Text

Forwards marked-up Pages from First Draft Tech Specs, Reflecting Understanding of Agreements Reached During 860527-30 Meetings at Site & 860605-06 Meetings in Bethesda, MD
	ML20199E872
Person / Time
Site:	Vogtle
Issue date:	06/17/1986
From:	Bailey J; GEORGIA POWER CO.
To:	Youngblood B; Office of Nuclear Reactor Regulation
References
	0537V, 537V, GN-948, NUDOCS 8606240134
	Download: ML20199E872 (194)
	v • d • e

e Georgia Power Company Route 2. Box 299A Waynesboro. Georgia 30830 Telephone 404 554-9%1 404 724-8114 Southere Company Service 4, Inc.

Post Office Box 2625 Birmingham, Alabama 35202 Telephone 205 870-6011 Vogtle Pro.!ect June 17, 1986 Director of Nuclear Reactor Regulation File: X7N16 Attention: Mr. B. J. Youngblood Log: GN-948 PWR Project Directorate #4 Division of PWR Licensing A U. S. Nuclear Regulatory Commission Washington, D.C. 20555 NRC DOCKET NUMBER 50-424 CONSTRUCTION PERMIT NUMBER CPPR-108 V0GTLE ELECTRIC GENERATING PIANT - UNIT 1 TECHNICAL SPECIFICATIONS

Dear Mr. Denton:

Enclosed for your staff's consideration are marked-up pages from the first draft of the Vogtle Unit 1 Technical Specifications. The revisions marked on these pages reflect our understanding of the agreements reached during the May 27-30, 1986, meeting at the site and the June 5-6, 1986, meeting in Bethesda.

If your staff requires any additional information, please do not hesitate to

contact me.

Sincerely,

<f:.

J. A. Bailey Project Licensing Manager JAB /caa Enclosure xc: R. E. Conway G. Bockhold, Jr.

R. A. Thomas NRC Resident Inspector J. E. Joiner, Esquire D. C. Teper B. W. Churchill, Esquire W. C. Ramsey M. A. Miller (2) L. T. Gucwa (w/o enclosure)

B. Jones (w/o enclosure) Vogtle Project File

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1.0 DEFINITIONS *

, The defined terms of this section appear in capitalized type and are applicable

, throughout these Technical Specifications, j ACTION 1.1 ACTION shall be that part of a Technical Specification which prescribes remedial measures required under designated conditions.

ACTUATION LOGIC TEST t

1.2 An ACTUATION LOGIC TEST shall be the application of various simulated 3 input combinations'in conjunction with each possible interlock logic state and verification of the required logic output. The ACTUATION LOGIC TEST shall include a continuity check, as a minimum, of output devices.

, ANALOG CHANNEL OPERATIONAL TEST l 1.3 An ANALOG CHANNEL OPERATIONAL TEST shall be the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY of alarm, interlock and/or trip functions. The ANALOG CHANNEL

,h OPERATIONAL TEST shall include adjustments, as necessary, of the alarm, inter-lock and/or Trip Setpoints such that the Setpoints are within the required range and accuracy.

f,.g 4 g ., g i AXIAL FLUX DIFFERENCE *#

8""Mek al pkunfy/")e.

1.4 AXIAL FLUX DIFFERENCE shall be the difference in ==:M::d flux signals between the top and bottom halves of a two section excore neutron detector.

CHANNEL CALIBRATION 1.5 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the

, channel such that it responds within the required range and accuracy to known values of input. The CHANNEL CALIBRATION shall encompass the entire channel including the sensors and alarm, interlock and/or trip functions and may be performed by any series of sequential, overlapping, or total channel steps

such that the entire channel is calibrated.

CHANNEL CHECK 1.6 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where i

possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.

t f

V0GTLE - UNIT 1 1-1 APR 2 41996 t

[ - - - . - . - -

. . . - - . - . - . - - , . - . . - . . . - . + . . . - -

DEFINITIONS CONTAINMENT INTEGRITY 1.7 CONTAINMENT INTEGRITY shall exist when:

a. All penetrations required to be closed during accident conditions are either:

b 1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in Tab 1e' [3.6-1] of Specification [3.6.3].

b. All aquipment hatches are closed and sealed, h- c. Each air lock is in compliance with the requirements of Specification 3.6.1.3, 3

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and

e. The sealing mechanism associated with each penetration (e.g., welds, bellows, or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE

, 1.8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.

$ CORE ALTERATIONS 1.9 CORE ALTERATION'S shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of C0RE ALTERATIONS shall not preclude completion of movement of a component to a safe conservative position.

, DOSE EQUIVALENT I-131 1.10 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic 5

i mixture of I-131, I-132, I-133, I-134, and I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of NRC Regulatory Guide 1.109, Revision 1, October 1977.

E - AVERAGE DISINTEGRATION ENERGY 1.11 -E :h:1W th:-everagedweighted. in-proportion-to-theecr-tution-of-each-eadiendide=inethe@)=of the-sum-of-the ever;ge4+ tad;:- :-

E rgi:: p:- disintegret4en-4HeV/d)hthe-radienvel41 rh th; ::71; E S4.r// Le Mc Gemgi e, we)4ko' inpfooi//o') 4 Me catceni%/4,; of eacf ractu>nuc/ic/c >> /4e rector costans' af Me /Mw WSan,tshi , 0 f Mt .sm of /ges, N!GV%Q bth 4 /uff /MJQ/?Qaorptsy y W w U w es7ey/g 10//75fes , opprakiN2,4es ait b f/ 7 p ys c//T

M 2 4 '906 or& Na/ i7a-iscbu ec/nwy41 Me an/cxd

es 4 s ,w . u o4 = :A m ~ m ol e s c e a cw s;;m s g:;c. m :a _ gs m ;s,g m e i DEFINITIONS QUADRANT POWER TILT RATIO 1.25 QUADRANT POWER TILT RATIO shall be the ratio of the maximum upper excore

! ' detector calibrated output to the average of the upper excore detector cali-j brated outputs, or the ratio of the maximum lower excore detector calibrated

output to the average of the lower excore detector calibrated outputs, whichever is greater. With one excore detector inoperable, the remaining three detectors shall be used for computing the average.

RATED THERMAL POWER 1.26 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant'of 3411 MWt.

REACTOR TRIP SYSTEM RESPONSE TIME 1.27 The REACTOR TRIP SYSTEM RESPONSE TIME shall be the time interval from l when the monitored parameter exceeds its Trip Setpoint et the channel sensor until loss of stationary gripper coil voltage.

REPORTABLE EVENT 1.28 A REPORTABLE EVENT shall be any of those conditions specified in I Section 50.73 of 10 CFR Part 50.

U SHUTDOWN MARGIN 1.29 SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which the reactor is subcritical or would be subtritical from its present condition assuming all full-length rod cluster assemblies (shutdown and control) are fully inserted except for the single rod cluster assembly of highest reactivity worth which is assumed to be fully withdrawn.

SITE BOUNDARY I

-hfe exc/wint bCWWa!*t) ANl N 3b*? Al y

1.30 The SITE BOUNDARY shall be th:t i h bey:nd which/the ': W ir ne+ther

) = :d,.nor 1 Med, ~ -ethe m be=centrokl:d by the l k:ncet.

i

) 29 nn 5./-l.

SLAVE RELAY TEST

}

1.31 A SLAVE RELAY TEST shall be the energization of each slave relay and

, verification of OPERABILITY of each relay. The SLAVE RELAY TEST shall include a continuity check, as a minimum, of associated testable actuation devices.

SOLIDIFICATION 1.32 SOLIDIFICATION shall be the conversion of wet wastes into a form that meets shipping and burial ground requirements.

V0GTLE - UNIT 1 1-5 APR 24 5

._ _ _ . . . . _ , . n- _. a r.~s .. .- . ~ . _ - . - - -

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TABLE 2.2-1 (Continued) f.

y REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS [.

e. c E SENSOR r-e TOTAL ERROR '

7 g FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE 71 Z 13. Steam Generator Water 17.0 12.18 1.5 >17.0% of narrow >15.3% of narrow [

s Level Low-Low range instrument Fange instrument p span span :-

14. Undervoltage - Reactor [2.0] [1.28] >9660 volts >9522 volts Coolant Pumps C p

c

15. Underfrequency - Reactor [7.5] O [0.1] >57.2 Hz >57.1 Hz 7 Coolant Pumps $

5

16. Turbine Trip g.

W m

a. Low Fluid Oil Pressure N.A. N.A. N.A. >600 psig >500 psig ["

97.fo % 976 %

di b. Turbine Stop Valve N.A. N.A. N.A. >ttM open >tt M open ;d Closure y s-

17. Safety Injection Input N.A. N.A. N.A. N.A. N.A. ti from ESF a

v a .:

+

D.

= wi f t@ LT :'

- g, w

w

f/:

il

,t k

TABLE 2.2-1 (Continued) [i.

< TABLE NOTATIONS W g-8 h m

NOTE 1: OVERTEMPERATURE AT s~

[

i'i AT 3 (3 f 7s

3) $ AT, [Ki-K2 f1 '

[T (7 f 3) - T'] + K3 (P - P') - f (AI)} f l'

s v

Where: = Measured AT by RTD Manifold Instrumentation; AT g 1 = Lead-lag compensator on measured AT; -

t.

Q, ;p It. T2 =

Timejonstants utilized in lead-lag compensator for AT. I t 78s, f T 2 g 3 s;

- (%

yf 3

= Lag compensator on measured AT; -i-f

} T3 = Time constants utilized in the lag compensator for AT, r3 / s; AT, = Indicated AT at RATED THERMAL POWER; h.

p Kg = 1.10; h

K2 [ {-0 01003/"I.h8O/2 #F .

I =

The function generated by the lead-lag compensator for T,yg dynamic compensation; h 2

= T T4. Is on tants utilized in the lead-lag compensator for T,yg,4 1 d8 s, T = Average temperature, *F; {-

1 a

y.

=

Lag compensator on measured T,yg; CD f

, Ts =

Time constant utilized in the measured T,yg lagcompensator,is[<0s; j' R Q '

p E4 .

w w

_ _ _ . .2, . c.

, a . .a -n. ~ .. - s . =

.9:

i yn TABLE 2.2-1 (Continued) [

5 TABLE NOTATIONS (Continued) $

h Q

E NOTE 1: (Continued) ;:

H j

i c T' <

588.5*F (Nominal T,yg at RATED THERMAL POWER);

K3 M0.0^^0?C';dO; f 0,00056ps, ;,

j ,

P = Pressurizer pressure, psig; ;

4 P' = 2235 psig (Nominal RCS operating pressure); .

Variable. 1 S = Laplace transform epeceber, s 1; {

=

and f t(AI) is a function of the indicated difference between top and bottom detectors of the i power-range neutron ion chambers; with gains to be selected based on measured instrument e response during plant startup tests such that: %

to C.

5 (1) For qt ~9b between -33.5% and + 6.5%, f (AI) = 0, where qt and q arebpercent RATED THERMAL jI'l POWER in the top and bottom halves of the core respectively, and qt + 9b is total THERMAL (

POWER in percent of RATED THERMAL POWER; L (2) For each percent that the magnitude of qt ~C b exceeds - 33.5%, the AT Trip Setpoint shall fi be automatically reduced by 1.27% of its value at RATED THERMAL POWER; and

{

(3) For each percent that the magnitude of qt ~Ab exceeds + 6.5%, the AT Trip Setpoint shall be automatically reduced by O.00% of its value at RATED THERMAL POWER. ii 0.83 f NOTE 2: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.6%. b;:

e g p

=

w y 8

21 J#

. .c _

. _. , . . a _ .ce u. c. . _ .: _, , _ _ , .

=..w __x . . . . . . . . . _ _ . , _ .

if p

9 D

L.

~

< TABLE 2.2-1 (Continued) 3' w

8 TABLE NOTATIONS (Continued) k,.n E,

N NOTE 3: OVERPOWER AT U e- n AT f1 f ta s 1 AT, {K 4 -Ksfy T3 7

fy f Ts 3 T-KsUfy,1 Ts3 T 1 - f 2(AI)}

w -

./.

[

Where: AT = Measured AT by RTD manifold instrumentation;

l f = Lead-lag compensator on measured AT; It.T2 = Time constants utilized in lead-leg compensator p, for AT, 11 = 8 S, r2 = 3 s; $u 1

3, 3

= Lag compensaton on measured AT; i'l M

ro 13 = Time constants utilized in the lag compensator for AT, e 13=0s; {.?

,o

W

= Indicated AT at RATED THERMAL POWER; AT, ;

K4 5 [ 1.089, i

> 0 Ks h P' O.02/'F for increasing average temperature and E for decreasing average k.[

temperature, "

y '7f3 7

=

The function generated by the rate-lag compensator for T,yg dynamic compensation, (c.

2 -

i

=

17 Time constants utilized in the rate-lag compensator for T,yg, 1 F10 s, 7

1

((

=

$:o 1 + TsS Lag compensator on measured T,yg; g.. .

N A -

lE.

s w

_,m c . .. _ , ,-.e m . .

4 L

C h

s6 TABLE 2.2-1 (Continued) (-

u.

8 TABLE NOTATIONS (Continued) K m f/,

e g

' NOTE 3: (Continued) k n-Z Ts =

Time constant utilized in the measured T,yg lag compensator, Q g Tsd0 s; -

2 :

Ks h F 0.0013/ F for T > T" and Ks #0 for T 5 T",

o T = Average Temperature, *F; z.

T" = Indicated T avg at RATED THERMAL POWER (Calibration temperature for AT %

b' -

instrumentation, < 588.5*F), e S = Laplace transform epeeeeep, s -1; and varialole (f t.

F f 2(al) = 0 for all AI. v n R.-

i v o NOTE 4: The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 3; 3.6% of AT span. n n

p.

p da e

e L .

f- '

D ..

w ,

.n

. 1.

e t,

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2.1 SAFETY LIMITS

l BASES 2.1.1 REACTOR CORE l The restrictions of this Safety Limit prevent overheating of the fuel I and possible cladding perforation which would result in the release of' fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and reactor coolant temperature and pressure have been related to DNB through the W-3 correlation. The W-3 DNB correlation has been l developed to predict the DNB flux and the location of DNB for axially uniform l

and nonuniform heat flux distributions. The local DNB heat flux ratio (DNBR) is defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux and is indicative of the margin to DNB.

The minimum value of the DNBR during steady-state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95% probability at a 95% confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions. g 3. .j ,,g The curves of Figure 2. -1 show the loci of points of THERMAL POWER, Reactor Coolant System pres,sure and average temperature for which the minimum DNBR is no less than 1.30,for the-ever:;: enth:1;7 :Lthe=v::::1 :::it b :p:1-te t' enth:1py Of ::tu.at.d 1 4 :J. +-Insuf & %c 6 2-/ /Jere.

These curves are based on an enthalpy hot channel factor, FN f 1.55 and a reference cosine with a peak of 1.55 for axial power shape. AbH,llowanceis a

l includedforanincreaseinFhatreducedpowerbasedontheexpression:

n o F

aH = 1.55 [1+ re(1-P)]

e.s Where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f t(AI) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature AT trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

l V0GTLE - UNIT 1 8 2-1 APR 2 41986 w-- r- - "

Insert for Page B 2-1 the combination of THERMAL POWER, Reactor Coolant System pressure, and average temperature are outside of the range for which the W-3 correlation applies.

The curves of Figure 2.1-1 are a result of the intersection of three lines which have the same Reactor Coolant pressure but have variable THERMAL POWER and average temperature. The most limiting line segments developed

$ provide the curves of Figure 2.1-1. From right to left along the curves of Figure 2.1-1 the line segments provide the following THERMAL POWER, Reactor Coolant System pressure, and average temperature loci of points of which:

A. The W-3 correlation applies Whed:

" - the minimum DNBR is no less than the limit value (far right line segment);

1 B. The W-3 correlation does not applya>heq:

l

- the exit quality of the hottest channel enthalpy is not greater than 15% (middle line segment on Reactor Coolant System pressure curves, 2400 psia and 2250 psia; and the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid (far left line segment).

(

i

~

N 1

- t .1., s t utmwMm .n.:.ddSMJ ' ges;& _ : .w.,gwy<, twgua gg,2 ange i

LIMITING SAFETY SYSTEM SETTINGS BASES Undervoltage and Underfrecuency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide core protection against DNB as a result of complete loss of forced coolant r flow. The specified Setpoints assure a Reactor trip signal is generated

! before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips h from momentary electrical power transients. For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following th i ;'tr.:::: trip of tu: n ...; reactor coolant pump bus circuit q breakers sha not. exceed 1.2 seconds. For underfrequency, the delay is set so that the ime required for a signal to reach the Reactor trip breakers i after the U derfrequency Trip Setpoint is reached shall not exceed 0.3 second.

, , I loss of f otver WO N)e .

i d

I f

i l

l E

l d

1 1

1 V0GTLE - UNIT 1 B 2-7 APR 2 41986

-_---.rm.----m____--,_ . _ _ - _ . - - . ,

qW w. _ a ~ R & & .dhki+ A & k A u ;& w -La & % % h u , wi_ A M s % w n l LIMITING SAFETY SYSTEM SETTINGS ,

BASES Undervoltage and Underfrequency - Reactor Coolant Pump Busses (Continued)

On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10%

of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately

< 10% of full power equivalent); and on increasing power, reinstated automatically by P-7.

1 Turbine Trip .

t

< A Turbine trip initiates a Reactor trip. On decreasing power the Reactor

', trip from the Turbine trip is automatically blocked by P-9 (a power level of 3 approximately 50% of RATED THERMAL POWER); and on increasing power, reinstated automatically by P-9.

Safety Injection Input from ESF If a Reactor trip has not already been generated by the Reactor Trip 1 System instrumentation, the ESF automatic actuation logic channels will initiate t a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

s Reactor Trip System Interlocks The Reactor Trip System interlocks perform the following functions:

\ P-6 On increasing power P-6 allows the manual block of the Source Range trip (i.e., prevents premature block of Source Range trip), and g deenergizes the high voltage to the detectors. On decreasing power, H Source Range Level trips are automatically reactivated and high (i voltage restored.

A H P-7 On. increasing power P-7 automatically enables Reactor trips on low

. flow in more than one reactor coolant loop, reactor coolant pump i bus undervoltage and underfrequency, pressurizer low pressure and i pressurizer high level. On decreasing power, the above listed trips are automatically blocked.

h P-8 On increasing power, P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power, the L P-8 automatically biccks the above Mohsl trips.

?

k V0GTLE - UNIT 1 B 2-8 E _.___...~ _.- __ _

pAhnem v. o aiWdgiA6Esassisum.adhs.mgasawg.sya_ges l C ~

j

} l 3/4.1 REACTIVITY CONTROL SYSTEMS .

3/4.1.1 BORATION CONTROL P SHUTDOWN MARGIN - T, GREATER THAN 200*F LIMITING CONDITION FOR OPERATION u

)[ 3.1.1.1 The SHUTDOWN MARGIN shall be greater than or equal to 1.3% ak/k j for four loop operation, y[ APPLICABILITY: MODES 1, 2*, 3, and 4.

(( ACTION:

. With the SHUTDOWN MARGIN less than 1.3% ak/k, immediately initiate and continue l boration at greater than or equal to 30 gpm of a solution containing greater

? than or equal to 7000 ppm boron or equivalent until the required SHUTDOWN MARGIN 4 is restored.

y SURVEILLANCE REQUIREMENTS A$.$

4.1.1.1.1 The SHUTOOWN MARGIN shall be detarmined to be greater than or equal to 1.3% Ak/k:

I '

y a. Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after detection of an inoperable control rod (s) and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter while the rod (s) is inoperable.

If the inoperable control rod is immovable or untrippable, the above

}R 9,d required SHUTDOWN MARGIN shall be verified acceptable with an increased

allowance for the withdrawn worth of the immovable or untrippable

$ control rod (s);

h p b. When in MODE 1 or MODE 2 with K,ff greater than or equal to 1 at

$ least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that control bank withdrawal is l g within the limits of Specification 3.1.3.6; j

c. . . . . . ..___ . Yith K,ff less than 1, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to l

[T achieving reactor criticality by verifying that the predicted y] critical control rod position is within the limits of Specification g 3.1.3.6; W:., d. Prior to initial operation above 5% RATED THERMAL POWER'after each

,1 fuel loading, by consideration of the factors of Specifica-i tion 4.1.1.1.le. below, with the control banks at the maximum inser-T tion limit of Specification 3.1.3.6; and dn

! *See Special Test Exceptions Specification 3.10.1.

I h

7* V0GTLE - UNIT 1 3/4 1-1 i APR 2 4 Gg 1 m

$ l 1

I

y w:wwxw w- , . w,s acasi+~m A muni.ww w % s&Lu m 1 REACTIVITY CONTROL SYSTEMS

i 3/4.1.2 BORATION SYSTEMS FLOW PATH - SHUTDOWN

) LIMITING CONDITION FOR OPERATION j 3.1.2.1 As a minimum,.one of the following boron injection flow paths shall

be OPERABLE and capable of being powered from an OPERABLE emergency power j source:

i

a. A flow path from the boric acid storage tank via either boric acid transfer pump and a charging pump to the Reactor Coolant System if the boric acid storage tank in Specification 3.1.2.5a. is OPERABLE, or

b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank in Specification 3.1.2.5b. is OPERABLE.

APPLICABILITY: MODES 5 and 6.

1

ACTION.

1

With none of the above flow paths OPERABLE or capable of being powered from an j OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

O

! SURVEILLANCE REQUIREMENTS i

I 4.1.2.1 At least one of the above required flow paths shall be demonstrated j OPERABLE:

) a. 'nt k; t ma n g ,- 7 17: 4 S fyia p that e t :tuee.o m - l Th 7.;.ir icestere:n m-:;21544FO*7: 7.;i.+flewzpatAJeamh d t:-S iMt. -t.. 6;r =d,_: d 1 b. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, l

sealed, or otherwise secured in position, is in its correct

position.

i

} h hadevice er 7ch .s when Be borit ac'dsi's o kA u Me i k /Ygvir4l We W Sowt ves*ify /4sf N& t,?4p//cabk f//otu of /Ae. 1 aitfi//ary hulYCllQ {TESL /EV/o or 7ZSL ,/29//, /ZY/2 0/~ TZSL/2V/.f l TE.5L /29/y of TESL /Z'//S, 7*DL. /Z<//h or TDL/2Y/& YBL Zofbe or

~

v zoyoz or 7.rst coros., and TIst 20m /rTDL z "8 rzst zogoi 7ZsL om//Ae !

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- 1 l

1

. . wap M.-e ow m m=-+m s

ma===**mp y=eg e wn , . r,,<, . , . , ,

m- .-

.yt7,

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REACTIVITY CONTROL SYSTEMS ,

FLOW PATHS - OPERATING i

LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two* of the following three boron injection flow paths shall be OPERABLE:

a. The flow path from the boric acid storage tank via a boric acid transfer pump and a charging pump to the Reactor Coolant System (RCS), and s b. Two flow p.aths from the refueling water storage tank via charging pumps to the RCS.

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

i ACTION:

With only one of the above required boron injection flow paths to the RCS OPERABLE, restore at least two boron injection flow paths to the RCS 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 STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at 200*F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I SURVEILLANCE REQUIREMENTS E

L 1

4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

l I

I a. At h t =ce ; r ? d:y: by =ifying-thet4hr m er:tur: of-the.

l A'V 'N fi:n ;;th from-the4ori: ::id teoks isqc;;t:r- then=or-equakte.

US# $g 7 55'F .;h= it-famrWred=weter seeret;

, 3)'y l- b At least once per 31 days by verifying that each valve (manual, b.

[

power-optrated, or automatic) in the flow path that is not locked, I sealed, or otherwise secured in position, is in its correct position;docI

,r. At-lea st=once-per-183nonthe-duci ng-shutdown 4y=vertfHng4 hat-each-

-eutematic-valve _in thed4cw=patWee*-te4t: =rrect p;iti= on -

e safety-injection.tast-s4gna3+-end-l C #. At least once per 18 months by verifying that the flow path required by Specification 3.1.2.2a. delivers at least 30 gpm to the RCS.

JM, er.: hier, iafectiew-f1=;;th-ie-requiredatode-her-the vre-ef er,e-oe-more oFthz *GS ceM h;;; ': h:: thwer :;;;l to -

P I

A V0GTLE - UNIT 1 3/41-8 APR 2 41986 t

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Insert for Page 3/4 1-8 i

i I

At least once per 7 days when the boric acid storage tank is a required water source, verify that the applicable portions of the auxiliary building (TISL 12410 or TISL 12411, TISL 12412 or TISL 12413, TISL 12414 or 12415, TISL 12416 or TISL 12417, TISL 20900 or TISL 20901, TISL 20902 or TISL 20903,wid TISL 20904 or TISL 20905) and the portions of the flow path for which ambient temperature' indication are not provided are >65 F.

4 i

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1 -

1 REACTIVITY CONTROL SYSTEMS .

CHARGING PUMP - SHUTDOWN 3 1 S LIMITING CONDITION FOR OPERATION l 1

i 4/ /eas/

! 3.1.2.3Agne charging pump in the boron injection flow path required by i

? Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

1 APPLICABILITY: MODES 5 and 6.

4 ACTION: .

, With no charging pump OPERABLE or capable of being powered from an OPERABLE 5 emergency power source, suspend all operations involving CORE ALTERATIONS or i; positive reactivity changes.

.1

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A f

~

SURVEILLANCE REQUIREMENTS 4

1

, 4eir24.1 The-:bre~ required che@; p:;p :h:1-?~te-demenstrated-CAMASt;f-i>y

-'ff ;, =--re;irculation-??=, th:t -: diffe.. J hi p x:: = : n;; the pump-p:id4ssdevelope*whefFt::ted purn;;r.t--to

+ L gred er 4h:. m :;rl-t

<;;;ificetici. 0.0.5.

s 2 Mvh2rar2-MMharging-pumpspeme%dimy-the ab=: :;;ir-ed OPEM0tE=pm shall 5:d;;snstrateL1- --pe-91 -:t -leasWe=pec23L4.;y:,-erept t the-p] tenetems.v sseL-baad-4 .emauedWj v:ri*yimpthat-%e-meter civetrit-irreekers u

are=secu M J: " 77 7 -it4en.

l J/./.2.3 /// / emf onceper /8 mon /4 f4e a6 ore rqwrec/

l ihj C/larjij$

NW Rj@mp s/]al/ 4e clemonHrs&d J./. 2./OMg40tE 13 by otdpN ryi/Rerl by SHC/fts/MVj /

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REACTIVITY CONTROL SYSTEMS

[.HARGINGPUMPS-OPERATING L

LIMITING CONDITION FOR OPERATION h

I 3.1.2.4 At least two charging pumps shall be OPERABLE.

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

1; ACTICH:

}

? <

With only one charging pump OPERABLE, restore at least two charging pumps 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 STANDBY and borated to a SHUTDOWN MARGIN equivalent to at le.ast 1% Ak/k at 200'F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two charging pumps to OPERABLE status within the next

] 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

1 l

i SURVEILLANCE REQUIREMENTS i

a

< 4.1. 2.4.1-4t - ? ::&tr thaeging-paps:-she4 be-demons tra ted-OPERABLE- by L

variMesem-recimleuw-f w,= eau ueregi+4.meesee-eerees-eech- 1

;: f - :ste-tW'.ral-te- - ;;fd h d:velope.2 wherr-tested-pursuant ~ l 1 t: S;;;.ificnien-44.-h I Airer &2hrgh; ;r,.:, ::::;t +% d:/c allew;d4PERABLE;gesha01-M ^ :::t-eted '=;raMe-et-4eesMneep:- 31 d y: *n.er th; 4egerature sLee nf *" 8%r-Cee4snt-Systema ( '5) ecid hg: frhtheur_

eg!- t: [276387+by= vee 4fying-thetthe ;.ebe-c;h4breder:-:veMn

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V0GTLE - UNIT 1 3/4 1-10 APR 2 4198c I

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REACTIVITY CONTROL SYSTEMS BORATED WATER SOURCE - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.5 As a minimum, one of the following borated water sources shall be OPERABLE:

a. A Boric Acid Storage Tank with:

1) A minimum contained borated water volume of gallons,

2) A bo'ron concentration between 7000 ppm and 7700 ppm, and

3) A minimum solution temperature of 65'F.

b. The refueling water storage tank (RWST) with:

1) A minimum contained borated water volume of gallons,

2) A boron concentration between 2000 ppm and 2200 ppm, and

3) A minimum solution temperature of 50'F.

APPLICABILITY: MODES 5 and 6.

ACTION:

With no borated water source OPERABLE, 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: I

1) Verifying the boron concentration of the water, )

2) Verifying the contained borated water volume, and

3) ifyi~; the -5:ric-aeid-+t:r:g: u nt seletien2temperatur- d r.

ht f:=the-eestce=of-borated:watsc

/ 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 it

[ is the source of borated water and the outside air temperature is less than 50*F.

q)/yn & bsnt GCid s/om $2'/ $ U N & " ' " "

o,gj ,y,s. ambiW /cm,7ed c'se ot= M6 bdAC Ol'O' Tl#"9E N"A l d'"? /'l0*16 l4N '0l"'#7 7aum 5 ots, twify VAe bonC

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1

'l l REACTIVITY CONTROL SYSTEMS ,h i

l SURVEILLANCE REQUIREMENTS a i

L L

i 4.1.2.6 Each borated water source shall be demonstrated OPERABLE:

l' L a. At least once per 7 days by:

L L 1) Verifying the boron concentration in the water, a

W s 2) Verifying the contained borated water volume of the water r

4 source, and y .

j 3) -Ve,4fyb; thehic ^:5-d Ster;g 0 ;te; 3 aleth , t:: cratere.

- d en itM s=tb w o,ce ef-berated weter.

5

{ 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 outside air temperature is c'" r less than 50*F, er ;r:: tee tr.:n [1^^: T.,

1 k/hm /de boric acidsfora e / ant is ae sowce a,c hemfe/yafei-and /jfe s w bi g fe,,7 p ,4 y e o f fje boric cciW .rforeft /w;/c f&'n So/d}jon Ib<&5'*f,kri}'Mc hwic qc/c/ .5/srejr Jaffg

-fe m f ars f g/ c i r y g5op k

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3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT L NITING CONDITION FOR OPERATION 3.1. All full-length shutdown and control rods shall be OPERABLE a positi ed within i 12 steps (indicated position) of their group step / counter demand asition.

APPLICABI TY: MODES 1* and 2*.

ACTION: .

a.

Withogeormorefull-lengthrodsinoperabledue/nterferenceorto as a resylt of excessive friction or mechanical being immovable known to be untrippable, determine that the SHUTDOWN MARGIN require-ment of Sphcification 3.1.1.1 is satisfied w t'hin 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY ithin 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With more than ne full-length rod inope a le or misaligned from the group step ce nter demand position more than i 12 steps (indicated positi , be in HOT STANDB within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

c. With one full-lengt rod trippable t inoperable due to causes other than addressed ACTIONa./above,ormisalignedfrom its group step counter demand he,ight by more than i 12 steps (indicated position), P ER OP RATION may continue provided that within I hour:

1. The rod is restored to ERABLE status within the above r

alignment requiremejt's,

2. The rod is declarpd inoper le and the remainder of the rods in

. the group with the inoperabig rod are aligned to within i 12 steps of the inoperable rod while a intaining the rod sequence and insertion limits of Figures [3. -1] and [3.1-2]. The THERMAL POWERleveyshallberestricted .ursuant to Specification

[3.1.3.6 /during se sequent opera n, or

)

3. Therodisdeclaredinoperableandt(eSHUTDOWNMARGIN req t'ement of Specification 3.1.1.1 19 satisfied. POWER OP TION may then continue provided tN t:

a A reevaluation of each accident analysis of Table 3.1-1 is performed within 5 days; this reevaluh ion shall confirm that the previously analyzed results o these accidents remain valid for the duration of operati under these conditions; b) The SHUTDOWN MARGIN requirement of Specific ion 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

ffeeSpecialTestExceptionsSpecifications3.10.2and3.10.3. !

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I V0GTLE - UNIT 1 3/4 1-14 4

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REACTIVITY CONTROL SYSTEMS 3/4.1.3 MOVABLE CONTROL ASSEMBLIES GROUP HEIGHT _

LIMITING CONDITION FOR OPERATION -

3.1.3.1 All full-length shutdown and control rods shall be OPERABLE and positioned within

12 steps (indicated position) of their group demand position.

APPLICABILITY: MODES 1* and 2*.

AC, TION:

a. With one or more full-length rods inoperable due to being immovable es a result of excessive friction or mechanical interference or known to be untrippable, determine that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is satisfied within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one full-length rod trippable but inoperable due to causes ether than addressed by ACTION a., above, or misaligned from its group demand height by more than 2 12 steps (indicated position), POWER OPERATION may continue provided that within 1 hour:

1. The rod is restored to OPERABLE status within the above

alignment requirements, or

2. The rod is declared inoperable and the remainder of the rods in the group with the inoperable rod are aligned to within 1 12 steps of the inoperable rod while maintaining the rod sequence and insertion ilmits of Figure 3.1-1. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, or

3. The rcd is declared inoperable and the SHUTOOWN MARGIN requirement of Specification 3.1.1.1 is satisfied. POWER OPERATION may tuen continue provided that:

a) A reevaluation of each accident analysis of Table 3.1-1 is perferned within 5 days; this reevaluation shall confirm that the previously analyzed results of these accidents remain valid for the duration of operation under these conditions; b) THE SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is determined at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ; e Special Test Exceptions Specifications 3.10.2 and 3.10.3.

DIActa-car!Gii t" tits 1 .i 2- 3/4 1-15 VO6r TLE- UNIT f

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REACTIVITY CONTROL SYSTEMS

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LIMITING CONDITION FOR OPERATION JBfT

\

-r ACTJON (Continued) j c) A power distribution map is obtained from the movable k incore detectors and Fq(Z) and F" areverif*eItobe within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and

'd The THERMAL POWER level is reduced to Jess than or

. equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months th,e'High Neutron Flux ip Setpoint is reduced to less than or equal to 85%

o RATED THERMAL POWER.

3 l

SURVEILLANCE REQUIREMENTS

? 4.1.3.1.1 The position of eac/

h full

\

, length d shall be determined to be Z withinthegroupdemandlipitbyverifyingth ndividual rod positions at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time i (ervals when the rod position deviation monitor is inop'erable, then verify the g oup positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full / length rod not fully inserted in he core shall be determinedtobeOfERABLEbymovementofatleast10st s in any one direction at least once per 31 days.

/

1 s

i l

1 V0GTLE - UNIT 1 3/4 1-15 APR 21 TT* M- %_ :

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REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION ,,

ACTION (Continued) c) A power distribution map is obtained from the movable incoredetectorsandF(Z)andFhareverifiedtobe q

within their limits within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ; and d) The THERMAL POWER level is reduced to less than or equal to 75% of RATED THERMAL POWER within the next hour and within the following 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the High Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER. ,

c. With more than one rod trippable but inoperable due to causes other

1. hen suoresseQ Dy AL.;l10N a above, POWER OPERATION may continue provided that:

1. Within I hour, the remainder of the rods in the bank (s) with the inoperable rods are aligned to within i 12 steps of the inoperable rods while maintaining the rod sequence and insertion limits of Figure 3.1-la or Figure 3.1-lb, as applicable. The THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation, and )

2. The inoperable rods are restored to OPERABLE status within 72' hours.

d. With more than one rod misaligned from its group demand position by more than i 12 steps (indicated position), be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each full-length rod shall be determined to be within the group demand limit by verifying the individual rod positions at ,

least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months except during time intervals when the Rod Position I Deviation Monitor is inoperable, then verify the group positions at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.1.3.1.2 Each full-length rod not fully inserted in the core shall be deter-mined to be OPERABLE by movement of at least 10 steps in any one direction at least once per 31 days.

l

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TABLE 3.1-1 b

(

ACCIDENT ANALYSES REQUIRING REEVALUATION IN THE EVENT OF AN INOPERABLE FULL-LENGTH R00 Rod Cluster Control Assembly Insertion Characteristics Rod Clus Control Assembly Misalignment Loss of Reactor Coolant from Small Ruptured Pipes or from Cracks in

- Large Pipes Wh'ich Actuates the Emergency Core Cooling fstem Single Rod Cluste Control Assembly Withdrawal at Full Power

/

Major Reactor Coolan xSystem Pipe Ruptures (Loss-of-Coolant Acci, dent)

Major Secondary Coolant 5'ystem Pipe Rupture Rupture of a Control Rod D e Mechanism ousing (Rod Cluster Control Assembly Ejection) l 1

1 4

5 l l

V0GTLE - UNIT 1 3/4 1-16 APR 24

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i l

I :

TABLE 3.1-1 ACCIDENT ANALYSES REQUIRING REEVALUATION ,

IN THE EVENT OF AN INOPERABLE CONTROL OR SHUTOOWN ROD Rod Cluster Control Assembly Insertion Characteristics Rod Cluster Control Assembly Misalignment Decrease in Reactor Coolant Inventory .

