ML20058J297

Draft Tech Specs for Facility

ML20058J297
Person / Time
Site:	Wolf Creek
Issue date:	08/05/1982
From:	KANSAS GAS & ELECTRIC CO.
To:
Shared Package
ML20058J296	List: ML20058J292 ML20058J297
References
NUDOCS 8208090225
Download: ML20058J297 (105)

Text

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%Q TABLE 4.3-1 en O

80 REACTOR TRIP SYSTEM INSUllMENTATION SURVEILIANCE REQUIREMDirS o

88 mao TRIP E8" ANAIS ACTUATING MODES FOR CHANNEL DEVICE MIICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILIANCE EUNCTIONAL UNIT CHECK CALIBRATION TEST CHECK LOGIC 'IEST IS REQUIRED Manual Reactor Trip N/A N/A.

N/A R

N/A 1, 2, 3*,

4*, @

2.

Power Range A.

Neutron Flux S

D(2)M(3)

N/A N/A N/A 1,2 Q(5)R(4)

B.

High Setpoint N/A R

M N/A N/A 1,2 C.

Iai Setpoint N/A R

M (8)

N/A N/A 1999,2 w

D.

High Positive Rate N/A R

M N/A N/A 1,2 2

E.

High Negative Rate N/A R

M N/A N/A 1,2

,Y 3.

Intermediate Range, S

R(4)

S/U(1),M (B)

N/A N/A lift,2 Neutron Flux 4.

Source Range, Neutron Flux S

R(4)

S/U (1),M N/A N/A 288, 3, 4, 5 5.

Overtenperature a T S

R M

N/A N/A 1,2 i

6.

Overpower A T S

R M

N/A N/A 1,2 1

7.

Pressurizer Pressure A.

Iai S

R M

N/A N/A 1

B.

High S

R M

N/A N/A 1,2 8.

Pressurizer Water Level-High S

R M

N/A N/A 1

9.

Loss of Flow S

R M

N/A N/A 1

i

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TABLE 3.3-5 i

. 3

(_dh q

SEISMIC MONITORING INSTRUMENTATION MINIMUM i

MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE 1.

Triaxial Peak Recording Accelerographs a.

Radwaste Base Slab

+ 1.0g 1

b.

Control Room 1

1 1.0g fump Facility 1 1.0g c.

1 d.

STRUCTURE

+%@'. 0 g 1

e.

Auxiliary Bldg. SI Pump Suctions i 1.0g 1

f.

SGB Piping 19#.0g 1

g.

SGB Support i 1.0g 1

2.

Triaxial Time History and Response Spectrum Recording System, Monitoring the Following Accelerometers:

a.

Ctmt. Base Slab

~~ ~

i 1.0g 1

b.

Ctmt. Oper. Floor i 1.0g 1

c.

Reactor Support i 1.0g 1

d.

Aux. Bldg. Base Slab i 1.0g 1

e.

Aux. Bldg. Control Room Air Filters i 1.0g 1

f.

Free Field 1 0.5g 1

w

/

3., Triaxial Seismic Switches ACCELERATION LEVEL a.

OBE Ctmt. Base Slab 0.06s 1

b.

SSE Ctmt. Base Slab

.o. t s g 1

c.

OBE Ctmt. Oper. Fl.

1 uret d.

SSE Ctmt. Oper. Fl.

1 ung e.

System Trigger 1

un,

(

94 i

3%

l i

WCGS 3/43[

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INSTRUMENTATION CHLORINE DETECTION SYSTEMS LIMITING CONDITION FOR OPERATION 4.ls 3.3.t:t Two independent chlorine detection systems, with their alarm / trip setpoints adjusted to actuate at a chlorine concentration of less than or equal to 5 ppm, shall be OPERABLE.

i APPLICABILITY: ALL MODES i

ACTION:

i a.

With one chlorine detection system inoperable, restore the inoperable detection system to OPERABLE status within 7 days or within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.

b.

With both chlorine detection systems inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> initiate and maintain operation of the control room emergency ventilation system in the recirculation mode of operation.

c.

The provisions of Specification 3.0.4 are not applicable.

h.

T h. p isi. e o4 Sp.O.avien 3,0 3 avg. nov syliuble.

m m oo E los I.

