde the normal limit but witMn the allowable value specified in Table 4.4-2 is permitted fer up to 7 days.

During this 7-day period (1) the allowable values for el u trolyte

(

level ensures no.Shysical damage to the platas with an adequate'elutron transfer capability; (2) the allowable value for the average specific gravity l

of all the cells, not more than 0.020 below the manufactunr's recommended full l

charge specific gravity, ensures that the decrease in reting will be less than i

the safety margin provided in sizing; (3) the allowable value for an individual j

cell's specific gravity, ensures that an individual cell's specific gravity will not be more than 0.040 below the manufacturer's full charge specific l

l gravity and that the overall capability of the battery will be maintained within m 'cceptable limit; and (4) the allowable value for an individual j

cell's r aat witage, gnator than 2.10 volta, ensures the battery's capaility to perform ts design function.

bMTt*-- -t*!T 1-5 3/4 8-2

)

\\tGTLE LUJLT.S - N 3--

\\

e

ELECTRICAL, POVER SYSTEMS O,

IA$ES 3/4.8.4_ ELECTRICAL EQUIPMEKT PROTECTIVE DEVICES Containment electrical penetrations and penetration conductors are pro-tacted by either deenergizing circuits not required during reactor operation or by demonstrating the OPERASILITY of primary and backup overcurrent protec-tion circuit breakers during periodic surveillance. A list of containment penetration conductor overcurrent protective devites and feeder breakers to isolation transformers between 440 V class 11 busses and non-class 1E equipment is provided in Table 16.3-5 of the VEGP F5AA.

The, turve111ance Requirementa applicable to lower voltage circuit breakers provide assurance of breaker reliability by testing at least one regnsentative sasele of each manufactunr's brand of ciMutt breaker.

Each manufacturer's solded case and metal case circuit breakers are groused into representative samples wnich are then tasted on a rotating basis to ensun that all breakers are tested.

If a wide variety exista within any manufacturer's brand of circuit breakers, it is necessary ta divide that manufacturer's breakers into groups and treat each group as a separate type of breaker for surveillance purposes.

The bypassing of the actor-operated valves thermal overload protection except during periodic testing ensures that the thersal overload protection

(

will not pnvent safety-related valves free performing their function.

The (c

Surveillance Requineents for demonstrating the bypassing of the thersal overload protection continuously an in accordance with Aegulatory Guide 1.106, i

l "Thorsal Overload Protection for Electric Motors on Motor Operated Valves,"

Revision 1. March 1977.

'g N TLE - UNIT-t--

8 3/4 8-3 YCdrTLE. LLorT5 - l + A

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3/4.9 REFUEL,ING OPERAT!0N$

BA$($

3/4.9.1 80RfA CONCENTRATION The limitations on reactivity conditions during REIVELING ensure thatt (1) the reactor will remain subcritical during CORE ALTERATIONS, and (2) a unifors boron concentration is saintained for reactivity control in the water volume having direct access to the reactor vessel.

N locking closed of the mquired valves during nfueling operations precludes the possibility of uncontrolled baron dilution of the filled portions of the Reactor coolant System.

This action prevents flow to the RCS of unborated water by tlosing flowpaths frca sources of unborated water.

These limitations an consistent with the initial { h-vc1w ef 0.*: cr lese-for-tonditions assumed f safety analysis..

y g_j miudus a-M.w k beeeeenative-aMec. cc for wrecrt44*tieer-SietierirNe boron concentration value of 2000 ppa or greatergachdes-a-sensarvative-uncertalaty-any-me of w.

w.

h irg et M.

""'"'7 3/4.9.2 INSTRLt*EMTATION The OPERAAILITY of the Source Range Neutron Flux Aniters ensures that redundant monitoring capability is available to detect changes in the reactivity' condition of the core.

D.

3/4.9.3 DECAY TIM (

The minista requirement for teactor subcriticality prior to movement of irractiated fuel assemblies in the Ny: tor-vessel ensures that sufficient time has elaosed to allow the radioacV ve decay of the short lived fission products.

This decay time is consistant, with the assumptions used in the safety analyses.

3/4.9.4 CONTAINMENT SUILDING PENETRATION $

The requirements en containment building penetration closure and CPERASILITY ensun that a release of radioactive notarial within containment will be

' restricted free leakaos to the envirorment.

The OPEAA4!LITY and closun restrictions are suff< cient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pnssurization potential while in the RUl)(LING M)DE.

3/4.9.5 CCM4)MICAT10N$

The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATICMS.