Inadvertent Opening of a Pressurizer Safety or Relief Valve Break in Instrument Line or Other Lines from Reactor Coolant Pressure Boundary That Penetrate Containment Loss-of-Coolant-Accidents Increase in Heat Removal by the Secondary System hteam System Piping Rupture) !

Spectrum of Rod Cluster Control Assembly Ejection Accidents.

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. REACTIVITY CONTROL SYSTEMS , !

. 1

R00 DROP TIME LIMITING CO'NDITION FOR OPERATION

(

l 3.1.3.4 The individual f;"-?r;th shutdown and control rod drop time from I-the fully withdrawn position shall be less than or equal to 2.2 seconds from b beginning of decay of stationary gripper coil voltage to dashpot entry with:

a. T,yg greater than or equal to 551'F, and I

[ b. All reactor coolant pumps operating.

APPLICABILITY: MODES 1 and 2.

[

! ACTION:

I t

With the drop time of any f;?', k n;th rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to i MODE 1 or 2.

I v SURVEILLANCE REQUIREMENTS t; 4.1.3.4 The rod drop time of ' !' ';;;th rods shall be demonstrated through measurement prior to reactor criticality:

[ a. For all rods following each removal of the reactor vessel head, i

6- b. For specifically affected individual rods following any maintenance i on or modification to the Control Rod Drive System which could affect the drop time of those specific rods, and l c. At least once per 18 months.

I I

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i f V0GTLE - UNIT 1 3/4 1-19 . l

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XMiedhnh@$%%FM5WMMMhaCA ws.r udibbMWaw,MENWGb2nded POWER DISTRIBUTION LIMITS .

LIMITING CONDITION FOR OPERATION ACTION (Continued)

c. With the indicated AFD outside of the above required target band for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months of cumulative penalty deviation time during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and with THERMAL POWER less than 50% but greater than 15% of RATED THERMAL POWER, the THERMAL POWER shall not be increased equal to or greater than 50% of RATED THERMAL POWER until the indicated A is within the above required target ban Cwn nWirt- pt ch'MMon hts been fra'ucal-Jo kss b'esej (/' add 1 Itput W/L h1/Ze prexious 24 1]ou .

SURVEILLANCE REQUIREMENTS l

4.2.1.1 The indicated AFD shall be determined to be within its limits during POWER OPERATION above 15% of RATED THERMAL POWER by:

a. Monitoring the indicated AFD for each OPERABLE excore channel:

1) At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and -unH/

2) At least once per hour /f- the-fdr:t 21 hcces-afta r;;teing-the AFD Monitor Alarmg ot OPER

b. Monitoring and logging thetedi AFD di$ggs for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and at least once per 30 minutes thereafter, when the AFD Monitor Alarm is inoperable. The logged values of the indicated AFD shall be assumed i to exist during the interval preceding each logging.

4.2.1.2 The indicated AFD shall be considered outside of its target band when l two or more OPERABLE excore channels are indicating the AFD to be outside the target band. Penalty deviation outside of the above required target band shall be accumulated on a time basis of:

a. One minute penalty deviation for each 1 minute of POWER OPERATION

. outside of the target band at THERMAL POWER levels equal to or above 50% of RATED THERMAL POWER, and

b. One-half minute penalty deviation for each 1 minute of POWER OPERATION outside of the target band at THERMAL POWER levels between 15% and 50% of RATED THERMAL POWER.

! 4.2.1.3 The target flux difference of each OPERABLE excore channel shall be determined by measurement at least once per 92 Effective Full Power Days.

The provisions of Specification 4.0.4 are not applicable.

4.2.1.4 The target flux difference shall be updated at least once per 31 Effective Full Power Days by either determining the target flux difference

pursuant to Specification 4.2.1.3 above or by linear interpolation between the most recently measured value and 0% at the end of the cycle life. The provi-sions of Specification 4.0.4 are not applicable.

V0GTLE - UNIT 1 3/4 2-2 APR 24 N

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POWER DISTRIBUTION LIMITS ,

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - Fg b

c LIMITING CONDITION FOR OPERATION k

3.2.2 F (Z) shall be limited by the following relationships:

4 2 F9 (Z) $fk.30[K(Z)] for P > 0.5 g Fq (Z) 52..ST[K(Z)] for P 10.5

, and Wher'e: P _ THERMAL POWER RATED THERMAL POWER K(Z) = the function obtained from Figure 3.2-2 for a given core height location.

k APPLICABILITY: MODE 1.

ACTION

With Fg (Z) exceeding its limit:

a. Reduce THERMAL POWER at least 1% for each 1% qF (Z) exceeds the limit within 15 minutes and similarly reduce the Power L Range Neutron Flux-High Trip Setpoints within the next 4

., hours; POWER OPERATION may proceed for up to a total of 72

[ hours; subsequent POWER OPERATION may proceed provided the

. Overpower AT Trip Setpoints have been reduced at least 1%

y for each 1% F q (Z) exceeds the limit; and i b. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit re-

- quired by ACTION a., above; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be 9

g within its limit.

h i

4 e

APR 24 5 V0GTLE - UNIT 1 3/4 2-4 s

o r , .

._,y.,,r...,. ..

--..r,....,

% c.+assmus3y.smaavnwmay wm&m;svs;.awhm.smesm;.y,msm 4

l POWER DISTRIBUTION LIMITS .

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

' shall be evaluated to determine if F (Z) is within its limit by:

4.2.2.2 F xy q i a. Using the movable incore detectors to obtain a power distribution

/: map at any THERMAL POWER greater than 5% of RATED THERMAL POWE eM m exceedy 75% of RATED menww /wwtg 6//owin each neef ,

j b. Increasing the measured F xy component of the po er distribution ma

.a by 3% to' account for manufacturing tolerances and further increasing

] the value by 5% to account for measurement uncertainties,

> c. Comparing the F xy computed (Fx) btained in Specification 4.2.2.2b.,

above to:

1) The F limits for RATED THERMAL POWER (FRTP) for the appropriate xy x measured core planes given in Specification 4.2.2.2e. and f.,

- below, and

2) The relationship: j F =FRTP [1+0.2(1-P)],

x Where F ' is the limit for fractional THERMAL POWER operation express as a function of F RTP xy and P is the fraction of RATED xy was measured.

THERMAL POWER at which F xy according to the following schedule:

. d. Remeasuring F

1) When F is greater than the F xRTP limit for the appropriate l

measured core plane but less than the F*Y relationship, additional RTP power distribution maps shall be taken and F C compared to F xy xy and Fxy' either:

i a) Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding by 20% of RATED THERMAL C

POWER or greater, the THERMAL POWER at which F*Y was last determined, or i

b) At least once per 31 Effective Full Power Days (EFPD),

whichever occurs first.

9 4

V0GTLE - UNIT 1 3/4 2-6

'+'T*'9. 6i 7-yM', ,s*,, . t *tg- - TP9F" F w'2*'*"vePep***e 7m p -, _ -- . _ _ , , _

7 K_.

w =. _ m _wmm m.wnm, . .. ., c r . ,,~ n n., - u.mn,n ...,,,,~,c..- ~ ~ _ . , _ .- .,-

P 1

4 TABLE 3.3-1

{ p

. .g

, REACTOR TRIP SYSTEM INSTRUMENTATION jij

1 i;; b MINIMUM Y

} TOTAL NO. CHANNELS CHANNELS APPLICABLE N E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION y

... --e O A - 1. Manual Reactor Trip 2 1 2 1 2 a 1 E j 2 1 2 30' 4k' 5**' 11 h

r.

,jj

~

2. Power Range, Neutron Flux d b I:

^

a.

b.

High Setpoint Low Setpoint 4

4 2

2 3

3 1, 2/

,2 2T 2W y

g.

b t:

3. Power Range, Neutron Flux 4 2 3 1, 2 2F E High Positive Rate @

4. Power Range, Neutron Flux, 4 2 3 1, 2 2

' .) High Negative Rate j

d I;f

.j w

5. Intermediate Range, Neutron Flux 2 1 2 Id2 3

+

1

{

E 6. Source Range, Neutron Flux

a. Startup 2 1 2 F C,

4 ff

b. Shutdown 2 1 2 3,4,5 5 g

.I 7. Overtemperature AT a o F 2T 2I l

'J 4 hI Four Loop PlantN 4 M

N , Four Loop Operation 4 2 i :L, 2

8. Overpower AT T T I 2d [

UIir Loop P1 ii 6

]!':,i our Loop Operatio 4 2 3 1, 2 6F q b

9 (

=

, 9 u

.. eo p

$ b [.'

j L

.n ~ ~.-- _mv n-n- _w m.,,,,,- . n -.~-.....,n.n,-~.,,--,-

. ., ~ ~ ~ ~ ,.

F 1l !:

r i.

3

[l d <

TABLE 3.3-1 (Continued) ll 3 h REACTOR TRIP SYSTEM INSTRUMENTATION m

9]

j TOTAL NO. CHANNELS MINIMUM CHANNELS APPLICABLE y

Y b< .E.

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION ft j

w 9. Pressurizer Pressure--Low 4 2 3 1 f 6F b

k' j . [-

Pressurizer Pressure--High h ll

] 10. 4 2 3 . 1, 2 6F p Pressurizer Water Level--High # 6h

11. 3 2 2 I I is

12. Reactor Coolant Flow--Low b $

a. Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in h j any oper- each oper-1 6#* tj g

1 ating loop ating loop d d p b s i b. Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1 6F t7 4

g below P-8) two oper- each oper-q:j

ating loops ating loop f(

13. Steam Generato Water 4/sta. gen. 2/sta. gen. 3/sta. gen. 1, 2 6 H Level--Low-Low in any oper- each oper- p ating stm. ating stm.

"]i gen. gen. ,

'j 14. Undervoltage--Reactor Coolant j ,

Pumps 4-2/ bus 2-1/ bus 3 1

6#

b

$'A 15. Underfrequency--Reactor Coolant g b

]

Pumps 4-2/ bus 2-1/ bus 3 1 6F 3 16. Turbine Trip 6

[i a. Low Fluid Oil Pressure 3 2 2 l b 6Fg j b. Turbine Stop Valve Closure 4 4 1 1 12#6 J r y : *sa en Syn + 3a.s.s y. g w

.q $

2 .

-?

1 ;

Il TABLE 3.3-1 (Continued) ,

k REACTOR TRIP SYSTEM INSTRUMENTATION :

1 - t 1 MINIMUM l' j TOTAL NO. CHANNELS CHANNELS APPLICABLE $

FUNCTIONAL UNIT OF CHANNELS TO TRIP . OPERABLE MODES ACTION

]a -

j w 17. Safety Injection Input f i from ESF 2 1 2 1, 2 10 --

. 18. Reactor Trip System Interlocks h

-j a. Intermediate Range c ?.

3 Neutron Flux, P-6 2 1 2 2P 8 $ -

J j b. Low Power Reactor- ;f;

,s. Trips Block, P-7 -

t j P-10 Input 4 2 3 1 8 I A or M P-13 Input 2 1 2 1 8

{

'j T

c. Power Range Neutron

{c y Flux, P-8 4 2 3 1 8 d J H a d. Power Range Neutron 4 2 3 1 8 :

Flux, P-9 d e. Power Range Neutron N

'd Flux, P-10 4 2 3 1,2 8 b

)

N f. Turbine Impulse Chamber

': Pressure, P-13 2 1 2 1 8 a] '

I

! 19. Reactor Trip Breakers 2 1 2 1,a2 10, 13 3*f'4kSb

/

2 1 2 11

20. Automatic Trip and Interlock 2 1 2 1,a2 a a 10 4i SW Logic 2 1 2 3**, 11 M b q

l $ b

J

-j .

-Me6Mina=%W nh.;6:.ixh&Giu.JGL.scs2bikidaMhuuhnciM4L%M L%Miku s

TABLE 3.3-1 (Continued) ,

TABLE NOTATIONS

! G en the Reactor Trip System breakers are in the closed position and the

) Control Rod Drive System is capable of rod withdrawal. j ove the P-7 (Low Power Reactor Trip Block) setpoint.

! 6 d ove the P-9 (Reactor Trip on Turbine Trip Interlock) setpoint.

t -

b @The provisions of Specification 3.0.4 are not applicable.

C. low the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

d h low the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

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

} h Abox Me $-8(Si;tf Loof larr of flad)SeQond i ACTION STATEMENTS

) ACTION 1 - With the number of OPERABLE channels one less than the Minimum t- Channels OPERABLE requirement, restore 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, STARTUP and/or POWER OPERATION may proceed

! provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition g within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , 1

b. The Minimum Channels OPERABLE requirement is met; however, T the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months I for surveillance testing of other channels per Specification 4.3.1.1, and j c. Either, THERMAL POWER is restricted to less than or equal 5 to 75% of RATED THERMAL POWER and the Power Range Neutron

[ Flux Trip Setpoint is reduced to less than or equal to g 85% of RATED THERMAL POWER Githin 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; or, the t QUADRANT POWER TILT RATIO is monitored at least once per

[

1 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per Specification 4.2.4.2.

?

+

b.

V0GTLE - UNIT 1 3/4 3-5 APR 2 41986

,v--n- , er m, x ye r-over-*=7-=, . - . mm~m ypmy -ve mi- ~?---,,. g ym ng7~y s;,,; g~rmr*7,-, y~_ y sw3nyym.gs re y m;mm,

s.m _;.m=;gezu;.smagawa.lahkw::rugmaw;xhw.hpMu qww TABLE 3.3-1 (Continued) ,

ACTION STATEMENTS (Continued)

ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement and with the THERMAL POWER level:

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

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 )

Setpoint, and

b. Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the

inoperable channel to OPERABLE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, suspend all operations involving positive ieactivity changes. ,

1 ACTION 5 - With the number of OPERABLE channels one less than the Minimum ;

Channels OPERABLE requirement, restore the inoperable channel i to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip i System breakers, suspend all operations involving positive i reactivity changes and verify valves 1208-U4-175, 1208-U4-177, 1208-U4-183, and 1208-U4-178 are closed and secured in position within the next hour.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and 9

b. The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to @ hours for surveillance testing of other channels per Specification 4.3.1.1.

ACTION 7 - (Not used)

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months determine by observation of the associated permissive window (s) that the interlock is in its required state for the l existing plant condition, or apply Specification 3.0.3.

V0GTLE - UNIT 1 3/4 3-6 gg i -

I

. ~ . - - _ ,~ _. 7 .__7, w,m,.,,,.m

4-uMW - Am;cAann.aawm TABLE 3.3-1 (Continued) ,

ACTION STATEMENTS (Continued) k f ACTION 9 - (Not used)

$ ACTION 10 - With the number of OPERABLE channels one less than the Minimum T' Channels OPERABLE requirement, be in at least HOT STANDBY I

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to f 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

t ACTION 11 - With the number of OPERABLE channels one less than the Minimum k Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or open the Reactor Trip System breakers within the next hour.

ACTION 12 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 13 - With one of the diverse trip features (undervoltage or shunt trip

'. attachment) inoperable restore it to OPERABLE status within 48 i hours or declare the breaker inoperable and apply ACTION 10. The E

breaker shall not be bypassed while one of the diverse trip features j is inoperable except for the time required for performing mainte '

f nance to restore the breaker to OPERABLE status.

i h

i k

i i

r e

f :

i h

I t

i i t 1 V0GTLE - UNIT 1 3/4 3-7 l APR 2 41986 f

t .

1--

r . .

...t..,.- . . . _ _ _ , _ . . _ . , _ , _ , _ . , _ .

cr m a mm m m . m ,wa ww_ m m-i I

a ?

6 c)

TABLE 4.3-1 E

h

-4 I E REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS [

t c TRIP

. .) $

ANALOG

ACTUATING MODES FOR [

i CHANNEL DEVICE milch 0

")!

FUNCTIONAL UNIT CHANNEL CHECK CHANNEL CALIBRATION OPERATIONAL TEST OPERATIONAL TEST ACTUATION LOGIC TEST SURVEILLANCE IS REQUIRED I0

1. Manual Reactor Trip N.A. N.A. N.A. R(14) N.A. 1,2,3 4 5 i

2. Power Range, Neutron Flux

a. High Setpoint S D(2,4), Q(17) N.A. N.A. 1, 2 E

L M(3,4), k Q(4, 6), E R(4, 5) cl 2

b. Low Setpoint S R(4) SU(1) N.A. N.A. 1*#2 [.

E i R 3. Power Range, Neutron Flux, N.A. R(4) Q(17) N.A. N.A. 1, 2

{'

-

High Positive Rate {

c T b g 4. Power Range, Neutron Flux, N.A. R(4) Q(17) N.A. N.A. 1, 2 High Negative Rate C

'] d b

5. Intermediate Range, S R(4, 5) S/U(1) N.A. N.A. 1T2 1/

Neutron Flux b

, 6. Source Range, Neutron Flux S R(4, 5, S/U(1),Q(9,17) N.A. N.A. 3, 4, 5

7. Overtemperature AT S Q(17) N.A. N.A. 1, 2

(

R(g) g i 8. Overpower AT S R Q(17) N.A. N.A. 1, 2 b

., 9. Pressurizer Pressure--Low S R Q(17 N.A. N.A. 1 8

3 % lA E

=

to

10. Pressurizer Pressure--High S R Q(17,h)/ N.A. N.A. 1, 2 0 I 11. Pressurizer Water Level-- S R Q(17) N.A. N.A. 1" .

High '

b

12. Reactor Coolant Flow--Low S R Q(17) N.A. N.A. 1* J i

k

!<g TABLE 4.3-1 (Continued) 11 <

3 8 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 1 ;d .

"' TRIP

' ANALOG ACTUATING MODES FOR CHANNEL DEVICE WHICH

'.1' E SURVEILLANCE CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION Q CALIBRATION TEST TEST LOGIC TEST IS REQUIRED

,_. FUNCTIONAL UNIT CilECK

13. Steam Generator Water Level-- S R Q(17,18) N.A. N.A. 1, 2 t Low-Low i

e

14. Undervoltage - Reactor Coolant N.A. R N.A. #Q(l'7) N.A. l

~}

j Pumps (

0

15. Underfrequency - Reactor N.A. R N.A. E Q()7) N.A. 1

] Coolant Pumps f i:.

] ti*

>g R 16. Turbine Trip b

j b a. Low Fluid Oil Pressure N.A. R S/U,(1,10) N.A. N.A. 1 [

] h b. Turbine Stop Valve N.A. R N.A. S/U(1,10) N.A. Ib [P t Closure

! h

17. Safety Injection Input from N.A. N.A. N.A. R N.A. 1, 2 r

{ ESF i j

1 18. Reactor Trip System Interlocks -

d o a. Intermediate Range c h j Neutron Flux, P-6 N.A. R(4) R N.A. N.A. 2* t E

Mg E"}'"{T --

g,g, ,

._gg4, , _.

,,,n, _n,n, . __ g

!j b /. Power Range Neutron N.A.

7 33 jj Flux, P-8 N.A. R(4) R N.A. 1

% C/. Power Range Neutron N.A. N.A.

i

= Flux, P-9 N.A. R(4) R I a to '

' j] A

. LC 0

j

.; }

_ w v. c. . _m,, , _ m m.,, <. n .nm___,me., -_

i 9

l TABLE 4.3-1 (Continued)

< I 8 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS ,7 g

m U

TRIP

' MODES FOR k l ANALOG ACTUATING j E CHANNEL DEVICE WHICH h ACTUATION SURVEILLANCE L i Z CHANNEL CHANNEL OPERATIONAL OPERATIONAL j ,

s FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED ;.

y:

18. Reactor Trip System Interlocks (Continued) y N "

f!]

e. Powe: Range :i

Neutron Flux, P-10 N.A. R(4) R ,

N.A. N.A. 1, 2 i

f. Turbine Impulse Chamber N.A. N.A. 1 ;

i Pressure, P-13 N.A. R R

" a a cp

19. Reactor Trip Breaker N.A. N.A. N.A. M(7, 11) N.A. 1, ?, 3T 4T N A R 1,2,3y4pf,'/

d

20. Automatic Trip and Interlock N.A. N.A. N.A. N.A. M(7) , ,7j 1 Y Logic . x I

j

21. Reactor Trip Bypass Breaker N.A. N.A. N.A. M(15),R(16) N.A. 1, 2, 3% , 4p ,Q 4ii) 3 F 1 ,

.j i a

A W

l:s 0

'.I j j j =

=

P

.I 1 eo

p *

.! h.

e -

a i

p' l ti

-cagos.n:maagggggg g.gg g g ggg ;g g g TABLE 4.3-1 (Continued) .

TABLE NOTATIONS w hen the Reactor Trip System breakers are closed and the Control Rod Drive 1 System is capable of rod withdrawal.

b. Agy A 9(2cadWYnjo on 77vbn 7?ny LtWhek)SeQoMf; C..@ 5elow P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

cl.Melow P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

8/ock 6.(1)AhnIf not P-7performed

[ Low AwerKeacAr~

in previous Tydays. )sep;nf; F 31 (2) Comparison of calorimetric to excore power indication above 15% of RATED THERMAL POWER.

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

[ (3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE 5 above 15% of RATED THERMAL POWER. Recalibrate if the absolute difference is greater than or equal to 3%. The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

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

(5) Detector plateau curves shall be obtained, and evaluated and compared to manufacturer's data. For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

f (6) Incore - Excore Calibration, above'75% of RATED THERMAL POWER. The I provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

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

S j (8) Not used (9) Quarterly surveillance in MODES 3*, 4*, and 5* shall also include verifi-l cation that permissives P-6 and P-10 are in their required state for ll existing plant conditions by observation of the permissive window.

j Quarterly surveillance shall include verification of the Source Range High

& Flux at Shutdown Alarm Setpoint of less than or equal to 3.16 times

{ background.

s t

V0GTLE - UNIT 1 3/4 3-13 APR 2 41986 3- - - m ,,. .y, _y__.-..=.,.._. . . . . , . . ;., c .__ . .

m= __. . _ _ -4.m y - - - . . - - - - ~ - - - ~ ~ ~ - -

~=__m-_ - - - - - , - - - - , . , - ~ ~ - - , - - . - -

?

.t <

y TABLE 3.3-3

il is n Q ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION I

E ~

E

[ MINIMUM Y z TOTAL NO. CHANNELS CHANNELS APPLICABLE H FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION i:

e--

1 1. Safety Injection (Reactor ...

Trip, Feedwater Isolation, . 1 Component Cooling Water, Control Room j Emergency Mode Actuation, Start Diesel k

.j Generators, Containment Cooling Fans, 3

j Nuclear Service Cooling Water, Contain- j 1 ment Isolation, Containment Ventilation ti

) Isolation, and Auxiliary Feedwater 3

~j Motor-Driven Pumps). g i e.

' i; a. Manual Initiation 2 1 2 1,2,3,4 19 C

w a.

f D b. Automatic Actuation 2 1 2 1,2,3,4 14 [

'e- w Logic and Actuation :

y Relays L

}

j c. Containment 3 2 2 1,2,3 IS E lh Pressure--High-1 , '

cl gj d. Pressurizer 4 2 3 1, 2, 3$- 20 # (

-:1 Pressure--Low F li '

jj e. Steam Line Pressure-- 3/ steam line 2/ steam line 2/ stem line 1, 2, 3h 15 h !!

[

g Low any steam line

{)

$, . ~

d 4

}d q

=

?

f>

q

q +

SSee- also SpeciRcMien 33.5 6-

' l

h

([

r.

TABLE 3.3-3 (Continued) ;

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION iI p

[ MINIMUM ,;

< g TOTAL NO. CHANNELS CHANNELS APPLICABLE J/.

H FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION [.

s C

2. Containment Spray [;

c 3

a. Manual Initiation 2 1 with 2 ' 1, 2, 3, 4 19 E s 2 coincident i switches .

-J -i

b. Automatic Actuation 2 1 2 1,2,3,4 14 T Logic and Actuation Relays rf

s p 1' c. Contai ment Pressure-- 4 2 3 1,2,3 17 q High-3g- {

, * [.

, y 3. Containment Isolation

,{.

l a. Phase "A" Isolation /

1

}

1) 2)

Manual Initiation Automatic Actuation 2

2 1

1 2

2 1,2,3,4 1,2,3,4 19 14

{

Logic and Actuation [:

W Relays gg gfgf

{

l 3) Safety Injection See Mem-1. above for all Safety Injection initiating functions and [-

i requirements. [

4) Containment Area Radiat-jon(HighRange 2 1 2 1,2,3,4 bF .

{5 QRE-000S, R2-lO%)) 2] ,y y;

y 'f yc % . alw _Spaf wm E3.s. u.

v

[."

iV

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

TABLE'3.3-3 (Continued)

.<- t 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 1

~

r MINIMUM 4

, E TOTAL NO. CHANNELS CHANNELS APPLICABLE p

'i FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION k.

- E

b. Containment Ventilation E Isolation .[

1) Manur:1 Initiation 2 1 2 1,2,3,4,6 0 18,14 E I 2) Automatic Actuation 2 1 2 1,2,3,4,6 E 18,14 Logic and Actuation Relays 4 u

pnc+1malTinif .

4

3) Safety Injection See 14em.1. above for all Safety Injection initiating functions and i

]i R requirements. f; i

) Containment Area n Radiation (Low Range)* 1,2,3,4,6 0

~

Y (g,_pg4 p- 2 1 2 18, % E g 5) ontainment Ventilation I Radiaton b

1) Particulate (R6-WA) 1 1 El 1,2,3,4,6 C 18, % j.

I ii) Iodine ULE-Z5(58) 1 1 ti 1,2,3,4,6- 0 18,1(, [

1 1 El 1, 2, 3, 4, 6'- 18;1(o i lii) Gaseous (AE-6(6C) [f j - 4. Steam Line Isolation ,

a. Manual Initiation b,

j. 1) Individual 2/ steam line 1/ steam line 1/ operating 1, 2f 3E 24 2 steam line ;>

2) System 2 1 2 1, 2 3b 23 . !.

b. Automatic Actuation 2 1 2 1, 2 3 22 f'

< Logic and Actuation E Relays

, ?

ee U

e' i y E i 4 a , s ,c. u n s.m- ~Q

{p i

mm, f_--m , -mr,-, m.. .

% - * % w~ -m

!U

[

1,:

TABLE 3.3-3 (Continued) l

s, y-ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION ,

m i

{

[

z MINIMUM a, TOTAL NO. CHANNELS CHANNELS APPLICABLE 4 U FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION $.

~

4.

Steam Line Isolation (Continued) :7

c. Containment Pressure-- 3 2 2 1,2,3 15 High-2 F g j

. d. Steam Line Pressure--LowM 3/ steam 2/ steam 2/ steam 1,2,3

15 " kh line line any line %

7j steam line

e. Steam Line Pressure - 6f <l T.

3 Negative Rate--High 3/ steam 2/ steam 3# 15#

1 w line line any 2/ steam line h

33 A steam line j

{ ,

Y 5. Turbine Trip and b i 5 Feedwater Isolation I 1

a. Automatic Actuation i 2 1 2 1, 2 25 ?.

]j Logic and Actuation Relays i

0# t l.

'i b. Steam Generator 4/stm. gen. 2/sta. gen. 3/stm. gen. 1, 2 20 K i -

Water Level-- in any oper- in each i High-High (P-14) ating stm. operating ([

.i ~

gen. sta. gen.

c. Safety Injection See Item 1 above for all Safety Injection initiating functions and 4 requirements. }e p:

{

j P

6. Auxiliary Feedwater K '

~'-

Manual Initiation

a. 2 1 2 1,2,3 23 pl

);

b. Automatic Actuation Logic 2 1 2 1,2,3 22 and Actuation Relays .,

? ao

c. -

1 A c. Stm. Gen. Water Level-- '

O j g' Low-Low * ,(j

't.

l .

tSee also SpecifMilon L3* && V e

TABLE 3.3-3 (Continued) g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION l r-

' MINIMUM f TOTAL NO. CHANNELS CHANNELS APPLICABLE

{

j E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION ka l

w 6. Auxiliary Feedwater (Continued) [

1) Start Motor- d i Driven Pumps 4/stm. gen. 2/stm. gen. 3/stm. gen. . 1,2,3 20 # %

in any oper- in each 4 ating stm. gen. operating y stm. gen. [g

2) Start Turbine- d F.

.- ., Driven Pump 4/stm. gen. 2/stm. gen. 3/stm. gen. 1,2,3 20 " {j in any in each g j 2 operating operating g

-? stm. gen. stm. gen. 4

., w d. Safety Injection -

g.

i D Start Motor-Driven Pumps See Item 1. above for all Safety Injection initiating functions and F

.i w requirements. ?

4 e e. Loss-of_0. W. W.r .,W.. 2N kV Esf Esa Volf7 c.

0C ?;

.} r, i

Start Motor-Driven 4/ train 2/ train 3/ train 1, 2, 3, 19 f

} Pumps T l 11 Turbine-Driven Pump 4/ train 2/from 3/ train 1, 2, 3, 19 6

j , either train -

i , f. Trip of All Main [

j Feedwater Pumps, j Start Motor- k; j Driven Pumps and 9 j Tudi= ^ 4=rPg 2 1 1 1, 2 19 9 l -j; . ! -- - - c- == =11__l , -? - ? - 'l 3 _ i _=ti,=e.. Trcm; 6-m- --_ ' g.

p.'

7. Semi-Automatic Switchover to ?

, Containment Emergency Sump ;

4 t@

a. Automatic Actuation 2 1 2 1,2,3,4 14 ,

)f l g Logic and Actuation 9 c Relays .

g L

. s _

u,z , ,- . _. __ _

l 25

{'

TABLE 3.3-3 (Continued) )

8

$ ENGINEERED SAFETY FEATURES ACTUATION SYSTEM:INSTRUMEHTATION h Y

, HINIMUM c .

TOTAL NO. CHANNELS CHANNELS APPLICABLE ,

2' z FUNCTIONAL UNIT OF CHANNELS 10 TRIP OPERABLE MODES ACTION R

-4 7. Semi-Automatic Switchover to i H Containment Emergency Sump (Continued) -

U

b. RWST Level--Low-tow &'fvehr$f#2 3 1,2,3,4 17 h Coincident with Safety See 44em .1. above for all Safety Injection initiating functions 4 Injection and requirements. :~

/C6 incident With!

Containment Sump Level--High 4 2 3 1,2,3,4 17 m d fff j afety Injection See Item 1. above for all Safety Injection initiating functions ,'

and requirements. -

8. Loss of Power /o '//4/NM fer f.. -

CsF d "

); , a. 4.16 kV Ex:;;;ncy Bus 4/ bus 2/ bus 3/ bus 1, 2, 3, 4 20# g.

x

Undervoltage-Loss of Voltage . *;.

1 :.

1 w b. 4.16 kV Ex:r_ sF:;:n:y Bus -

(

? m

" Undervoltage-Grid Degraded . d p 20 "

~

Voltage j 9. Engineered Safety Features 4/ bus .

2/ bus 3/ bus 1, 2, 3, 4 -

{F ii Actuation System Interlocks k l a. Pressurizer Pressure, 3 2 2 1,2,3 21 -

l P-11 i.

v

b. Reactor Trip, P-4 2 1 2 1,2,3 23 [

f 10. Control Room Emergency Mode

Actuation Manual Initiation i,z 444, 4, ge, (,e , {

{

j a. 2 1 1 26 j

5

b. Automatic Actuation Logic and Actuation Relays 2 1 1 1,2,5,4,5"j ls

R E

26 g (e i to L z 3 4,5"jbe.

s a ~f

?

c. ntake Radio 2 1 aM : 26 p RE-82nb'nje. Ltn hhond %if*1 ction) bas Monitor 1 RE

]

{ d. afety I See 44em 1 above for all Safety -Injection initiating functions and requirements.

,p

[

i o t K '

3 in

p . m ._ _ s . . . . - cum _ eaA ,.ws.)hhem%gguagudagba,gggaw i

TABLE 3.3-3 (Continuea) .

, TABLE NOTATIONS 4

cl.dhe provisions of Specification 3.0.4 are not applicable.

- TH p n.n;t ka ::r t: thd:d 6 S': "^% b; h th: )

I-t:M d) 5:tp4.t- - F 12 (':c R T *-

1 L

I C. Dvem movem~4of ItradWed -rveI ormovemsst af Iwds ouw Wac/laN' bel i wifADi cA1fetm enf, I b. " Wip function automatically blocked above P-11 and may be blocked below i

g. P-11 when Safety Injection on low steam line pressure is not blocked. ;

Q.dipfunctionmaybeblockedinthisMODEbelowtheP-11(Pressurizer Pressure Interlock) Setpoint.

l ACTION STATEMENTS

', ACTION 14 - With the number of OPERABLE channels one less than the Minimum l Channels OPERABLE requirement, be in at least HOT STANDBY

, within 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 ; however, one channel may be bypassed for up to 2-hours for surveillance testing per Specification 4.3.2.1, provided

( j the other channel is OPERABLE.

1 ACTION 15 - With the number of OPERABLE channels one less than the Total

), Number of Channels, operation may proceed until performance of 3 the next required ANALOG CHANNEL OPERATIONAL TEST provided the f inoperable channel is placed in the tripped condition within i

' /2 /sce wYh X r f Y '

9-1 I'Gn .161-wig a channel associated with an operating loop inoperable, restore the inoperable channel to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 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 at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . One channel associated with an operating loop may be bypassed for f

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.l g

[ ACTION 17 - With the number of OPERABLE channels one less than the Total i

Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condition and the Minimum Channels OPERABLE requirement is met. One additional channel ,

may be bypassed for up to.2' hours for surveillance testing per !

Specification 4.3.2.1. 1

?

ACTION 18 - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge supply

( and exhaust valves are maintained closed.

n1OVer+1enf of EWGdi&l[vek of MW8m8M/ 0b htb OV"f iL fe..DutInhie/

' irad fue).

f No+ appliutLk if one main S}eam hold &1 &NL awlassDcinks/ byfa.u isolaHon vah'Ljur skamRM il chrad V0GTLE - UNIT 1 3/4 3-23 APR 2 4 %

t '

1 i

,;. Insert to page 3/4 3-30

.g g3 ACTION 16 - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, comply with the ACTION t- requirements of Specification 3.9.9 (Mode- 6).

i a

+

i 4

i i

(

i

'f J

l I '

l

.}

U

, 1

^

l 1

-ew,*,-us,e4---, r r-,-, .. ,r 4.-~pr-e.- --,-.,---e--e, ,- e +,y,e-r.- + - ,-yw- , - n-,--, r-, ww-, c -c ,,e aw-, y , ww-- - r m,v - v - y,

%s %_m m agm cbkWwadq;.gfwgim n&;um,,,awgammem,mam,, ,sm TABLE 3.3-3 (Continued)

ACTION STATEMENTS (Continued)

ACTION 19 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore 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 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 .

ACTION 20 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , and

b. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for s6rveillance testing of other channels per Specification 4.3.2.1.

, Newther" of CRMBLE Cl4MeE (E4?If**1 W ACTION 21 - WithAless than the Minimum " t f Channels OPERABLF4 within 1hourdeterminebyobservationoftheassociatedpermissive.Sh/v.s I

h3 ht _ - . _ . ..._.. ,. that the interlock is in its required state for the existing plant condition, or apply Specification 3.O.3.

ACTION 22 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following G hours; however, one channel may 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 OPERABLE.

/ 691/J)StLYTffM i d \

MWM'd1 1 ACTION 23Humbee-of-

- With theChannels, number restore of[PERABLE channels the inoperable channel oneto less than the -DeterP OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least HOT STANDBY within i 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least. HOT SHUTDOWN within the following l 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

ACTION 24 - With the number of OPERABLE channels one less than the Total Number of Channels, restore the inoperable channel to OPERABLE l status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or declare the associated valve inoperable and take the ACTION required by Specification [3.7.1.5].

1 ACTION 25 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may 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 OPERABLE.