SURVELLIANCE REQUIREMENTS q.fo 5%N 4.3.+-t-Each chlorine detection system shall be de[nstrated OPERABLE by

?

performance of a CHANNEL CHECK at least once per M L..,, an ANALOG CHANNEL OPERATIONAL TEST at least once per 01 1 -, and a CHANNEL CALIBRATION at least once per 10

.~....u montM et.O dinhe 1

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q FIRE DETECTION INSTRtniFNTS

_ Detection Instrutnants I

ROOM ZONE' HEAT

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TABLE 3.3rif i

RADIOACTIVE LIQUID EFFLUENT HONITORING INSTRUMENTATION e

I o

tr V'

HINIMUM CHANNELS INSTRUMENT OPERABLE ACTION 1.

CROSS RADIOACTIVITY MONITORS PROVIDING AUTOMATIC TERMINATION OF RELGASE Liquid Radwaste Discharge Monitor (RE-18)

(1) 28 a.

b.

Steam Generator Blowdown Discharge Monitor (RE-52)

(1) 29 c.

Turbine Building Drain Monitor (RE-59)

(1) 30 d.

Secondary Liquid Waste System' Monitor (RE-45)

(1) 31 us 4

4 VJ 2.

FLOW RATE HEASUREMENT DEVICES a.

Liquid Radwaste Discharge Line (1) Waste Monitor Tank A Discharge Line (1) 32 (2) Waste Monitor Tank B Discharge Line (1) 32 i.s t

c__is

,._.._ni_...___.

1 s_-

v-w

- _ - - - o 2 2= -.= : :r-

z. = : -

b. +

Steam Generator Blowdown Effluent Lines (1)

'32 '

c. A Secondary Liquid Waste Syst em Discharge Line (1) 32 e

~-

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' f'y TABLE 4.3-g o

[

RADIOACTIVE LIQUID EFFI.UENT HONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS J

~

CHANNEL CIIANNEL SOURCE CHANNEL FUNCTIONAL 1

INSTRUMENT CllECK CllECK CALIBRATION TEST j

i 1.

CROSS BETA OR CAMMA RADIOACTIVITY MONITORS g

PROVIDING ALARM AND AUTOMATIC TERMINATION OF RELEASE L

f a.

Liquid Radwaste Discharge Monitor (RE-18)

D P

R(2)

Q(1) b.

Steam Generator Blowdown Discharge l

Monitor (RE-52)

D H

R(2)

Q(1) e c.

Turbine Building Drain Monitor (RE-59)

D M

R(2)

Q(1) 4-d.

Secondary Liquid Waste System.

I::

e Monitor (RE-45)

D P

R(2)

Q(1)

I '.

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-4 2.

FLOW RATE HEASUREMENT DEVICES a.

Liquid Radwaste Discharge Line D(3)

N.A.

R Q

. (1) Waste Monitor Tank A Discharge Line D(3)

N.A.

R Q

i.

(2) Waste Monitor Tank B Discharge Line D(3)

N.A.

R Q

l 1

b.

Steam Generator Blowdown Discharge Line D(3)

N.A.

R Q

M'7

, Q ny q

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c,. _ i.

, ~.

= i -. n. -. 1...

u a

=

l d.

Secondary Liquid Waste System Discharge Line D(3)

N.A.

R Q

l; l

1 l

l

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I q7 TABLE 4.3-14F(Continued)

TABLE NOTATION

....m,

.......u.

ANA Loco-CH AM"Gt-OP G MT* *"A L (1) ThepN"CL TU:CTICHAL-TEST shall-also demonstrate that automatic isolation INSERT-09 of this cathway pund control room alarm annunciation occurs if any of the i

rollowing concitions exists:

D

1 1.

Instrument indicates measured levels above. the alarm / trip setpoint.---]

i INS'ERT-11

2. I Circuit failure.

INSERT-12 INSERI-10 l

3.

Instrument indicates a downscale failure.--INSERT-12 4

Instrument controls not set in operate mode--DRSERT-12

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(2) j$ry The initial CHANNEL CALIBRATION shall be performed using one or more of the reference standarcs certified by the National Bureau of Standarcs or using stancards that have been obtained from suppliers that participate in measurement assurance activities with NBS.

These stancards shall permit calibrating the system over its intended range T

mos.