V0GTLE UNITS - 1 & 2 8 3/4 9-1 i

O INSERT AR ensures a Keff of 0.95 or less and includes a conservative allowance f or calculational uncertainties of 100 ppm vf boron.

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I FIGURE 5.1-2 LOW POPULATION ZONE V0GTLE - UNITS 1 & 2 5-3

Ots!GN FEATURf 5

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5. 6 FUEL STORAGE CRIT!CALI[linM1.

5.6.1.1 ho spent uel storage escks are designed and shall be maintained g

with:

A k,ff equivalent to less than or equal to 0.95 when flooded with s.

unborated water, which includes a conservative allowance of 0.91X D/k for uncertainties as described in Section 4.3 of the FSAR, and b.

A nominal 10.6 inch center-to-contar distance between fuel assemeltes placed in the storage racks.

5f6.f2 k,

fo n fue)I e[rstcrelo ing red dry I the' 7

erat, ton /

p spept f I sta a e ac sh4f1 exceed O.

who aqu us foas A spse

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5. 6.1. (Ded w thW,peht fuel storage racks are designed and shall be We4 s sai l

A k,ff equivalent to less than or equal to 0.95 den floode tP~

a.

unborated water, which includes a conservative allowance o

- I13 %

Ak/k for uncertainties as described in Section 4.3 of the FSA4, h b.

A nominal spacing of 10.54 inches in the North-South direction and 10.4 inches in the East Wst direction between fuel assemelies laced in the storage rocks, d.N

,5.6.1.[y7hek for new fuel for the first core loading stored dry in the 44 4 5", ws fueif[torage racks shall not uceed 0.94 den aqueous foaa

-r fation is assumed, i

ORAINAGE 5.6.2 The spent fuel storage pools are designed and shall be maintained to prevent inadvertent draining of the pool below elevation 194' 1Y.

CAPACITf 5.6.3 The spent fuel storage pools are designed to contain sufficient storage rack locations for long-tars storage.

Currently, the Unit 1 pool contains two storage racks with a comeined capacity of 284 fuel assemeltes.

The Unit 2 pool say contain up to 20 storags racks with a coseined capacity of 2094 fuel assemeltes.

l 5.7 CCMPONEXT CYCLIC OR TRAN5!ENT t.!MIT i

l 5.7.1 '7he coaconents identified in Table 5.7-1 are designed and shall te saintained within the cyclic or transient limits of Table 5.71.

YOGTLE UNITS - 1 & 2 5-5 i

"-'s g.,

g (P

--f TABLE 5.7-1 COMPONENT CYCLIC 08 TRANSIENT t1MITS c

h CYCLIC OR MSIGE CYCtf g

COMPONENT TRANSIENI tIMIT OR TRAltSIENT seector Caelant. System 298 heatg cycles at < 300T/h Heate le - T from $'200T and 200 coeldown cycles at to >

M i

$ 188*F/h.

CooTeensi cycle - T from p

1 550 T to $ 200 W.

200 pressurizer coeldsun cycles Presserizer coeldsun cycle et < 200*F/h.

temperatures free > 650*F te

$ 200T.

SS less of lead cycles, without

> 15E of RATES TIENtM. POWER to 4

lamedlete Turtles or Aewter trip.

k of AATED THENtM. POER.

e 40 cycles of 1sss-of-offsite Less-of-offsite A.C. electrical A.C. electrical power.

ESF Electrical System.

A cycles of less of flew la one Less of only one reacter 4

reeter coolant leep.

coolant,pemy.

400 Anector trip cyC.:

1eeK to M of AATES TENGE. POWER.

30 esmalliery spray Spray water temperature differential j

ectmetles cycles.

> 320*F and < 625*F.

f 200 leek tests.

Pressurized to 1 2'ae5 psig.

14 hydrostatic pressure tests.

Pressurized to 1 3347 pals.

Secondary Cae? ant System I steen line break.

Break in e > e.-lach steen lies.

le hydrostatic pressure tests.

Pressurized to 1 1481 psig.

1 1

.___.-__.,,.,_--n.

,-_,--,n-_

,n.

n,

ADN!NISTRATIVE CONTR0i.S 6.1 RESPONSIBILITY The General Manager - kogue., lea,l' hld nd-Nut 6.1.1

-Mucieer-eparWm-(awW shall be respon-sible for overall plant operation and shall delegate in writing the succession onsibilit to this re -cretd h%y during his absence.%Aef - Nt.udCLt*

ICu&

6.1. 2 The will annv411y reissue a directive that emphasizes the pri<ar/

sanagesent responsibility of the onshift Operations Supervisor (or during his absence frca the control roca, the individual designated to assume the coevaand functions) for safe operation of the plant under all conditions on his shift and that clearly establishes his command duties.