SEE Teen f& y .ElY J-2f V0GTLE - UNIT 1 3/4 3-24 APR 2 4 -

F

"- me- -.v. ,, - . - , . , an e.=y~e,~qq:y m m . e e_ .,,..n,.m,_

, , ,7 ,__ ,

Thsert* S ?"Y 5/Y5~EY ACTION F - With the number of OPERABLE channels one less than the.

2G total number of channels, restore the inoperable channel to OPERABLE status within 7 days or within the next F hours initiate and maintain operation of the Control Room Emergency Ventilation System in the Emergency mode. With two channels inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of the Control Room Emergency Ventilation System in the Emer-gency mode.

ACTION JS - With the number of channels OPERABLE one less than the gej Minimum Channels OPERABLE requirement, restore' the number of OPERABLE channels to the Minimum Channel 0PERABLE require-l ment within 7 days, 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 , in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

mm, <~w

. , , n c -. -, -v , n - a -m p

(

e t

N b

TABLE 3.3-4 (Continued) &

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

/

, SENSOR f TOTAL ERROR 4 E FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE $

. --e ,

w 3. Containment Isolation P

8 j a. Phase "A" Isolation .

[

, 1) Manual Initiation N.A. N.A. N.A. N.A. N.A. Ic

'i 2) Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. F j and Actuation Relays @

]j 3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

i h 1 4) Containment Area Radiation P

$ (High Range) (RE-0005 /4-MM(N. A. ) (N.A.) (N.A.) (N.A.) (N.A.)

2 Y b. Containment Ventilation 6 i S Isolation I

$ h j 1) Manual Initiation N.A. N.A. N.A. N.A. N.A. Bf i E j 2) Automatic Actuation N.A. N.A. N.A. N.A. N.A.

i Logic and Actuation @

i n Relays ;

i 3) Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and -

1 Allowable Values. 4.

er I

5

.4) Containment Area Radioactivity (Low later Later later Later -

h:

$ h. Range)(RE-ooot,Al-@3) f(

$ 7 5) Containment Ventilation (,!

j Radiation , -(44-2SG54)

[F

{ e i ParticulateActivity[5d later Later Later Later '

S i ii Iodine Activity [#E ~2S Later Later Later

) iii Gaseous Activity (46.zS6se)later Later Later #

Later Later

..?

_ m l Li u_ . _ _ _ _ _ _ _ _ _ _ _ _ _ _

..:= = . - --

k i

i, l

TABLE 3.3-4 (Continued) g.

8 i.

S ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 1 r

! SENSOR ,

i [

z TOTAL ERROR [

.l FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE *

,: -4 e

w 4. Steam Line Isolation

[:.

a. Manual Initiation N.A. N.A. N.A. N.A. N.A. $

1 k

j b. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. It j and Actuation Relays i {

i c. Containment Pressure--High-2 3.0 0.71 1. 5 515.0 psig <15.8 psig [g A

h

3 d. Steam Line Pressure--Low 13.1 [10.71] [1.5] 1585 psig 1567 psig* p ij e. Steam Line Pressure - 5.0 0.5 0 -<100 psi /s <123 psi /s** f g

~ ~

g Negative Rate--High

f w 5. Turbine Trip and Feedwater b

F-j $ Isolation v

[

V 1 a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A. D j Actuation Relays fr i E i b. Steam Generator Water 5.1 2.18 1. 5 <78.0% of ~<79.8% of narrow j level--High-High (P-14) ange instrument [-

, yarrowrange ;p

instrument span. r' 3 i span. F

i

, %cdlerni1tn it .

c. Safety. Injection See Mee 1 above for all Safety Injection Trip Setpoints and gf -

. Allowable Values. . 6 4'

i. 6. Auxiliary Feedwater If

= $

a. Manual Initiation N.A. N.A. N.A. N.A. N.A. g; to :,

3 i N

b. Automatic Actuation Logic and Actuation Relays N.A. N.A. N.A. N.A. N.A. '

IF l '7 E

i I J f V.

- t t N

w - .~.u. ..._s w ,~ .w. - ~ a. -- , . n . v. .

..+._e ,n_.._. _

F:

[F

{

TABLE 3.3-4 (Continued) !

g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS L r- (r, 7

TOTAL SENSOR ERROR

[

I E FUNCTIONAL UNIT ALLOWANCE (TA) Z_ (S) TRIP SETPOINT ALLOWABLE VALUE

! -.e

(.

f s 6. Auxiliary Feedwater (Continued) -

z

c. Steam Generator Water f

. J Level--Low-Low y.

D

1. Start Motor-Driven 17.0 12.18 1.5 >17.0% of >15.3% of Bi

Pumps Harrow range Earrow range E instrument instrument span. L span. g

2. Start Turbine- 17.0 12.18 1. 5 >17.0% of >15.3% of ;;

Driven Pump Earrow range Earrow range &f

-] instrument instrument g 1 w span. span. p h Fmetimal MaiY '*i w d. Safety Injection See Mem-1. above for all Safety Injection Trip Setpoints and bi j h Start Motor-Driven Pumps Allowable Values.

[

k 1 e.

orQcded%/6 kVEsf Bn 6thye-Loss-of-Off;!_ Tv-er fg t

j i Start Motor-Driven Pumps N.A. N.A. N.A. 3598 volts with >3548 volts with r j 20 s time delay <20 s time delay [

o e I ii Turbine Driven Pump N.A. N.A. N.A. 3598 volts with >3548 volts with I a

20 s time delay 720 s time delay Ei e.

-j f. Trip of All Main Feedwater N.A. N.A. N.A. N.A. N.A. j/

q Pumps, Start Motor-Driven

> Pumps and Turbine-Driven

' [c 1 Pump _ _

j:

i g. Suction Transfer on Low N.A. N.A. N.A. $ [442] ft ~< [441] ft F 3

h, Pressure [.5 c

j ^ '

.h ? 'N' t K' i i, i :,.

3- M

n w ,, .. ,, n n .,-.,,.. _ m .o ,

~, ,

-- m_memm ~ ..w g4 W

U F

TABLE 3.3-4 (Continued) i g F g ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS @;

E F SENSOR ,

h E z

TOTAL ERROR $

FUNCTIONAL UNIT ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE s j

] w t I s 7. Semi-Automatic Switchover to .

Containment Emergency Sump [y

a. Automatic Actuation Logic N.A. N.A. N.A. N.A. N.A.

. and Actuation Relays ,

p

b. RWST Level--Low-Low N.A. N.A. N.A. 274 in. from 254 in. from tank bas?

Coincident With tank base 35.6% 32.5% of instrument

~f Containment Sump Level af instrument span sp_an (ib j N.A. N.A. N.A. < [30] in. < [32.5] in, 1

and- Fundfo6Al biit-ibove [680] ft ibove [680) ft k afety Injection See Mem-1. above for all Safety Injection Trip Setpoints and

?)( y Allowable Values.

s

[

e

8. Loss of Power /0'/.ll M ASF 8'O Esf

[{C j U a. 4.16 kV .;;c;,; .;y- Bus N.A. N.A. N.A. 2912 volts >2912 volts I':

j Undervoltage-Loss of Voltage with a with a < 0.8 k

) $ 0.8 second time [

second time delay. r.

j delay. I

^

Y

'} - b. 4.16 4 kV E .. gc.;f ESPBus N.A. N.A. N.A. 3598 volts 3598 volts k Undervoltage-Grid Degraded with a 20 with a 20 h lj Voltage second time second time [

] delay. delay. t i N

9. Engineered Safety Features '

j g Actuation System Interlocks D[

ii :e - D ee a. Pressurizer Pressure, P-11 N.A. N.A. N.A. < 1970 psig 5 1981 psig 4

i Q $

g b. Reactor Trip, P-4 N.A. N.A. N.A. N.A. N.A. ,

j 21 jo i .

Su

. . _ . _ .x , _ w ex = m a- ~ . --- u-~ x.a.~ < .- - - . . - - .a . . . . - -

.9; ,

b; y

TABLE 3.3-4 (Continued) ..

r

< b 8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

" SENSOR

?

TOTAL ERROR ALLOWANCE (TA) Z (S) TRIP SETPOINT ALLOWABLE VALUE E FUNCTIONAL UNIT a .'

.. 10. Control Room Emergency Mode d A-tiv; tier, j fjcfR4fIM N.A. N.A.- N.A. h-

a. Manual Initiation N.A. N.A S'

b. Automatic Actuation s Logic and Actuation Relays N.A. N.A. N. A.' N.A. N.A. [.*:

&ncfiom/ Z4sr/+- .

c. Safety Injection See .Mem-1 above for all Safety Injection Trip Setpoints M and Allowable Values. $

,? .

,c Intake Radiogas monitor (N.A.) (N.A.) (N.A.) (N.A.) (N.A.) g.

R d.

g t

[ (/26-l21/G,128- 12117) i

$ t

[

0-4

! 'h s'-

!. c h T.

s l [

' e

?;

i @

],

sq i = f:'

! w E 1

C J

e i n

n in % n s,n M ;:, M ip P ,

.- a .u.M e MimicW A 4 av.mn&.s s 'i-m o.wr!nciwaighu TABLE 3.3-5 .

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATION SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

1. Manual Initiation

a. m Safety Injection (ECCS) N.A.

b. Containment Spray N.A.

c. Phase "A" Isolation N.A.

d. Auxiliary Feedwater Motor-Driven Pumps N.A.

m caataiment "enuinthIseletion =- " . A.-

Cf' Steam Line Isolation N.A.

pF::1:trhchti= -- ~ ~%.A.

tr. - Axilie;;yJ::t tter - - - -

N.A.

3 "-ateiment-t'=tihth R:hti= N.A.

j. Centaiment4seleti-- - - - - - - " .- A:

fX. Control Room Emergency Mode Actuation N.A.

b VMf//4 fior7

$ J. Reactor Trip N.A.

)3 p(. Start Diesel Generators N.A. I n: p==L Coomg " ter - ;, . A.-

C om6ontainment-CoeLi~; c og- __wp 4 Mm.le#=$eYvice-Chikwy4ater -

-N.A.

?. Containment Pressure--High-1

a. Safety Injection (ECCS) 1 29(1)/12(5) ,

b. AfReactorTrip(fromSI) 12 L C- Jf' dwat* < 7(3) 1 d, 8 .= s et:i==t *[" Isolation Isolation _

< 2(6)

e. .4)' Containment Ventilation Isolation 1 6.5 E .57 Auxiliary Feedwater Motar-Driveg Pumps < 60 -

04 d O'Y'4 88T

3. Jrf Nuclear Service ooling Water i 100(1)/86(2)

Fans -< 40(1)/28(2) l

h. Jf Containment Coolinhi/4*/eq

. r Wo A g Control Room 4mergency Mode Actuation N.A.

y 8 Start Diesel Generators 1 [10]

10) Coyan=t CerFng ';hter - -[ ]

nw'mk Isohfion N.4

~ Oyed as/9 Wake n,A.

Conht@errf GM/hA SMJ AhA.

theles/ Jew,u CWh'y 4 n cA1/e- g,g, Ck/obunmf ArtMa/gn Esokf4 y,4, V0GTLE - UNIT 1 3/4 3-32 ggg

___ . -,- m ,-_,,,.. .. ..- _ _._ _

M@dMGMM%r a / ' - - , 9,fS M n h d M U n a Jiggw-s_-..4 a e M

h kr.d ~

l i _

! TABLE 3.3-5 (Continued) ,

n( ENGINEERED SAFETY FEATURES RESPONSE TIMES :

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

]

0 3. Pressurizer Pressure--Low

a. Sa ety Injection (ECCS) 1 29(1)/12(5) b ,2- Reactor Trip (from SI) 12

c. N Feedwater Isolation < 7(3) cl. 2f Isolation 2(6) l G. A h Containment Ventilation Isolation 6.5

$ f. Auxi,11ary Feedwater ator-Driven Pumps -< 60 5-im ent-

9. S)j Nuclear ServicetCool ng Water i 100(1)/86(2) i Fans < 40(1)/28(2)

)

h..

A. ET

.//_i>Containment Coolinblation ControlRoom@me%rgency Mode Actuation

~

N.A.

j .i . [ Start Diesel Generators 1 [10]

-10) C- ; .a.a M i.,3 Am - - - -

i-[ ]

4. Steam Line Pressure--Low

[ a. Safety Injection (ECCS) 1 12(5)/24(4)

{ bM Reactor Trip (from SI) 12 j 6- .2N Feedwater Isolation ~< 7(3) a A "

i d. a f f_"hast'-- at Isolation 1 2(g) f e. _4 f Containment Ventilation Isolation 1 6.5 j f J f Auxiliary Feedwater otor-Driven Pumps < 60 h

3. .&f Nuclear Service ko$11nh M W 100(1)/86(2)

h. 8 Containment Coolin Fans < 40(1)/28(2)

I j. .S f Control Room b e cyMIdeActuation N.A.

I d. S f Start Diesel Generators 1 [10]

-1G}--Cr ; n:nt CoeM.., .._ _. 3t ]'

b. Steam Line Isolation 9](3) 1 [7.0 p 5. Containment Pressure--High-3

a. Containment Spray 1 88(2)/100(1) l

6. Containment Pressure--High-2 Steam Line Isolation <g3)

,o e

7. Steam Line Pressure - Negative Rate--High Steam Line Isolation < 7(3) r 1 0

7 .

g V0GTLE - UNIT 1 3/4 3-33 M 24 w, l 2

y n

E ,

(m__---_ _ #- .

m m www.- v- ~~en-rws sse :--r

- ~ + r ro . -mm trw- *. m aw- -

w

em% dew.MWV eMiuwh>2RiF3am ams h e e 4 den;4adAn63 TABLE 3.3-5 (Continued) .

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

8. Steam Generator Water Level--High-High

a. Turbine Trip i 2.5

b. Feedwater Isolation 1 7(3)

9. Steam Generator Water Level--Low-Low

a. Motor-Driven Auxiliary Feedwater, Pumps 1 60

b. Turbine-Driven Auxiliary l Feedwater Pump < 60

~

10.

or doga&d Loss-of-Off;i.c M sYlaItY 23/* Bn> k'o//cp Auxiliary Feedwater 1 60

11. Trip of All Main Feedwater Pumps

Motor Driven Pumps N.A.

. t. T rii=-D i ;= ' -' -

-- -- ~u-A

12. RWST Level--Low-Low (o/ffc/abf //)/4 Me/y 5/ec//n/

e. _Cni mit z.t wit M Safety-injectica -W.*-

b. oincident'witn containment SVamp Level--High and Safety Iniegtion /

- utomatic Switchover to Containment Emergency Sump 146(2)/60(1)

13. Loss of Power E.s F -k

a. 4.16 kV Emergeney Bus Undervoltage- 1 2.0 Loss of Voltage Star.V Sipc /

b.

1 4.16 kV Errp.r.cy/sBus&e'se/ Gefjen/o,~ tR 1 21.2 Undervoltage - Grid Degraded Voltage

i. S/ntf *fn/ fo /)iesc/ Guer;rfor,

14. Control Room Intake Radiogas -

Control Room Ventilation Emergency Mode Actuation 1 10.0 (Isolation)

15. Containment Radioactivity I

a. Area Radiation Low Range-Containment Ventilation Isolation i 15.0

b. Containment Ventilation Radiation-Containment Ventilation Isolation 1 15.0

c. Area Radiation High Range-Containment Isolation Phase A 1 15.0 V0GTLE - UNIT 1 3/4 3-34 g H 96

PM?%s&McM n% i & _- - 3.r th u:. m46t & - '91 ;;s. , a :@ nukw , %:<.phWWurhMick -

l ~

).

u TABLE 3.3-5 (Continued) ,

TABLE NOTATIONS

[

( myl smsny, (1) A Diesel generator starting and j sequence loading delays included.

[ (2) Diesel generator starting and sequence loading delay not included.

y Offsite power evailable.

1 P (3) Electrohydraulic valves.

[ hr,a/ Sensieg, (4)/ Diesel generator starting and j sequence loading delay included. RHR j pumps not included.

~

$ (5) Diesel geherator starting and sequence loading delays not included.

d RHR pumps not included.

l (6) Does not include valve closure time.

-R The or1Sc -/ hie S}M4/ jnc/trolL NL l/m' Cldd4 asso c4/k wiFit e htc # valla 9e retys as c/c/emihm/ $h 7]L/e LJ-Y 1

pks m addi&s/ /.2 .seco,ds suocided w/M iy/er,puy RI snd ciycon eyemfik "T/ie asponse fina slau inc/ude /4 % io,

}

onM He udewdhp relayi as defamtoed ,y Tibia.1-yy a p/as e eddH/snd I,2 .seco,dr osssca24d nam t cired opersfien.

interpong rety ain i

J l

V0GTLE - UNIT 1 3/4 3-35 APR 241986

_ , , . , . . - _ sm w ym - - -.<-,,--.-.-w-..wym-4,.e -

e., = - . es swe g . -_.- ,_ ,,,

~~- _ ._ :. ~ - , . . , . . .. . a . a .w. _ y_n_

. .. m. .. . .- . ~ . - .-

.i

! s TABLE 4.3-2 ((

8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION E

-SURVEILLANCE REQUIREMENTS fs 1 TRIP ,

i E ANALOG ACTUATING MODES .

Z CHANNEL DEVICE MASTER SLAVE FOR WHICH f

I g CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

1. Safety Injection (Reactor Trip, -

l Feedwater Isolation, Component Cooling Water, Control Room .

1} Emergency Mode Actuation, Start l' Diesel Generators, Containment Cooling Fans, and Nuclear Service Cooling ';

Water, Containment Isolation, O Containment Ventilation Isolation, and fi t'

Auxiliary Feedwater Motor-driven Pumps). [ .

1 w a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3,4 i[

} h b. Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 I.!

' Logic and Actuation 1 i

Relays 4 ;

} c. Containment Pressure- S R H N.A. N.A. N.A. N.A. 1, 2, 3 g

]

High-1 y j d. Pressurizer Pressure S R H N.A. N.A. N.A. N.A. 1,2,3 ,

j Low

e. Steam Line S R H N.A. N.A. N.A. N.A. 1, 2, 3 l; Pressure-Low d 6 b

2. Containment Spray [

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2,'3, 4 ,

g

b. Automatic Actuation Logic and Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4 [.

j :o Relays .J Z c. Contai nt Pressure- S R H N.A. N.A. N.A. N.A. 1, 2, ,

f;.

High-3 F

,5 E

$ &f G/Jo Cr f/CAflM N'S N S 6 e

di

r g- ;. --~ , , , ,, - -m m, . m . m- .-,m,. - , o- >.

m.%. _. . - _ . m _

si i

Td

E A b 1 TABLE 4.3-2 (Continued) 8

} Di

'j h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION i j #

m SURVEILLANCE REQUIREMENTS f a

}l '

TRIP s

? E ANALOG ACTUATING MODES -

i Q CHANNEL DEVICE MASTER SLAVE FOR WHICH P g CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANC:i FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST ISREQUIRED'j

3. Containment Isolation .

4 i N I

a. Phase "A" Isolation if

.! 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A 1, 2, 3, 4 [

l 1 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 i i Logic and Actuation ![

{ knc/M4 N^Y

] w 3) Safety Injection See-leem 1. above for all Safety Injection Surveillance Requirements. y

4) Containment Area Radiation S R H N.A. N.A. N.A. N.A. 1, 2, 3, 4 h

((Hich Range)/f-QvS, AE-cooC) ' Q, j b. Containment Ventilation Isolation ,

.]) 1) Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1, 2, 3, 4 C 3

1 2) Automatic Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1, 2, 3, 4 7' j Logic and Actuation Relays ,

g .g [l:

3) Safety Injection Seeh 1. above for all Safety Injection Surveillance Requirements. p.

g

4) Containment Area E.

Radiation (Low Range) S R M N.A N.A N.A N.A 1,2,3,4,b

% (A6-oo'b /E- 00*3) k

= r Z C"lll b E I h,3 1

E

. m

, pm mm .: . .. ' -

,. ~. 'rw' ~ - -w ' -w- wa~q y

n li:

Ik Ek l! D j .:-

TABLE 4.3-2 (Continued) k%

8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION R pf SURVEILLANCE REQUIREMENTS m - g%

3 '

TRIP 5

).

! c ANALOG ACTUATING MODES r' 1

h CHANNEL DEVICE MASTER SLAVE FOR WHICH

" CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE r FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED d Containment Isolation (continued) , f

5) Containment Ventilation b Radiation Ek i E Particulate Activity ( & @ ; R H N.A. N.A. N.A. N.A. 1,2,3,4,6 ii Iodine G

,,, ActivitfMMS)S R M N.A. N.A. N.A. N.A. 1,2,3,4,6h 3

y iiiGaseous(( FEN)S Activity

4. Steam Line Isolation R M N.A. N.A. N.A. N.A. 1,2,3,4,6 [i r

[ a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 f O

b. Automatic Actuation Logic and Actuation N.A. N.A N.A N.A. M(1) M(1) Q 1, 2, 3 $

Relays h

p h c. Containment Pressure- S R M N.A. N.A. N.A. N.A. 1,2,3 i j High-2 * &c l

d. Steam Line S R M N.A. N.A. N.A. N.A. 1, 2, 3 Pressure-Low i.

4 J e. Steam Line Pressure- S R H N.A. N.A. N.A. N.A. 3

~

l Negative Rate-High b 1

4 5. Turbine Trip and Feedwater Wg.

1 Isolation '

{.-

a. Automatic Actuation N.A. N.A. N.A. N. A. M(1) M(1) Q 1, 2 6

] g Logic and Actuation C f.

}

9 :=

7 Relays

b. Steam Generator Water S R N.A N.A. N.A. N.A. N.A. 1, 2 Q,.,

{

. Level-High-High (P-14)

c. Safety Injection See Item 1 above for all Safety Injection Surveillance Requirements.

Q gg o

[

j kSee /)ho Spuifiutfk 9J 3 6- lp?

g

_ -= x . .=. . - ; w- w a- ~ .g- . + - . + .: ,

i K

]

4 y

j p w

TABLE 4.3-2 (Continued) [

l ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION m SURVEILLANCE REQUIREMENTS r i %

1 c TRIP i:

5 ANALOG ACTUATING MODES

-4 CHANNEL DEVICE MASTER SLAVE FOR WHICH L H CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLAN2 FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST ISREQUIRET

6. Auxilia'y Feedwater 1

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3 f

b. Automatic Actuation N.A. N.A N.A. N.A. M(1) M(1) 0 1, 2, 3 E and Actuation Relays '

c. Steam Generator Water

! Level-Low-Lowk

] 1) Start Motor-Driven [.

J q Pumps S R M N.A. N.A. N.A N.A 1,2,3 9

4

, y g

2) Start Turbine Driven Pump S R H N.A. N.A. N.A N.A 1,2,3

{g e

d. Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

f}

e. Los s-o f ^" 'g "-- ---or ik adec/

N.A. 'f.%kV R

E3f 5v1 9 VeH)e-N.A. M N.A. N.A. N.A 1,2,3 4

h j ,

g. Trip of All Main Feed N.A. N.A. N.A. R N.A. N.A. N.A 1, 2 s water Pumps p-

g. Suction Transfer on Low S

.}

j Pressure R H N.A. .N.A. N.A. N.A.

1,2,'3]fp j / -

6 4

7. Semi-Automatic Switchover to e j Containment Emergency Sump j )

a. Automatic Actuation Logic and Actuation N.A. N.A. N.A. N.A. M(1) M(1) Q 1,2,3,4d '

j g' Relays i

{e j WSpe /}/ro ],WeifIatfhn <rC 3 .T.G .

m i o

, b

_ .. _. ~ .- _. _

[$

\

2

, TABLE 4.3-2 (Continued) h ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

, ;, SURVEILLANCE REQUIREMENTS y

{ , 4 j c TRIP ff:

) 5 H

ANALOG ACTUATING MODES @

, CHANNEL DEVICE MASTER SLAVE FOR WHICH PJ i - CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANQf, FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST IS REQUIREGh l 7. Semi-Automatic Switchover to -

h j Containment Emergency Sump (continued) %

! :s

b. RWST Level-Low-Low S R H N.A. N.A. N.A. N.A 1,2,3,4 ?

nt With Saiety Su fmcHsm) fft101 ObM S/OllNhEfWi@ .Surve///a/ict /qviremQ _

[l o M ent Containment with Sump Level-High S R M N.A. N.A. N.A. N.A. 1, 2, 3, [

{ S_afety Infection See Item 1. above for all Safety Injection Surveillance Requirements.

{

o

8. Loss of Power /a %/44'V ED: 4kr Esf if f

a. 4.16 kV L .....;j N.A. R N.A. M N.A. N.A. N.A. 1,2,3,4 %

Bus Undervoltage-Loss ;.~e of Voltage 1 ESP 4

j b. 4.16 kV fassgency N.A. R. N.A. M N.A. N.A. H.A. 1, 2, 3, 4 fr i Bus Undervoltage-Grid [

Degraded Voltage y

9. Engineered Safety $

} Features Actuation %' ff

.i System Interlocks %g "f, 3

! % a. Pressurizer N.A. R M N.A. N.A. N.A. N.A. 1,2,3 h j [ Pressure, P-11 ij-i e 2-i g b. Reactor Trip, P-4 N.A. N.A. N.A. R N.A. N.A. N.A. 1,2,3, p

? v t:

- {:

c

t-i J

! E

. . e . .. -.

_ _ . . . . _. .w .+__ - - ~ -

w d

v 1 .

V f TABLE 4.3-2 (Continued) E y .: $

8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION $

l

{'

SURVEILLANCE REQUIRENENTS

{

TRIP 5 1 E ANALOG ACTUATING MODES y I

I Z CHANNEL DEVICE MASTER SLAVE FOR WHICH sj w CHANNEL CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST TEST TEST ISREQUIRED$

I 10. Control Room Ventilation Emergency Mode $,

Actuation 2 T

a. Manual Initiation N.A. N.A. N.A. R N.A. N.A. N.A. All Modes I;5 1 #

1 b. Automatic Actuation Logic and Actuation f; Relays N.A. N.A. N.A. N.A. M(1) N.A. N.A. All Modes f-w .-

A c. Intake Radiogas w )tonitor S R M N.A. N.A. N.A. N.A. All Modes -

' L ( " ' ' " ') " ~'*"N Fmcftaru/74ff :.

" Safety Injection See.34mr 1 above for all Safety Injection Surveillance Requirements

, d.

O TABLE NOTATION M (1) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS. %

i E k i' I f i w l ~

i, di N

3 t

? tr I i

.' N V j.:

a

. - - . . . - - - - - - - . ._ . . , . , . __ . ~ - . - .

a _ ,- -, .

s.

lY 5

?!.

TABLE 3.3-6 j n

RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS g a b H

MINIMUM h 1 E CHANNELS CHANNELS APPLICABLE ALARM / TRIP ff JJ Z FUNCTIONAL UNIT TO TRIP / ALARM OPERABLE MODES SETPOINT ACTION Ai

" r-

1. Containoent c.

See 744/r 3,3-f M lA,4 i a. Containment Area (Low Range) 1 2 *M- .

d2]' M. -Nr-586 750/f3d-3 I RE-0002, RE-0003 f, g,3, cf ggg 7;g/g JJa/ (g

b. Containment Area (High Range) 1 2 4 15- ST -Qfr) SrC 7dI'// l'N i RE-0005, RE-0006 s

c. RCS Leakage Detection f

1) Gaseous Activity RE-2562C 1 1 1,2,3,4 5 2 x back- MZ8 9 ground -

2) Particulate Activity RE-2562A 1 1 1,2,3,4 < 2 x back- 29 Z 8

, ground 1 2. Containment Ventilation

$ T a. Gaseous Activity RE-2565C 1 2 e /23,44 f7 f 4 sh M 26 5" M ~~

h

b. Particulate Activity RE-2565A 1 2 e/2,y4Cf f

** ' 56 I f f 7[4 /d J 3 -5 I)

c. Iodine Activity RE-2565B 1 2 Mt/ /f/

z 5 4/ C #

- Idl##$ See Tih/c J,J_J fh j 3. Control Room Air Intake g l Radiation Level RE-12116, 2 471/f2 ,fj V sMe)=um#W W See ~G4/c 2 3-3 k See Tabk 3,y-4 RE-12117 54y$*k [1

! 4. Volume Reduction Room Selective '

3 Cubicle Monitors -

},

l /

a. Gaseous Activity ARE-13133C 1 1 [] 27 -

r k

'i b. Particulate Activity ARE-13133A 1 1 [] 27 4

! c. Iodine Activity ARE-13133B 1 1 [] 27 b i %,

i b.

$ to 7

e w , Y: :

t.

5

h

, E e f

cc

-l b

_c 2 .

cy _

., J _ ,

.# _..,,. em s_ .ws_;-- = . w .g i g l G i .:

h TABLE 3.3-6 (Continued) f RADIATION HONITORING INSTRUMENTATION FOR PLANT OPERATIONS j l

E ilm MINIMUM E E CHANNELS APPLICABLE ALARM / TRIP MEASUREMENT I i M FUNCTIONAL UNIT OPERABLE MODES SETPOINT RANGE ACTION i:

I e {

5. Process Monitors 's N .

y r-Noble Gas Effluent Monitors fl N / Y.

i. Radwaste Building \ / 6 Exhaust System 1 1, 2, 3 & 4 (/) rad /hr 1-102 uCi/cc 30 h

! 's / p i ii. Auxiliary Building i Exhaust System 1 1, 2, 3 & 4 ( ) rad /hr 1-108 uCi/cc 30 b.

R iii. Steam Safety Valve Discharge 1/ valve

\ 1,N [3&4 ( ) rad /hr 1-108 uCi/cc 30 j:l

! T i $ iv. Atmospheric Steam [i' Dump Valve Discharge 1/ valve / 1, 2, 3 & 4 ( ) rad /hr 1-108 uCi/cc 30 g

<, u

v. Shield Building h Exhaust System 1 1, 2, 3 & 4 ( ) rad /hr 1-104 uCi/cc 30 g a

A vi. Containment Purge & q I Exhaust System 1 1, 2, 3 & 4 ( ) rad /hr 1-105 uC1/cc 30 I {k?

vii. Condenser Exh:"st V j System 7 1 1, 2, 3 & 4 ( ) rad /hr 1-1 5 uCi/cc 30

(

I

Alarm /TripSetpoin[of10 rad /hrisacceptable. h i r

%. \ kc to .

h & '

l j

g e

M\ x

?

v

} i

(-

1 L

e WkesMMMMNMwrngjsdeb - / - mSanDM&M ArMXEmehruA

=

TABLE 3.3-6 (Continued) .

TABLE NOTATIONS

- Mth-4epadiated Tve / wi/vn c f aly parYam$U"' "9 WN55Y *bhNY YffoYbe n -the e dviasc see'r-eress-D rip movmmt ar jrt;ra%bd fue/ or snowry' af /Mdr ovviWadi</p/

= -With-ful in the fuel terage-poo4-ereas:-

ACTION STATEMENTS g jg g ,jfg Y nort her of- t ACTION N - With less-then the M'-i r- Ch; . .;is OPERABLE [ requirement, opggggtf p,wivm Cla/w/.1 opecotion e:y cc.tia w previded the centainment poi exh=+- :hes are= maintained-cle;;d. --M//.Vy /de./7C77B4/ r- end fTf vittnidrfi a f .5)H'ciftbsfi&1 3, '/ 6. /. '

IACTION 27 - ~Vith the number W UPERA Ei channels one less than the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months isolate the I Control Room Emergency Ventilation System and initiate operation of the Control Room Emergency Ventilation System in the recirculation mode.

ACTION 28 - With less than the Minimum Channels OPERABLE requirement, opera-tion may c'ontinue for up to 30 days provided an appropriate 4

portable continuous monitor with the same Alarm Setpoint is provided in the fuel storage pool area. Restore the inoperable l monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel storage pool areas.

ACTION 29 - Must satisfy the ACTION requirement for Specification 3.4.6.1.

ACTION 30- With the number of OPERABLE Channels less than required by the Minimum Channels OPERABLE requirements, initiate the preplanned

/- alternate method of monitoring the appropriate parameter (s),

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and-i

1) either restore the inoperable Channel (s) to OPERABLE status within 7 days of the event, or ;

2) prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within 14 days following the event outlining the action taken, the cause of the

) inoperability and the plans and schedule for restoring the system to OPERABLE status.

l 3 l A l V0GTLE - UNIT 1 3/4 3-45 APR 241986

-w-e= m.. , eg w -w pp.eem em- - - =,* &m.ae <=e+ war ew-or= + + e= . _ _. +--==~c a,i+ 9 eg . w-

~3e * == y s .e,e =#rw.e -

~ _ =- --_ .- - . . . ~ - _ - = - . _ _ _ n ,.:

a

{

by

.. TABLE 4.3-3 Y 8 RADIATION MONITORING INSTRUMENTATION FOR PLANT kp El OPERATIONS SURVEILLANCE REQUIREMENTS

] d E

l . ANALOG s e CHANNEL MODES FOR WHICH y 5

CHANNEL CHANNEL

OPERATIONAL SURVEILLANCE 4.

i FUNCTIONAL UNIT CHECK CALIBRATION TEST IS REQUIRED Y 1 H Z

1. Containment g gp y,y_g L

g

a. Containment Area (Low Range) S R -

All g

i l RE-0002, RE-0003 g ggfg g3 7 .

Containment Area (High Range) --~ 2 - -- "

b. S All 1.[

RE-0005, RE-0006 2 w

c. RCS Leakage Detection l,j

1) Gaseous Activity RE-2562C S R M 1,2,3,4 e

2) Particulate Activity'RE-2562A S R M 1, 2, 3, 4 0 y

$ 2. Containment Ventilation bN 4,1--2 L Y a. Gaseous Activity RE-2565C f -2 -M AH, %

sr< Table 4,3'-L M N b. Particulate Activity RE-2565A =S e-R "-M AH-Sn %Ne 4,3-2.

c. Iodine Activity RE-2565B -S -~n P All-(

j 3. Control Room Sec 754/e y,S-2_ j-t Air Intake Radiation Level --

R M - A11 - d

4. Volume Reduction Room Selective N b j Cubicle Monitors i a. Gaseous Activity ARE-13133C S R M i

b. Particulate Activity ARE-13133A S R M h

c. Iodine Activity ARE-131338 S R M TABLE NOTATIONS h.

i pl

] g *During operation of the Radwaste Solidification System. :

1 :so b

se k,

b

$ I c

i:

t

. .x = . . = =_ - ---

- .. = = = = = = = - - - - r===-- -y I*

h N TABLE 4.3-3 (Continued) {

j 8 RADIATION MONITORING INSTRUMENTATION FOR PLANT '

# OPERATIONS SURVEILLANCE REQUIREMENTS -

/

's,

) '"

/ T j ANALOG

.. E 1

E CHANNEL MODES-FOR WHICH M Q CHANNEL CHANNEL OPERATIONAL SURVEILLANCE 9 g FUNCTIONAL UNIT CHECK CALIBRATION TEST p /IS REQUIRED E 4

S. Process Monitors \ N /

f i Noble Gas Effluent Monitors $

. i. Radwaste Building s y Exhaust System S N R M 1, 2, 3 & 4 7 T

11. Auxiliary Building N'N '

E Exhaust System S R H 1, 2, 3 & 4 L R

' N 7 i iii. Steam Safety Valve R 1

Y Discharge S R H 1, 2, 3 & 4 5 to j

iv. Atmospheric Steam Dump Valve N E Fi i Discharge S R M 1, 2, 3 & 4 j 1

Shield Building 1

v. b

! Exhaust System S R H 1, 2, 3 & 4 j.f j <.

1 vi. Containment Purge'& {

! Exhaust Systep S R M 1, 2, &4 sf

] / f 1 vii. CondenseFExhaust i Syste / S R -M 1, 2, 3 & 4 . h'

I *Acce t le criteria for calibration are provided in Table II.F.1-3 of NUREG-0737.

, . . l

,4 /* '

E i b

t.