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For subsequent CHANNEL CALIBRATION, sources that have ms..

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been related to the initial calibration shall be used.

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.m-CHANNEL CHECK shall consist of verifying indication of flow curing perlocs s,

of release.

CHANNEL CHECK shall be made at least once daily on any day on which continuous, periodic, or batch releases are made.

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. Tea.LE.3.3 4.. (Continued)

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RADIDACTIVE GASEOUS EfflUENI HONITORING INSTRUMENTATION

.e v -

i HINIMUM CllANNELS d4 INSTRUHENI OPERABLE APPLICABillTY ACT10ll h%#.n t.1 Y.n e r V 1 r t.i r T. T K2, t' n T 9 A f*

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Flow Rate Honitor (1) 36 Sampler flow Hate Honitor (1) e.

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CONIAINHLNI PURGE SY$1EH 38 Hohle Gas Activity Honitor INSERT-19 (1) 7 a.

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INSERT

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INSERT - 19 s.

Page 3/4 3-Table 3.3,-gli "providing Alarm and Automatic Termination of Release (RE-22. RE-33. RE-31, and RE-32)"

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s, g os o l!ADl0 ACTIVE GA5l005 [FFLUENT HONITORING INSTRUMENIA110N l

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HINIMUM CllANNELS IN51RUMENI OPERAlllE APPLICABILITY ACTION 4.

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INSERT - 20

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1 "RACWASTE BUILDING VENT" t-l' i

I INSERT - 21 60 it Page 3/4 3,g, Table 3.3-g "Providing Alarm and Automatic Termination of Waste Gas Holdup System Release (RE-10)"

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INSERT-23

" GASEOUS RADWASTE TREATMENT SYSTEM" INSERT-24

" Waste Gas Holdup Tanks" INSERT-25 ACIION 39 - With, no channels OPERABLE, addition of waste gas to the GASEOUS RADWASTE TREATMENT SYSTEM may continue provided a grab sample is taken from the on-service vaste gas decay tank and analyzed once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and the oxygen concentration remains less than 1%.

(; '

l INSERT-26 ACTION 40 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, except for testing, isolate the oxygen supply to the affected recombiner.K "i;h neu ch.....ela 02076 1.0, addition of waste gas.co the GASEOUS RADWASTE TREATMENT SYSTEM may continue provided a grab sample is taken from the on-service vaste gas decay tank and analyzed once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and thg oxygen concentration remains less than 1%.

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,i RADIDACTIVE CASEQUS EFFLUENT H0!!!TORING INSTRUllENTATI0ff SURVEllLANCE REQUIREMENTS a

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gyggAYt 0NA L

'N ANALOG 9

CilANNEL H0 DES IN WillCli r

CllANNEL SOURCE CllANNEL TU::CTi^,;'oi.

SURVEILLANCE INSTRilllENT CilECK CllECK CAllDRATION lEST REQUIREli 1.

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WASTE GAS Il0LDUP SYSTEM EXPLOSIVE o

GAS HONITORING. SYSTEM f

INSERT-27

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PAGE 3/4 3-g k6 I

INSERT-27 i

a.

Inlet Hydrogen Monitor b.

Outlet Hydrogen Monitor l

c.

Inlet Oxygen Monitor d.

Outlet Oxygen Monitor e-e

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pw.

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TABLE,t 3_- M C.,ntinued) l ocE

?0 ItADIOACIIVE(.ASIOUSIFFLUEH1HONIIORINGINSTRUMEllIATIONSURVEILLANCEREjUIREHENTS re u re m

'g ANALOG CilANNEL H0 DES IN WillCll CllANNEL SOURCE CllANNEL

-fHNC;;0;;AL SURVEILLANCE IN51RUHLHI CllECK CilECK CALIBRATION TESI REQUIRED CCNUL NSE n E"AC U *.i:G N S'l51C W

e. Neh!c C;; Acti.ity ;;e..i to.

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N.A.

R Q

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INSERT-28

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c. Particu!:te 5 5pler V

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N.A.