6.2 ORGAN!ZATION

@s55374-C/Eu6 MD ChW CRCrMIZATICM 3

g g %.2.1 (.The-effsite-orgentaatf**-for-plant-sanagement-and-tech

'6 1LM a; showrtrrigure 6.Z-1.

g 9LMT STAFF l

6.2.2 The plant organization shall be as shown in Figure 6.2-2 and:

a.. Each on-duty shif t shall be composed of at least the minisua shif t I

crew composition shown in Table 6.2-1; b.

When fuel is in either reactor at least one operator licensed on the applicable unit shall be in the control roce.

Itradditigergile D

Jicense,Qt is in MODE 1, 2, 3 or 4 at least eqe,1antog spergter either r e

cn the applicable unit (s) shall be in f

e, c.

An individual

who'has successfully coeoleted the Initial Technician Training portion of the Health Physics Training Program or its equivalent shall be on site when fuel is in either reactor; d.

All CORE ALTERATIONS thall be observed and directly supervised by either a licensed senior Operator or Itcensed senior operator Limited ta Fuel Handling who has no other concurrent responsibilities during this operetten; j

e.

Administrative procedures shall be developed and implemented to limit the working hours of plant staff in performance of safety-related l

functions (e.g., licensed Senior Operators, licensed Operators, key Health Physics Technicians, key non-licensed cperators, and key maintenance personnel).

I 4f a single Senior Operator does not hold a Senior Operator's license on both l

units, two or more senior Operators who in concination are licensed as Senior l

l Operators on both units may fulfill this requirement.

I "This individual may be abserit for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in l

orcer to accesumodate unexpected absence, provided irwooiate action is taken to fill the required position.

V0GTLE UNITS - 1 & 2 6-1 i

TNSER.T AS

[2.1 Onsite ano offsite organizations shall be established for plant operation and corporate management, respectively.

The onsite and offsite organizations snan include the positions for activities affecting the safety of the nuclear power plant.

4.

t.ines of authority, responsibility, and cosununication shall be estaa-lished and definea for the highest management levels through inter-mediate levels to anc including all operating organization positions.

These relationships shall be cocumented and updated, as appropriata, in the form of organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent fenes of documentation.

These requirements shall be documented in the F5AR.

b.

The General Manager - Nuclear Plant shall be responsible for overall plant safe operation and shall have control over those onsite activ-ities necessary for safe operation and maintenance of the plant.

c.

The vice President - Nuclear shall have corporate responsibility for overall plant nuclear safety and shall take any sensures needed ta ensure acceptable performance of the staff in operating, maintaining, anc providing technical support to the plant to ensure nuclear safety, d.

The individuals who train the operating staff and those who carry out health physics and quality assurance functions may report to the appropriate onsite manager; however, they shall have sufficient organizational frescos to ensure their indepencence free operating pressures.

t

i O..,',.

ADMIN!$TRATIVE CONTROLS t

PLANT STAFF (Continued)

Adequate shift coverage shall be maintained without routine heavy use of overtise.

The obje:tive shall be ta have operating personnel work a nominal 40-hour week while the plant is operating.

(This work week may consist of 12-hour shift schedules.) However, in the event that unforeseen problems require substantial. amounts of evertime to be used, er during extended periods of shutdown for refueling, major maintenance, or major p'. 3t modification, en a toeporary basis the fallowing guidelinea shall, 4 followedi 1.

An individual should not be permitted te work core than 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months straight, excluding shift turnover time.

2.

An individual should not be pensitted t4 work acre than 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 44 hour5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months period, nor more than 71, hours in any 7 day period, all excluding shift turnover ties.

3.

A break of at least 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months should be allowed betwen work periods, including shift turnover time.

4 f.xcept during axtended shutdown periods, the use of overtime should be considered en an individual basis and not for the e.ntire staff on a shift.

Any deviation from the above guidelines shall be authorized by the applicable department superintendent, er higher levels of manage-sent, in acesrdance wit % established precedures and with documenta-tion of the basis for granting the deviatten.

Centrols shall be included in the procedures swh that individual ancess overtime shall be reviewd monthly Dy the General Manaer aiWeeth 5:1u.- Oi inny or his designee to asture that excessive hovrs were authorized and that they de not become routine.

N& CAL $

DA

's..