TABLE 3.3-9 REMOTE SHUTDOWN SYSTEM Minimum Readout 1

Channels Channels Available Operable Instrument Function Location

1. Source Range Neutron Flux A 1 NI 31E 1

2. Extended Range Neutron Flux B 1 NI 13135 CAD 1 A, B 1/ Loop 1/ Loop

3. RCS LP Cold Leg Temperature Loop 1 TI 04130 (Panel A)

} Loop 2 TI 0423D (Panel B f Loop 3 TI 04330 (Panel B Loop 4 TI 0443D (Panel A 2 2

4. RCS Hot Leg Temperature A k Loop 1 TI 0413L Loop 4 TI 0443C 2 2

5. Core Exit Themocouples B Loop 2 Core Quadrant TI 10055 Loop 3 Core Quadrant TI 10056 A, B 2 2

6. RCS Wide Range Pressure

- Panel A PI 405A b- Panel B PI 403A A, B 1/ Loop 1/ Loop 7

_ k . Steam Generator Level Wide Range Loop 1 LI501B(PanelA)

Loop 2 LI 502B (Panel B)

Loop 3 LI 503B (Panel B)

A Loop 4 LI 504B (Panel A)

L A, B 2 2

8. Pressurizer Level Panel A LI 459C Panel B LI 460C 3

L 1 LI 0990C 1

9. RWST Level L 1 PI-101152 13

10. BAST Level L 2 23

11. CST Level Tank 1 g LI 5100 Tank 2 LI 5115 J

!?

1. A - Remote Shutdown Panel PSDA B - Remote Shutdown Panel PSDB L - Local Indication

2. Graph will be provided to detemine level from pressure reading.

3. Alternate local level indication may be established to fulfill the minimum channels OPERABLE.

(

! V0GTLE - ONIT 1 3/1 3-58 l

[

L

l

g TABLE 4.3-6 m

N REMOTE SHUTDOWN MONITORING INSTRUMENTATION 7 SURVEILLANCE REQUIREMENTS E

U CHANNEL CHANNEL

- INSTRUMENT CHECK CALIBRATION

1. Source Range Neutron Flux M R

2. Extended Range Neutron Flux M R l

l 3. RCS LP Cold Leg Temp M R

4. RCS Hot Leg Temp M R

, . 5. Core Exit Thermocouples M R l h 6. RCS Wide Range Pressure M R

! 7. Steam Generator Level Wide Range M R

8. Pressurizer Level M R

9. RWST Level M R

10. BAST Level M R

11. CST Level M R l

l 1

1

c.- .:--.: -

n.w. a w. .- - - - - a w u.. - _ - - ~ w w.- m _ - - - -

TABLE 3.3-13 (Continued) f

.:: f-8 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRtMENTATION .

d m

[i v

' [~

MINIMUM CHANNELS E INSTRUMENT OPERABLE APPLICABILITY ACTION b.

Z [

s 4. Plant Vent

a. Noble Gas Activity Monitor 1

47, 48 h (RE-12442C or RE-12444C) ,

e

b. Iodine Sampler' 1

51 !-

(RE-124428 or RE-124448) l.

y

c. Particulate Sampler 1 51 l (RE-12442A or RE-12444A)

Flow Rate Monitor

R

d. 1 46 (FI-12442) ;

Y ,. .

e. Sampler Flow Rate Monitor *

! O 1 46 L (FI-12442 or FI-12444) k.'

5. Radwaste Solidification Building ~ D

a. Noble Gas Activity Monitor 1

47 h (ARE-0026C) .

{.

b Iodine Sampler 1 *' 51 (ARE-00268) j.;

c. Particulate Sampler 1

51 p+

(ARE-0026A) .

1

= d. Flow Rate Monitor 1

46 ;

w (AFI-0026) e

((

E e. Sampler F1'ow Rate Monitor 1

46 F-E (AFI-0026) .

~

w th tre.

it w.

TABLE 3.3-13 (Continued)

TABLE NOTATIONS i

At all times.

- During GASEOUS WASTE PROCESSING SYSTEM operation.

*During radioactive releases via this pathway.

ACTION STATEMENTS ACTION 45 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment provided that prior to initiating the release:

a. At least. two independent samples of the tank's contents I

are analyzed, and

b. At least two technically qualified members of the facility staff independently verify the release rate calculations and discharge valve lineup.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION 46 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION 47 - With,the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue provided grab sampics are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 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 than required by the ;

Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

= ACTION = 49=- With=the= number-ofmchannels= OPERABLE less than required =by-the-

== Minimum-Channels-OPERABLE-7:;uirementy-;;,;, etier, ef thia CASE 0"O-

= WASTE = PROCES S ING = SYST EM = mayr conti nue= provi ded = grab = sampl es=a re -

=co11ected=at=least=once=per=4= hours =and= analyzed-within-the -

=fol=1owing+ hours ~

-ACTION 50 - With th; n;.ccr -cf-channeels-OPERABLE :n: 11ss-than-required-by -

th: Minir Channels-OPERABLE-requirementreperation=of=this___-

-syst:- :y : ntine: pr:vid:d grch-samples-ere=taken-and= analyzed- 1

-et-leas t-once-pe r-24-hoursMi th-beth-c hannels-i noperabl e , - l

-operation ::y continue-prov4ded-grab-samples-are-taken-and-analyzed-

-et-least-ones per 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s-dueing degassing-operat-iens-and-at-least-One: per 21 5:er: & ning a thermoperations,.

V0GTLE - UNIT 1 3/4 3 NNN I 1

l l

Insert for page 3/4 3-7 ACTION 4 - a. With the outlet oxygen monitor channel inoperable, operation of the system may continue provided grab samples are taken and rnalyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the oxygen concentration remains less than 1 percent,

b. With the inlet oxygen monitor inoperable, operation may cor.tinue if inlet hydrogen monitor is OPERABLE.

c. With both oxygen channels or 00th of the inlet oxygen and inlet hydrogen monitors inoperable, suspend oxygen supply to the recombiner. Addition of waste gas to the system may continue provided grab samples are taken 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 concentra-tion remains less than 1 percent.

Action 50-60idh #t iwmber of c/wa OMod "' Im %

np<s ayw usiwum %ets onnisa qu&cmmt 5us,0 enc / 0X 4 1 J Uf/ fW f d N L / t C M 1 l]/7/ P/ ", /N//Nsf sf warte yas vo' ne .sysw imy emne pmdel ginb wyles We fahn asd' andl y eel af /MJf t

cw e p u y h u a ciu, 9 d y eis & y y u s w o,-

af b5/mcefW 2f /nvrr chwy oMer eperpfs'wy sm'kieoxy i ekolf. ym c&icegafwi iami;is less ja,,

f

(

TABLE 3.3-13 (Continued)

TABLE NOTATIONS (Continued)

ACTION 51 - With the number of channels OPERABLE less than required by the Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue provided samples are contin-uously collected with auxiliary sampling equipment as required in Table 4.11-2.

4+ ION With th: TM:7 f h:rn:1: OPERABLE :n: less-threquired-by-th: "inf er Ch:nnel: OPEP..".SLE r::;;fr;;;nt, :::;: d exy cr 2??l y-

-te the rc::d iner V0GTLE - UNIT 1 3/4 3-77 APR 2 41986

3 f

. i

.i TABLE 4.3-9 (Continued) [

8 RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS i n

m r

ANALOG t E CHANNEL MODES FOR WHICH {

Z CHANNEL SOURCE CHANNEL OPERATIONAL SURVEILLANCE g INSTRUMENT CHECK CHECK CALIBRATION TEST IS REQUIRED g'

4. Plant Vent ,

a. Noble Gas Activity Monitor D M R(3) Q(2) *

(RE-12442C or RE-12444C) j

b. Iodine Sampler W N.A. N.A. N.A.

i (RE-12442B or RE-124448)

N.A. N.A. *

c. Particulate Sampler W N.A. ;

t' (RE-12442A or RE-12444A) .

.* j Y d. Flow Rate Monitor D N.A. R Q

+

a (FI-12442)

e. Sampler Flow Rate Monito.' D N.A. R Q (FI-12442 or FI-12444) 5

5. Radwaste SolihBEllillng ~

w N

a. Noble Gas Activity Monitor D M R(3) Q(2) *

(ARE-0026C)

(

b. Iodine Sampler W N.A. N.A. N.A. *

\

,/ (ARE-00268) ( :

c. Particulate Sampler W N.A. N.A. N.A. *

(ARE-0026A) :.

5, d. Flow Rate Monitor D N.A. R Q

I

(AFI-0026) :

o .~

2 e. Sampler Flow Rate Monitor D N.A. R Q p-(AFI-0026) +

( f t.6 5

I::

a .m. _ t . -..a < . .s . . _ . m_ . -..-u. a w. n Au.C C,unELEM i

l TABLE 4.3-9 (Continued) l TABLE NOTATIONS _

At all times.

During GASEOUS WASTE PROCESSING SYSTEM operation. l During Releases via this pathway.

t

, (1) The ANALOG CHANNEL OPERATIONAL TEST shall also demonstrate that automatic l isolation of this pathway and control room alarm annunciation occurs it i any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint, or

b. Circuit failure, or

$ c. Instrument indicates a downscale failure, or

d. Instrument controls not set in operate mode.

If '

(2) The ARALOG CHANNEL OPERATIONAL TEST shall also demonstrate that control r room alarm annunciation occurs if any of the following conditions exists: )

e ,

j a. Instrument indicates measured levels above the Alarm Setpoint, or l j

be

b. Circuit failure, or

c. Instrument indicates a downscale failure, or

d. Instrument controls not set in operate mode.

(3) The initial CHANNEL CALIBRATION shall be performed using one or more of

{ the reference standards certified by the National Bureau of Standards (NBS) y or using standards that have been obtained from suppliers that participate

] in measurement assurance activities with NBS. These standards shall permit 1 calibrating the system over its intended range of energy and measurement i'l range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

(4) The CHANNEL CALIBRATION shall include t yse o standard gmples in anerd@6 sent+w~~ - -" - u/Hb M14 ma'11 a ttu 3 RW

{ v/ Vo vm ! f k tk/f' tiarJS.

f & ticek<!/at?(fili'a ft

. a .shnds/

. n__ ,ys 39_ __ _<.m__,,

._ - _ _ j/> d F r?smhu/

u.3, __2 g_$

} .dnt) be esad liI'f5 N(ib[5//c'n] /

05 $< /1l1 h1andly2t/:

3 n ---v.ar=w =w=-+ m-~;~ y - , wn

>t j (5) The CHANNEL CALIBRATION shall include +he use of standard gas samples les 4(cardsted-

) ychnt~ ni~ - -%1 - p# fx k)fe. manubecft/ter'r teamnrenaWi2ns. Jn caWiffe,y4 y .s/ d ttf As S sam,0k ikul fur' volume finM/ CXy en , br/ect ryiy, f a. One volume percent oxygen, balance nitrogen, and 3

j Sht//Four

b. be ssed volume it] percent

/k cs//brWins oxygen, de C4eek balance /iherify nitrogen. orG/fe oygen ang/ w, 1

l

i

] Naa41986

{!

V0GTLE - UNIT 1 3/4 3-80 r

% ~ . - . . . .a . ... . a . ._ . w.. . e ..--. .. u . -- .....-. .a.. .a. J : : .e . . . N z w :..::ux...

INSTRUMENTATION .

HIGH ENERGY LINE BREAK ISOLATION SENSORS LIMITING CONDITION FOR OPERATION 3.3.3.12 The high energy line break instrumentation listed in Table 3.3-14 shall be OPERABLE.

APPLICABILITY: As noted in Table 3.3-14.

ACTION:

a. With the number of OPERABLE electric steam boiler isolation instruments less than required by Table 3.3-14, restore the inoperable instru:nents to OPERABLE status within 7 days or suspend operation of the electric steam .

PERABLE status. 7/N//UtfM"5 b.

D"/AS"JLM? MWBh'?#55)###.st red t With the number of OPERABLE steam generator blowdown line isolation instru-ments or letdown line isolation instruments less than required by Table 3.3.14, restore the inoperable instruments to OPERABLE status within 7 days or be in 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 with the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I SURVEILLANCE REQUIREMENTS 4.3.3.12 Each of the above high energy line break isolation instruments shall be demonstrated OPERABLE by the performance of an ANALOG. CHANNEL OPERATIONAL TEST at least once per 18 months.

i t

i l

)

l V0GTLE - UNIT 1 3/4 3-81 APR S4 E E

)

1 I

[ .

t

p.

c a_. - us._ mm._.._ -. w __ w. _.__ __ - - _u..2_..: _ t. u d,_.2 5 50.3.bm b.A s

l,i /

t TABLE 3.3-14 /

HIGH-ENERGY LINE BREAK INSTRUMENTATION

}- /

i /

Isolation Minimum Applicable j Function Channels Operable Modes

1. EleegricSteam 1/ channel !

q Boiler solation

2. Steam Generator 1/ channel / 1,2,3,4 h Blowdown line /

Isolation j i

3. Letdown Line 1/ channel / 1,2,3,4 Isolation

) -

/

h /

q l

n I,

d n

)

1 iI 3

b 5

d B

d E.

i f) a A

d

.i 1

j *Requiredddringallmodeswhenelectricsteamboilerisinoperation.

i \

1 h V0GTLE - NIT 1 3/4 3-82 APR 24 M a

r,

'?

_.__s,.,_,........_,y__-_...

/1 TABLE 3.3-16*

HIGH-ENERGY LINE BREAK INSTRUMENTATION Isolation Instrument Minimum Applicable Function Channel Chanaels Operable Modes

1. Electric Steam TE19722A(R052) 1 Boiler Isolation TE 19723A (RD52)

TE 19722B (RC41) 1 TE 197238 (RC41)

TE 19722C (RC64) 1 TE 19723C (RC64) ,

TE19722D(RC66) 1 TE 19723D (RC66)

FT 19722 1 FT 19723 TE 19722E RC95 1 TE 19723E RC95

2. Steam Generator TE15212A(RB08) 1 1,2,3,4 Blowdown Line TE 15216A (RB08)

Isolation TE 152128 RC106 1 1,2,3,4 TE 15216B RC106 TE 15212C (RC107) 1 1,2,3,4 TE 15216C (RC107)

TE 152120 (RC108) 1 1,2,3,4 TE 15216D (RC108)

FT 15212A (Loop 1) 1 1,2,3,4 FT 15216A FT 15212B (Loop 2) 1 1,2,3,4 FT 152168 FT 15212C (Loop 3) 1 1,2,3,4 FT 15216C FT 15212P'(Loop 4) 1 1,2,3,4 FT 15216P

3. Letdown Line TE 15214A (A07 1 1, 2, 3, 4

. Isolation TE 15215A (A07 TE 152148 (A08) 1 1,2,3,4 TE 15215B (A08)

TE 15214C (A09) 1 1,2,3,4 TE 15215C (A09)

PT 15214 1 1,2,3,4 PT 15215

Required during all mndes when electric steam boiler is in operation.

V0GTLE - UNIT 1 3/4 3-86

p h~2.~:.a. a. w.w .a e...s .. w.l hw _.>..:. w .. u...,,~.A.'d.; M dh..ya w a i l

1 REACTOR COOLANT SYSTEM

, HOT STANDBY

[ LIMITING CONDITION FOR OPERATION

$ 3.4.1.2 At least two of the reactor coolant loops listed below shall be

( OPERABLE with at least two reactor coolant loops in operation when the Reactor Trip System breakers are closed and at least one reactor coolant loop in opera-tion when the Reactor Trip System breakers are open:*

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

{ reactor coolant pump, i b. Reactor Coolant Loop 2 and its associated steam generator and

) reactor cbolant pump,

( c. Reactor Coolant Loop 3 and its associated steam generator and i reactor coolant pump, and

d. Reactor Coolant Loop.4 and its associated steam generator and reactor coolant pump.

4 APPLICABILITY: MODE 3.**

L ACTION:

)'

a. With less than the above required reactor coolant loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-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 .

b. With only one reactor coolant loop in operation and the Reactor Trip System breakers in the closed po.sition, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months open the Reactor

( Trip System breakers.

h c. With no reactor coolant loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the

[ required reactor coolant loop to operation.

t SURVEILLANCE REQUIREMENTS

{

a 4.4.1.2.1 At least the above required reactor coolant pumps, if not in e operation, shall be. determined OPERABLE once per 7 days by verifying correct breake alignments and indicated power availability.

h]'

}

4.4.1.2.2 The required steam generators shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 17%jat least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

gggy 4.4.1.2.3 The required reactor coolant loops shall be verified 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 .

s s

All reactor coolant pumps may be deenergized for up to I hour provided:

j (1) no operations ar'e permitted that would cause dilution of the Reactor 1 Coolant System baron concentration, and (2) core outlet temperature is j maintained at least 10*F below saturation temperature.

a **See Special Test Exception Specification 3.10.4.

i! APR 241986 1 V0GTLE - UNIT 1 3/4 4-2 d

1 1

Vhrsi s .a i ..-a , a a c .. :n>;,.. .-

.u...,_ . .... L ;,a ~.. . .. .... . . - . . .. . . . .s - ,n.lihw Ga h.1:.sh.

REACTOR COOLANT SYSTEM HOT SHUTDOWN SURVEILLANCE REQUIREMENTS

}

4.4.1.3.1 The required reactor coolant pump (s), if not in operation, shall be

determined OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

L

4.4.1.3.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary. side water level to be greater than or equal to 17% at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Ofw/cle mg e.

T 4.4.1.3.3 At least one reactor coolant or RHR train shall be verified 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 .

l t

l 1

0 1

)

k i

,f .

~

io I

i i

i

'i

)

E V0GTLE - UNIT 1 3/4 4-4 i

ll

%: h . & d ..La. hila s . a = _ a. .a R a .. ~ -.. . .~.... =.

. :.2 R nhs4 l

}

REACTOR COOLANT SYSTEM

}

COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) train shall be OPERABLE and in operation *, and either:

I a. One additional RHR train shall be OPERABLE **, or 4

I

b. The secondary side water level of at least two steam generators

[ shall be greater than 17%F#ev/derap APPLICABILITY: M00'E 5 with reactor coolant loops filled ***.

1 ACTION:

I

[ a. With one of the RHR trains inoperable or with less than the required 4 steam generator water level, immediately initiate corrective action 1 to return the inoperable RHR train to OPERABLE status or restore the required steam generator water level as soon as possible.

(

~

b. With no RHR train in operation, suspend all operations involving a i reduction in boron concentration of the Reactor Coolant System and i

immediately initiate corrective action to return the required RHR

[ >

train to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.4.1.1 The secondary side water level of at least two steam generators a when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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 .

O

The RHR pump may be deenergized for up to I hour provided: (1) no operat. ions are permitted that would cause dilution of the Reactor Coolant System bcron concentration, and (2) core outlet temperature is maintained at least 10*F 2 below saturation temperature.

I;

- 0ne 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 testing g provided the other RHR train is OPERABLE and in operation.

- - A reactor coolant pump shall not be started unless the secondary water temperature of each steam generator is less than 50 *F above each of the e

l Reactor Coolant System cold leg temperatures.

H l

d N V0GTLE - UNIT 1 3/4 4-5 APR 2 41986

[

h

%. . m m ., _ , . . . _ __. .

. . . : u.a ..~. dk L~=,Nik.a :.

Gwsk&10%ia'd.&

...a a..w = . u.. .:.i., .a ...n au . .:.=~. .

~

I REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED t

, LIMITING CONDITION FOR OPERATION t

3.4.1.4.2 Two residual heat removal (RHR) trains shall be OPERABLE

and at l least one RHR train shall be in operation.**

} APPLICABILITY: MODE 5 with reactor coolant loops not filled.

I ACTION:

}

h a. With less,than the above required trains OPERABLE, immediately i -

initiate corrective action to return the required RHR trains to j-. OPERABLE status as soon as possible.

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

1 e v i

j SURVEILLANCE REQUIREMENTS .

i 5

4.4.1.4.2.1At least one RHR train shall be determined to be in operation and I

circulating reactor coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

h 4.4./.4.2.2 Vilfe 120 E - ud- 175, / 208-uy-/76 1208-k4- 177, od.f zoa-ky-183

\ s);d/ ft Veri,Gec/ clasec/ G4c/ securer / fiffas/VfDr1 by niecdaijim/ s/ opt of

/en..} once.per Tf clay 5.

I i

0ne 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 testing provided the other RHR train is OPERABLE and in operation.

i **The RHR pump may be deenergized for up to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months provided: (1) no operations f are permitted that would cause dilution of the Reactor Ccalant System boron !

concentration, and (2) core outlet temperature is ma,intained at least 10'F l below saturation temperature. I e

Y8 kM$W heZ Afff" O L Ak fhY duc/taye t h //tu (f w a - 2 4 - fij75, froj'/8Q-ftog 7s - y 9 -j 7 7'

,' av 12os-uy-iss) of ched ad se cwa /in ,oawor,

? immeclie/e/ c/ose ax/securc m fos 0/ost Me AMWST clacharft valfes i

) V0GTLE - UNIT 1 3/4 4-6 APP 2 /146

L. m a o- '

+- &,ci d,.usc wa.m &..ma .u .....a:. t.a.2.m..__J._..: El_1_E .

1

.h

~

REACTOR COOLANT SYSTEM

]

l 3/4.4.4 RELIEF VALVES i

LIMITING CONDITION FOR OPERATION 3.4.4 All power-operated relief valves (PORVs) and their associated block j valves shall be OPERABLE.

j APPLICABILITY: MODES 1, 2, and 3.

1 ACTION:

] ,

a. With one or more PORV(s) inoperable, because of excessive seat i 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 OPERABLE status 1

or close the associated block valve (s); otherwise, 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 HOT SHUTDOWN within the j following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

t

.{ b. With one or more PORV(s) inoperable due to causes other than exces-l 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 OPERABLE status or close the associated block valve (s) and remove power from the block valve, and a 1. With only one Class 1 PORV OPERABLE, restore at least a total i

of two Class 1 PORVs to OPERABLE status within the following 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-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 and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or

2. With no Class 1 PORVs OPERABLE, restore at least two Class 1 PORVs to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 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 and HOT 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 i the block valve (s) to OPERABLE status or close the block valve (s) and j remove power from the block valve (s) or close the PORV and remove power from its associated solenoid valve; and (2) apply ACTION b above, as

]j appropriate, for the isolated PORV(s),

d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS

[

j 4.4.4.1 -I. Oddition t; the 7;;;ir;;;nte f hm"+d-- ' ^ "

EichPORV j shall be domonstrated OPERABLE at least once per 18 months by:

a. Operating the valve through one complete cycle of full travel, and ,

l

b. Performing a CHANNEL CALIBRATION.of th: :-+"-+4aa 4--tr -;nt:t h ;

I I a

!! ~

V0GTLE - UNIT 1 3/4 4-10 g H M6 a

y ._ ._ . __ . . . . _ . . _. _ ~ ..

d.w..; = u w ::c d & & & ~ g :a.a. = _ w . _ . .u. a .._ a a w _ _ .,.a.

REACTOR COOLANT SYSTEM a 3/4.4.5 STEAM GENERATORS I

LIMITING CONDITION FOR OPERATION 3

i j 3.4.5 Each steam generator shall be OPERABLE.

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

ACTION:

1 1 With one or more st6am generator's inoperable, restore the inoperable generator (s) j to OPERABLE status prior to increasing T,yg above 200*F.

SURVEILLANCE REQUIREMENTS a

l 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of f the following augmented inservice inspection program,:nd th; 7;;uirc.T. nt: cf -

j.I Sp::ificott.. ^ 1 S-1 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting and inspecting at least the minimum number of steam generators specified in Table 4.4-1.

4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the fre-quencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.4.5.4. The tubes selected for each inservice inspection shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except:

a. Where experience in similar plants with similar water chemistry 1 indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas;

]

b. The first sample of tubes selected for each inservice inspection (subsequent to the preservice inspection) of each steam generator shall include:

-1

'g

.i 1

[

1.

V0GTLE - UNIT 1 3/4 4-12 APR 24126 L 1 l

1.___ _ _ . . - . _ _ _ . . . , . .._. ._ .. .

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REACTOR COOLANT SYSTEM 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS, LIMITING CONDITION FOR OPERATION 1

4 j 3.4.6.1 The following Reactor Coolant System Leakage Detection Systems shall

)

be OPERABLE: g, fica /a/e

a. The Containment Atmosphere Gaseous dioactivity Monitoring System,

b. The Containment Normal Sumps Level, Reactor Cavity Sump Level and Jl Flow Monitoring System, and C,asec w o r l Either the

c. ntainment air cooler condensate flow rate or a Containment l Atmosphere articulate Radioactivity Monitoring Systergr"/W h/en cadd ra a a. u .i.a.

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

t ACTION:

With only two of the above required Leakage Detection Systems OPERABLE, t operation may continue for up to 30 days provided grab samples of the contain-ment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the required Gaseous or Particulate Radioactive Monitoring System is inoperable; otherwise, 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

4.4.6.1 The Leakage Detection Systems shall be demonstrated OPERABLE by:

and Pw Heulag

! a. Containment Atmosphere Gaseousr[ Monitoring Systems performance of

CHANNEL CHECK, CHANNEL CALIBRATION, and ANALOG CHANNEL OPERATIONAL j TEST at the frequencies specified in Table 4.3-3,

b. Containment Normal Sumps Level, Reactor Cavity Sump Level and Flow Monitoring System performance of CHANNEL CALIBRATION at least once j

per 18 months, and

c. Containment Air Cooler Condensate Monitoring System performance of CHANNEL CALIBRATION at least once per 18 months.

l b

i V0GTLE - UNIT 1 3/4 4-19 M 24 h H

l. -

l __ _ _ _ _ - -._ . . _ , - . _ . . . _. ,

mm _ z_ _u- e e J v a / .mL .i.,.x t .n.__..l.. .ll.1 Z .T.Z .1 REACTOR COOLANT SYSTEM -

2 OPERATIONAL LEAKAGE

)

! LIMITING CONDITION FOR OPERATION

! 3.4.6.2 Reactor Coolant System leakage shall be limited to:

'j j a. No PRESSURE BOUNDARY LEAKAGE,

b. 1 GPM UNIDENTIFIED LEAKAGE, i c. 1GPMtotal[ - o secondary leakage through all steam generators

j and 500 gallons per day through any one steam generator, 1

] d. 10 GPM IDENTIFIED LEAKAGE from the Reactor Coolant System, l; e. GPM CONTROLLED LEAKAGE at a Reactor Coolant System pressure of j HIS i 20 psig.

f. 0.5 GPM leakage per nominal inch of valve size up to a maximum of 5 GPM at a Reactor Coolant System pressure of 2235 1 20 psig from any Reactor Coolant System Pressu e Isolation Valve,g ciff t S 3ekh :. " ~ ^ The muimum alniaNe bador 1 Cxc fret.wrt Dolation Valetaleakagt shst/ bewaift.sperjR&

a th i APPLICABILMDlES1, M :fagt @ Q*2, 3, and 4.

ACTION:

1 a. With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT STANDBY j within 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 .

.i

-j.

b. With any Reactor Coolant System leakage greater than any one of the

, above limits, excluding PRESSURE BOUNDARY LEAKAGE and leakage from

Reactor Coolant System Pressure Isolation Valves, reduce the leakage

., rate to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-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 1 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . .

g gj l

j

c. With any Reactor Coolant Systre ressure Isolation Valve leakage greater than the ebewe limit? isolate the high pressure portion of

') the affected system from the low pressure portion within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by i use of at least two closed manual or deactivated automatic valves, i 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 , l

q

Test pressures less than 2235 psig but greater than 350 psig are allowed.

)i Observed leakage shall be adjusted for the actual test pressure up to i 2235 psig assuming the leakage to be directly proportional to pressure I differential to the one-half power.

i l V0GTLE - UNIT 1 3/4 4-20 1

~j APR 241986 i

' ^

gia.un.L = ~ = xwni ::.am._ R&a.s;&db..;iC..La_.

1 i

] REACTOR COOLANT SYSTEM 3/4.4.8 SPECIFIC ACTIVITY d

j LIMITING CONDITION FOR OPERATION 1

1 1

3.4.8 The specific activity of the reactor coolant shall be limited to:

]

j a. Less than or equal to 1 microcurie per gram DOSE EQUIVALENT I-131, and

]

1 j ,

b. Less than or equal to 100 4 microcuries per gram of gross radioactivity.

APPLICABILITY: MODES 1, 2, 3, 4, and 5.

.]

3 ACTION:

MODES 1, 2 and 3*:

a. With the specific activity of the reactor coolant greater than

1 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with T less than 500 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; and o f ge n s A d/oac h Wy 1 b. With the specific acti y of the reactor coolant greater than 100 4 microCuriespergram!beinatleastHOTSTANDBYwithT avg less than 500 F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

MODES 1, 2, 3, 4, and 5:

ofgrau Nd/BacMdd/

1 ivity of the reactor coolant greater than d Withthespecifi[c 1 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100 d micro-j Curies per gram, perform the sampling and analysis requirements of Item 4.a) j j

of Table 4.4-4 until the specific activity of the reactor coolant is restored to within its limits.

l 9

ij SURVEILLANCE REQUIREMENTS il a

4.4.8 The specific activity of the reactor coolant shall be determined to be within the limits by performance of the sampling and analysis program of Table 4.4-4.

d 1

1

With T,yg greater than or equal to 500*F.

~V0GTLE - UNIT 1 3/4 4-26

! APR 241986 t

o 4

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

ke .:. .ww w., , x.w - 22:_. z_u==a.c.Lu.aw:hb u 511 526

TABLE 4.4-4 (Continued)

? , MAIT I TABLE NOTATIONS l I4 l *Agrossradioactivityanalfsisshallconsistofthequantitativemeasurement 3

. of the total specific actisity of the reactor coolant except for radionuclides l with half-lives less than A minutes and all radioiodines. The total specific

! activity shall be the sum of the degassed beta gamma activity and the total of all identified gaseous activities in the sample within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after the sample

~

is taken and extrapolated back to when the sample was taken. Determination of the contributors to the gross specific activity shall be based upon those 5 energy peaks identifiable with a 95% confidence level. The latest available data may be used for pure beta emitting radionuclides.

- A radiochemical hnalysi,s or'l shall consist of the quantitative measurement of the specific activi$y for each radionuclide, except for radionuclides with half-lives less than 40-minutes and all radiciodines, which are identified in the reactor coolant. The specific activities fgr these individual radio-nuclides shall be used in the determination of_E for the reactor coolant sample. Determination of the contributors to E shall be based upon those energy peaks identifiable with a 95% confidence level.

- - Sample to be taken after a minimum of 2 EFPD and 20 days of POWER OPERATION have elapsed since reactor was last subtritical for 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or longer.

Until the specific activity of the Reactor Coolant System is restored f within its limits.

t

[

i i

l l

1 i

4 i

i V0GTLE - UNIT 1 3/4 4-29 APit 241986 L ._. , .

j

-- by -

Nic . -

. mii 2;iawi, i G ,_,.. a_. ..a c e m d n, REACTOR COOLANT SYSTEM PRESSURIZER a

j LIMITING CONDITION FOR OPERATION

-1 .

j I

! 3.4.9.2 The pressurizer temperature shall be limited to:

a. A maximum heatup of 100*F in any 1-hour period,

. b. A maximum cooldown of 200'F in any 1-hour period, and awiliary s c. A maximum 4 spray water temperature differential of 625'F.

APPLICABILITY: At all times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition I on the structural integrity of the pressurizer; determine that the pressurizer I remains acceptable for continued operation or be in at least HOT STANDBY within l the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and reduce the pressurizer pressure to less than 500 psig )

within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . l 1

SURVEILLANCE REQUIREMENTS j 4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown. The I spray water temperature differential shall be determined to be within the l 4 limit at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during auxiliary spray operation. '

~

. \

l 1

V0GTLE - UNIT 1 3/4 4-34 APR 2i ggg 4

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1 4

$ REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS A

LIMITING CONDITION FOR OPERATION

( !

I 3.4.9.3 At least one of the following Overpressure Protection Systems shall

! i I

be OPERABLE: !

I i a. Two power-operated- relief valves (PORVs1 with + lift settingsef- l ggg.. . . , r u n , t .

wb ;

QN/wMd 20 Is/ EceidQNE'755)r esid wy wiHt Ryof!

lah/hdedM RC5 'funpemfurt-3.'/% of l 4 '

b. Two residual heat removal (RHR) suction relief valves each with a setpoint of 450 psig i 1%, or

c. The Reactor Coolant System (RCS) depressurized with an RCS vent ef-g t n -:;:r: Mehew (b M4 /4 #/E//t'*j of l

r: :t:m'11en least 670 SPM sva = p+ler fyra af f70 g. f APPLICABILITY: MODES 4 2:.; th: tc;. cretra-T,fw ey .uohHeg-4sdest ttlan-

-er_e; _' '- ;2753 I, :^^ : and MODE 6 with the re actor vessel head on.

l

> ACTION:

and$ !

a. With one PORV and one RHR suction relief valve inoperable, either e restore two PORVs or two RHR suction relief valves to OPERABLE status within 7 days or depressurize and vent the RCS thr:_;h et as specieeCl f

g //13,e/.1.3.cgbe 227 2 -

"^" N 9:L

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

l With both PORVs and both RHR suction relief valves inoperable, b.

I depressurize and vent the RCS th m gt M :st 2 --:;:r+--inch as 3/ec/Od

[ jn 3 </...j

. 9 5 c abreventr within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

c. In the event either the PORVs, the RHR suction relief valves, or the f RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to f Specification 6.9.2 within 30 days. The report shall describe the circumstances initiating the transient, the effect of the PORVs or RCS vent (s) on the transient, and any corrective action necessary to '

prevent recurrence.

$ d. The provisions of Specificatfon 3.0.4 are not applicable.

i V0GTLE - UNIT 1 3/4 4-35

%24 586

.)

' ^

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_ 5, w. qa- - . - . .. .

. . . w_a.u :a i EMERGENCY CORE COOLING SYSTEMS 1

3/4.5.2 ECCS SUBSYSTEMS - T, GREATER THAN OR EQUAL TO 350*F BRAFT q

LIMITING CONDITION FOR OPERATION j

3.5.2 Two independent Emergency Core Cooling System (ECCS) subsystems shall be OPERABLE with each subsystem comprised of:

a. One OPERABLE centrifugal charging pump,

b. One OPERABLE Safety Injection pump U: r h :; ph : 'd ,

c. One OPERABLE RHR heat exchanger, 1

d. One OPERABLE RHR pump, and I e. An OPERABLE flow path capable of taking suction from the refueling water storage tank on a Safety Injection signal and semi-automatically I

transferring suction to the containment emergency sump during the recirculation phase of operation. l APPLICABILITY: MODES 1,2,and3b ACTION:

a. With one ECCS subsystem inoperable, restore the inoperable subsystem 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 STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following j 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

I y b. In the event the ECCS is actuated and injects water into the Reactor r

Coolant System, a Special Report shall be prepared and submitted to j the Commission pursuant to Specification 6.9.2 within 90 days describ-

? ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor f

> for each affected Safety Injection nozzle shall be provided in this

[

Special Report whenever its value exceeds 0.70.

t (Ol$ $ O $C$ l'[AY Oh *he $* * *Y b l l Se edry hfo Mode 3 & Me Saf)/y IfecHe /bg dec/and%,ea4/e Pcau/ s S,seci/jupn 3.S.3 2. proria/ 6e 34/2/> Zirjec/N /smy h 4ff it.s/0tfd /c 0PEAfIBL6 S/4fus aj//f}; y,fpyp- , p ,/jy np p e

) 1%perufu,c of one or enore of Me RLJ cs/d/p etceediy J75'f' ww ouan ,qw.

B V0GTLE - UNIT 1 3/4 5-3

! APR 2([1986 l ,

I.___ . _ _ _ _ _ _ _ _ , . _ _ _ _ . _ . .

kgisk_Rc. :2, %: , , wa_g: ~ n:.254%.iha _u .L. . . ; _ -.x&ams i

EMERGENCY CORE COOLING SYSTEMS j SURVEILLANCE REQUIREMENTS l 4.5.2 Each ECCS subsystem shall be demonstrated OPERABLE:

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

! are in the indicated positions with power to the valve operators removed:

l Valve Number Valve Function Valve Position HV-8835 SI Pump Cold Leg. Inj. OPEN

! HV-8840 RHR Pump Hot Leg. Inj. CLOSED HV-8813 - SI. Pump Mini. Flow Isol. OPEN HV-8806 SI Pump Suction from RWST OPEN i HV-8802A, B SI Pump Hot Leg Inj. CLOSED HV-8809A, B RHR Pump Cold Leg Ini OPEN W- BE03A, 6 CCP Daawy 4o Dxdn.J.%k OPM

b. At least once per 31 days by:

1) Verifying that the ECCS piping is full of water by venting the

ECCS pump casings and accessible discharge piping high points, j and f 2) Verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

c. By a visual inspection which veriffes that no loose debris (rags, trash, clothing,'etc.) is present in the containment which could be transported to the Containment Emergency Sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1) For all accessible areas of the containment prior to establish-ing CONTAINMENT INTEGRITY, and

2) Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established. I r d. At least once per 18 months by:

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

a) With a simulated or actual Reactor Coolant System pressure signal greater than or equal to 425 psig the interlocks

prevent the valves from being opened, and b) With a simulated or actual Reactor Coolant System pressure signal'less than or equal to 750 psig the interlocks will I cause the valves to automatically close.