R Q

e. Sampler Flow Rate Monitor D

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INSERT -28 j

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" UNIT VENT" INSERT -29 b3 Page 3/4 3,JW "Providing Alarm (RE-21)"

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8 TABLE 4,3,W(Continued) gg 3

p HAl)l0ACilVI. GAS [005 l~ll LUENI H0tilI0littlG itiSIRilfl[NIAll0N SURVEILLANCE RI.QtilRLilENTS 2,

p NAL

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CllANN[L 50llRCE CllANiiEL

-TU;;CI'0;;AL SURVEILLAtlCE i

Ill51 Rut 1Elli CllECK CNECK CAllBilATION TEST REQUIRt0

3. CONIAltlHEtti PURGL SYSiltl

,6 INSERT-31

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INSERT - 31

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"Providing Alarm and Automatic Termination of Release (RE-22. RE-33. RI-31 and RE-32)"

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TAtkE 4.3;M (Continued) i

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RADIDACIIVE CASEOUS EFFLUENT HONI10 RING INSTRUMENIAll0H SURVEILLANCE REQUIREMENTS l

gggATieMAL AMA LOG CllANNEL H00ES IN WillCll CllANNEt SOURCE CilANNEL f0H6tt0NAt SURVEILLANCE I

INSTRUMEHI

- CllECK CllECK CAllullA110H IEST REQUIRED

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l e us s s su a a ss s uirsp E a su n.H.

t s.,

t a

ra_

rs.

na

?

ga. e sum na tu a suss s tus 58 p.H.

M y

r_

_i___

rs-r_.

na_ _ _ =.

n

^

s.

J uurp g E s a sum nutU I futi l LU 5 EJ H.H.

no y

_ ~. -

o INSERT -32 l

s.

Page 3/4 3 "RADWASTE BUILDING VENT" INSERT - 33

%d Page 3/4 3-ff "Providing Alarm and Automatic Termination of Waste Gas Holdup System Release (RE-10)"

INSERT -34 d

Page3/43g "and P" INS $RT - 35 W

Page 3/4 3.SC

" (1) "

8 1

t i

C S

s T ABLE 4. 3-12'(Conti nued)

TABLE NOTATION

• At all times.

INSERI-36 During ;;;; ; : 5:'du; :;~'-- operation (treatment for primary system

^^

offgases).

INSERT-37 g4tg o p g g A.n eg4 g, I

(1) TheACHANNEL;FL'" T:;aAi TEST shall also demonstrate that automatic isolation of this patnway ehuf control room alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm / trip setpoint. INSERT-38 2.

Circuit failure. INSERT-39 3.

Instrument indicates a downscale failure. INSERT-39 4

Instrument controls not set in operate mode. INSERT-39 AgALOW OfttArieMAL CHANNEL TUNCTICNAL TEST shall also demonstrate that control room (2) TheA A

alarm annunciation occurs if any of the following conditions exists:

1.

Instrument indicates measured levels above the alarm / trip setpoint.

2.

Circuit failure.

C-3.

Instrument indicates a downscale failure.

4.

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 or using standards that have been obtained from suppliers that participate in measurement assurance activities with NB5.

These standards shall permit calibrating the system over its intended range.;f en;rgy.nf u.e a a m c.;n. ren;;.

For subsequent CHANNEL CALIBRATION, sources that have been relatec to the initial calibration shall be used.

'C;; ;;in; ;';nt;

, :;r::i t. : ; ;..;.;,.;te;'..n...',,;..;'.... p.;;;;... fec n.s

_..,.<-.... s (4) The CHANNEL CALIBRATION shal'l include the use of standard gas samples containing a nominal:

1.

One volume percent hydrogen, balance nitrogen, and 2.

Four volume percent hydrogen, balance nitrogen.

(5) The CHANNEL CALIBRATION shall include the.use of standard gas samples containing a nominal:

1.

One volume percent oxygen, balance nitrogen, and 2.

Four volume percent oxygen, balance nitrogen.

INSERT-40 W c.Cr* / cal-g

' ' :~: ;-

3/4 3-

.7---

94 PAGE 3/4 3-J!r" 1

INSERT -36

" GASEOUS RAIMASTE TREATMENT SYSTEM" t

INSERT-37

-"and /or" INSERT-38

"(1 solation and alarm)"

INSERT-39

"(alarm only)"

INSERT-40

" (6 ) The CEANNEL CHECK. 11 consist of verifying that the iodine cartridge and particulate filter are in' stalled in the sample holder."

b I

l I

4 8

i i

PLANT SYSTEMS 3/4.7.5 ULTIMATE HEAT SINK ':

e n ;

r i

LIMITING CONDITION FOR OPERATION 3.7.5 The ultimate heat sink shall be OPERABLE with:

~

s%6.7 FSt*

A minimum water level at or above elevation k^4 Mean Sea Level, j

a.