V0GTLE UNIT 5 - 1 & 2 6-2 e'

/ '

6lbkh o

FIGURE 6.2-1 OFFSITE ORGANIZATION YOGTLE UNITS - 1 & 2 6-3 h

E.

-e.

O ELETE3 FIGURE 6.2-2

'y PLANT ORGANIZATION V0GTLE UNITS - 1 & 2 6-4

.I

TA8LE 6.2-1 MINIMUM SHIFT CREW COMPOSITION TWO UNITS WITH A COMN CONTROL ROOM POSITION NUMBER OF INDIVIOUALS REQUIRED TO FILL POSITION BOTH UNITS'IN BOTH UNITS IN ONE UNIT IN MODE 1, 2 MODE 1, 2, 3, MODE 5'or 6 3, or 4 AND ONE UNIT IN or 4 OR DEFUELED MODE 5 or 6 or DEFUELFD 05 1

1 1

SRO 1

none**

1

~

R0 3*

28 3*

NLO 38 38 38 STA 1***

none 1***

~

05 - 0parations Supervisor with a Senior Operator license SR0 - Individual with a Senior Operator license R0 - Individual with an Operator license l

NLO - Non-Licensed Operator STA - Shift Technical Advisor The shift crew composition say be one less than the mini $da requirements of fabis 6.2-1 for A period of time not to exceed 2 hors in oNar to a :,;ceeWatt unexpected absence of oneduty shift crew members provided f amediata action is taken to restore the shift crew ccsoosition to within the etnisus requirements of, Table 6.2-1.

This provision @as not permit any shift crew positten to be unmanned upon shift change due to 3.n encoming shift crewman being lata or absant.

During any absence of the Operation Supervisor' free the centrol roca while either Jnit is in MODE 1, 2, 3, or 4, an individual with a valid Senior Oporater license shall be designated to assues the control rnos consand function. During any absence of the Operations Supervisor from the contro' roce while either unit is in MODE 5 or 6, an individual with a valid Senior Operator license or Operator license shall be designated to assume the control room command function.

"At least one of the requiMd individuals must be assigned to the designated position for each unit. -

- At least one ifcons bn erator or Licensed Senior Operator Limited to Fuel Handling who ha othe concurrent responsibilities must be present during CORE ALTERATI

$ ne her unit.

- - The STA position shall be manned in MODES 1, 2, 3, and 4 unless the Operations Supervisor or the individual with a Senior Operator license meets the qualifications for the STA as stated in the Policy Statement on Enginesring Expertise on Shift, dated Octobe.r 28, 1985.

V0GTLE UNITS - 1 & 2 6-5 o

l ADMINISTRATIVE CONTROLS 6.2.3 INDEPENDENT SAFETY ENGINEERING GROUP (ISEG)

FUNCTION 6.2.3.1 The ISEG shall function to examine plant operating characteristics, l

NRC issuances, industry advisories, Licensee Event Reports, and other sources of plant design and operating experience information, which may indicate areas for improving plant safety.

The ISEG shall sake detailed re dations for revised procedures, equipment modifications, maintenance a ivitie operations activities, or other means of improving plant safety to th Sn4 + g ice President-Nuclear. 4perethr.: t h ;h tM A nes., h ieer 7. ferre., : and m a alagice! Mfety y 0

COMPOSITION 6.2.3.2 The ISEG shall be composed of at least five, dedicated, full-time engineers.

Each shall have a bachelor's degree in engineering or related science and at least 2 years professional level experience in his field, at least 1 year of which experience shall be in the nuclear field.

RESPONSIBILITIES 6.2.3.3 The ISEG shall be nsponsible for maintaining surveillance of plant activities to provide independent verification

that these activities are performed cornetly and that human errors are reduced as such as practical.

RECORDS 6.2.3.4 Records of activities performed by the ISEG thall be propand, sain-tained, and forwarded each calendar month to the Senior Vica President -

Nuclear, eeratter; timugn tne kn.ger--Eclesr Perfor coe: :nd h4f c1csk l

--Sef y ~ %

~

l 6.2.4.SHIFf 71ECHNICAL ADVI$0R l

6.2.4.1 The Shift Technical Advisor shall provide advisory technical support i

to tne Shift Supervisor in ths areas of thermal hydraulica, nactor engineering, and plant snalysis with regard to the safe operation of the plant.

The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline arid shall have received specific training in 2.he l

responsa and analysis of the plant for transients and accidents, and in plant l

design and layout, including the capaht11 ties of instrweentation and controls in the control rcos.

l 6.3 TRAINING 6.3.1 A retraining and replacement training program for the plant staff shall be maintained under the direction of the Plant Training and Emergency Prepared-ness Manager.