2) Avisualinspectionofthebntainment umpaNerIfyingthat i the subsystem suction inlets are not restricted by debris and L

that the sump components (trash racks, screens, etc.) show no evidence of structural distress or abnormal corrosion.

V0GTLE - UNIT 1 3/4 5-4 APR 24 aos I

l -

L= x.z;. u.a =_ . & _= / .; u. a . w . .x z.x. . w ,

EMERGENCY CORE COOLING SYSTEMS i

l 3/4.5.3 ECCS SUBSYSTEMS - T, LESS THAN 350*F o

LIMITING CONDITION FOR OPERATION

)

3.5.3.1As a minimum, one ECCS subsystem comprised of the following shall be OPERABLE:

, a. One OPERABLE centrifugal charging pump, j i b. One OPERAS'LE RHR heat exchanger, l

1 c. One OPERABLE RHR pump, and

.El d. An OPERABLE flow path capable of taking suction from the refueling

l water storage tank upon being manually realigned and transferring )

i suction to the Containment Emergency Sump during the recirculation !

phase of operation.

] '

c APPLICABILITY: MODE 4. (

I ACTION: f

a. With no ECCS subsystem OPERABLE because of the inoperability of either the centrifugal charging pump or the flow path from the y refueling water storage tank, restore at least one ECCS subsystem to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in COLD SHUTDOWN within the next

~

o. , 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months .

b. With no ECCS subsystem OPERABLE because of the inoperability of l J either the residual heat removal heat exchanger or RHR pump, restore

at least one ECCS subsystem to OPERABLE status or maintain the Reac-tor Coolant System T avg less than 350*F by use of alternate heat .

removal methods,

) c. In the event the ECCS is actuated and injects water into the Reactor Coolant System, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 90 days describ-ing the circumstances of the actuation and the total accumulated actuation cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided in this Special. Report whenever its value exceeds 0.70.

l

' ~ 4-"- ' -

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lj .srheM=6*-APERABhE whermat.hed==ture-Ofeomrr- ;1:f t.'a "CS c'd j h,G e.das- Guuuinn ==W+*4569F+-

':l

i. V0GTLE - UNIT 1 3/4 5-7 f)l APR 24 tcg U

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tf_a- .s <-- n . _ . . _ -:e .2- AP. 4 ca;m ;a. _. _.2m c_cu.e e s -, c2 -

EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS

4.5.3.1 The ECCS subsystem shall be demonstrated OPERABLE per the applicable requirements of Specification 4.5.2.

" s. A2-Ai; c .ergi..; pu ;s-- e+Sefety4..jnti: ;-- 3, excepht%ve-i allowed =4PEAASLC ;;;;;;;, :h:H-be ....w..etieted ir.:;;r_ti; ty. mifyi.g 4t=t.-

3 t'.: ;te. ci,cait br;:h.; cr: :::;r:d i- t' :;:r ;;;iti: :t != :t- r.;: per-2e < - u___..__ .<_

77 ,; _7 _ _ ._ _ ; f gy,; = _ -; _ ;pg_

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i V0GTLE - UNIT 1 3/4 5-8 .

APR 2 4 w.-

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EMERGENCY CORE COOLING SYSTEMS 0

3/4.5.3 ECCS SUBSYSTEMS - Tavg5 350 F ,

SAFETY INJECTION PUMPS LIMITING CONDITION FOR OPERATION 3.5.3.2 All Safety Injection pumps shall t* inoperable.

APPLICABILITY: Modes 4, 5, and 6 with the reactor vessel head on.

ACTION:

With a Safety Injection pump OPERABLE, restore all Safety Injection pumps to an inoperable status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.5.3.2 All Safety Injection pumps shall be demonstrated inoperable

by '

verifying that the motor circuit breakers are secured in the open position within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entering MODE 4 from MODE 3 prior to the temperature 0

of one or more of the RCS cold legs decreasing below 325 F, and at least once per 31 days thereafter.

(

An inoperable pump may be energized for testing or for filling accumulators provided the discharge at the pump has been isolated from the RCS by a closed isolation valve with power removed from the valve operator, or by a manual isolation valve secured in the closed position.

(

V0GTLE - UNIT 1 3/4 5-9

x.q . . M_ .4 e ..

MQLu. ,

E%

y .. .

l BORON INJECTION SYSTEM 3/4.5.4 REFUELING WATER STORAGE TANK vl 3 LIMITING CONDITION FOR OPERATION

-Ji

'j ;. 3.5.4 The refueling water storage tank (RWST) shall be OPERABLE with:

1

a. A minimum contained borated water volume of 631,478 gallons.

b. A boron concentration of between 2000 ppm and 2100 ppm of boron,

c. A minimum, solution temperature of 50*F, and a

,d d. A maximum solution temperature of 120*F. l

\

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

low-1cvel. !

. 0{ ;

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

~ ACTION:

a. With the RWST inoperable except for the Sludge Mixing Pump Isolation I Valves, 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 1 least HOT STANDBY within 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. With a Sludge Mixing Pump Isolation Valve inoperable, restore the valve

"..L '

to OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or isolate the sludge mixing train by

,.' either closing the manual isolation valves or deenergizing the OPERABLE solenoid pilot valve within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and maintaind closed.eder-2 :^ C^^t d .

. SURVEILLANCE REQUIREMENTS 1

j 4.5.4 The RWST shall be demonstrated OPERABLE:

,- a. At least once per 7 days by:

.] 1) Verifying the contained borated water volume in the tank, and

2) Verifying the baron concentration of the water.

1

'. 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 1 the outside air temperature is less than 50*F. l

'] c. At least once per 18 months by verifying that the sludge mixing pump q isolation valves automatically close upon an RWST low-level test

.i signal.

'l V0GTLE - UNIT 1 3/4 5-9 I

.a , ,

a APR t 41986

..i 3 -

.4

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

y &Q ,r:n & 9, 5- .o_.. _ _ _

_; h). .

[

t -

5)

! CONTAINMENT SYSTEMS CONTAINMENT LEAKAGE j l LIMITING CONDITION FOR OPERATION f

1 3.6.1.2 Containment leakage rates shall be limited to:

.i

/ a. An overall integrated leakage rate of less than or equal to L ,

l [0.203%byweightofthecontainmentairper24hoursatapr$ssure not less than P,, 45 psig, or j b.

] A andcombined valves subject leakage to Type B and rat'e C tests, of when lesspressurized than 0.60 to k ,L, for

'##' all pen j

1 45 psig.

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

ACTION:

]

With either the measured overall integrated containment' leakage rate exceeding

, 0.75 L, or the measured combined leakage rate for all penetrations and valves i subject to Types B and C tests exceeding 0.60 L,, restore the overall integrated leakage rate to less than 0.75 L,, and the combined leakage rate for all pene-trations subject to Type B and C tests to less than 0.60 L, prior to increasing the Reactor Coolant System temperature above 200*F.

i j SURVEILLANCE REQUIREMENTS 1

! 4.6.1.2 The containment leakage rates shall be demonstrated at the following

] test schedule and shall be determined in conformance with the criteria speci-fied in Appendix J of 10 CFR Part 50 using the methods and provisions of ANSI l, N45.4-[1972):

q

a. Three Type A tests (Overall Integrated Containment Leakage Rate) 1 shall be conducted at 40 1 10 month intervals during shutdown at j a pressure not less than P,, 45 psig, during each 10 year service 1 period. The third test of each set shall be conducted during the

] shutdown for the 10 year plant inservice inspection; i

d 4

1 fj 4

V0GTLE - UNIT 1 3/4 6-2 APR 2 41986 l

,y-. . . . . - - . - ~. . . -

_m.c_m. . . . . .

. . _ .: ~ :.~ . .... c s a a. u_ x_.ve.t:a :n. ._a..:-.mc_.,c.. .m.i ll ~

CONTAINMENT SYSTEMS i

i AIR TEMPERATURE l

LIMITING CONDITION FOR OPERATION i

l

, 3.6.1.5 Primary containment average air temperature shall not exceed 120 F.

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

ACTION:

With the containment average air temperature greater than 12 O*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 n 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 SHUTOOWN within the following i 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I SURVEILLANCE REQUIREMENTS l

i 4.6.1.5 The primary containment average air temperature shall be the arith-metical average of the temperatures at the following locations and shall be d

determined at least once per 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

Location Tag Numbers

a. Level 2 TE-2563 l b. Level B TE-2613

, c. Level C TE-2612 i

l 3 l l :

MOc local S*fk "f " #'Y " S'9 l#'"H#" ' 1

. l l l l

L

[ ,

V0GTLE - UNIT 1 3/4 6-7 APR 2 4198b j l

l i

~

ma..iam.m.a. ._.m u.uw MwA_ saeu.md_ab dL... 1.Li mace.. . . .u ..a j

'5 4 .

l CONTAINMENT SYSTEMS CONTAINMENT STRUCTURAL INTEGhITY LIMITING CONDITION FOR OPERATION 1

) 3.6.1.6 The structural integrity of the containment shall be maintained at a level consistent with the acceptance criteria in Specification 4.6.1.6.1 i

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

a. With more than one tendon with an observed lift-off force between the predicted lower limit and 90% of the predicted lower limit or

_]

g with one tendon below 90% of the predicted lower limit, restore the 1 tendon (s) to the required level of integrity within 15 days and perform an engineering evaluation of the containment and provide a

i; Special Report to the Commission within 30 days in accordance with l Specification 6.9.2 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 .

i

.} b. With any abnormal degradation of the structural integrity other than I

ACTION a. at a level below the acceptance criteria of Specification j 4.6.1.7, restore the containment to the required level of integrity 4

within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and perform an engineering evaluation of the contain-1 ment and provide a Special Report to the Commission within 15 days in accordance with Specification 6.9.2 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 l 4.6.1.6.1 Containment Tendons. The containment tendons' structural integrity 1 shall be demonstrated at the end of 1, 3, and 5 years following the initial i containment vessel structural integrity test and at 5 year intervals thereafter, j The tencons' structural integrity shall be demonstrated by:

a. Determining that a random but representative sample of at least 13 tendons (4 inverted U and 9 hoop) each have an observed lift-off force within predicted limits for each. For each subsequent inspection one tendon from each group may be kept unchanged to develop a history and to correlate the observed data. If the observed lift-off force of any one tendon in the original sample population lies between the predicted lower limit and 90% of the predicted lower limit, two tendons, one on each side of this tendon should be checked i for their lift-off forces. If both cf these adjacent tendons are found to be within their predicted limits, all three tendons should be restored to the required level of integrity. This single deficiency j

~

may be considered unique and accceptable. Unless there is abnormal degradation of the containment during the first three inspections,

the sample population for subsequent inspe tions shall include at least 5 tendons (2 inverted U and 3 hoop) . Th/1/e,/ s/u// /4 chm 6 essen//a//q ofe.u fe'rk /c deferonsu ifRnf 5'andr 7 de awptele de/emianm)d broken er Wd m ay e j} V0GTLE - UNIT 1 3/4 6 W 24 h 1

a .-- a.a.m a a. . - : . = = :.: E G L i. u. G i l b A d G d G 2 .O a G w i-.. & ,

~

! l CONTAINMENT SYSTEMS

.)

SURVEILLANCE REQUIREMENTS (Continued)

I j b. @:r':=.ing t=d= d:t=:':9;, %:;=thne, =d ::t;.-bl t::t: ;r ;-

<j px.h .;)y :t xced t=d=-fre ; =;h gr=p (in ;rted '.' rd heep).

< [tg/dceW74 :

A c; .dcl3 n '= t;d t = dr 'r = =:h gr =p shall b; x-; h t:ly-inse<f/0 d : t : r i = M e r d = t; i d = t ?-fy b = 'a n ; r d : ;0- r i r = = d d; te r -

/ age 3 (d-9 ^ ':; that .... th; .r.t! = ! = ;th ef t h . - , a win er :tr= d 4 / M

1

^4 q 1) The. tendon wires or strands are free of corrosion, cracks, and

, damage, J

]1 2) There are not changes in the presence or physical appearance of the sheathing filler grease, and

3) A minimum tensile strength of [240,000] psi (guaranteed ultimate

strength of the tendon material) for at least three w

) strand samples (one from each end and one at mid-length) cut from each removed.wwe=w strand. Failure of any one of the

! -w4ee=oc-strand samples to meet the minimum tensile strength test is evidence of abnormal degradation of the containment

,! structure.

c. Performing tendon retensioning of those tendons detensioned for inspection to their observed lift-off force with a tolerance limit

~

of +6%. During retensioning of these tendons, the changes in load and elongation should be measured simultaneously at a minimum of three approximately equally spaced levels of force between zero and

]1 the seating force. If the elongation corresponding to a specific load differs by more than 5% from that recorded during installation, j an investigation should be made to ensure that the difference is not j related to wire failures or slip of wires in anchorages; j d. Assuring the observed lift-off strxx:. forces u;;;dart-b behesp;.;madyum e s inimum andmM/ent 1 dr i s ; sMr given.bekw, wWrcheradhstetFtcEae.ceuntJacekstsc-

) h . u .;7654. Vah<U gh'en in J/L fedory suNtill@cb pr4Cedtge..

1 4.... . 0 % {12^] hs.

9=p. -e;";:a tim he

, same-- il34] hi-4

e. Verifying the OPERABILITY of the sheathing filler grease by:

] '

1) No voids in excess of 5% of the net duct volume,

2) Minimum grease coverage exists for the different parts of the j anchorage system, and a

3) The chemical properties of the filler material are within the

, tolerance limits as specified by the manufacturer.

j V0GTLE - UNIT 1 3/4 6-9 APR 2 4 boo

.rd f* * *

$m.gg

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

Insertforpage3/46-WT I

~

Removing a strand from one hoop tendon and one inverted U tendon and performing an inspection and material test. It shall be deter.nined that over the entire length of both removed strands:

1 1

. .._.1.m.,m ..;.,_ m a A.wm,t...a.m. .,,,, . _ .z.u... m h ..w.s ... & _ .._. .,. a.s : w

CONTAINMENT SYSTEMS -

1 CONTAINMENT COOLING SYSTEM l

) LIMITING CONDITION FOR OPERATION i

? 3.6.2.3 Two independent groups of containment cooling fans shall be OPERABLE i with four fans to each group. [ Equivalent to 100% cooling capacity.]

APPLICABILITY: PODES 1, 2, 3, and 4. ;

i l L ACTION:

9 I a. With one group of the above required containment cooling fans inoper-able and both Containment Spray Systems OPERABLE, restore the inoper-s able group of cooling fans to OPERABLE status within 7 days or be in 1 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 , f b. With two groups of the above required containment cooling fans inoperable and both Containment Spray Systems OPERABLE, restore at

j least one group of cooling fans 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 STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD L SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore both above required

! groups of cooling fans to OPERABLE status within 7 days of initial

loss 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 .

j c. With one group of the above required containment cooling fans inoper-

! able and one Containment Spray System inoperable, restore the inoper-able Spray System 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 STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLE SHUTDOWN

!. within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Restore the inoperable group of containment cooling fans to OPERABLE status within 7 days of initial loss 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 L COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

I SURVEILLANCE REQUIREMENTS

[ 4.6.2.3 Each group of containment cooling fans shall be demonstrated OPERABLE:

a. At least once per 31 days by:

1) Starting each ::r :; = ting fan from the control room, and '^^

verifying that each fan group operates at low speed -

for at least 15 minutes, and

2) Verifying a cooling water flow rate of greater than or equal to MM!& gpm per pair of containment fan coolers.

1359

b. At least once per 18 months by verifying that each fan group starts

, automatically and operates at low speed W,50^ -fe)-on a Safety Injection test signal.

1 1

4 V0GTLE - UNIT 1 3/4 6-15 i . APR 2 41986 1

L...._....w.-.,7 . . . . . . . - -

. . m a. w im_v w m m e.u a w x a.u s m w r c us w.Aan. _. .. .. G.RM- swN

~

i y CONTAINMENT SYSTEMS -

3/4.6.6 ELECTRICAL PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM .

'l

) /

J LIMITING CONDITION FOR OPERATION

)l

/

3.6.6 shall be\wo independent Electrical Penetration Room Exhaust Air Cleanup 0PERABLE. / Systems

\

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

/

ACTION:

h@ . /

i With one Electricgl Penetration Room Exhaust Air Cleanup.S stem inoperable, i restore the inoperable system to OPERABLE status within/7 days 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 . p/

i /

j SURVEILLANCE REQUIREMENTS /

g.

N /

4.6.6 Each Electrical Penetrat'on Room Exhaust Air Cleanup System shall be 3 demonstrated OPERABLE: /

/

a. At least once per 31 days n a STAGGERED TEST BASIS by initiating, from the cor. trol room, fidw\through the HEPA filters and charcoal adsorbers and verifying /that'the q system operates for at least 10

{, continuous hours withj the heaters operating; 5 / N

b. At least once per 18 months or (1)\pfter any structural maintenance i on the HEPA filter or charcoal adsorber housings, or (2) following painting, fire,for chemical release ih any ventilation zone,communi-cating with the system by:

o /

Verify,ing that the cleanup system sat 6fies the in place pene-d 1) tratjon and bypass leakage testing acce'ptance criteria of le'ss

- than [*]% and uses the test procedure guidance in Regulatory a Positions C.5.a, C.5.c, and C.5.d of Regulhtory Guide 1.52,

j Re' vision 2, March 1978, and the system flow Yate is cfm

: 610%;

2), Verifying, within 31 days after removal, that a aboratory analy-sis of a representative carbon sample obtained in'eccordance with d

/ Regulatory Position C.6.b of Regulatory Guide 1.52 evision 2, March 1978, meets the laboratory testing criteria of egulatory g Position C.5.a of Regulatory Guide 1.52, Revision 2, q for a methyl iodide penetration of less than [**]%; an,rch 1978,

1

t 4

M

!! V0GTLE - UNIT 1 3/4 6-24 i APR 2 41986

\

w a -

,7***9'.** Pam***'?* *P hM a ** **" M*9 7 se *'*"d**"NC'

,,- , *'2*W9**,9**- -'co-e==, w m m m ys.*#my,gger ., += eep % -w-

W..a. d m.LJEhb.AhnmEdaEEEwSEEbcam6w h.m amm p e k -

/

/ ,

, CONTAINMENT SYSTEMS

,/

I i SURVEILLANCE REQUIREMENTS (Continued) ,

3) Verifying a system flow rate of cfm i 10% during system operation when tested in accordance with ANSI N510-1975.

{ c. A(terevery720hoursofcharcoaladsorberoperation,byverifying,

wi sin 31 days after removal, that a laboratory analysis of,a repre-

< sent'ative carbon sample obtained in accordance with Regulatory i Positi'on C.6.b of Regulatory Guide 1.52, Revision 2, March 1978, meets th'e s laboratory testing criteria of Regulatory Position C.6.a '

of Regulat'ory Guide 1.52, Revision 2, March 1978, for'a methyl iodide penetration of .less than [**]%;

i d.

N At least once per 18 months by:

1) Verifying that the s pressure drop across .the combined HEPA filters and charcoal adsorber banks is/less than [6] inches

- Water Gauge while operating the syster.,. at a flow rate of cfm i 10%,

( p/

2) Verifying that the s gtem starts on a Safety Injection test j signal, l 3) VerifyingthatthefiltercooIingbypassvalvescanbemanually i

opened, and /

4) Verifying that the heater's issipate .

i kW when tested in accordance with ANSI N510-137E e.

/ \

After each complete or partial repla 2 ment of a H$PA filter bank, by E

verifying that the cleanup system sati'sfies the in place penetra-tion and bypass leakag'e testing acceptance criteria of less than [*]%

in accordance with ANSI N510-1975 foraDDRtestaerosolwhileoperat-t cfm 1.10%, and ingthesystema/aflowrateof \

f. After each complete or partial replacement of 'a scharcoal adsorber bank, by veri

' penetration a,fying nd bypassthat leakage the cleanup testing system satisfies acceptance the inofplace t:riteria less f

than[*]%jnaccordancewithANSIN510-1975 for a h'alogenated

.g hydrocarbon refrigerant test gas while operating the ' system at a 5

4 flow rate of cfm i 10%. N

0.05% value , applicable when a HEPA filter or charcoal adsorber e ciency of 99% is assumed, or 1% when a HEPA filter or charcoal adsorber efficiency t of 95% or/less is assumed in the NRC staff's safety evaluation. (Use'the 3 value ass'umed for the charcoal adsorber efficiency if the value for the\

3 HEPAf)1terisdifferentfromthecharcoaladsorberefficiencyintheNRC\

j staff)s safety evaluation.) s j **Valie applicable will be determined by the following equation: \

i P f1 E

, when P equals the value to be used in the test requirement

(%), E is efficiency assumed in the SER for methyl iodide removal (%), \

and SF is the safety factor to account for charcoal degradation between \

j tests (5 for systems with heaters and 7 for systems without heaters). '

3 y V0GTLE - UNIT 1 3/4 6-25 ;

.}

APR 2 41986

..,--e-m... < ., - n.e -.-4. - . .y n gs .-..-e -a.. ,y- ec --e n , , . .-

,_x- .:asw. : . ... ca.au a.h x20i.m=&aiuk@2d Ltd15kia.hL.hiwhuw2::ai1,.a:-..u r

PLANT SYSTEMS k SURVEILLANCE REQUIREMENTS (Continued)

A 3) Verifying that each non-automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in

) its correct position; and f

R

4) Verifying that each automatic valve in the flow path is in the fully open position whenever the Auxiliary Feedwater System is pl: ::d 'rstomeMc catat:r t.;c ;L;40 HATES THEAMAL 99WEk in slandkf /dr ar<nMeg &<d'cwfu aukvia, tic. truyr*a far, or 4thM abut /0 9a RATE 6 T/fiRMAL fopygg*

b. At least once per 18 months during shutdown by:

1) Verifying that each automatic valve in the flow path actuates

- to its correct position upon receipt of an Auxiliary Feedwater i Actuation test signal, and 1 2) Verifying that each auxiliary feedwater pump starts as designed 1

automatically upon receipt of an Auxiliary Feedwater Actuation

! test signal.

i .

I I

I i

l

.1 J

I 4 l I

4 1

'l l

.k l 3/4 7-5 j V0GTLE - UNIT 1 m 2.: me ;

> l A i 1

1 i

-. _.md-.: 1%a :~. . s.aa . 2:., t.u:.M2... ' . hrs 2:a.a:hGik.s: .. .=.< ~ w - asa+:... .,a l

PLANT SYSTEMS

! CONDENSATE STOR*'!iE TANK i

LIMITING CONDITION FOR OPERATION

s///001 3.7.1.3 sash Condensate Storage Tank (CST)[shall be OPERABLE with a contained water volume of at least 340,000 gallons of water.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

t y v100I _

! With +ne CSTFinoperable, within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

I" VVm/

a. Restore th: 'q:r:ble CST to OPERABLE status 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 HOT SHUTDOWN within i the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , or

( ,y400I y v4002.

b. Demonstrate the/0PERABILITY of O: - :::' ':;; CSTrand restore %e j ' :;;r;t'.; CST to#OPERABLE status within 7 days or be in at least 3 HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the

{ following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

f i

SURVEILLANCE REQUIREMENTS l-t l \ftlool

4.7.1.3.1 .Eeek CST [shall be demonstrated OPERABLE at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by g verifying the contained water volume is within its limits when the tank. is the

supply source for the auxiliary feedwater pumps.

l lj,q, j, 3. z. CST WMz. Sits // be clem ons/rahd' OPEPJ)Bl.2 af NBf once per i2. hovts veri iny N1L 6per/shed'W4t'er to/ tune iJ Gf l84Jf

, se,000 gaa s s-- u i, m sm s- & a auay m e a w p ~yas.

i V0GTLE - UNIT 1 3/4 7-6 APR 2 41986

u.. s i ux G ua n d u s a n a bSas k 5 % au.LG L a.::c m. 1,w w & w ~

f . a ac.u aJ d L s .

1 Os TABLE 4.7-1 .

1

' SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY h

y SAMPLE AND ANALYSIS PROGRAM L!

I TYPE OF MEASUREMENT SAMPLE AND ANALYSIS AND ANALYSIS FREQUENCY

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

2 Determination *

2. Isotopic Analysis for DOSE a) Once per 31 days, when-EQUIVALENT I-131 Concentration ever the gross radio-activity determination i indicates concentrations k greater than 10% of the

!; allowable limit for radioiodines.

b) Once per 6 months, when-ever the gross radio-activity determination indicates concentrations less than or equal to 10%

of the allowable limit

for radioiodines.

I '

i . M h *A gross radioactivity analysis shall c sist of the quantitative measurement Il of the total specific activity of the secondary coolant except for radio-f nuclides with half-lives less than W minutes. Determination of the contributors to the gross specific activity shall be based upon those energy peaks identifiable with a 95% confidence level.

}

h L

!i V0GTLE - UNIT 1 3/4 7-8 24 M a '

l

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PLANT SYSTEMS Systems (Consistin1 ef a.mn.in sMon l MAIN STEAM LINE ISOLATION VALVES i.s e /s.Ro n valve.(MsiV) so,4. i t.s

' ersociadeel bypass vo.lve(Msie v) pee LIMITING CONDITION FOR OPERATION ,un, ff,, , .

7%so 3.7.1.5 4een main steam line isolation ;;in '"0".') shall be OPERABLE.&

APPLICABILITY: MODES 1, 2, and 3.

ACT ON:

M00E-h Mi+2 er: MSIY ineperel but egen. Peuge gpgoa779u ... 33 4 ..

psevi &d th- 4 crerable valve 1: restered t: OPE"^?LE :t:t : ith'-

' h;;.;r:; :th;r.;i:: 5; in "0T ST'."0"Y within th; a xt S h ur: ;

"^T S""

rith'- th: ' ll rin; 5 5: r:.

"00E5 ? :nd 3:

With :n: "SIY in:p;r:51:, : h::; : t :p;r:ti:n in "00E O ;r : 23 pr;;;;d p ='id^d +ha ise'atica valve 15 --i-trined c1:::d. Oth r.;i::, :,; in l;0T ST""02Y with*- th: :xt S h: r: :nd in "0T S"L"005" within th: f;11,, ia; h.

ge.p/ ace w/M inser+ fa $9 e S/4 1-9 Y SURVEILLANCE REQUIREMENTS n.nnf MS/8V 4.7.1.5 Each MSI shall be demonstrated OPERABLE by verifying full closure within 5- seconds when tested pursuant to Specification 4.0.5. The provisions of Specification 4.0.4 are not applicable for entry into MODE 3.

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l An EMU /lf main .mdeam line ise/<fien syshm ca.n consut of an 0/ EAU 18tf MstV and an in opersh/e bud closed. associated his /6 Y provided the inoperabk psis y i.s ma.intained. clerel.

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Insert for Page 3/4 7-9 MODE 1:

a. With two main steam line isolation systems in any steam line inoperable; POWER OPERATION may/ continue provided each MSIV in the affected steam line is open and at least one main steam line isolation system in the affected steam line is restored to OPERABLE status within._ hours. Otherwise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,

b. With one main steam line isolation system inoperable, power operation r.;ay continue provided the MSIV in the affected isolation system is open and the inoperable system is restored to OPERABLE status within _ days. Other-wise, reduce power to less than or equal to 5% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

MODES 2 and 3:

a. With two main steam line isolation systems in any steam line inoperable, subsequent operation in MODES 2 or 3 may proceed provided at least one main steam line isolation system in the affected steam line is maintained closed.

The provisions of Specification 3.0.4 are not applicable. Othemise, be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With one main steam line isolation system inoperable but closed, subsequent operation in MODES 2 or 3 may proceed provided that the isolation system in

( the affected steam line is maintained closed. The provisions of Specification 3.0.4 are not applicable. -

c. With one main steam line isolation system inoperable but open, either close the OPERABLE isolation system or restore the inoperable system to OPERABLE status within days. Othemise be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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PLANT SYSTEMS 3/4.7.3 COMPONENT COOLING WATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.3 Two independent component cooling water trains snall be OPERABLE with at least two pumps per train.

APPLICABILITY: H0 DES 1, 2, 3, and 4.

ACTION:

With only one component cooling water train OPERABLE, restore the inoperable trains 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 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

4.7.3 At least two component cooling water trains shall be demonstrated OPERABLE:

, a. At least once per 31 days by verifying that each valve 'r::::',- Or pr-:r ;p;reted)-eenairf ;_re' *; . ;';t;d ;,.:r. i.; that is not locked, sealed, or otherwise secured in position is in its correct position; and

b. At least once per 18 months during shutdown, by verifying that each Component Cooling Water System pump starts automati: ally on a Safety Injection test signal.

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3/4.7.6 CONTR'OL ROOM EMERGENCY FILTRATION SYSTEM LIMITING CONDITION FOR OPERATION y

3.7.6 Two independent Control Room Emergency Filtration Systems shall be a

OPERABLE.

f APPLICABILITY: -AH=N99E-S. Nidd6S j j I, Z .3 4&/. NO8/J 6 dad 6 MMp tr>0yemenf afirYaclid/td { e/ or n1svtmestf af /Sads aves

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irradh/el fue/.

MODES 1, 2, 3 and 4':

]

With one Control Room Emergency Filtration System inoperable, restore 41 the inoperable system to OPERABLE status within 7 days 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 SHUI]OWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

darur movent essf of iritt/ikfea' fue/ Or* mortinenf af /Mb ochy MODES 5and6[irr. afe/ /ue/

, a. With one Control Room Emergency Filtration System inoperable, restore the inoperable system to OPERABLE status within 7 days or 9, initiate and maintain operation of the remaining OPERABLE Control Room Emergency Filtration System in the reti nal:t h a mode.

e m erryen ci fdffol"'1

b. With both Control Room Emergency Filtraticn 59ste i or with the OPERABLE Control Room Emergency p System, - Ch::popIab 1 '##d *'7 required to be in the reciEchtka mode by ACTION a., not capable of being powered by an OPERABLE emergency power source, suspend all iMT-I- m_

-l operations involving CLmow)Mf -/ 0f frfdsfidEU ?rc/

e,7f sfornrorm,Jth~

/asa') r;;;;,;;isity ch:n;;:D& i 1 ke /.

SURVEILLANCE REQUIREMENTS Sl 4.7.6 Each Control Room Emergency Air Cleanup System shall be demonstrated OPERABLE:

i a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months by verifying that the control room air temperature is less than or equal to [80]*F;

, b. At least once per 31 days on a STAGGERED TEST BASIS by initiating, from the control room, flow through the HEPA filters and charcoal 4

10 continuous hours with the heaters ;;;mth;; ca,s-/n/ C/ra*

adsorbersandverifyingthatthesystemoperatesforatleast,/(MCA/l'C /

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PLANT SYSTEMS SURVEILLANCE REQUIREMENTS (Continued)

3) Verifying a system flow rate of 16,000 cfm 1 10% during system F

operation when tested in accordance with ANSI N510-1980.

I After every 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of charcoal adsorber operation, by verifying, e c.

within 31 days after removal, that a laboratory analysis of a

)

representative carbon sample obtained in accordance with Regulatory

s Position C.6.b of Regulatory Guide 1.52, Revision 2, March 1978,

[ meets the laboratory testing criteria of Regulatory Position C.6.a i of Regulatory Guide 1.52, Revision 2, March 1978, for a methyl k iodide penetration of less than 99.8%;

4 .

j d. At least once per 18 months by:

1

I 1) Verifying that the pressure drop across the combined HEPA filters and charcoal adsorber banks is less than [6] inches l Water Gauge while operating the system at a flow rate of 16,000 a ,

cfm + 10%,

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Cmhiwn/-%N/A/b/ Isola //en 1 2) Verifying that the system starts on a-Cefit; % ..: n test

j. signal, 3

3) Verifying that the system maintains the ECCS pump room at a negative pressure of greater than or equal to 1/8 inch Water Gauge relative to the outside atmosphere,

4) Verifying that the filter cooling bypass valves can be manually opened, and i

5) Verifying that the heaters dissipate 80 1 4 kW when tested in accordance with ANSI I i.0-1980.

If i e. After each complete or pai al replacement of a HEPA filter bank, j~

by verifying that the cleanup system satisfies the in place pene-tration and bypass leakage testing acceptance criteria of less than H

99.5% in accordance with ANSI N510-1980 for a DOP test aerosol while i operating the system at a flow rate of 16,000 cfm i 10%; and l

j f. After each complete or partial replacement of a charcoal adsorber I bank, by verifying that the cleanup system satisfies the in place ,

j j penetration and bypass leakage testing acceptance criteria of less than 99.5% in accordance with ANSI N510-1980 for a halogenated i

I

hydrocarbon refrigerant test gas while operating the system at a flow rate of 16,000 cfm i 10%.

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PLANT SYSTEMS -

u SURVEILLANCE REQUIREMENTS (Continued)

b. Stored sources not in use - Each sealed source and fission detector

's shall be tested prior to use or transfer to another licensee unless

': tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use; and i

[ c. Startup sources and fission detectors - Each sealed startup source

. and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair 3: or maintenance to the source.

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TABLE 3.3-13 8

Ej RADI0 ACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION n;

d MINIMUM CHANNELS g INSTRUMENT OPERABLE

APPLICABILITY ACTION N 1. GASEOUS WASTE PROCESSING SYSTEM

a. Noble Gas Activity Monitor - .

Providing Alarm and Automatic ',

Termination of Release (ARE-0014) 1 45

b. Effluent System Flow Rate Measuring Device (AFI-0014) 1 *** 46

2. GASE0US WASTE PROCESSING SYSTEM Explosive Gas Monitoring System .

a. Hydrogen Monitor 1/recombiner *: 49, 5 %

b. "y i ;;; r-0xygen Monitor 2/recombiner **

49

3. Condenser Air Ejector and Steam Packing Exhauster System

a. Noble Gas Activity Monitor 1 *** 47 (RE-12839C)

b. Iodine Sampler 1 *** 51 (RE-12839B)

c. Particulate Sampler 1 *** 51 (RE-12839A)

d. Flow Rate Monitor 1 *** 46

3. (FI-12839) g e. Sampler Flow Rate Monitor 1 *** 46 m (FI-12839)

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'i ELECTRICAL POWER SYSTEMS I A.C. SOURCES 1

1 SHUTDOWN l

1 LIMITING CONDITION FOR OPERATION l

3.8.1.2 As a minimum, the following A.C. electrical power sources shall be OPERABLE:

)

I

a. One circuit between the offsite transmission network and the Onsite Class 1E Distribution System, and 1

b. One diesel generator with:

1) A day tank containing a minimum volume of 750 gallons of fuel,

2) A fuel storage system containing a minimum volume of 64,000 gallons of fuel,

3) A fuel transfer pump,

--4) u Lubricat4ng4i.3-storagWaining-e-eie4 mum-teteWhee-of-p' ions-oHWienting-e64 r-and-

-5) -- C@ility:to-transfee Jubricating4ti-4 rom-sto. + t; the~

Ji:::1 gene #ah r N A -

g APPLICABILITY: MODES 5 and 6.

i ACTION:

{

I With less than the above minimum required A.C. electrical power sources OPERABLE, immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, movement of irradiated fuel, or crane operation with loads over the fuel storage pool, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the Reactor Coolant System through a greater than or equal to square inch vent. In addition, when in MODE 5 with the reactor coolant loops not filled, or in MODE 6 with the water level less than 23 feet above the reactor vessel flange, immediately initiate corrective action to restore the required sources to OPERABLE status as soon as possible.

SURVEILLANCE REQUIREMENTS 4.8.1.2 The above requ' fred A.C. electrical power sources shall be demonstrated OPERABLE by the performance of each of the requira ment of Specifications 4.8.1.1.1, 4.8.1.1.2 (except for Specification 4.. .1.1.2a.5)), and 4.8.1.1.3.

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i 3/4.8.2 D.C. SOURCES l

I OPERATING q LIMITING CONDITION FOR OPERATION l

q l 3.8.2.1 As a minimum, the following D.C. electrical sources shall be OPERABLE:

J

a. 125 V-dc Battery bank 1ADlB, and one of its associated full-capacity chargers.

b. 125 V-dc Battery bank 1BD1B, and one of its associated full-capacity chargers.,

c. 125 V-dc Battery bank ICD 1B, and one of its associated full-capacity chargers.

d. 125 V-dc Battery bank 1D018, and one of its associated full-capacity chargers.