USGS datum, and b.

An average water temperature of fess than or equal to 4 FF.

APPLICABILITY:

MODES 1, 2, 3 and 4.

ACTION:

With the requirements of the above specificationsnot satisfied, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIRMENTS

,4.7.5 The ultimate heat sink shall be determined OPERABLE at least once per om h by verifying the average water temperature and water level to be within their limits.

l 1

I wc.ss --

t 575 -

3/4 7-13 S3 p I y g.,3 _

l

i o

[

PLANT SYSTEMS to 3/4.7.)4 FIRE SUPPRESSION SYSTEMS FIRE SUPPRESSION WATER SYSTEM LIMITING CONDITION FOR OPERATION i

.f so i

3.7.)(.1 The fire suppression water system shall be OPERABLE with:

[

2 3300

,(Two), fire suppression pumps, each with a capacity of $25889 gpm, a.

with their discharge aligned to the fire suppression header, eee:. i tt. e.T.i ni.T.uT. eentein;d;;ir; pf

)K Sur re'- j'^?{-?;li;3, me-en.nsi c.a anen.s

b. b(

An OPERABLE flow path capable of taking suction from the,

^ ri wn c

u v.eand ^--

transferring the water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrant curb valves, the last valve ahead of the water flow alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each deluge or spray system required to be OPERABLE per Specifications 3.7. M.2, - and 3.7.1A.E.

/O. 4.

10

,p APPLICABILITY:

At all times.

ACTION:

n With one pump and/or one water supply inopgragle, restore the'inoper-a.

able equipment to OPERABLE status within t= caps or, in lieu of any other report required by Specification 6.9.1, prepare and submit a Special Report to the Commission pursuant to Specification 6.3.2 s

withinthenextSSEEEEsoutliningtheplansandprocedurestobe used to restore the inoperable equipment to OPERABLE status or to provide an alternate. backup pump or supply.

The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.

b.

With the fire suppression water system otherwise inoperable:

1.

Establish a backup fire suppression water system within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 2.

'In lieu of any other report required by Specification 6.9.1, submit a Special Report in accordance with Specification 6.9.2:

l a)

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

l b)

Confirmed by telegraph, mailgram or facsimile transmission no later than the first working day following the event, g

i and 1

I to t.Gs r -

Il h' :TS '

3/4 7-29

""" ? * "M5L.

l 4

l 4

- PLANT 5YSTEMS ACTION:

(Continued)

2. -e4 4 c)

In writing within u following the event, outlining the

=y:

action taken, the cause of the inoperability and the plans and schedule for restoring the system to OPERABLE status.

SURVEILLANCE REQUIREMENTS I

to 4.7. M.l.1 The fire suppression water system shall be demonstrated OPERABLE:

a ew mar i ave e is 2. s oir pr.

a.

At least once ;:

;* by verifying ^ u - "" : "

L --

m.

w b.

At least once per 9

=y--

7 r - --- n r t by starting eestirac electric motor driven pump and operating it for at least 15 minutes on recirculation flow.

-m c.

At least once per ^'

'y5 by verifying that each valve (manual, power operated or automatic) in the flow path is in its correct position.

'*'*" ^*"'

/At least ---- = ?""-L by performance of a 9 ---- T M sine man, ewi4.

d.

- a K

At 1:::t One p;r 12 enth: by cyclir.; :;.ch t:stc.ble ;;1v: i.. ;..%

fl:a p;th thr;;;5 at 1:::t :.. :,pict: j1

f ful' tr

;:1.

nsusu 4

c. A At least once per h, m. by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

1.

Verifying that each automatic valve in the flow path actuates to its correct position, 33oo 2.

Verifying that each pump develops at least $28009 gpm at a system head of $298$. feet, wt turnsw 3.

Cycling each, valve in the flow path teshsnehA --

L m

-f =:'

--p-M '

through at least one complete cycle of full travel, and 4.