Personnel will meet the minimum education and experience recom-mendations of ANSI N18.1-1971 and, for licensed staff, bpeedh ? Of 10 CFR 55 59 and the supplemental mquitants specified in Sect 4ns A and C of Enclost.re 1 of the March 28, 1980 NRC letter to all licensees, befon they an considered qualified to perfore all duties independently.

Prior to meeting the neommend-ations of ANSI N18.1-1971, personnel any be trained to perfors specific tasks and will be qualified to perfore tnose tasks indersndently. A "Not nsponsible for sign-of f function.

h gp q V0GTLE UNITS - 1 & 2 6-6

e INSERT AT Personnel who complete an accredited program which has been endorsed by the NRC shall meet the requirements of the accredited program in lieu of the above.

1 I

1 l

i r

l l

i l

)

l 4

i i

i 1

l

)

l.

s-e-rmew

-%w

ADMINISTRATIVE O NTROL$

0

.4 6

REVIEV AND AUDIT 6.4.1 PLANT REVIEW BOARD (PRB)

FUNCTION Qg g gg 6.4.1.1 The PRS shall function to advise the on all matters related to nuclear safety, COMPOSITION 6.4.1.2 The PR8 shall be composed of Depar'aent 5 terintendents or Managers, or supervisory personnel aporting directly to Detaitment Superintendents or Managers from the departments listed below:

Operations Maintenance Quality Centrol Health Physics Nuclear Safety and Compliance Engineering Support A senior health physicist is acceptable for the Health Physics Department PDS' repiesentative.

The chairman, his alternate and other members and their alternates of the PR8 shall os designated by the =%. em ALTTRNATES

~

6.4.1.3 No more than two alternates shall participata as voting members in PRS activities at any one time, MEETINGFREQUENQ i

l

6. 4.1. 4 The PRS shall meet' at least once per calendar month and as c.onvened l

by the ?RS Chairman or his designated alternata.

QUORUN l

6. 4.1. ?,

The oorur of the PRS necessary for that perforosace of the PRA l

responsibil.ity tM authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four sesbers including alternates.

RESPONSIBILITIES 6.4.1.6 The PR8 shall be responsible for:

a.

Review of 1) procedures which establish plant-wide administrative controls to implement the QA program or Technical Specifications surveillance program, 2) procedures for changing plant operating modes, 3) emergency and abnormal operating procedures, 4) procedures for effluent releases of radiological consequences, and 5) fuel handling procedures.

V0GTLE UNITS - 1 & 2 6-7 i

! "2 ~

ADMINIS'RiTIVE CONTROLS T

RESPONSI51LITIES (Continued) b.

Review of 1) prograss required by Specification 6.7.4 and changes thereto, and 2) proposed procedures arid changes to piccadures which involve an unreviewed safety question as per 10 CFR 50.59.

Review of all proposed tests and experiments that affect nuclear c.

safety; Review of all proposed changes to the Technical Specifications; d

Review of all propos.ed changes or modifications to plant systems or e.

equipment that affect nuc1 ty, including proposed changes to Chapter 16.3 of the Vogt1 5afety alysis Report (FSAA);

3Fined f.

Investigation of all viol the Technical Specifications, including the prep d forwarding of reports covering evalua-tion and recomme t'on to revent recurrence, to the Senior Vice President-Nuc1 r/Operati+ne d to the Safety Review Board; g.

Review of all REPC A8LE ENT5; h.

Review of plant operations to detect potential hazards to nuclear safety; 1.

Performance of special reviews, inve jgations, or analyses and reports thereon as nquested by the or the Safety Review Board; j.

Revi w of tha Security Plan and implementing pru eduren and submittal of ncesesnded changes to

@ and the Saf6ty Review Board; A

i[ Review of thw Emergancy Plan and lesenting procedures and submittal CO M of reconcended changes to the

' d the Safety Review Board; an M S.AO e-

- --/

ki d

(

Reviw of any ace.idental, emplar.ned, or uncontrolled radioactive 1.

c 9

release including the preparation of reports covering on.

Ogg ncesasadations, and sposition of the carnctive a ion to event,

n eur n % a ard th orweting^qf these reports to 44M6r ice President Hucles h e at,d to the Sefaty Revi card-m.

Review of changes to SS CONTROL PROGRM, the QFFSITE DOSE CALCULATION MANUAL, and the Radwesta Tnatment Systems; and n.