APPLICABII.ITY: MODES 1, 2, 3, and 4.

aHusfone ACTION:

With one of the required battery banks and/orNe"oM b -

restore the inoperable battery bank and/ord'"" - :w'ity charger chargers inoperable, to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 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 .

R I

j SURVEILLANCE REQUIREMENTS

04g. associaled

)r. 4.8.2.1 Each 125-volt battery bank anddharger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1) The parameters in Table 4.8-2 meet the Category A limits, and q 2) The total battery terminal voltage is greater than or equal to n 126 volts on float charge.

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TABLE 4.8-2 BATTERY SURVEILLANCE REQUIREMENTS

)

d CATEGORY A(1) CATEGORY B(2) i i PARAMETER LIMITS FOR EACH LIMITS FOR EACH ALLOWABLE (3) l 0

DESIGNATED PILOT CONNECTED CELL VALUE FOR EACH CELL CONNECTED CELL

~

Electrolyte > Minimum level > Minimum level Above top of y Level indication mark, and < %" above indication mark, and < %" above plates, and not maximum level maximum level overflowing j] indication mark indication mark Float Voltage > 2.13 volts > 2.13 volts (6) > N volts 4

Not more than 0.020 below the l -

/, /90 average of all l Specific > -4i400{5) -> %,406 connected cells Gravity (4) /, /9.5 Average of all Average of all connected cells connected cells

> tir896- 3 (5) 1.200 1:

TABLE NOTATIONS (1) For any Category A parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months all the Category B measurements are taken and found to be within their allowable values, and

[ provided all Category A and B parameter (s) are restored to within limits within the next 6 days.

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(2) For any Category B parameter (s) outside the limit (s) shown, the battery may be considered OPERABLE provided that the Category B parameters are

, within their allowable values and provided the Category B parameter (s) are restored to within limits within 7 days.

. (3) Any Category B parameter not within its allowable value indicates an

inoperable battery.

(4) Corrected for electrolyte temperature and level, j j (5) Or battery charging current is less than 2 amps when on float charge.

(6) Corrected for average electrolyte temperature.

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D.C. SOURCES flS-V W/& k kJ AMdB ar/ fDDiB) God 01L SHUTDOWN j g /

LIMITING CONDITION F0 OPERATION i

t 3.8.2.2 As a minimum, % : 125 veltttt: yh2 ..J -it :::::Teted-fttW

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shall be OPERABLE.

APPLICABILITY: MODES 5 and 6. gg ACTION: bo//) /

i i With the required bhttery bank and/or%11 ::p::ity- / char ger5 /

inoperable, ,

immediately suspend all operations involving CORE ALTE$ATIONS, positive l

3 reactivity changes, or movement of irradiated fuel; irfitiate corrective ;

action to restore the required battery bank and fvM-capeetty charger to OPERABLE status as soon as possible, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and i vent the Reactor Coolant System through a square inch vent.

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SURVEILLANCE REQUIREMENTS 4.8.2.2 The above required 460/12 olt battery bank and full-capacity charger shall be demonstrated OPERABLE in accordance with Specification 4.8.2.1.

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j 3/4.8.3 ONSITE POWER DISTRIBUTION j OPERATING LIMITING CONDITION FOR OPERATION l

I j 3.8.3.1 The following electrica' busses shall be energized in the specified

, manner with tie breakers open [both] between redundant busses within the unit

[and between units at the same station]:

a. A.C. Emergency Busses consisting of: i

{ 1. Train A a) 4160 volt switchgear 1AA02 b) '

480 volt switchgear 1AB04 Ll 1) MCC 1ABE l c) 480 volt switchgear 1AB05

1) MCC 1ABA J) 2) MCC 1ABC 1 3) MCC 1ABF d) 480 volt switchgear 1AB15

1) MCC 1ABB

2) MCC 1ABD 3 2. Train B 3 a) 4160 volt switchgear 1BA03 s b) 480 volt switchgear 18B06

] 1) MCC IBBE

c) 480 volt switchgear 18807 5

1) MCC 1BBA

{ 2) MCC 1BBC

3) MCC 1BBF d) 480 volt switchgear 1BB16

1) HCC 1BBB l 2) MCC 1880

b. 120 volt A.C. vital Busses

! 1. Associated with Train A i a) Channel I e

1) Panel 1AY1A energized from inverter 1AD111 connected to switchgear 1AD1*

2) Panel 1AY2A energized from inverter IA01111 connected e to switchgear 1AD1*

I b) Channel III i Panel ICY 1A energized from inverter 1C01I3 connected 1) h to switchgear 1C01*

2. Associated with Train B f a) Channel II

$ 1) Panel 1BY1B energized from inverter 1801I2 connected l to switchgear IBD1* .

(in a sin aSSocia/ed sel/ck9 ear f "Twoinverterstmaybehaje.j,s;gsconnected from their-Di[@i=4us for up to 24 necessar). for the purpose of pl 5:t'- ; i d provided: (1) heir .vit:1 cri .;'r-::eq#9; are energizedchcr;@; ; . th;ir :::::kted

2) the v44e4

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a ONSITE POWER DISTRIBUTION h

3 LIMITING CONDITION FOR OPERATION 1

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h 2) Panel 18Y2B energized from inverter 1801I12 connected 4 to switchgear 1801*

j b) Channel IV

]

1) Panel 10Yl8 energized from inverter 1D01I4 connected to switchgear 1D01*

q 7 c. 125 volt D.C. Busses consisting of:

L

1. Associated with Train A

] a) System A

) 1) Switchgear IAD1 energized from battery 1AD1B

2) MCC 1AD1M energized from switchgear IAD1 4

3) Distribution panel 1AD11 energized from switchgear 1AD1

d. 4) Distribution panel 1AD12 energized from switchgear i 1AD1

'I b) System C

1) Switchgear ICD 1 energized from battery 1C018

2) MCC ICD 1M energized from switchgear 1CD1

3) Distribution panel ICD 11 energized from switchgear 1CD1 f 2. Associated with Train B a) System B

1) Switchgear 1811 energized from battery 1801B

2) MCC IBD1M energized from switchgear 1801

3) Distribution panel 18011 energized from switchgear 18D1

!i 4) Distribution panel 18012 energized from switchgear i 1801 5 b) System D

1) Switchgear 1001 energized from battery 10018

2) Distribution panel 10011 energized from switchgear 1D01 Q APPLICABILITY: MODES 1, 2, 3, and 4.

ACTION:

g. /M

, a. With one of the required h of A.C. emergency busses not fully energized, reenergize the d4e6e6en 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 5 the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

halicr hank or Wer/er malatendnce,

I" f^ OY assodated swikhgear "Two inverters'may be disconne ed from their Oste =4es for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months as necessary, for the purpose of crfer ' ; n_ n di;' ; 9--;; :n th:' aseoefuted'

"'^

_., _ J provided: (1) their,y%s4.buseee- are energized 6 and (2) the v66a4 ,

gere/5 busses

.,___1..2 associated with the other l iattery bank ,7; x; ;' ;d fix th;ir gWr/td /lEe

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{ l ONSITE POWER DISTRIBUTION LIMITING CONDITION FOR OPERATION

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iI

.I ACTION (Continued) /2D s'0//

b. With one#A.C. vital bus either not energized from its associated inverter, or with the inverter not connected to its associated M-du//cMgw h (1) reenergize the A.C. vital bus within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 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 ; and (2) reenergize the A.-C. .it:1juve/

]a h from its associated inverter connected to its associated D.C.

bus within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-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 , c.

f-125 Volf With onerD.C. bus not energized M ik -fn-3 it:,3 ec//idamer

!:td bettery L. .;s ,

- reenergize the " " ' - ' - - - -

" - ---- ' ^ " '-"--" ' - ' "i thi n 2 ho u rs or be in at le it 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 with the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . l SWll& car-or' ckrfdb v}janp1/75/ M/k SpecjRu/n&ider SURVEILLANCE REQUIREMENTS 4.8.3.1 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and indicated voltage on the busses.

i -

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l i.

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ONSITE POWER DISTRIBUTION

SHUT 00WN LIMITING COwnITION FOR OPERATION a

1 3.8.3.2 As a minimum, the following electrical bussesNshall be energized j' in the specified manner:

a. One train of A.C. emergency busses consisting of one 4160-volt switchgear, three 480-volt A.C. switchgear, and six 480-volt A.C. !

Motor Control Centers,

b. One train' of 120-volt'A.C. vital busses energized from their asso-ciated inverters connected to their respective D.C. busses, and

c. One train of 125-volt D.C. switchgear and associated distribution equipment energized from its associated battery bank.

APPLICABILITY MODES 5 and 6.

ACTION:

With any of the above required electrical busses not energized in the required

} manner, immediately suspend all operations involving CORE ALTERATIONS, positive

) reactivity changes, or movement of irradiated fuel, initiate corrective action j to energize the required electrical busses in the specified manner as soon as j possible, and within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , depressurize and vent the RCS through at least a a square inch vent.

I l

9.

SURVEILLANCE REQUIREMENTS f

) 4.8.3.2 The specified busses shall be determined energized in the required manner at least once per 7 days by verifying correct breaker alignment and q

indicated voltage on the busses.

?.

9 1

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i V0GTLE - UNIT 1 3/4 8-17 APR 2 4 M d

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t ELECTRICAL EQUIPMENT PROTECTIVE DEVICES CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT PROTECTIVE DEVICES AAID FEE. DER BREAKEF.i ;

TO I50L-ATiokJ TEAh!S FORMERS BETWEEM 480V CLMs LE Btt.tSES AklO NON CLAH__12 LIMITING CONDITION FOR OPERATION EGWPMEA)[

k 3.8.4.1 All containment penetration conductor overcurrent protective devices onclf given in Table 3.8-1 shall be OPERABLE.

(feccler break'ers /c /so/af tm APPLICABILITY: MODES 1, 2, 3, and 4. M 4rkte n Le/ deen +'So V ACTION-W eyym ent.

With one or more of the containment penetration conductor overcurrent protective device (s)[given in Table 3.8-1 inoperable:f-fr/c'eder Chelween '/80 Y Class JE kus.s brea/cer.t anc/raugemuy yo' iso /af/q/non

-e. -- " ;t., #;ut;;;.he i. ' 4}Nst:'" er f:4megitem efvrpov,rf-

-th; k: ^ it(:)zby- trippifig=thehated backup-cim. .M -. .hrr

< er r;-thst-or-menevin;; the-iWM ei c M br:r' cr rithifi I gg/.o / ace, ?! t r:, fri::: th: ;ff;;t;d ; jet =d-- c7c : .t ' .;;; ret,le,-

O//g7 M '; r"y th: b r hr;.civ 't irr h:r t i: tri;;:d er th; he,p;r-M# :th ;isc.it ;,, __.. . - rh:f ::t :rc---"* '- :t- - r ;:: i73-4 h /d4 4heeee84ee; -th: ;;;;hi:;; ef I;;:i::ti: 2. 0. 4 2 7; :: t ;p pl ic ;;, k I

]y 3/f8 j /8 t =:r=rr;rL f:;hn t :'r;;it: dich-in; their ixi.;. ci, _;i.-

untr:-tri;;;f;%ir hapetae-c47::it tr; .h;. ; r . ;r ., .r-r==: ,- :.

b. Be in at least HOT STANDBY within the next 6 haurs and in COLD l SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

)

SURVEILLANCE REQUIREMENTS 4.8.4.1 All containment penetration conductor overcurrent protective devices 4/k/ Oca'er-given in Table 3.8-1 shall be demonstrated OPERABLE: beteba his/4// / rem /eare b/weM 980 V u JE bene.a s

a. At least once per 18 months: and nog C,4n ft (gv/ pmegf.

13.8 KV

1) By verifying that the r fi r n'.^:;: D '; LC circuit breakers j are OPERABLE by selecting, on a rotating basis, at least 49(of i the circuit breakers W rh n T L i, and performing the ,

following:

a) A CHANNEL CALIBRATION of the associated protective relays, 1 b) An integrated system functional test which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits ;

function as designed, and l n

)

1 i 3

V0GTLE - UNIT 1 3/4 8-18 APR 24 5 1

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(

a. Restore the protective device or feeder breaker to OPERABL'E status or deenergize the circuits by racking out or removing the inoperable circuit breaker or protective device and tripping the associated backup

? circuit breaker within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the effected system or compo-nent inoperable, and verify the inoperable circuit breaker or protective device racked out or removed at least once per 31 days thereafter; the i provisions of Specification 3.0.4 are not applicable to overcurrent l

devices or feeder breakers in circuits which have their backup circuit breakers tripped, their inoperable circuit breaker racked out or removed, or

! b. Deenergize the circuits by racking out or removing the inoperable cir-l cuit breaker or protective device within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected system or component inoperable, and verify the inoperable circuit breaker or protective device racked out or removed at least once per 7 days thereafter; the provisions of Specification 3.0.4 are not applicable to

) overcurrent devices or feeder breakers in circuits which have their inoperable circuit breaker racked out, or removed, or

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ELECTRICAL EQUIPMENT PROTECTIVE DEVICES I

SURVEILLANCE REQUIREMENTS (Continued) f one For each circuit breaker found inoperable during these c) functional tests, an additional representative sample of at least tet f of eM the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type i h have been fanctionally tested. l By selecting and functionally testing a representative sample 2) of at least 10% of each type of lower voltage circuit breakers.

Circuit breakers selected for functional testing shall be j selected on a rotating basis. Testing of these circuit j breakers shall consist of injecting a currentf.t'. : ::::: l

,_p 7 n- - : ; trC';7 tip '

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L r:--t g ; _ _ at't 3 _;:.._.

,,_ . _ _dy _ _ . .. _c,i,7 3 1-;;;;;d. Circuit breakers found inoperable during functional testing shall be restored to )

E OPERABLE status prior to resuming operation. For each circuit '

? breaker found inoperable during these functional tests, an

additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be function-ally tested until no more failures are found or all circuit breakers of that type have been functionally tested; and

.: b. At least once per 60 months by subjecting each circuit breaker to an

] inspection and preventive maintenance in accordance with procedures

] { prepared in conjunction with its manufacturer's recommendations.

1 l

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ELECTRICAL EQUIPMENT PROTECTIVE DEVICES 64FiTV-RELATED APOTOR-OPERATED VALVES THERMAL OVERLOAD PROTECTION M BYPASS DEVICES 1

i LIMITING CONDITION FOR OPERATION f ~se&fy-refeteif mehr 9ecak) i 3.8.4.2 The thermal overload protection end bypass devices, 't:; 1.dth th

estem.aeee6ee of eachuvalve listed in Table 3.8-2 shall be PERABLE.

d APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE.

Si 1 ACTION:

i i With c: :r __ Ithe thermal overload protection anddee bypass devices q inoperable, declare the affected valve (s) inoperable and apply the appropriate q ACTION Statement (s) of the affected valve (s).

1 5 Yr 04tf Orte ofmW SQklZj-RldY 1919/DW oj]eg*g/4//g.

] SURVEILLANCE REQUIREMENTS /

~i 4.8.4.2 The above Nguired thermal overload protection and bypass devices j shall be demonstrated OPERABLE:

-; a. At least once per 18 months, by the performance of a TRIP ACTUATING 1 DEVICE OPERATIONAL TEST of the bypass circuitry for those thermal 1

overload devices which are either:

l 1. Continuously bypassed and temporarily placed'in force only when the valve motors are undergoing periodic or maintenance testing, or

2. Normally in force during plant operation and bypassed under accident conditions.

b. At least once per 18 months by the performance of a CHANNEL CALIBRATION f of a representative sample of at least 25% of:

I 1. All thermal overload devices which are not bypassed, such that

each non-bypassed device is calibrated at least once per 6 years.

'l 2. All thermal overload devices which are continuously bypassed and temporarily placed in force only when the valve motors are j undergoing periodic or maintenance testing, and thermal overload j devices normally in force and bypassed under accident conditions J such that each thermal overload is calibrated and each valve is

.j cycled through at least one complete cycle of full travel with a the motor-operator when the thermal overload is OPERABLE and not bypassed, at least once per 6 years.

1 fep/occ wMG Z<ued6 ye 3/9 g-zi a

I V0GTLE - UNIT 1 3/4 8-21

'd APR 2 41986 l j

( INSERT TO PAGE 3/4 8-M El 4.8.4.2 The above required thermal overload protection bypass devices shall be verified to be OPERABLE.

h a. Following maintenance on the valve motor starter, and

b. Following any periodic testing during which th'a thermal overload device was temporarily placed in force.

c. At least once per 18 months, during shutdowr..

l:

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s j 3/4.9 REFUELING OPERATIONS k

j 3/4.9.1 BORON CONCENTRATION i

, LIMITING CONDITION FOR OPERATION 1

J 3.9.1 The boron concentration of all filled portions of the Reactor Coolant i System and the refueling canal shall be maintained uniform and sufficient to j ensure that the more restrictive of the following reactivity conditions is met; J

either:

a

a. A K,ff of 0.95 or less, or l

] b. A boron concentration of greater than or equal to 2000 ppm.

APPLICABILITY: MODE 6.

d ACTION:

1 With the requirements of the above specification not satisfied, immediately )

suspend all operations involving CORE ALTERATIONS or positive reactivity i changes and initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm boron or its 4 equivalent until K is reduced to less than or equal to 0.95 or the boron i

concentrationisr$NoredtogreaterthanorequaltoJ20007 ppm,whicheveris the more restrictive, j SURVEILLANCE REQUIREMENTS .

o I

1 4.9.1.1 The more restrictive of the above two reactivity conditions shall be !

j determined prior to:

a. Removing or unbolting the reactor vessel head, and i

b. Withdrawal of any full-length control rod in excess of 3 feet from its fully inserted position within the reactor vessel.

4.9.1.2 The boron concentration of the Reactor Coolant System and the refueling j canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

i 3, 4.9.1.3 Valves 1208-U4-175, 1208-U4-177, 1208-U4-183, and 1208-U4-176 shall be verified closed and secured in position by mechanical stops :- i; 7;nceal Of cir r :!:: trim ;r:r at least once per 31 days.

i fl

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9 APR 24 h V0GTLE - UNIT 1 3/4 9-1 l

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REFUELING OPERATIONS

.i 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS

] LIMITING CONDITION FOR OPERATION I

l A

8l 3.9.7 Loads in excess of pounds shall be prohibited from travel over fuel assemblies in the storage pool.

s APPLICABILITY: With fuel assemblies in the storage pool.

k ACTION: , ///dMNhk

)

3 a. With the requirements of the above specification not satisfied, place the crane load in a safe condition.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

J

't k

i SURVEILLANCE REQUIREMENTS 4.9.7 Crane interlocks and physical stops which prevent crane travel with loads in excess of pounds over fuel assemblies shall be demonstrated OPERABLE within 7 days prior to crane use and at least once per 7 days thereafter during crane operation.

l b -

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4 1

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APR 24 Etib a

l V0GTLE - UNIT 1 3/4 9-7 i

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l REFUELING OPERATIONS 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION HIGH WATER LEVEL LIMITING CONDITION FOR OPERATION i

3.9.8.1 At least one residual heat removal (RHR) train shall be OPERABLE and

in operation.*

f APPLICABILITY: MODE 6,whenthewaterlevelabovethe$opofthereactor vessel flange is grbater than or equal to 23 feet.

ACTION:

With no RHR Seep OPERABLE and in operation, suspend all operations involving an increase in the reactor decay heat load or a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to i return the required RHR train to OPERABLE and operating status as soon as j possible. Close all containment penetrations providing direct access from

the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

? -

l SURVEILLANCE REQUIREMENTS i

4.9.8.1 At least one RHR train shall be verified in operation and iirculating reactor coolant at a flow rate of greater than or equal to [2800] gpm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

l i

l -

I

*The RHR train may be removed from operation for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of the reactor vessel i hot legs.

l

! APR 2 41986 I V0GTLE - UNIT 1 3/4 9-8 i

/

REFUELING OPERATIONS 3/4.9.10 WATER LEVEL - REACTOR VESSEL LIRITING CONDITION FOR OPERATION l 3.9.10\t1:: cst 23feetofwatershallbemaintainedoverthe'topofthe

/ '

3 l

reactor vessel flange. '

APPLICABILI During movement of fuel assemblies or cont 1 rods within the

, containment when either the fuel assemblies being moved'or the fuel assemblies l seated withire the r.eactor vessel are irradiated while'in MODE 6. !

' ACTION: /

/ ;

Withtherequirements,oftheabovespecificatioNnotsatisfied,suspendall operations involving mbyement of fuel assemblies or control rods within the t reactor vessel. / !

/ l 4

/ ,

,/ :

/ i

/ ,

m /

/

I

/

SURVEILLANCE REQUIREMENTS / l

'\

3 4.9.10 The water level shall be determined to be at least its minimum required

., depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior'to the start 'of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during movement of fuel assembifes or control rods.

A i

APR 141986 V0GTLE - UNIT 1 3/4 9-11'

\

REFUELING OPERATIONS

(

3/4.9.10 WATER LEVEL - REACTOR VESSEL FUEL ASSEMBLIES LIMITING CONDITION FOR OPERATION J

3.9.10.1 At least 23 feet of water shall be maintained over the top of the reactor vessel flange.

4 APPLICABILITY: During movement of fuel assemblies within the containment when .

y the fuel assemblies being moved are irradiated, j ACTION J

With the requirements of the above specification not satisfied, suspend all operations involving movement of fuel assemblies within the reactor vessel.

b SURVEILLANCE REQUIREMENTS 4.9.10.1 The water level shall be determined to be at least its minimum required 3 depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months there-( after during movement of fuel assemblies.

s it f

l V0GTLE - UNIT 1 3/4 9-11

REFUELING OPERATIONS WATER LEVEL- REACTOR VESSEL CONTROL RODS LIMITING CONDITION FOR OPERATION 3.9.10.2 At least 23 feet of water shall be maintained over the top of the

{a irradiated fuel assemblies within the reactor pressure vessel.

n APPLICABILITY: During movement of control rods within the reactor pressure

, vessel while in MODE 6.

ACTION:

With the requirement of the above specification not satisfied, suspend all operations involving movement of control rods within the pressure vessel.

SURVEILLANCE REQUIREMENTS 4.9.10.2 The water level shall be determined to be at least its minimum required depth within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during movement of control rods within the reactor vessel.

{

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I REFUELING OPERATIONS 3/4.9.11 WATER LEVEL - STORAGE POOL

)

j LIMITING CONDITION FOR OPERATION 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are in the storage pool.

ACTION:

4

a. With the requirements of the above specification not satisfied, I; suspend all movement of fuel assemblies and crane operations w6e cvu #1C L

j iuf-f vel pel lostitsM-the fed :terage-erees and restore the water level to within limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

f b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS B

i 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.

F s

L a

i

, V0GTLE - UNIT 1 3/4 9-12 APR 24 W

') .

L,.w .E- : G awc::O'< l.stuux. it.a.sh5hniSMS&Lik.=hNiM 3/4.10 SPECIAL TEST EXCEPTIONS

[ 3/4.10.1 SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 4

3.10.1 The SHUTOOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and SHUTDOWN MARGIN provided reactivity equivalent to at least the highest estimated control rod worth is available for trip insertion from OPERABLE control rod (s).

APPLICABILIlY: MODE 2.

ACTION:

c

a. With any 611-i r;th control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion, immediately initiate and continue boration at greater than or equal to 30 gpm of a solution containing greater than or equal to 7000 ppm I boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.

b. With all f ' b'...;th control rods fully inserted and the reactor subcritical by less than the above reactivity equivalent, immedi-ately initiate and continue boration at greater than or equal to

{ 30 gpm of a solution containing greater th.an or equal to 7000 ppm

baron or its equivalent until the SHUTDOWN MARGIN required by j Specification 3.1.1.1 is restored.

SURVEILLANCE REQUIREMENTS i

' 4.10.1.1 The position of each-fulbi...;th control rod either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

I t

4.10.1.2 Each doH W control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position i within _*' h n:s prior to reducing the SHUTDOWN MARGIN to less than the limits of Specifica ion 3.1.1.1.

7chy i

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V0GTLE - UNIT 1 3/4 10-1 i

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1 1 SPECIAL TEST EXCEPTIONS 3/4.10.3 PHYSICS TESTS a

] LIMITING CONDITION FOR OPERATION 1

1 a i 3.10.3 The limitations of Specifications 3.1.1.3, 3.1.1.4, 3.1.3.1, 3.1.3.5, i and 3.1.3.6 may be suspended during the performance of PHYSICS TESTS provided:

a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,

b. The Reactor Trip Setpoints on the OPERABLE Intermediate and Power J

Range channels are set at less than or equal to 25% of RATED THERMAL POWER, and c.

The Reactor is greater Coolant than System or equal to lowest operating loop temperature (Tavg) 3'F.

APPLICABILITY: MODE 2. WI j ACTION:

I j a. With the THERMAL POWER greater than 5% of RATED THERMAL POWER, j immediately open the Reactor trip breakers.

b.

With a React 13 less than c'CF, lant System restore T operating loop temperature (Tavg) a to within its limit within

' 15minutesorbeinatleastMOTSTANDSYwithinthenext 15 minutes.

SURVEILLANCE REQUIREMENTS 4

J 1

4.10.3.1 The THERMAL POWER shall be determined to be less than or equal to 5%

of RATED THERMAL POWER at least once per hour during PHYSICS TESTS.

4.10.3.2 Each Intermediate and Power Range channel shall be subjected to an ANALOG CHANNEL OPERATIONAL TEST within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months prior to initiating PHYSICS q TESTS.

1

] 4.10.3.3 The Reactor Coolant Systen temperature (T,yg) shall be determined to i be greater than or equal to fEu]'F at least once per 30 minutes during PHYSICS

j t

TESTS.

/51/

1

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I i ,

SPECIAL TEST EXCEPTIONS l 3/4.10.5 POSITION INDICATION SYSTEM - SHUTDOWN l LIMITING CONDITION FOR OPERATION i

1 1 l 3.10.5 The limitations of Specification 3.1.3.3 may be suspended during the performance of individual f Pt;th shutdown and control rod drop time

measurements provided;

a. Only one shutdown or control bank is withdrawn from the fully inserted 1 position at a time, and l

b. The rod position indicator is OPERABLE during the withdrawal of the

, rods.*

APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time measurements.

ACTION:

l With the Position Indication Systems inoperable or with more than one bank of p

rods withdrawn, immediately open the Reactor trip breakers.

I l

3 SURVEILLANCE REQUIREMENTS i

4.10.5 The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and at least once per 2 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during rod drop time measurements by verifying the Demand j Position Indication System and the Digital Rod Position Indication System agree:

a. Within 12 steps when the rods are stationary, and

b. Within 24 steps during rod motion. _

w "This requirement is not applicable during the initial calibration of the Digital Rod Position Indication System provided: (1) K is maintained lessthanorequalto0.95,and(2)onlyoneshutdownoff[ontrolrodbank j is withdrawn from the fully inserted position at one time.

) APR 2 4 km V0GTLE - UNIT 1 3/4 10-5 4

't

.. . .____,.m - .-

TABLE 4.11-1 (Centinuid)

TABLE NOTATIONS (Continued)

(3)The principal gamma emmiters for which the LLD specification applies include the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, 2n-65, Mo-99, Cs-134, Cs-137, Ce-141, and Ce-144. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Semiannual Radioactive Effluent Release Report pursuant to Specification 6.9.1.4 in the format cutlined in Regulato.ry Guide 1.21, Appendix B, Revision 1, June 1974.

(4)A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen that is representative of the liquids released.

(5)A continuous release is the afscharge of liquid wastes of a nondiscrete volume, e.g., from a volume of a system that has an input flow during the continuous release. A I(0)To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of flow of the effluent stream. Prior to analyses, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

nd This wi//,M1fcl,be is a CCA h comic /e/ec/ a ca/ihuna re/eene9n3/oryWMoe arforarf */s sece/g/wp /tsf. (ANest friere. tr a ,Or/>71 sty lo secmchwi re/ecs a re Mird/ 77sae su% i/ wiv/ ee% cu/enous rehsse w/,7 Ale,WPrved/ace w ee !y ccm z ,0.

, p c W e s a e,9 /es adsd no sc/M/y ogs,,

1 l

I 1

l V0GTLE - UNIT 1 3/4 11-4 APR 24 EN 1 l

l RADIOACTIVE EFFLUENTS -

DOSE LIMITING CONDITION FOR OPERATION 3.11.1.2 The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive i materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-1) shall be limited:

a. During any calendar quarter to less than or equal to 1.5 mrems to the whole body and to less than or equal to 5 arems to any organ, and

b. During any calendar year to less than or equal to 3 mrems to the whole body and to less than or equal to 10 arems to any organ.

APPLICABILITY: At all times.

ACTION:

a. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare !

and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) for exceeding the limit (s) and defines the corrective actions that have hen taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. -TMs-Special Deport -sheH-else-ineled::- 01) tM 44sults=of-sadiologisabana4yses-ofhdrinking War-source rand- v

13) the radioloipcal--impast=en-HafehedMaking w;ter-sg;;li:: with-

_r:;;:-d te the :quieements-of-40-CFR P;rt -141, SMe=DMak4@ tee--

-Act.*

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.1.2 Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days. _

V0GTLE - UNIT 1 3/4 11-5 APR 2 41986

RADIOACTIVE EFFLUENTS -

g,UIDRADWASTET9EATMENTSYSTEM LIMITING CONDITION FOR OPERATION 3.11.1.3 The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the projected doses due to the liquid effluent, from each unit, to UNRESTRICTED AREAS (see Figure 5.1-1) would exceed 4i4& mrem to the whole body or 0.2 mrem to any organ in a 31-day period. LO,06 APPLICABILITY: At all times.

ACTION:

a. With radioactive liquid waste being discharged without treatment and

. in excess of the above limits and any portion of the Liquid Radwaste Treatment System not in operation, prepare and submit to the Commis-sion within 30 days, pursuant to Specification 6.9.2, a Special Report that includes the following information:

1. Explanatien of why liquid radwaste was being discharged without !

treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action (s) taken to prevent a recurrence.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not app 11 cable.

SURVEILLANCE REQUIREMENTS 4.11.1.3.1 Doses due to liquid releases from each unit to UNRESTRICTED AREAS l shall be projected at leas.t once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized.

4.11.1.3.2 The installed Liquid Radwaste Tr<!atment System shall be considered OPERABLE by meeting Specifications 3.11.1.1 and 3.11.1.2.

V0GTLE - UNIT 1 3/4 11-6 M 24 W

i.wi=kauwx. .-...&...;.a -w. -. .- _ u: i b.a u 1 +.1 i :

5 i .

RADICACTIVE EFFLUENTS

[ LIQUID HOLOUP TANKS i

LIMITING CONDITION FOR OPERATION l

3.11.1.4 The quantity of radioactive material contained in each outside f temporary tank shall be limited to less than or equal to 10 Curies, excluding j tritium and dissolved or entrained noble gase @

APPLICABILITY: At all times.

i ACTION:

oa}nde le9R04)

a. With the quantity of radioactive material in any of the i .: -i bt d /

tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank

contents to within the limit, and describe the events leading to this condition *in the next Semiannual Radioactive Eff'uent Release Report, pursuant to Specification 6.9.1.4.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS f

/ 6Hher 4.11.1.4 The quantity of adioactive material contained in each of the above j listed tanks shall be de rmined to be within the above limit by analyzing a

' representative sample of the tank's contents at least once per 7 days wh.en I

radioactive materials are being added to the tanV or eac// /4/c// d/ 2 mdisac}Ne MahWn/ frip^ b ih a&I/ Hon b 6e M i

t 3.

k

)

r.

1 l

i k

I i

1 il l V0GTLE - UNIT 1 3/4 11-7 APR 2 4 W6 a

. , + = , y, ...3 yw.. *

, . .. _.__ _ :. ._ m - ,2 m _ - -

.. . . :_ a e n . a -. e _,

. q,

f. ~

[:

9 TABLE 4.11-2 8 RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Ir

'~f E-m p

MINIMUM LOWER LIMIT OF E SAMPLING ANALYSIS TYPE OF DETECTION (LLD)(1) {.

Z GASEOUS RELEASE TYPE FREQUENCY FREQUENCY ACTIVITY ANALYSIS (pCi/ml) h

1. Waste Gas Decay P P Tank Each Tank Each Tank Priricipal Gamma Emitters (2) 1x10 4 Grab Sample -

p i

2. Containment Purge P P ??

24" or 14" Each PURGE (3) Each PURGE (3) Principal Gama Emitters (2) 1x10 4 {'

Grab Sample [

M H-3 (oxide) Ix10 8 I.

3. a. Plant Vent M(3),(4),(5) Principal Gamma Emitters (2) 1x10 4 I R Grab Sample M(3)

H-3 (oxide) 1x10 8 j h b. Condenser Air M(5) Principal Gamma Emitters (2) 1x10 4 i

Ejector & Grab Sample ;

Steam Exhaust H-3 (oxide) 1x10 8 .h

c. Solidification M PrincipaT Gamma Emitters (2) 1x10 4 c2 - E

[ > Bldg, Radwaste Grab Sample M l rea, Others '

4. All Release Types Continuous (6) y(7) I-131 1x10 12 (

as listed in 1., 2.,

Charcoal !

and 3. above 3 j, Continuous (6) y(7) Principal Gama Emitters (2) 1x10 11 Particulate f

> Sample !

E Continuous (6) M Gross Alpha 1x10 12 f

" Composite Par-ticulate Sample -

l' Continuous (6) Q Sr-89, Sr-90 , 1x10 11 Composite Par-ticulate Sample }I c

L.

l l:

l .

i TABLE 4.11-2 (Continued) .

TABLE NOTATIONS (Continued) ,

1 t

(2)The principal gamma emitters for which the LLO specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m,

, Xe-135, and Xe-138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, CS-137, Ce-141 and Ce-144 in Iodine and particulate releases. This list does not mean that only these nuclides are to be considered. Other gamma peaks that are identifiable, together with ,

those of the above nuclides, shall also be analyzed and reported in the l Semiannual Radioactive Effluent Release Report pursuant to Specification )

6.9.1.4 in the format outlined in Regulatory Guide 1.2, Appendix B, l

! Revision 1, June 1974. !

I (3) Sampling and analysis shall also be performed following shutdown, startup,

! or a THERMAL POWER change exceeding 15% of RATED THERMAL POWER within a

1-hourperiod.g i j II4)Tritiymgrabsamplesshallbetakenatleastonceper24hourswhenthe

refueling canal is flooded.

(5) Tritium grab samples shall be taken at least once per 7 days from the ventilation exhaust from the spent fuel pool area, whenever spent fuel is i in the spent fuel pool.

) I6)The ratio of the sample flow rate to the sampled stream flow rate shall be

known for the time period covered by each dose or dose rate calculation

! made in accordance with Specifications 3.11.2.1, 3.11.2.2, and 3.11.2.3.

I ) Samples shall be changed at least once per 7 days and analyses shall be 4

completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing, or after removal from sampler.

Sampling shall also be performed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for at least 1 7 days following each shutdown, startup, or THERMAL POWER change exceeding i

15% of RATED THERMAL POWER within a 1-hour period and analyses shall be l completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of changing. When samples collected for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months

are analyzed, the corresponding LLDs may be increased by a factor of 10.

l This requirement does not apply if: (1) analysis shows that the DOSE

, EQUIVALENT I-131 concentration in the reactor coolant has not increased more than a factor of 3; and (2) the noble gas monitor shows that j effluent activity has not increased more than a factor of 3.

T/1/s apu&tm od cht got ag/ if (f) wM/ sLC 9 3boult $17f ///e, l 0"E 8guihkW Z-/3/ cmeffa//;q m Mir7% m n, dos / d

{ 0 /xs it6/ AlCWdW i'lm >M7/1 416cfor cf,5j //yl(g)

}//c inMe pu m m hrsbws Nd ef/w;/ g.!p,yp

/yr ryf arcprm f mdk Aari a Nchrof J.

l V0GTLE - UNIT 1 3/4 11-11 gpg 34 5

)

4

_ _ _ _ _ ___ _ _ _-_ E _ _._

RADI0 ACTIVE EFFLUENTS '

GAS DECAY TANKS LIMITING CONDITION FOR OPERATION 3.11.2.6 The quantity of radioactivity contair. J in ed h gas decay tank shall be limited to less than or equal to 2.0 x 105 Curies of noble gases (considereo as Xe-133 equivalent).

APPLICABILITY: At all times.