Verifying that each fire suppression pump starts (sequentially) to maintain the fire suppression water system pressure greater than or equal to 80 psig.

4, %

At leort once per 3 years by performing a flow test of the system in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14th Edition, published by the National Fire Protection Association.

C u.> c c.i,: ~3

' '.;T 3 -

3/4 7-30 Y !5 1980

= - -

~~

• PLANT SYSTEMS SURVEILLANCE REOUIREMENTS (Continued) to

4. 7. M. l. 2 The fire pump diesel engine shall be demonstrated OPERABLE:

m w ns a.

At least once per ^' 2 ;= by verifying:

a.som.is nr umr W m.a. avaa.

1.

The fuel storage tank y ^ ' - __

c-and s si e t

2.

The diesel, starts from ambient conditions and operates for at least 30 minutes on recirculation flow.

game.

b.

At least once per ^^ : y= by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-65, is within the acceptable limits specified in Table 1 of ASTM 0975-74 when checked for viscosity, water and sediment.

rus % a u x At least once per M, 2-- ' ; ' ---. by subjecting the diesel c.

to an inspection in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

to

4. 7. M. l. 3 The fire pump diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE:

wsu.

a.

At least once per W by verifying that:

1.

The eTectrolyte level of each battery is above the plates, and s

2.

The overall battery voltage is greater than or equal to 24 volts.

<a nsu b.

At least once per ^^ f-i by verifying that the specific gravi.ty is appropriate for continued service of the battery.

esmawes At least once per " -- "- by verifying that:

c.

Th' batteries, cell plates and battery racks show no visual 1.

e indication of physical damage or abnomal deterioration, and 2.

The battery-to-battery and terminal connections are clean, tight, free of corrosion and coated with anti-corrosion material.

1

\\o c.ud m&T-S-

• 3/4 7-31

' nUY I C I550 -

1_

{

I t

t ELECTRICAL POWER SYSTEMS W

3/4.8./1 ONSITE POWER DISTRIBUTION W smes OpetststetG A. c. b15710 8artons - oPERATwG.

LIMITING CONDITION FOR OPERATION 3.8.$.1 The following electrical busses shall be energized " th: :p::i'i:d maeece with tie breakers open C n::L between redundant busses:#'

(_ > ' _.--_.

r-na'

• f:f r "I ^.C. E cr;;ng C[nn ::n;i: ting Of:

/a.

i I

4160) Volt Emergency Bus # N 8os.

/

Volt Emergencyg # go os., ac,os, we.or E b.

Division #

C. Emergency Busses consistin 1.

(4160) vol mergency Bus #

2.

(480) volt Em ency Bus #

c.' ~ (120) volt A.C. Vital dnergizedfromitsassociated SEE Remnus iriverter connected to D..

Pm.e.

d.

-(120) volt A.C. Vital B energizedfromitsassociated

. inverter connected t

.C. Bus # N 1

e.

(120) volt A.C.

' cal Bus #

en ized from its associated inverter con ed to D.C. Bus #

l f.

(120) vo A.C. Vital Bus #

energize om its associated O

inver r connected to D.C. Bus #

h g.

0/125) volt D.C. Bus #1 energized from Battery k #1.

(250/125) volt D.C. Bus #2 energized from Battery Bank APPLICABILITY: MODES 1, 2, 3, and 4.

4 ACTION:

4,sss ma ma eswa ews.enswr oe mi.o m.4 no A.c. 6 mas a'

With =

L-

'"'E-

-- ' i

~

tuffy. energized, re'-energize the# M 'wt h N houriod be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within I

the followin0 v30 hours.

na vo l

b.

With one,A.C. Xital s h not energized from its associated 3,.

nyerter, or with' inverter not connected to its associated D.C. f 4usaM-usr, (MI re-energize the A.C. Vital 4us+within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at

• • "^

• E least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLDJ g DOWN i

within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> i

Eus from its assccinted inver? asn6m4d4 re energize the A.C. fital ter connected to ts associated D.C.

hBus within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in et lent HOT ST"iNDGY within the next l

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the fellowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

WithoneD.C.BusnotgergizedfromitsassociatedIatterySnas, c.

re-energize =-- :.1 - s from its associated %ttery ennt. witkin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

. A x

m s.aaec inverterx may be disconnected from therD.C. Bw for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1sse i

n- _:n for the purpose of performing an eq associated battery las** provided,f.1) -46mRe vital busame =,u,aggghgo, l ana ) the vital busses associated with the other battery banks arTfM8I'fjiYd.k;; thcir i 2*N ";;rt:r: r d c: m erted te their u n cini.ed C.C. R ' 7 3/48-810 WCM

j kerno.s 9n E i t C ~3/4. 8 I ce stTf. PbwER., bisTT.s6u m 4 h3 EMS a

3. 8. 7.. I a..