Review of the Fire Protection Program and Implementing procedures and submittal of neoenended changes to the

~1 6.4,1. 7 The PRS shall:

a.

Recommend in writing to the approval or disapproval of items considered under Specification 6.4.1.6a. through e. prior to their implementation; 1

V0GTLE UNITS - 1 & 2 6-8 1

ADMINISTRATIVE CONTROLS O

RESPONSIBILITIES (Continued) b.

Render determinations in writing with regard to whether or not each ites considered under Specification 6.4.1.6a. through f. constitutes an unrwiewed safety question; and c.

Provide written ficatto ithin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to the Senior Vice President Huclea J:r:ti:n; d the Safety Review Board of dis-agreesent between d the-9846; however, the M shall have responsibility for resolution of such disagreements pursuaat to Specification 6.1.1.

g RECORDS Q

phnt-

6. 4.1. 8 The PR8 shall maintain written minutes of each PR8 aceting that, at a minimum psynt the results of all P vities performed under the respon-sibili gions of these Technt ifttations.

Copies shall be provided i

to ce President-Nuclea 5:r;ti:

and the Safety Review Board.

6.4.2 SAFETY REVIEW BOARD (SRS)

FUNCTION 6.4.2.1 The SRS shall function to provide independent review and audit of

[,'

designated activities in the areas of:

a.

Nuclear power plant operations, b.

Nuclear engineering, c.

Chemistry and radiochemistry, d.

Metallurgy, e.

Instrumentation and control, f.

Radiological safety.

g.

Met.hanical and electrical engineering, and h.

Quality assurance practices.

The SRB shall report to and advise r a..irajice Pnsident-Nuclea era on those areas of responsibility spec 4M^111 Specifications 6.4.2.7 6.4.2.8.

COMPOSITION 6.4.2.2 The SRS shall be organized as one board for all GPC Ntclear power plants. The SAS shall be composed of a minious of five persons who, as group, provide the exne

( W Toview and audit the operation of a n sa powar plant.

The cha 1 and oth members shall ba appointed by th /.-fer S

-u f-Vice President-Nuclea The composition of the betitm other such person as he may desi tW r

hk t the requirements of ANSI N18.7-1976.

V0GTLE UNITS - 1 & 2 6-9

_ ADMINISTRATIVE CONTROLS AUDITS (continued) g.

The Radiological Environmental Monitoring Program and the results thenof at least once per 12 months; h.

The OFFSITE DOSE CALC')LATION MANUAL and implementing procedures at least once per 24 months; i.

The PROCESS CONTROL PROGRAM and implementing procedures for processing and packaging of radioactive wastes at least once per 24 months; j.

The perfomance of activities requimd by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months; k.

Yhe Emergency Plan and implementing procedures (at least once per 12 months);

O 1.

The Secu Plsn and implementing procedures (at least once per 12 mon s).

RECORDS 6.4.2.9 Records of SR8 activities shall be prepand, approved, and distributed as indicated below:

a.

Nint f each SR8 meeting shall be rope proved,' and forwarded to th 4eMerfice President-Nuclea.per;tica: ithin 14 days follow-ing eac

'ering; b.

R f reviews enecapassed by 5 at n 6.4.2.7 shall be pre-pa p ved, and forwarded to th Anfr ice President-Nuclear e H retiene ithin 14 days following 1

on of the review; and c.

ports e MYffcatTiiP$r4

.8 shall be forwarded to the Geocv.tve-44ee-Ar eid::t, Sonig, ice President-Nuclear hhe%erf;;d671mitions responsible for the areas audited within 30 days aftsr cogletion of the audit by the j

auditing organization.

{

6.5 REPORTA8LE EVENT ACTION 6.5.1 The following actions shall be takan for REPORTA8LE EVENTS:

1 l

a.

The commission shall be notified and/or a nport submitted pursuant to the requirements of Section 50.72 and Se: tion 50.73 to 10 CFR Part 50,.

and b.

Each REPORTA8LE EVENT shall.be mviewed by the PR8, results of this review s PtI47sbeitted to the SR8 and th

(+P ice President-Nuclea s i.1 w n b l

V0GTLE UNITS - 1 & 2 6-12 i

ADMINISTRATIVE CONTROLS 6.6 SAFETY LIMIT VIOLATIOJ 6.6.1 The following actions shall be taken in the event a Safety Limit is violated:

In accordance with 10 CFR 50.72, the NRC Operations Center shall be a.

notified by telephon soon as practical and in all thin one hour after

"'- g has been determined.

fr'r ice o

President-Nuclea e'ified within 24W, the SRS, PRS, and the s all be

/

G e e e rca h er-b.