ACTION:

a. With the quantity of radioactive material in any gas decay tank exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Semiannual Radioactive Effluent Release Report, pursuant to Specification 6.9.1.4.

b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.11.2.6.lThe quantity of radioactive material contained in each gas decay tank shall be determined to be within the above limit at least once per M7dgs hours when radioactive materials are4eing added to the tan //o /& haf /

/we hen of;nMy nefnew ') days <

% //, Z G 2 . Tit & not/ cfccs/%ec/ivyory&e/4,yw' c (pj9

/ Wrandschr msfers/ cm/anid&1 esc & wr/e Jle guss///y/asf ska// be oWerninid /c & w%Gai nakse ggsdaspi hiif af /en/ onceper 24 bed a/ ben iedoache iwfends han han afM /s Me fank nif/fefreeaw Ef AoWL V0GTLE - UNIT 1 3/4 11-16 APR 2 4 m

II _ :u. < _ .. . .__ w m ._ . m.a w i.w a .c.a .u....a u:* M:c.:sz. u E M n:.a + 2.2..

)l j 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING -

l 1

3/4.12.1 MONITORING PROGRAM N

Jg LIMITING CONDITION FOR OPERATION t

j 3.12.1 The Radiological Environmental Monitoring Program shall be conducted j as specified in Table 3.12-1.

1 j APPLICABILITY: At all times.

i

] ACTION:

1 .

1 a. With the Radiological Environmental Monitoring Program not being j conducted as specified in Table 3.12-1, prepare and submit to the

Commission, in the Annual Radiological Environmental Surveillance i Report required by Specification 6.9.1.3, a description of the

] reasons for not conducting the program as required and the plans for -

, preventing a recu r nce.

! ccn Hrme

b. With the d evel of radioactivity as the resul1. of plant effluents in an environmental sampling medium at a specified location exceeding the reporting levels of Table 3.12-2 when averaged over any calendar quarter, prepare and submit to the Commission within 30 days, pursuant to Specification 6.9.2, a Special Report that identifies the cause(s) i' for exceeding the limit (s) and definer the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose

to a MEMBER OF THE PUBLIC is less than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, or 3.11.2.3. When more than j

one of the radionuclides in Table 3.12-2 are detected in the sampling 1

medium, this report shall be submitted if:

l concentration (1) concentration (2) l t reporting level (1) + reporting level (2) + .. 1 1.0 I

When radionuclides other than those in Table 3.12-2 are detected and f

are the result of plant effluents, this report shall be submitted if i the potential annual dose

to a MEMBER OF THE PUBLIC from all radio-i nuclides is equal to or greater than the calendar year limits of

. Specification 3.11.1.2, 3.11.2.2, or 3.11.2.3. This report is not I required if the measured level of radioactivity was not the result I

of plant effluents; however, in such an event, the condition shall j be reported and described in the Annual Radiological Environmental Surveillance Report required by Specification 6.9.1.3.

I i

The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

reana 16Me ona ins /, adyItab, wnew s~yA tr'e,su//; a f ./Ar w,;/y,s,cAnj w W i f

%)

s/f4// beemno maybe cffsir CeyleNc/ 9l n/af/>/c @ Y:p fylh cW r/0!Y.

/ @S kJ/ l buf irl My Mjetti/4/s) .50 c/a J.

1 APR 2 41986 l V0GTLE - UNIT 1 3/4 12-1 '

... -. _ . _ .Ia.I i....i.~. .A _.I.i &._ . _..~. m.__ s a.v_ s.a..;u._1 aLaaa..u.;.Lwscum. 0:.mn . '

RADIOLOGICAL ENVIRONMENTAL MONITORING .

LIMITING CONDITION FOR OPERATION l ACTION (Continued) l

c. With milk orve#;qrkJim

.1: ;fy ;;g teleasamples unavailable from one or l more of the sample locations required by Table 3.12-1, identify I specific locations for obtaining replacement samples and add them I within 30 days to the Radiological Environmental Monitoring Program I given in the 00CM. The specific locations from which samples were

! unavailable may then be deleted from the monitoring program. Pursuant to Specification 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflecting the new location (s) with cupporting information identifying the cause of l the unavailability of samples and justifying the selection of the new location (s) for obtaining samples.

d. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

I SURVEILLANCE REQUIREMENTS l

4.12.1 The radiological environmental monitoring samples shall be collected i pursuant to Table 3.12-1 trom the specific locations given in the table and

] figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 3.12-1 and the detection capabilities required by Table 4.12-1.

l l'

1 I

l l

}

l V0GTLE - UNIT 1 3/4 12-2

TABLE 3.12-1 h RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE E EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY Z AN0/OR SAMPLE SAMPLE LOCATIONS II) COLLECTION FREQUENCY OF ANALYSIS

1. Direct Radiation (2) Thirty lsix routine monitoring Quarterly. Gamma dose quarterly.

stations -(0"10"?O) either with ,

two or more dosimeters or with one instrument for measuring and recording dose rate continuously, placed as follows:

An inner ring of stations, one in

each meteorological sector in the general area of the SITE BOUNDARY R

(""1 ""15);

fo U ne in cI.

Anouterringofstations[in.

each meteorological s Jecr the 0 i: "-E .(3 to @ milef range from the site (""17 ""22); and The balance of the stations

-499959RW)- to be placed in special interest areas such

as population centers, nearby residences, schools, and in one ?

or two areas to serve as control stations. .

=

te A

B D'l

] i I.

! TABLE 3.12-1 (Continued)

, 8 RADIOLOGICAL ENVIR0fMENTAL MONITORING PROGRAM i

i NUBSER OF

) REPRESENTATIVE i E EXPOSURE PATHWAY SAMPLES AND SAMPLING AND TYPE AND FREQUENCY j Q AND/OR SAMPLE SAMPLE LOCATIONS (3) COLLECTION FREQUENCY OF ANALYSIS

" 2. Airborne

}

Radiolodine and Samples from five locations Continuous sampler oper- Radioiodine Cannister:

j Particulates ation with. sample collec- I-131 analysis weekly.

i tion weekly, or more

! Three samples -MlWW) from frequently if required by i close to the three SITE dust loading. Particulate Sampler:

) BOUNDARY locations, in Gross beta radioactivity different sectors; e H he analysis following u _m _ . __i...i.._2 - - - - '

i 1 ~ ; _ z_- ---- ; a:

. i s ' . filter change;(3) and g R*

--- - '- ----' - ' ' ' ' gamma isotopic analysis

of composite (by M One sample from the vicinity location) quarter 1:

4 of a community having the

] highest calculated annual average groundlevel D/Q; and ,

4 One sample M from a control location, as for example M - w of wf u b/f3ff Ced/er of -iMMel10 to 20 mile $' distant and in the least prevalent wind direction.

3. Waterborne ,

Gamma isotopic analysisI4) 1

a. Surface (5) One sample upstream -(Wett. Composite sample over One sample downstream (Welt. 1-month period.(O [ . C

,,a ua erly

% t. ::=:.: L-;1:^ ";; =; :: '= :; ;;;; ^=.t::;,. T-- : i=%;; :.:. -

' := ;=, =;T -dy-;f=unety=te=w -- --

tritiemmysim-w:y.

7 _ , ,_ _ . _ M+}-

- 7 .

I

, i e

TABLE 3.12-1 (Continued) .g gg7 ,gg RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM [ M N 1 M4 /rW/#7M

?

plad NUMBER OF REPRESENTATIVE E EXPOSURE PATHWAY SAMPLES AND gy) SAMPLING AND TYPE AND FREQUENCY Z AN0/0R SAMPLE SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS

" 3. Waterborne (Continued) -

3,3lje

\ ho af bg. Orinking 4me samples d each of one to Composite sampleIover I-131 analysis on each three (" 1- M ) of the nearest p (6) *"

- ::!t when the dose

/

/ M wate:V:r;;?' : that could be affected by its discharge.

~]'

_ y n s is er-calculated for the con-sumption of the water formed; monthly com-7"'" af posite otherwis g. afd is greater than'1 mrem h samplesf-cen- a control location Ode 4). f/46 Sam / /e- of' //n/54ec/

ypp ay gjgf ,4ffy -

per year N. Composite f

for gross beta and gamma

-frea/med$ /Mt eWery isotopic analyses (4) g st/ee/<5 or mcn//ff ,f monthly. Composite for R. P/

//'f'#'j*

. g tritium analysis quarterly.

i ro E

Cf. Sediment One sample from downstream area Semiannually. Gamma isotopic analysis I4)

from with existing or potential semiannually.

Shoreline recreational value M .

4. Ingestion

a. Milk Samples from milking animals Semimonthly when Gamma isotopicI4) d i ree locations (!:1 :.;;- 'r!! :r: er p::t =-; I-131 analysis semi-O Sn7elt withi?6-4em- distance having the :::th!y at :tkr .i . ;.- monthly *-- - ' '-

highest dose potential. If cre er p::ture; :: thly there are none, then one at :th ; t!a ;.

sample from milking animals in each of three areas *

(?:! - : .2; between ee. S fo e/w/ Sm//cf

> 4-4m-distant where doses E are calculated to b reater se than 1 arem per yr. Oneg

" samp1e irom mi1 king animals at a control location g M to miles ch/et er 4 7r/qld m'cnabb / 7 in I . ._ ,a u 21_ .g ;,e , _ . u _

_!:::t pr:nk .t wind directioha/%5 serf ///4/e/7Co.

~

t TABLE 3.12-1 (Continued) 8 RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM N

~

NUMBER OF REPRESENTATIVE E EXPOSURE PATHWAY SAMPLES AND Iy)

SAMPLING AND TYPE AND FREQUENCY Z AND/0R SAMPLE SAMPLE LOCATIONS COLLECTION FREQUENCY OF ANALYSIS w

4. Ingestion (Continued A lM f an Gamma isotopic analysis f4)

b. Fish -and A One sample of commercially Sample in season, or Irrerte- and recreationally important semiannually if they on edible portions.

bretes- species in vicinity of plant are not seasonal.

ll NW one ofge h),%h*[9*'T***a,,f Ib-)

04* b ous cic> "f I'c/4') A0ne sample of w species in o/p/aer/ da ' fjc. y areas not influenced by plant d scharge-(:L:: Ib_b y,g ,., 3 g,y, ,,,,,,

) c. Feed -One-sample-of-each-principal- -At-thof MrveMN. Gamma isotopic analyses I4) g Predeets -class-of- feed-products-from- on edible portion.

'? C,m55 o(

LeAq ' Iab -any-area-that-is-irrigeted-by-water-in-which-liquid-

-plant,-wastes-have bee.. -

7 4 g, -discharged (!:1 !;_) .

o15//< /oca//NJ ded/ /42 -Samples-of-three-different- Monthly during Gamma isotopicI4) and I-131 5/7E 60u4/&/f[ /4 -kinds--of bre d -leaf vegeta- growing season. analysis.

c/jffueer/ Sec-/ ors. -t4 n grie r. nearest. each-of-

-two-different offsttv-toca-

-tions-of-highest-predicted-annual-average greur.d-level B/Q-if-mHk-sampling is not-performed '!; 2 --Ell).

- 4)ne-sample-ef-each-ef-the- Monthly during Gamma isotopic (4) nd I-I3I Ulf 5#'7 5-M* 4 -simHar-broad-leaf-vegeta- growing season. analysis.

2 co,ffa/ foar/ de7 of- / -t4cn gr wn 15-to 30 h dis-E ageu/ f5 gu7u tanHn-the-least-prevalent a j wind-direction ?f = ilk-sam-

cl6f aer # - p H6g is not perfor.cd ':c20 ~

h

p EacA syg lacasnm wil/ he chsgrWe/ by 4nurnbsr, ispirte , WM ob /4&/

f a pohr /ma/s*yy TABLE 3.12-1 (Continued) t,, /,scerf //Jec4/ero/

Me /~o TABLE NOTATIONS (1) Specific parameters of distance and direction sector from f th c:nt:M in:

" -- reactoci and additional description where pertinent, shall be pro-vided for each and every sample location in Table 3.12-1 in a table and figure (s)ln the ODCG Refer to NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants," October 1978, and to Radiological Assessment Branch Technical Position, Revision 1, November 1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous conditions, seasonal unavailability, and malfunction of-ew W

-eeMe sampling equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environ-mental Surveillance Report pursuant to Specification 6.9.1.3. It is recog-nized that, at times, it may not be possible or practicable to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be ,

chosen for the particular pathway in question and appropriate substitutions e,j tam /-

made within 30 days in the Radiological Environmental Monitoring Program pi//

given in the ODCM. Pursuant to Specification 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change ' 1 in the ODCM including a revised figure (s) and table for the ODCM reflect-ing the new location (s)gwith supporting information identifying the cause of the unavailability of samples for the pathway and justifying the selec-tion of the new locatiop(s) for obtaining samples j or//y afant/kJ////

f d/

jiray suikble rieu /O Wfiins.

(2) One or more instruments, such as a pressurized fon chamber, for measuring and recording dose rate continuously may be used in place of, or in addi-tion to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLJ) is considered to be one phosphor; two or more phosphors in a packet are considered as two or more dosimeters.

Film badges shall not be used as dosimeters for measuring direct radiation.

(3) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

V0GTLE - UNIT 1 3/4 12-7: ggg

m , c.. n - -

3 TABLE 3.12-1 (Continued) .

TABLE NOTATIONS (Continued) l (4) Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

(5) The " upstream sample" shall be taken at a distance beyond significant l influence of the discharge. The " downstream" sample shall be taken in an l area beyond but near the mixing zone.

(6) ^ :::;::it: :r 71: i: ::: in d ich th: ;r :: t ity-(* Mque &)-ef-14qu i d-s ampl ed-f: pr;;;rti ::1 te th: ;r::tity of fleeing 'izeid : d '- dich th: ::thod - l

--of-sampling pley:d r::elte in : :p::in:r th:t-1: 7:presentat4ve-of-the - )

-14 quid-f4ew,-In-thi: pr;;ramdompositesamplealiquotsshallbecollected at time intervals that are very short (e.g. , hourly) rel Ative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample. .

)

47 )-Groundwater-semplesn hen-be-ta ken-when-this-source-is-tapped-for-dri nki ng=. )

r irrigatiempurposes-in-areas-where-the-hydraulic gradicat er recharge

-properties-ar: :uit el: for contamination e

}- The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM. 6

/

(9) If harvest occurs more than once a year, sampling shall be performe during each discrete harvest. If harvest occurs continuously, samplin i

~(' , shall be monthly. Attention shall be paid to including samples of tuberous and root food products, f x

(8) /V milkuyr ami>14/ is a cm or wp'frduc, m ,/ f 5 , b u m s. ,

Cmso<>p/iy,,

l

($ 5 ymm4 i loh}s/c qi7s si3 h 17d/ M 'Y'W 6'10f1 lD W!/ -Nt /.sutr Lii,,y' cir $c/ech &7 /de T-131, n 5yata/t an/pn Air ]~ 131 W M be p err % e/l V0GTLE - UNIT 1 3/4 12-8 APR S 41986

j TABLE 3.12-2 i y REPORTING LEVELS FOR RADI0 ACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES

[ REPORTING LEVELS E

WATER AIRBORNE PARTICULATE FISH MILK F000 PRODUCYS

" (pCi/1) OR GASES (pCi/m3 ) (pCi/kg, wet) (pCi/1) (pCi/kg, wet)

ANALYSIS H-3 20,000*

. Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000

{

M Zn-65 300 20,000 S

Z -N -95 400 Nb - S @

I-131 2 0.9 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,0GO 70 2,000 lYO 200 30 0 a-140 200 MNZ

For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/1 may be used. .

N

=

to A ,

TABLE 4.12-1 DETECTION CAPABILITIES FOR ENVIRONMENTAL SAMPLE ANALYSIS (1) (2)

LOWER LIMIT OF DETECTION (LLD)(3) h I blYT 7)

'!ATER

. AIRBORNE PARTICULATE FISH MILK \F4MH&4#994 NIM SEDIMENT

" (pC1/1) OR GASES (pCi/m3 ) (pCi/kg, wet) (pCi/1) (pCi/kg, wet) (pCi/kg, d y)

ANALYSIS Gross Beta 4 0.01 H-3 2000*

Mn-54 15 130 Fe-59 30 260 R

Co-58 15 130 Co-6D 15 130 U Zn-65 30 260

} O j

j["5

I-131

[5 1g 0.07 1 60 Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 ~

ch-/10 60 GO

( Q (a-140 15 & % 15

If no drinking water pathway exists, a value of 3000 pCi/1 may be used.

M Tr tw cln/7b>y c+Ycrf4M? eKisf.r , a &/v't 0/ /S g lll />n & ZrJW -

=

a

.~ . ._

TABLE 4.12-1 (Continued) .

TABLE NOTATIONS as pian / e/T/nt!5 bsenuclidesaretobeconsidered.

(1)Thislistdoesnotmeanthatonly?togetherwiththoseoftheabove Other peaks that~are identifiable nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Surveillance Report pursuant to Specification 6.9.1.3.

(2) Required detection capabilities for thermoluminescent dosireters used for environmental measurements shall be in accordance with the recommenda-tions of Regulatory Guide 4.13.

(3)The LLD is defined, for purposes of these specifications, as the smallest concentration of. radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which mty include radiochemical separatton:

4.66 s D LLD = ,

E -

V -

2.22 -

Y - exp(-Aat)

Where:

LLD = the "a prior!" lower limit of detection (picoCuries per unit mass or volune),

s b

= the standard deviation of the background counting rate or of the counting 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 = the number of disintegrations per minute per picocurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec.2),and At = the elapsed time between sample collection, or end of the sample collection period, and time of counting (sec).

Typical values of E, V, Y, and at should be used in the calculation.

4 V0GTLE , UNIT 1 3/4 12-11 ffR24 W

RADIOLOGICAL ENVIRONMENTAL MONITORING .

3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION t

3.12.2 A Land Use distance .of.&4srd)Censusgs' hall be 5 miles-) the location conducted in each of the 16 and shall identify within a meteorological sector garden of *tfdf the nearest greater than 48am8milk(500 animal, the nearest ftz)jproduct g br residence, and the nearestd lea ici}hin 4be kManall kWdr^ llanf May k fr No% $HJ S UiV

APPLICABILITY: At all times.

ACTION:

a. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in Specification 4.11.2.3, pursuant to Specifica-tion 6.9.1.4, identify the new location (s) in the next Semiannual Radioactive Effluent Release Report. /

//.5sm/ // an 4 /4//04/c

b. With a Land Use Census identifying a ocation(s) that yields a calculated dose or dose commitment la the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with ecification 3.12.1, add the new location (s) within 30 days to the adiological Environmental Moni-toring Program given in the ODCMj The sampling location (s), exclud-ing the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after [0ctober 31] of the year in which this Land Use Census was conducted. Pursuant to Specification 6.14, submit in the next Semiannual Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new location (s) with informa-tion supporting the change in sampling locations.

c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

l l

V0GTLE - UNIT 1 3/4 12-13 APR 2 41986 c , .- -

- n.

, 4, y;. v'e ~ -

RADIOLOGICAL ENVIRONMENiAL MONITORING ((j-/ [ / uv/r SURVEILLANCE REQUIREMENTS /

4.12.2 The Land Use Census shall be conducted uring the growing season at l 1 east once per 12 months using that informatio that will provide tr.: i::t good l results, such as by a door-to-door survey, : ri:1 :.rvey,--or by consulting l local agriculture authoritiesg The results of the Land Use Census shall be 1 included in the Annual RadioTogical Environmental Surveillance Report pursuant ,

to Specification 6.9.1.3.

y b p,ns com bh7qfiG1 Of lNJP IUCl bib as lfcagible.

e V0GTLE - UNIT 1 3/4 12-14 APR 24 m

~

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7 mggg.4 g433-7 ,. - - - -- ]

RADIOLOGICAL ENVIRONMENTAL MONITORING -

1 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 3.12.3 Analyses shall be performed on all radioactive materials, supplied as part of an Interlaboratory Comparison Program that has been approved by the Commission,thatcorrespondtosamplesgrequiredbyTable3.12-1.

APPLICABILITY: At all times. [

ACTION:

a. With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Anriual Radiological Environmental Surveillance Report pursuant to Specification 6.9.1.3.

b. The provisions of Spect'ications 3.0.3 and 3.0.4 are not applicable.

SURVEILLANCE REQUIREMENTS 4.12.3 The Interlaboratory Comparison Program shall be described in the ODCM.

A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Surveillance Report pursuant to Specification 6.9.1.3.

V0GTLE - UNIT 1 . 3/4 12-15 M # 4 E86

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REACTIVITY CONTROL SYSTEMS ,

BASES 1

1 1 MODERATOR TEMPERATURE COEFFICIENT (Continued)

! involved subtracting the incremental change in the MDC associated with a core

] condition of all rods inserted (most positive MDC) to an all rods withdrawn J condition and, a conversion for the rate of change of moderator density with j temperature at RATED THERMAL POWER conditions. This value of the MDC was then j transformed into the limiting MTC value -3.9 x 10 4 Ak/k/*F. The MTC value

] of -3.0 x 10 4 ok/k/*F represents a conservative value (with corrections for i burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -3.9 x 10 4 Ak/k/*F.

i The Surveillance Requirements for measurement of the MTC at the beginning and near the end of the fuel cycle are adequate to confirm that the MTC remains :

within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 HINIMUM TEMPERATURE FOR CRITICALITY 51.

f This specification ensures that the reactor will not be made ritical

.. with the Reactor Coolant System average temperature less than-ES+1-) F. This l

limitation is required to ensure: (1) the moderator temperature coefficient l

1s within it analyzed temperature range, (2) the trip instrumentation is within i its normal operating range, (3) the pressurizer is capable of being in an j OPERABLE status with a steam bubble, and (4) the reactor vessel is above its e minimum RT temperature.

HDT

~

R 3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is I' avail.ble during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, l (3) separate flow paths, (4) boric acid transfer pumps, and (5) an amergency 1

power suppi; from OPERABLE diesel generators.

With the RCS average temperature above 200*F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The i boration capability of either flow path is sufficient to provide a SHUTDOWN i

l i

V0GTLE - UNIT 1 B 3/4 1-2 l i APR 2 41996 1

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REACTIVITY CONTROL SYSTEMS ,

t6

.! BASES

'i l BORATION SYSTEMS (Continued)

MARGIN from expected oporating conditions of 1.3% Ak/k after xenon decay and cooldown to 200 F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions and requires

[5106] gallons of 7000 ppm borated water from the boric acid storage tanks or [52,622] ga lons.of 2000 ppm borated water from the refueling water storage tank (RWST). nsable. volwe.

With the RCS temperature below 200 F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity ccndition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The Mmitatiofr foc-awnaximoe,of-car setrifuge4-eharging. pump =to be-1 0^9?l".Maethe-Stervei41 encFRequkement- to iverify4Mrcha rg4ag-pumps-excep t-the r:';" ired 4PERABLExpump-te;be_ineperable=below=027&3*F providessassurance that+meseaditir peessure4ransient cambe-reli'eved4y-the-operatiefrof-e-l W e-PORVr The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1% Ak/k af ter xenon decay and cooldown from 200 F to 140*F. This condition requires either gallons f [7000] ppm borated water from the boric acid storage tanks or gallon of 2000 ppm borated water from the RWST. gfg gg The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 10.5 for the solution recirculated

. within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on i mechanical systems and components.

The OPERABILITY of one Boron Injection System during REFUELING ensures that this system is available for reactivity control while in MODE 6.

1 3/4.1.3 MOVABLE CONTROL ASSEMBLIES 1

1 The specifications of this section ensure that: (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) the potential effects of rod misalignment on associated accident analyses are limited. OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits. Verification that the Digital Rod Position

Indicator agrees with the demanded position within i 12 steps at 24, 48, 120, V0GTLE - UNIT 1 B 3/4 1-3 APR 241986 1

' ? n --.- . . - . .. . . . . .-

. . . . . - .a a..:. ... - . . i w , = w. x a. ;. .. a. . .:. u .. w. w . u. - w e - - .w.n w.

l 3/4.2 POWER DISTRIBUTION LIMITS .

BASES The specifications of this section provide assurance of fuel integrity j during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) j events by: (1) maintaining the minimum DNBR in the core greater than or equal i to 1.30 during normal operation and in short-term transients, and (2) limiting i the fission gas release, fuel pellet temperature, and cladding mechanical l

properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance i criteria limit of 2200'F 1s not exceeded.

I l The definitions of certain hot channel and peaking factors as used in j these specifications are as follows:

Fq (Z) Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the I average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods; Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of e

a Fh the integral of linear power along the rod with the highest integrated

power to the average rod power; and F

xy(Z) Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

a 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fq (Z) upper '

s bound envelope of 2.30 times the normalized axial peakSg factor is not exceeded i

{ during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions.

The t il- k.eth rods may be positioned within the core in accordance with i their respective insertion limits and should be inserted near their normal l position for steady-state operation at high power levels. The value of t.he target flux difference obtained under these conditions divided by the fraction of RATED THERMAL POWER is the target flux difference at RATED THERMAL POWER for the associated core burnup conditions. Target flux differences for other THERMAL POWER levels are obtained by multiplying the RATED THERMAL POWER value i by the appropriate fractional THERMAL POWER level. The periodic updating of I the target flux difference value is necessary to reflect core burnup I considerations.

h i

V0GTLE - UNIT 1 B 3/4 2-1 P l V .

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3/4.3 BASES INSTRUMENTATION 'W/ u"1 (yQ'/Jor) fet/t/fM <

l l 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES N '

ACTUATION SYSTEM INSTRUMENTATION The OPERA 81LITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint (2) the specified coincidence logic h = h t M , (3) sufficient redundancy is main-

. tained to permit a channel to be out-of-service for testing or maintenance and (4) sufficient, system functional capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall 3

reliability, redundancy, and diversity assumed available in the facility 1 design for the protection and mitigation of accident and transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses. The Surveillance Requirements speci-fled for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveil-lance tests performed at the minimum frequencies are sufficient to demonstrate I this capability. See Icerf & 8 3/4 S-f The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables io are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4. Opera-tion with Setpoints less conservative than the Trip Setpoint but within the j Allowable Value is acceptable since an allowance has been made in the safety 4 analysis to accommodate this error. An optional provision has been included

.i for determining the OPERABILITY of a channel when its Trip Setpoint is found j to exceed the Allowable Value. The methodology of this option utilizes the 4 "as measured" deviation from the specified calibration point for rack and sensor components in conjunction with a statistical combination of the other

, uncertainties of the instrumentation to measure the process variable and the

.' uncertainties in calibrating the instrumentation. In Equation 3.3-1, j Z + R S < TA, the interactive effects of the errors in the rack and the j sensor, and the "as measured" values of the errors are considered. Z, as 1 specified in Table 3.3-4, in percent span, is the statistical summation 1 of errors assumed in the analysis excluding those associated with the sensor.

and rack drift and the accuracy of their measurement. TA or Total Allowance

]b is the difference, in percent span, R or Rack Error is the "as measured"

^

deviation, in the percent span, for the affected channel from the specified ;

., Trip Setpoint. S or Sensor Error is either the "as measured" deviation of d

4 V0GTLE - UNIT 1 B 3/4 3-1 APR 241986 a

p

, , . - . .... , , - , ,..-.c n. ,, s ~,c., ,, 4 _ __

..-._ --a_

~

3 Insert to Page B 3/4 3-1

(

~

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System", and supplements to that report. Surveillance intervals and out of

) service times were determined based upon maintaining an appropriate level of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

b n

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INSTRUMENTATION ,

] BASES 1

i REMOTE SHUT 00WN SYSTEM (Continued) dirl 9 control, and ;r:r- cf crit xd- transfer switches ng[essary to eliminate effects

.! of the fire and allow operation of instrumentation,fcontrol x d ;:x:r circuits

.[ required to achieve and maintain a safe shutdown condition are independent of 1 areas where a fire could damage systems normally used to shut down the reactor.

1 This capability is consistent with General Design Criterion 3 and f;;xdh ".

OFR ht 50. CM48 9. 5./.

l

.{ 3/4.3.3.6 ACCIDENT MONITORING INSTRUMENTATION 1 .

The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor n, and assess these variables following an accident. This capability is consis-j tent with the recommendations of Regulatory Guide 1.97,-Asattsh4, "Instrumen-j tation for Light-Water-Cooled Nuclear Power Plants to Assess Plant Conditions

.]

During and Following an Accident," Revision 2, December 1980 and NUREG-0737,

" Clarification of TMI Action Plan Requirements," November 1980.

3/4.3.3.7 CHLORINE DETECTION SYSTEMS The OPERABILITY of the Chlorine Detection Systems ensures that sufficient

, capability is available to promptly detect and initiate protective action in the event of an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, rem :irr t " Protection of Nuclear Power Plant Control room Operators Against an Accidental Chlorine Release," February C .

3/4.3.3.8 FIRE DETECTION INSTRUMENTATION d The OPERABILITY of the fire detection instrumentation ensures that both

.i adequate warning capability is available for prompt detection of fires and that i Fire Suppression Systems, that are actuated by fire detectors, will discharge

! extinguishing agents in a timely manner. Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility Fire Protection Program.

.j Fire detectors that are used to actuate Fire Suppression Systems represent

.j a more critically important component of a plant s Fire Protection Program

_i than detectors that are installed solely for early fire warning and notifica-l} tion. Consequently, the minimum number of OPERABLE fire detectors must be

1 greater.

The loss of detection capability for Fire Suppression Systems, actuated by fire detectors, represents a significant degradation of fire protection for 1*

1 i

V0GTLE - UNIT 1 B 3/4 3-5 NE

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A Ti BASES d

N 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION l 5 '

The plant is designed to operate with all reactor coolant loops in operation and maintairi DNBR above 1.30 during all normal operations and antici-j pated transients. In MODES 1 and 2 with one reactor coolant loop not in !

J operation this specification requires that the plant be in at least HOT STANDBY l within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . {

. In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal &ccident can be prevented, i.e., by j opening the Reactor Trip System breakers.

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR train provides sufficient heat removal capability 3 for removing decay heat; but single failure considerations require that at y :/t=h: (either RHR or RCS) be OPERABLE.

. i least two-hh:is//, w ops hj In MODE 5 with reactor coolant loops not filled, a single RHR train i provides sufficient heat removal capability for removing decay heat; but single

'A failure considerations, and the unavailability of the steam generators as a t heat removing component, require that at least two RHR trains be OPERABLE. The

'l locking closed of the required valves in Mode 5 (with the loops not filled) j precludes the possibility of uncontrolled boron dilution of the filled portion of the Reactor Coolant System. This action prevents flow to the RCS of unborated water by closing flowpaths from sources of unborated water. These limitations are consistent with the initial conditions assumed for the boron dilution accident in the safety analysis.

The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System. The reactivity change rate associated with boron reduction will, Li therefore, be within the capability of operator recognition and control.

1.1, d The restrictions on starting an RCP with one or more RCS cold legs less j than or equal to 350 F are provided to prevent RCS pressure transients, caused "1 by energy additions from the Secondary Coolant System, which could exceed the l limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by i restricting starting of the RCPs to when the secondary water temperature of

, each steam generator is less than 50 F above each of the RCS cold leg i temperatures. l G l I

i i

V0GTLE - UNIT 1 B 3/4 4-1 APR 2 41986 Y

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1, REACTOR COOLANT SYSTEM .

> BASES i SPECIFIC ACTIVITY (Continued) appropriately small fraction of 10 CFR Part 100 dose guideline values following a steam generator tube rupture accident in conjunction with an assumed steady-I f'M'{ state %eee4er-to-secondary steam generator leakage rate of 1 gpm. The values i for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative i

in that specific site parameters of the Vogtle site, such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

J j The ACTION statement permitting POWER OPERATION to continue for limited time periods with the reactor coolant's specific activity greater than 1 microcurie / gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which j may occur following changes in THERMAL POWER.

$ The sample analysis for determining the gr ss specific activity and E can

? exclude the radioindines because of the low rea'ctor coolant limit of 1 microcurie /

gram DOSE EQUIVALENT I-131, and because, if/he limit is exceeded, the radiciodine level is to be determined every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . If the gross specific activity level and radioiodine level in t,he reactor coolant were at their q limits, the radioiodine contribution woy'id be approximately 1%. In a release 1 of reactor coolant with a typical mixtyre of radioactivity, the actual radio-1 iodine contribution would probably be/about 20%. The exclusion of radio-nuclides with half-lives less than 44> minutes from these determinations has

[ been made for several reasons. The first consideration is the difficulty to 1 identify short-lived radionuclides in a sample that requires a significant

[ time to collect, transport, and analyze. The second consideration is the g

predictable delay time between the postulated release of radioactivity from L the reactor coolant to its release to the environment and transport to the The

g SITE BOUNDARY, which is relatable to at least 30 minutes decay time.

I choice ofWminutes for the half-life cutoff was made because of the nuclear l

characteristics of the typical reactor coolant radioactivity. The radionuclides a in the typical reactor coolant have half-lives of less than 4 minutes or half-lives of greater than 14 minutes, whiebeHows=a-diat,inctiaa Mtween4 he

-sadie.W "a e ^^"- -"-below,4-han-14fe410h- For these reasons the radionuclides that are excluded from consideration are expected to decay to very low levels before they could be transported from the reactor coolant to the SITE BOUNDARY under any accident condition.

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l REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Following the g'eneration of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as

. follows. A composite curve is constructed based on a point-by point comparison

, cf the steady-state and finite heatup rate data. At any given temperature, the i allowable pressure is taken to be the lesser of the three values taken from the l curves under consideration.

! The use of the composite curve is necessary to set conservative heatup j 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 i

to the outside and the pressure limit must at all times be based on analysis of t,he most critical criterion.

Next, the composite curves for the heatup rate data and the cooldown rate

j data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

Finally, the new 10CFR50 Appendix G Rule which addresses the metal temperature of the closure head flange and vessel flange regions is considered. This rule states that the minimum metal temperature of the closure flange regions should be at least 120*F higher than the limiting RT NDT f r these regions when the pressure

! exceeds 20 percent of the preservice hydrostatic test pressure (621 psig for

! Westinghouse Plants). For Vogtle Unit 1 the minimum temperature of the closure fisnge and vessel flange regions is 140 F, since the limiting RT NDT is 20 F (see Table B 3/4-4.1). The Vogtle Unit I heatup curve shown on Figure 3-4.2 is not impacted by the new 10CFR50 rule. However, the Vogtle Unit 1~cooldown curve shown in Figure 3-4.3 is impacted by the new 10CFR50 rule.

1 Although the pressurizer operates in temperature ranges above those for which

! there is reason for concern of nonductile failure, operating limits are provided i to assure compatibility of operation with the fatigue analysis performad in i accordance with the ASME Code requirements, t

COLD OVERPRESSURE PROTECTION SYSTEMS

! The OPERABILITY of two PORVs, the RHR suction relief valve or an RCS ' vent 1

opening of at least square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR g Part 50 when one or more of the RCS cold legs are less than or equal to 350 F.

j Either PORV has adequate relieving capability to protect the RCS from overpressur-t ization when the transient is limited to either: (1) the start of an idle RCP j with the secondary water temperature of the steam generator less than or equal

to *F above the RCS cold leg temperatures, or (2) the start ofn-;MLp p rd W injection into a water-solid RCS. %// /&<e. cdugag awhubsegaf ,

The Maximum Allowed PORV Setpoint for the Cold Overpressure Protection System (COPS) is derived by analysis which models the performance of the COPS assuming various mass input and heat input transients. Operation with a PORV Setpoint less i, than or equal to the maximum Setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV j V0GTLE - UNIT 1 B 3/4 4-15 gg

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I COLD OVERPRESSURE PROTECTION SYSTEMS (Continued)

I Setpoint which can occur as a result of time delays in signal processing and valve

'l opening, instrument uncertainties, and single failure. To ensure that mass and 7j heat input transients more severe than those assumed cannot occur, Technical Specifications require lockcut of all be6-ene safety injection pumps:d :' ht e r =. trit + M rg W ; z;, while in MODES 4, 5, and 6 with the reactor vessel head installed and disallow start of an RCP if secondary temperature is more than 50*F above primary temperature.

The Maximum Allowed PORV Setpoint for the COPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance

, specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance t with the schedule in Table 4.4-5.

'.T 3/4.4.10 STRUCTURAL INTEGRITY

{ {i i The inservice inspection and testing programs for ASME Code Class 1, 2, j and 3 components ensure that the structural integrity and operational readine.is of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by

'j 10 CFR 50.55a(g) except where specific written relief has been granted by 9 the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, Edition and Addenda through the 1970 Winter Addenda.