A Licensee Event Report shall be prepared in acc ance with M GM' GM 10 CFR 50.73.

c.

The Licensee Ersnt Report shall be submitted to the ission in ance wita 10 CFR 50.73, PRS, SR8, th and the denie=

ce President-Muclea eathn within 30 days after dis-m cove of the event.

d.

Critical operation of the affected unit shall not be resumed until authorized by the Nuclear Regulatory Commission.

6.7 PROCEDURES AND PROGRAMS 6.7.1 Written proceduns shall be established, implemented, and maintained covering the activities referenced below; The applicable procedures ncommended in Appendix A of Regulatory a.

Guide 1.33, Revision 2, February 1978; b.

The energency oport.Mng procwdums nquind to implement the require-sents of NUf!EG-0737 and Supplement 1 to NUREG-0737 as stated in Generic letter No. 82-33; c.

Security Plata implementation; d.

Energency Plan tuplementation; e.

PROCESS CONTROL PROGRAM implementation; f.

OFFSITE 00SE CALCULATION MANUAL implementation; g.

Quality Assurance for effluent and environuental sonitoring; 1

h.

Fire Protection Prograar Implementation; and 1.

Technical Specifications Improvement Program implementation.

(FSAR Chapter 16.3) 6.7.2 Each procedure of 6.7.1 above, and changes thereto, shall be reviewed as set forth in administrative procedures and approved by either theM or the department head of the responsible department prior to ieglementation with the exception of the following which shall be approved by the@NNG:

V0GTLE UNITS - 1 & 2 6-13 b<neRU MQrr e r - Nu.dCo#

9Io-n1'

ADMINISTRATIVE CONTROLS D.s

~~'

PROCEDURES AND PROGRAMS (Continued) 1)

procedures which establish plant-wide administrative controls (which implement the quality assurance program and the Technical Specifica-tions surveillance program),

2) unit operating procedures (UOPs)

~

3) emergency operating procedures (E0Ps) 4)

abnormaloperatingprocedures(AdPs) 5)

procedures for iglementing the security plan, energency plan, and the fire protection program, and 6) fuel handling procedures.

PRS responsibilities for procedums are delineated in 6.4.L 6.7.3 Temporary changes to proceduns of Specification 6.7.1 say be made pro-vided:

The intent of the original procedun is not altered; a.

b.

The change is approved by two members of the plant management staff,

{f at least one of whom holds a Senior Operator license; and The change is documented, reviewed in accordance with Specifica-c.

tion 6.7.2 and approved by the or department head of the responsible department wi 4 days of isolementation.

GCnern) Mct - e e Mticl W b d>t P 6.7.4 The following prograss shall be established, imp 1 ted, and saintained:

a.

Primary Coolant Sources Outside Containment A program to nduce leakage from those portior.s of systems outside 3

containment that could contain highly radioac,tive fluids during a

, serious transient or accident to as low as practical levels.

The systaurg include the following:

1)

Residual Heat Removal Systae 2)

Containment Spray Systas (excluding Ma0H Subsystas) 3)

Safety Injection (excluding Boron Injection & Accumulators) 4)

Chemical and Volume Control System (Letdown, Boron Recycle, and Charging Pumps) 5)

Post Accident Processing System 6)

Gaseous Wasta Processing System 7)

Nuclear Sampling System (Pressurizer steam and liquid sample lines, Reactor Coolant sample lines, RHR sample lines, CVCS Dominera112er and Letdown Heat Exchanger sample lines only)

V0GTLE UNITS - 1 & 2 6-14 i

.n-

C ADMINISTRATIVE CONTROLS O

SEMIANNUAL RADI0 ACTIVE EFFLUENT REL

.E REPORT (Continued)

The Semiannual Radio etive Effitent Release Reports shall also include the following:

an expla ation as to why the inoperability of liquid or gaseous effluent monitoring inst umentation uas not corncted within the time specified in Specification 3.3.3.

or 3.3.3.@, respectively; and description of the events leading to liquid holdup tanlEs or gas storage tanks exceeding the limits of Specification 3.11.1.4 or 3.11.2.6, respectively.

MONTHLY OPERATING REPORTS 6.8.1.5 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or safety valves, t

shall be submitted on a monthly basis to the Director, Office of Resource Management, U.S. Nuclear Regulatory commission, Washington, D.C. 20555, with a copy to the Regional Administrater cf the Regional Office of the NRC, no later than the 15th of each month following the calendar month covered by the report.