3/4.4.11 REACTOR COOLANT SYSTEM VENTS U

Reactor Coolant System vents are provided to/ exhaust noncondensible gases and/or steam from the Reactor Coolant System tliat could inhibit natural circulation core cool,ing. The OPERABILITY ofC east one Reactor Coolant System ventpathfromtheJteactorvesselheadf,'ensuresthatthecapabilityexiststo perform this function.

j The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply, or control system does not

-t prevent isolation of the vent path.

y l The function, capabilities, and testing requirements of the Reactor Coolant 4

System vents ara consistent with the reqeirements of Item II.B.1 of NUREG-0737,

" Clarification of TMI Action Plan Requirements," November 1980.

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ECCSSUBSYSTEMS(ContinuedQ/ / ,.,gaf/c The limitation for t x = ef-one centrifugal-charging pump-and-one safety injection pumpsto beCC'."OC ad "=rStmveiMence D*Thr nt to--

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OPC25 e~nai>>rPumF-t0TiLUyn dI6 below 350*F provides assurance that of a mass a single addition pressure transient can be relieved by the operation PORV.

~

5 5

1 The Surveillance Requirements provided to ensure OPERABILITY of each l

9 component ensures that at a minimum, the assumptions used in the safety !

i analyses are met and that subsystem OPERABILITY is maintained. Surveillance j Requirements for throttle 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 system 4 to each injection point is necessary to: (1) prevent total pump flow from j exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and l 1 (3) provide an acceptable level of total ECCS flow to all injection points j equal to or above that assumed in the ECCS-LOCA analyses and (4) to ensure that centrifugal charging pump injection flow which is directed through the seal injection path is less than or equal to the amount assumed in the safety analysis.

4 4

3/4.5.s REFUELING WATER STORAGE TANK 1 The OPERABILITY of the Refueling Water Storage Tank (RWST) as part of the j ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident, or a steam line rupture.

In the event of a LOCA, the limits on RWST minimum volume and boron concentration ensure that:

to permit recirculation cooling (1) sufficient flow to tha water core,isand available (2) thewithin containment reactor will i

]

remain subcritical in the cold condition following mixing of the RWST and the RCS water volumes with all control rods inserted except for the most reactive i control assembly.

These assumptions are consistent with the LOCA analyses.

4 i The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics,.

I j The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 8.5 and 10.5 for the solutf.on recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine andsystems mechanical minimizes the effect of chloride and caustic stress corrosion on and components.

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3/4.6 CONTAINMENT SYSTEMS

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BASES i

3/4.6.1 PRIMARY CONTAINMENT 3/4.6.1.1 CONTAINMENT INTEGRITY Primary CONTAINMENT INTEGRITY ensures that the release of radioactive materials from the containment atmosphere will be restricted to those leakage paths and associated leak rates assumed in the safety analyses. This restriction, in conjunction with the leakage rate limitation, will limit j the SITE BOUNDARY radiation doses to within the dose guideline values of E1 10 CFR Part 100 during accident conditions.

1 cj 3/4.6.1.2 CONTAINMENT LEAKAGE The limitations on containment leakage rates ensure that the total

.i containment leakage volume will not exceed the value assumed in the safety analyses at the peak accident pressure, P3. As~an added conservatism, the

-l measured overall integrated leakage rate is further limited to less than or equal to 0.75 La durin9 performance of the periodic test to account for possible degradation of the containment leakage barriers between leakage tests.

The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10 CFR Part 50.

3/4.6.1.3 CONTAINMENT AIR LOCKS Tne limitations on closure and leak rate for the containment air locks are required to meet the restrictions on CONTAINMENT INTEGRITY and containment

'4 leak rate. Surveillance testing of the air lock seals provides assurance that j the overall air lock leakage will not become excessive due to seal damage

.; during the intervals between air lock leakage tests.

l 3/4.6.1.4 INTERNAL PRESSURE y

The limitations on containment internal pressure ensure that: (1) the j; containment structure is prevented from exceeding its design negative pressure

] differential with respect to the outside atmosphere of 3 psig, and (2) the

+

.1 containentpeakpressuredoesnotexceedt[hedesignpressureof52psig during ktther steam line break conditions The max' mum peak pressure expected to be obtained from a ^C^ r steam i line break event is 41.9 psig assuming an initial containment pressure of 0.3 psig. The limit of 3 psig for initial positive containment pressure will limit the total pressure to 45 psig, which is less than design pressure and is consistent with the safety analyses.

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{ 3/4.6.1.5 AIR TEMPERATURE u

1 The limitations on containment average air temnerature ensure that the over-all containment average air temperature does not exceed the initial temperature condition assumed in the safety analysis for aM^C? r steam line break

, accident 9/ Measurements shall be made at all listed locations, whether by fixed 1 or portable instruments, prior to determining the average air temperature.

)l 3/4.6.1.6 CONTAINMENT STRUCTURAL INTEGRITY

1) This limitation ensures that the structural integrity of the containment will be maintained comparable to the original design standards for the life of the facility. Structural integrity is required to ensure that the containment will withstand the maximum pressure of 41.9 psig in the event of a steam line 1 break accident. The measurement of containment tendon lift-off force, the

tensile tests of the tendon waveseee strands, the visual examination of tendons, j anchorages and exposed interior and exterior surfaces of the containment, and ;

-?

the Type A leakage test are sufficient to demonstrate this capability. (The j j tendon h strand samples will also be subjected to stress cycling tests i and to accelerated corrosion tests to simulate the tendon's operating conditions and environment.)

The Surveillance Requirements for demonstrating the containment's structural integrity are in compliance with the recommendations of Revision 2 of Regulatory Guide 1.35, " Inservice Surveillance of Ungrouted Tendons in Prestressed Concrete Containment Structures," and proposed Regulatory Guide 1.35.1, " Determining

~

Prestressing Forces for Inspection of Prestressed Concrete Containments,"

April 1979.

The required Special Reports from any engineering evaluation of contain-1 ment abnormalities shall include a description of the tendon condition, the i condition of the concrete (especially at tendon anchorages), the inspection 1

procedures, the tolerances on cracking, the results of the engineering evalua-i tion, and the corrective actions taken.

3/4.6.1.7 CONTAINMENT VENTILATION SYSTEM j The 24-inch containment purge supply and exhaust isolation valves are q required to be sealed closed during plant operations since these valves have not i been demonstrated capable of closing during a [LOCA or steam line break accident].

i Maintaining these valves sealed closed during plant operation ensures that exces-

! sive quantities of radioactive materials will not be released via the Containment j Purge System. To provide assurance that these containment valves cannot be inad-f vertently opened, the valves are sealed closed in accordance with Standard Review Plan 6.2.4 which includes mechanical devices to seal or lock the valve closed, or

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BASES e COMBUSTIBLE GAS CONTROL (Continued)

The Hydrogen Mixing Systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent g localized accumulations of hydrogen from exceeding the flammable limit.

3/4.6.6 ELECTRICAL PENETRATION ROOM EXHAUST AIR CLEANUP SYSTEM

]

I The OPERABILITY of the Electrical Penetration Room Exhaust Air Cleanup System ensures that radioactive materials leaking from the containment atmo- G_

sphere through containment penetrations following a LOCA are filtered and /

adsorbed prior to reaching the environment. Operation of the system with the 8 heaters operating for at least 10 continous hours in a 31-day period is suffi I cient to reduce the buildup of moisture on the adsorbers and HEPA filters. The k operation of this system and the resultant effect on offsite dosage calculations was assumed in the LOCA analyses. ANSI N510-1975 and Generic Letter 83-13,

" Clarification of Surveillance Requirements for HEPA Filters and Charcoal Adsorber Units in Standard Technical Specifications on ESF Cleanup Systems" are used as procedural guides for surveillance testing.

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) BASES 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that i the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70*F and 200 psig are based on a steam generator RT NDT of 60*F and are sufficient l to prevent brittle fracture.

f 3/4.7.3 COMPONENT. COOLING WATER SYSTEM 0 The OPERABILITY of the Component Cooling Water System ensures that suf-l ficient cooling capacity is available for centinued operation of safety-related

$ equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the 3

assumptions used in the safety analyses.

3/4.7.4 NUCLEAR SERVICE COOLING WATER SYSTEM i

i The OPERABILITY of the Nuclear Service Cooling Water System ensures that sufficient cooling capacity is available for continued operation of safety-related equipment during normal and accident conditions. The redundant cooling

capacity of this system, assuming a single failure, is consistent with the u assumotions used in the safety analyses. -

3/4.7.5 ULTIMATE HEAT SINK s

] The limitations on the ultimate heat sink level, temperature, and minimum j required number of OPERABLE fans ensure that sufficient cooling capacity is !

j available either: (1) provide normal cooldown of the facility or (2) mitigate i j the effects of accident conditions within acceptable limits. l

2'7 The limitations on minimum water level, maximum tempera e, and inimum i i required number of OPERABLE fans are based on providing a 20 ty ccoling water '

supply to safety-related equipment without exceeding its design basis temperature and is consistent with the recommendations of Regulatory Guide 1.27, " Ultimate

] Heat Sink for Nuclear Plants," March 1974.

a 3/4.7.6 CONTROL ROOM EMERGENCY FILTRATION CLEANUP SYSTEM l 8 The OPERABILITY of the Control Room Emergency Air Cleanup System ensures l j that: (1) the ambient air temperature does not exceed the allowable temperature l for continuous-duty rating for the equipment and instrumentation cooled by this system during accident conditions, and (2) the control room will remain habitable i for operations personnel during and following all credible accident conditions.

, Operation of the system with the heater control circuit energized for at least 2 10 continuous hours in a 31-day period is sufficient to reduce the buildup of j

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The Surveillance Requirement for demonstrating the OPERABILITY of the N station batteries are based on the recommendations of Regulatory Guide 1.129,

" Maintenance Testing and Replacement of Large Lead Storage Batteries for

)j Nuclear Power Plants," February 1978, and IEEE Std 450-1980, "IEEE Recommended

{ Practice for Maintenance, Testing, and Replacement of Large Lead Storage a Batteries for Generating Stations and Substations," and 484-1975 " Recommended 1 Practice for Installation Design and Installation of Lead Storage Batteries for Generating Stations and Substations."

, Verifying average electrolyte temperature above the minimum for which the battery was sized, total battery terminal voltage on f.loat charge, connection resistance values, and the performance of battery service and discharge tests ensures the effectiveness of the charging system, the ability to handle high discharge rates, and compares 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 gravity. The limits for the designated pilot cells float voltage and specific 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 specific gravity, greater i than 2.13 volts and not more than 0.020 below the manufacturer's full charge specific gravity with an average specific gravity of all the connected cells not more than 0.010 below the manufacturer's full charge specific gravity, i i

ensures the OPERABILITY and capability of the battery.

r Operation with a battery cell's parameter outside the normal limit but within the allowable value specified in Table 4.8-2 is permitted for up to 7 days. During this 7-day period: (1) the allowable values for electrolyte level ensures no physical damage to the plates with an adequate electron transfer capability; (2) the allowable value for the average specific gravity

of all the cells, not more than 0.020 below the manufacturer's recommended full e charge specific gravity, ensures that the decrease in rating will be less than the safety margin provided in sizing; (3) the allowable value for an individual 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 gravity and that the overall capability of the battery will be maintained within an acceptable limit; and (4) the allowable value for an individual cell's float voltage, greater than email volts, ensures the battery's capability to perform its design function. f

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[ 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES l 1

Containment electrical penetrations and penetration conductors are pro-tected by either deenergizing circuits not required curing reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protec-

)' tion circuit breakers during periodic surveillance.

4 The Surveillance Requirements applicable to lower voltage circuit breakers

and-dwees provide assurance of breaker-end=fece reliability by testing at

least one representative sample of eatn manufacturer's brand of circuit breaker 1 -: nd/:r fwee. Each manufacturer's molded case and metal case circuit breakers in&cr fr::: are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers :nd/;r fr;:: are tested. If a wide variety exists within any manufacturer's brand of circuit breakers end/ m Jeeee, it is necessary to divide that manufacturer's breakers.amMee-f4mes-

,. into groups and treat each group as a separate type of breaker e fwees for surveillance purposes.

The of the motor-operated es he m 1 overload protection ~

he thermal overload protection will

]3 +4 ' ^:;;r:! O;;; d: i;;; ensures that [rforming their function. The Surveil-

" not lanceprevent safety-related Requirements valves fromthe for demonstrating pg4PEAMHtH4-of the thermal overload protectionfare in accordance with Regulatory Guide 1.106, " Thermal Overload Protection' for Electric Motors on Motor Operated Valves," Revision 1, March 1977.

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RADI0 ACTIVE EFFLUENTS BASES 3/4.11.2.5 EXPLOSIVE GAS MIXTURE t This specification is provided to ensure that the concentration of poten-tially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. [ Automatic control features are includad in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits.] Maintaining the concentration of hydrogen and oxygen below their flanmability limits provides assurance that ,

the releases of radioactive materials will be controlled in conformance with I the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

3/4:11.2.6 GAS STORAGE TANKS The tanks included in this specification are those tanks for which the quantity of radioactivity contained is not limited directly or indirectly by another Technical Specification. Restricting.the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem.

This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases Due to a Waste Gas System Leak or l Failure," in NUREG-0800, July 1981. Since only the gamma body dose factor (DFB$ )

is used in the analysis, the Xe-133 equivalent is determined from the DFB5 value for Xe-133 as compared to the composite DFB 5 for the actual mixture in the tank.

3/4.11.3 SOLID RADIOACTIVE WASTES This specification implements the requirements of 10 CFR 50.36a and' General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to, waste type, waste pH, waste / liquid / SOLIDIFICATION agent / catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times.

3/4.11.4 TOTAL DOSE 4

This specification is provided to meet the dose limitations of b CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report when- l ever the calculated doses due to releases of radioactivity and to radiation from l uranium fuel cycle sources exceed 25 mrems to the whole body or any organ, except l the thyroid, which shall be limited to less than or equal to 75 mrems. For sites containing up to four reactors, it is highly unlikely that the resultant dose to a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the individual reactors remain within twice the dose design objectives of Appendix I, and if direct radiation doses from the units and from outside l

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3/4.12 RATI0 LOGICAL ENVIRONMENTAL MONI'TORING ,

BASES 3/4.12.1 MONITORING PROGRAM N #79 Y# N#

/

The Radiological Environmental Monitoring Program required by this specification provides representative measurements of radiation and of radio-active materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposure of MEMBERS OF THE PUBLIC resulting from the plant operation. This M nitoring program implementsSection IV.B.2 of Appendix I to 10 CFR Part 50 and thereby supplements the ,

Radiological Effluent Monitoring Program by' cifyin; that th: ::::r:b!:-meAsun nj /

concentrations of radioactive materials and levels of radiation r : net 'i;MrN Ahon expected on the basis of the effluent measurements and the modeling of the environmental e'xposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position _on_ Environ- ,

mental Monitoring._JThe initially specified monitoring program will be effective C for at~ least the first 3 years of commercial operation. Following this period,

/(D, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of e lower limits of detection (LLDs). The LLDs required by Table 4.12-1 are considered optimum for routine environmental measurements in industrial laboratories. It should be recognized that the LLD is defined as an a_ priori (before the fact) limit representing the capability of a measure-ment system and not as an a posteriori (after the fact) limit for a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A. , " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K. , " Detection Limits for Radicanalytical Counting Techniques" Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

3/4.12.2 LAND USE CENSUS This specification is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater 5 2

. thangt0=m provides assurance that significant exposure pathways via leafy vegetables will be identified and monitored since a garden of this size is the minimum required to produce the quantity (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following assumptions were made: (1) 20% of the garden was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/m2, V0GTLE - UNIT 1 B 3/4 12-1 APR 2 4 %

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DESIGN FEATURES .

AR

5.3 REACTOR CORE l

FUEL ASSEMBLIES

5.3.1 The core shall contain 193 fuel assemblies with each fuel assembly

j containing 264 fuel rods clad with Zircaloy-4. Each fuel rod shall have j a nominal active fuel length of 144 inches.:nd ::nt:i : : ? r- t;t;l eight.

j / 1% , ,, _,a c:ni =? The initial core loading shall have a maximum enrichment d of 3.2 weight percent U-235. Reload fuel shall be similar in physical design j to the initial core loading and shall have a maximum enrichment of 3.5 weight j

percent U-235.

j CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 53 full-length and no part-length control rod assemblies. The full-length control rod assemblies shall contain a nominal 142 inches of absorber material. The composition shall be 95.5% natural ;

All control rods shall be clad with !

balff'umand4.5%naturalzirconium.

j stainless steel tubing.

I A 5.4 REACTOR COOLANT SYSTEM

! DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a. In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements,

b. For a pressure of 2485 psig, and 1

1 c. For a temperature of 650 F, except for the pressurizer which is 680*F.

VOLUME 1

5.4.2 The total water and steam volume of the Reactor Coolant System is j 12,240 + 100 cubic feet at a nominal T,yg of 588.5 F. i 1

1 i 5.5 METEOROLOGICAL TOWER LOCATION l

l 5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1.

4

=l

1 1

V0GTLE - UNIT 1 5-5

.! APR 34 Ng

'1 ll "

i!

. -...---. .. ;-...... . . . ~ ~. .. a.. . . . -. =.:.n...... ..

TABLE 5.7-1 i <

8 COMPONENT CYCLIC OR TRANSIENT LIMITS .

m CYCLIC OR DESIGN CYCLE E COMPONENT TRANSIENT LIMIT . OR TRANSIENT Z

s Reactor Coolant System 200 heatup cycles at < 100*F/h Heatup cycle - T from < 200*F avg -

and 200 cooldown cycles at to > 550*F. [.

-< 100 F/h. CooTdo in cycle - T avg from p

> 550*F to $ 200*F 200 pressurizer cooldown cycles Pressurizer cooldown cycle at < 200'F/h.

temperatures from > 650'F to p<

5 200*F.

. {e

, 80 loss of load cycles, without > 15% of RATED THERMAL POWER to 0,'

immediate Turbine or Reactor trip. D% of RATED THERMAL POWER. fl S'

" 40 cycles of loss-of-offsite Loss-of-offsite A.C. electrical {x A.C. electrical power. ESF Electrical System. h h'-

80 cycles of loss of flow in one Loss of only one reactor reactor coolant loop. coolant pump.

400 Reactor trip cycles. 100% to 0% of RATED THERMAL POWER. (. .

p 10 auxiliary spray Spray water t erature differential [,

actuation cycles. > 320*Fbut < 25' F q

(;

200 leak tests. Pressurized to > 2485 psig.

t 10 hydrostatic pressure tests. Pressurized to > 3107 psig. f:

k 3

Secondary Coolant System I steam line break. Break in a > 6-inch steam line. [-

10 hydrostatic pressure tests. Pressurized to > 1481 psig.

[e p

I .'

e

ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The General Manager - Vogtle Nuclear Operations (GMVNO) shall be respon-sible for overall plant operation and shall delegate in writing the succession to this responsibility during his absence.

~- -

6.1. 2 he 5 ft Su ervisor (or during his absence from the control room, 1

fn n ag t shall be reissued to all station personnel on an o

' ate management] Wg , ,, ,,, ,, ,

g L nua g g, g 6.2 ORGANIZATION

'^%4M&

W N"AN A A~, A4~==JMj

",^ ^ 'HW / 41 .

M'Af c=' + M D ^ 4 @

0FFSITE -44.

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6.2.1 The offsite organization YdE plant management and technical suppor't shall be as shown in Figure 6.2-1.

0WT .

4M W STAFF 6.2.2 The organization shall be as shown in Figure 6.2-2 and:

a. Each on-duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1;

b. At least one licensed Operator shall be in the control room when fuel is in the reactor. In addition, while the unit is in MODE 1, 2, 3, or 4, at least one licensed Senior Operator shall be in the control room;

c. A Health Physics Technician

shall be on site when fuel is in the reactor;

d. All CORE ALTERATIONS shall be observed and directly supervised by either a licensed Senior Operator or licensed Senior Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation;

e. A site Fire Brigade of at least five members

shall be maintained on site at all times. The Fire Brigade shall not include the Shift Supervisor and the [two] othat members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency; and

The Health Physics Technician and Fire Brigade composition may be less than the minimum requirements 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 order to accommodate unexpected absence, provided immediate action is taken to fill the required positions, V0GTLE - UNIT 1 6-1 f APR 24 BSc J

ADMINISTRATIVE CONTROLS UNIT STAFF (Continued) f.

Administrative procedures shall be developed and implemented to limit the working hours of plant staff in performance of safety-related functions (e.g. , licensed Senior Operators, licensed Operators, key HealthPhysicsTechnicians,/eynon-licensedoperators,andkey maintenance personnel).

of Adequate shift coverage shall be maintained without routine heavy

g-"- The objective shall be to have operating personnel

, use of overtim " 40-hour week while the plant is operating.c

. work a - -

However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shut-down for refueling, major maintenance, or major plant modification, on a temporary basis the following guidelines shall be followed:

c ike A & m wgia. M Q.uA.44. ) -

1. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time.

2. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24-hour period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48-hour period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7-day period, all excluding shift turnover time.

3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time.

4. Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift.

Any deviation from the above guidelines shall be authorized by the e,y / M {Phat %perint:nint] ;r his depaf, or higher levels of manage-W ment, in accordance with established procedures and with documenta-tion of the basis for granting the deviation. Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the General Manager - Vogtle Nuclear Operations or his designee to assure that excessive hours have not been assigned.

Routine deviation from the above guidelines is not authorized.

1 V0GTLE - UNIT 1 6-2 APR 24 W l

1 1

l l

i FIGURE 6.2-2 PIADT N ORGANIZATION V0GTLE - UNIT 1 6-4 APR E 41966 !

I

ADMINISTRATIVE CONTROLS Wc LEA R Ny 6.2.3 -IN9EPEW9EMT-SAFETY ENGINEERING GROUP ($SEG)

FUNCTION N N 6.2.3.1 The G shall function to examine plant operating characteristics, NRC issuances, industry advisories, Licensee Event Reports, and other' sources

't: Of :f-ilar of plant design and operating experience information, incWi5The =dSEG shall riacign which may indicate areas for improving plant safety.

make detailed recommendations for revised procedures, equipment modifications, maintenance activities, operations activities, or other means of improving plant safety to the Senior Vice President-Nuclear Operations through the Manager-Nuclear Performance and Analysis.

COMPOSITION J-6.2.3.2 The $5EG shall be composed of at least five, dedicated, full-time engineers. Meted = :ite. 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,N 6.2.3.3 The (S'EG shall be responsible for maintaining surveillance of va4p/anf activities to provide independent verification

that these activities are performed correctly and that human errors are reduced as much as practical.

. RECORDS d 6.2.3.4 Records of activities performed by the EG shall be prepared, main-tained, and forwarded each calendar month to fa-high 1:=100 p:::t Official im a = technicaMy< ori ente d 30si ti o n-who-i s-not-i n-the-managemen t-c hain-fo r-- ,

-power =productic&. MW Vice-t%ric/ent-/L}vc/w()xntfuns Mrouf7 M 2-N* I fdd h2Y4 SHIfTYEYHN ALADNSk 6.2.4.1 The Shift Technical Advisor shall provide advisory technical support to the Shift Supervisor in the areas of thermal hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the m 4. The Shift Technical Advisor shall have a bachelor's degree or equivalent in a scientific or engineering discipline and shall have received specific training in the response and analysis of the un K for transiente and accidents, and in design and layout, including the capabilities of instrumentation and contr s l in the control room. L--p/aaf-PUNT 6.3 UNW STAFF QUALIFICATIONS gg C. T / Y l

l

Not responsible for sign-off function.

V0GTLE - UNIT 1 6-6 g l

ADMINISTRATIVE CONTROLS , .

PtGNT /- P '* t ipeT STAFF QUALIFICATIONS (Continued)

! 6.3.1 Each member of the M staff shall meet or exceed the minimum qualifica-tions of ANSI /ANS N18.1971 for comparable positions, except for the [ Radiation Protection Manager] who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975. The licensed Operators and Senior Operators shall also meet or exceed the minimum qualifications of the supplemental requirements specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees.

6.4 TRAINING 6.4.1 A retraining and replacement training program for the plant staff shall be maintained under,the direction of the Superintendent of Nuclear Training and shall meet or exceed the requirements and recommendations of Section of ANSI /ANS N18.1971 and Appendix A of 10 CFR Part 55 and the supplemental require-ments specified in Sections A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational expe.rience.

6.5 REVIEW AND AUDIT 6.5.1 PLANT REVIEW BOARD (PRB)

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

COMPOSITION 6.5.1.2 The PRB shall be composed of the:

Chairman: [ Plant Superintendent]

Member: [ Operations Supervisor]

Member: [ Technical Supervisor]

Member: [ Maintenance Supervisor]

Member: [ Plant Instrument and Control Engineer]

Member: [ Plant Nuclear Engineer]

Member:

ALTERNATES

[ Health Physicis {# U 4 'S)-

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

f .(. l

4 pWy,s f*f',Y,(f N

Y 9.c

\

l V0GTLE - UNIT 1 6-7 APR 24198b

MAFT ADMINISTRATIVE CONTROLS MEETING FREQUENCY 6.5.1.4 The PRB shall meet at least once per calendar month and as convened by the PRB Chairman or his designated alternate.

QUORUM 6.5.1.5 The quorum of the PRB necessary for the performance of the PRB responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates..

05 RESPONSIBILITIES 6.5.1.6 The PRB shall be r'esponsible for:

t gg /

a. Review of: (1) all proposed procedures required by Specification T 8 and changes thereto, (2) all proposed programs required by Specification 6.8 and changes thereto, and (3) any other proposed procedures or changes thereto as determined by the GMVNO to affect nuclear safety; I

b. Review of all proposed tests and experiments that affect nuclear safety;

c. Review of all proposed changes to the Technical Specifications; lart+

d. Review of all proposed changes or modifications to (m44 systems or equipment that affect nuclear safety;

e. Investigation of all violations of the Technical Specifications, including the preparation and forwarding of reports covering evalua-tion and recommendations to prevent recurrence, to the Vice President and General Manager-Nuclear Operations and to the Safety Review Board; i

f. Review of all REPORTABLE EVENTS;

g. Review of plant operations to detect potential ha ards to nuclear safety;

h. Performance of special reviews, investigations, or analyses and reports thereon as requested by the GMVNO or the Safety Review Board; i.

Review of the Security Plan and implementing procedures and submittal of recommended changes to the GMVNO and the Safety Review Board; V0GTLE - UNIT 1 6-8 APR 241986 I

l

'DMINISTRATIVE CONTROLS A

AUDITS (Continued)

b. The performance, training, and qualifications of the entire plant staff at least once per 12 iconths; The results of actions taken to correct deficiencies occurring in !

c.

plant equipment, structures, systems, or method operation that affect nuclear safety, at least once per 6 months;

d. The parformance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix B, 10 CFR Part 50, at least once per 24 months;

e. The fire protection programmatic controls including the implementing procedures at least once per 24 months by qualified licensee QA personnel;

f. The fire protection equipment and program implementation at least once per 12 months utilizing either a qualified offsite licensee fire protection engineer or an outside independent fire protection consultant. An outside independent fire protection consultant shall be used as least every third year;

g. The Radiological Environmental Monitoring Program and the results thereof at least once per 12 months;

h. The OFFSITE DOSE CALCULATION MANUAL and implementing procedures at

. least once per 24 months;

i. The PROCESS CONTROL PROGRAM and implementing girocedures for processing and packaging of radioactive wastes at least once per 24 months;

j. The performance of activities required by the Quality Assurance Program for effluent and environmental monitoring at least once per 12 months;

k. The Emergency Plan and implementing procedures (at least once per 12 months);

1. The Security Plan and implementing procedures (at least once per 12 months); and (7 h ther area of unit operation considered appropriate by the SRB or the Vice President-Nuclear Operations.

N -

RECORDS 6.5.2.9 Records of SRB activities shall be prepared, approved, and distributed as indicated below:

a. Minutes of each SRB meeting shall be prepared, approved, and forwarded to the Senior Vice President-Nuclear Operations within 14 days follow-ing each meeting;

b. Reports of reviews encompassed by Specification 6.5.2.7 shall be pre-pared, approved, and forwarded to the Senior Vice President-Nuclear Operations within 14 days following completion of the review; and V0GTLE - UNIT 1 6-12 f

AMt 841806

ADMINISTRATIVE CONTROLS RECORDS (Continued)

c. Audit reports encompassed by Specification 6.5.2.8 shall be forwarded to the Executive Vice President-Power Supply, Senior Vice President-Nuclear Operations and to the management positions responsible for the areas audited within 30 days after completion of the audit by the auditing organization.

6.6 REPORTABLE EVENT ACTION 6.6.1 The following actions shall be taken for REPORTABLE EVENTS:

or

a. The Commission shall be notified an a report submitted pursuant to the requirements of Section 50.72 and Section 50.73 to 10 CFR Part 50, and

b. Each REPORTABLE EVENT shall be reviewed by the PRB, and the results of this review shall be submitted to the SRB and the Vice President and General Man'ager-Nuclear Operations.

6.7 SAFETY LIMIT VIOLATION 6.7.1 The following action,s shall be taken in the event a Safety Limit is violated:

a. In accordance with 10 CFR 50.72, the NRC Operations Center shall be notified by telephone as soon as practical and in all cases within one hour after the violation has been determined. The Vice President and General Manager-Nuclear Operations, the SRB, PRB, and the GMVNO shall be notified within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ,

b. A Licensee Event Report shall be prepared in accordance with 10 CFR 50.73.

c. The Licensee Event Report shall be submitted to the Commission in accordance with 10 CFR 50.73, and to the PRB, SRB, the GMVNO and the Vice President and General Manager-Nuclear Operations within 30 days after discovery of the event.

l d. Critical operation of the unit shall not be resumed until authorized by the Nuclear Regulatory Commission.

6.8 PROCEDURES AND PROGRAMS 6.8.1 Written procedures shall be established, implemented, and maintained covering the activities referenced below:

a. The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978;

b. The emergency operating procedures required to implement the require- ;

ments of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33; l

(

V0GTLE - UNIT 1 6-13 APR S 41986 l

ADMINISTRATIVE CONTROLS l

ANNUAL REPORTS (Continued) surveillance, inservice inspection, routine maintenance, special main-tenance [ describe maintenance], waste processing, and refueling). The '

dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeter (TLD), or film badge measurements. Small exposures totalling less than 20% of the individual ,

1 In the aggregate, at least 80% i total dose need not be accounted for. '

of the total whole-body dose received from external sources should be assigned to specific major work functions;

b. The results of specific activity analyses in which the primary coolant exceeded the limits of Specification 3.4.8. The following information shall be' included: (1) Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior

~-'

to the first sample in which tha limit was exceeded (in graphic and tabular format): (2) Results of the last isotopic analysis for radio- '

l iodine performed prior to exceeding the limit, results of analysis while limit was exceeded and results of one analysis after the radiciodine j activity was reduced to less than limit. Each result should include i date and time of sampling and the radiciodine concentrations; l

4 (3) Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded; (4) Graph of the I-131 concentration (pCi/ge) and one other radioidine isotope concentration (pCi/gs) as !

a function of time for the duration of the specific activity above the l

steady-state level; and (5) The time duration when the specific activity of the primary coolant exceeded the radioiodine limit, l

c. A report shall be prepared and submitted to the commission on an annual i

basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcuries of removable r contamination.

d. An annual data report on diesel generator reliability will be submitted and, in addition, the following information will be included:

1. A summary of all tests (valid and invalid) that occurred within the time [ period over which the last 20/100 valid tests were performed].

2. Analysis of failures and determination of root causes of failures.

, 3. Identification of all actions taken or to be taken to 1) correct i

the root causes of failures defined in b) above and 2) achieve a general improvement of diesel generator reliability.

4. An assessment of the existing reliability of electric power to <

engineered-safety-feature equipment.

ANNUAL RADIOLOGICAL ENVIRONMENTAL SURVEILLANCE REPORT

/M ;

</

L 6.9.1.3 Routine Annual Radiological Environmental Surveillance Reports coverIngew & L-

' tM ere--th- ef tM -% during the previous calendar year shall be submitted prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following initial criticality and shall include copies of reports of the preoperational Radiological Environmental Monitoring Program of the unit for,at least two years prior to initial criticality.

V0GTLE - UNIT 1 6-17 i

APR S4 W 6 e

+e ,- , - ,-,--.-,--.---m- -

- - - - - -,,----,n.,,,-- . , , , - - - . . - .n,- _ . . ,,,n, __-, -, , ,- ,,-----,,,-,.---,m,,

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( e m - & U m r i _ _____ __

ADMINISTRATIVE CONTROLS M-4 % er M W & lLE s-/ # A # l wmjug ~"" Y N S m m N ANNUAL RADIOLOGICAL ENVIRONMENTAL OPGATM REPORT (Continued)

The Annual Radiological Envir.amental Surveillance Reports shall include summaries, interpretations, and an analysis of trends of the results of the ; ,

radiological environmental surveillance activities for the report period, !

includinga comparison with preoperational studies, with operational controls, (1ili appropriaty, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. j The reports shall also include the results of the Land Use Census required by !

Specification 3.12.2. l )

The Annual Radiological Environmental Surveillance Reports shall include the results of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table and figures in the Offsite Dose Calculation Manual, as well as summarized and tabulated results of these analyses ar.d measurements in the format:of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. Tn the event that some inoivi-dual results are not available for inclusion with the report, the report shall The be submitted noting and explaining the reasons for the missing results.

e submitted as soon as possible in a supplementary report.

mi dga 1

The reports shall also include the following: a summary description of the Radiological Environmental Monitoring Program; ** 1=4-4we legible maps

vering all sampling locations keyed to a table giving distances and directions from ha centarlf ;; cf em reactor?, the results of licensee participation in the Interlaboratory Comparison Program and the corrective action taken if the specified program is not being performed as required by Specification 3.12.3; reasons for not conducting the Radiological Environmental Monitoring Program as required by specification 3.12.1, and discussion of all deviations from the sampling schedule of Table 3.12-1; discussion of environmental sample measure-ments that exceed the reporting levels of Table 3.12-2 but are not the result of plant effluents, pursuant to ACTION b. of Specification 3.12.1; and discussion of all analyses in which the LLD required by Table 4.12-1 was not achievable.

SEMIANNUAL RADIOACTIVE EFrLUENT RELEASE REPORT 6.9.1.4 Routine Semiannual Radioactive Effluent Release Reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days after January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality.

The Semiannual Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21,

" Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data

0ne map shall cover stations near the SITE BOUNDARY; a second shall include the more distant stations.

1 V0GTLE - UNIT 1 6-18 gpgg4 g l

l

E T INUW I 1 ADMINISTRATIVE CONTROLS 6.12 HIGH RADIATION AREA i C % L 6.12.1 Pursuant to paragraph 20.203(c)(5) of 10 CFR Part 20, in lieu of the

" control device" or " alarm signal" required by paragraph 20.203(c),each high radiation area, as defined in 10 CFR Part 20, in which the intensity of radia-tion is equal to or less than 1000 mR/h at 45 cm (18 in.) from the radiation source or from any surface which the radiation penetrates shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., Health Physics Technician) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radia-tion areas with exposure rates equal to or less than 1000 mR/h, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas. Any individual or group of individuals permitted to enter such areas shall be provided with or ccompanied by one or more of the following:

A radiation monitoring device which continuously indicates the a.

r.adiation dose rate in the area; or

b. A radiation monitoring device which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel have been made knowledgeable of them; or

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the

[ Radiation Protection Manager] in the RWP.

6.12.2 In addition to the requirements of Specification 6.12.1, areas accessible to personnel with radiation levels greater than 1000 mR/h at 45 cm (18 in.)

from the radiation source or from any surface which the radiation penetrates shall be provided with locked doors to prevent unauthorized entry, and the keys shall be maintained under the administrative control of the shift Foreman on duty and/or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP which shall specify the dose rate levels in the immediate work areas and the maximum allowable stay time for individuals in that area. In lieu of the stay time specification of .

the RWp, direct or remote (such as closed circuit TV cameras) continuous , \(

surveillance may be made by personnel qualified in radiation protection !

N procedures to provide positive exposure control over the activities being l

[

performed within the area. l For individual high radiation areas accessible to personnel with radiation : '

levels of greater than 1000 mR/h that are located within large arus, such as l

PWR containment, where no enclosure exists for purposes of locking, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded, conspicuously posted, and a flashing light shall be activiated as a warning device.

V0GTLE - UNIT 1 6-23 APR S 41986

--

ML20199E872

Text

Dear Mr. Denton:

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