RADIAL PEAXING FACTOR LIMIT REPORT limits for RATED THERMAL POWER (F,yRTP) shall be establishe

6. 8.1. 6 The Fxy for at least each reload core and shall be maintained available in the Control Room.

The limits shall be estabitshed and taplemented on a time scale g:.

consistant with normal procedural changes.

\\

The analytical methods used to generata the F limits shall be those xy previously reviewed and approved by the NRC*.

If changes to these methods are deemed necessary they will b6 evaluated in accordance with 10 CFR 50.59 and submitted to the NRr. for review and approval prior to their use if the change is determined to involve an unreviewed safety question or if such a change would require amendment of previously submitted documentation.

A report containing the F lialts for all core planes containing B'ank "D" xy control rods and all unrodded core planes along with the plot of predicted F.Pr,1 vs axial core height (with the limit envelope for comparison) shall be provided to the NRC Document Control desk with copies to the Regional Admints-

[

trator and the Resident-Inspector within 30 days of their implementation, t

SPECIAL REPORTS 6.8.2 Special reports shall be submitted to the Regional Administrator of the Regional Office of the NRC within the time period specified for each report.

"WCAP 8385 "Power Distribution Control and Load Following Procedures" and WCAP 9272.A "Westinghouse Reload Safety Evaluation Methodology."

i l

l V0GTLE UNITS - 1 & 2 6-20 i

t i

i

ADMINISTRATIVE CONTROLS 6.12 PROCESS CONTROL PROGRAM (PCP) (Continued) 3)

Documentation of the fact that the change has been reviewed and found acceptable by the PRS.

g b.

Shall become effective upon approval by the M.

6.13 0FFSITE DOSE CALCULATION MANUAL (00CN) 6.13.1 The 00CM shall be approved by the Commission prior to implementation.

6.13.2 Licensee-initiated changes to the 00CM:

a.

Shall be submitted to the Commission in the Semiannual Radioactive Effluent Release Report for the period in which the change (s) was made effective.

This submittal shall contain:

1)

Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental inforsation.

Inforsation submitted should consist of a package of those pages of the 00CM tt.,be changed with each page numbered, dated and containing the revision number, together with appropriate analyses or evaluations justifying the change (s);

,,(.

2)

A cietermination that the change will not reduce the accuracy or reliability of dose calculations or Setpoint determinations, and 3)

Documentation of the fact that the change has been reviewed and found acceptable by the PR8.

b.

Shall become effective upon approval by the 6>6H6.

U bG#

l 6.14, MAJOR CHALGES TO LIQUID. GASEOUS. AND SOLID RA0 WASTE TREATNENT SYSTEMS

6.14.1 Licensee-inititted major changes to the Radwesta Treatment Systems (liquio, gaseous, and solid):

a.

Shall be reparted to the Commission in'the Semiannual Radioactiye Effluent Release Report for the period in which the evaluation was reviewed by the PRS. The discussion of each change shall contain:

1)

A summary of the evaluation. that led to the detaraination that the change could be made in accordance with 10 CFR 50.59; 2)

Sufficient detailed information to totally support the reason f ar the change without benefit of additional or supplemental information; i

i "Licensees say choose to submit the information callbi for in this Specification s

as part of the annual FSAR update.

V0GTLE UNITS - 1 & 2 6-24

ADMINISTRATIVE CONTROLS

\\

6.14 MAJOR CHANGES TO LIQUID. GASEOUS, AND SOLID RA0 WASTE TREATMENT SYSTEMS (Continued) 3)

A detailed description of the equipment, components, and processes involved and the interfaces with, other plant systems; 4)

An evaluation of the change, which shows the predicted releases of radioactive materials in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto; 5)

An evaluation of the change, which shows the expected maximum egosures to a MEM8EA 0F THE PU8LIC in the UNRESTRICTED AREA and to the general population that differ froe those previously estimated in the License application and amendments thereto; 6)

A comparison of the predicted releasas of radioactive saterials, in liquid and gaseous affluents and in solid wasta, to the actual releases for the period prior to when the change is to be made; 7)

An estimate of the agosure to plant operating personnel as a result of the change; and 3

8)

Occumentation of the fact that the change was reviewed and 4

.8 found acceptable by the PR8.

(

b.

Shall become effective upon approval by the M.

l C-cremJ Mcenaper -

Neuaar PLP t

I l

i V0GTLE UNITS - 1 & 2 6-25 i

_

ML20154C545

Text

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