Mt6o deur Lau s-c. hs

• aao1 4 60 V.e Leus,ac t h ies W 46 e i, 4G es, Sm M 6 of IF.

4 12 o Q.wr A.C. Vi ms. 8%ssas

• N N. s n o NN.3 hatusee ma.m,,, wen.raa. N g gg

,,.. gu is, 11 7 dost b. C.. Mims 6%gg gg i N M.o t Awt MKo3 e-suise Fum cer:11R,tEs NK it a,a l N K 13. i i C Ae.eb b. 4 g i,o sf.e g ,w e.4 hs

• M 6 o 2.

4 Bo Vest haAwe.s b,e.S

• N G o2., MC, og, o?

MG ohE 12.0 Vowr A.C. Vims. 8%ues.

• NMat ** N4
• 9 pau.t%E6 F4ow saw&E TEd. MM 11 h4 M N 16(.

iz.r V.sr o.c. s,s., ax.4 a-o a w.s u ui.4 l ...,u es ai<.ss oic.it. l i 1

i 6 l ,e ' ELECTRICAL POWER SYSTEMS SURVEILLANCE REQUIREMENTS 2. osernaret CPERAME 4.8.Y.l yhe specified buss _gshall be i u-- "--7 3u-least once per - _, by verifying correct breaker alignment and ---m indicated voltage on the busses. I l t c e c. g w e.w ' ' ^> 61 'M 3/4 8,2( -- ' ; J.' ^. cot g_

j. ? ELECTRICAL POWER SYSTEMS A. C.. N DISTRIBUTION - hTb ow,J %mmenna LIMITING CONDITION FOR OPERATION

3. 8.k 2 As a minimum, the following electrical busses shall be energi, zed

- - t '.. g ; ; i f I--_ -- ~ c.- ( a. division of A.C. Emergency Busses c. ing of one (4160) volt l and one volt A.C. Emergenc ~ .SEE NG ## b. Two (120) volt A.C. ergized from their associated PpE inverters co d to their respecti Busses. ( c. M 50/125) volt D.C. Bus energized from its associatedh bank. APPLICABILITY: MODES 5 and 6. ACTION: een m m u m m ar *F With asp =f the above r--, ' _2 f __- : M busses andr energized L f' _ J - ---f m immediately suspend all operations involving CORE ALTERATIONS, positive reactivity changes, or movement of irradiated fuel, initiate corrective action to energize the required electrical busses in the specified manner as soon as possible, and within 8 hours depressurize and vent the RCS through a 4 23. square inch vent. 2 SURVEILLANCE REQUIREMENTS z. 4.8.1.2 The specified buss shall be determined energized L-t'.: _f--' esmer at least 'once per by verifying correct breaker alignment and indicated voltage on the busses. 1 4 l O " " ~ - "-:T b 3/48% =

i t her_s Pws. 3.b.1.1. l-A.

• {140 Vot:t*

6'.. a s se,,,,m S%g 3 MSol 1 AlBb Vest 6maat s.ec at 6%ut5 # NG eI N G o s, A,-+ 4 Gore 110 \\)owT A-C. Vi ewt Bassas

• nmo i, a~o Ndos G.,e G,M. 4%ED FAe w r ewSAfSAS-NeJ 11 Awt MM 43 12,7 dost D.C.

b SE1 N K.* % A e NW.01 s,.sas ig Eb S1 S AtTERA 1lLS M M 11 A a-o M K.t3, O R. b. 44 g so s).sr Leeso cy be NS o g,, 4B0 Vebt be a4e c. v hsa.s

• MG.s, gs.q, g4,, g gt,,o Q.wr A. C. d s r% %.

OMS %ES % Md *L A-6 N4OM b e at.411r6 6 fEfto% 3 otA3'ER,1 MM st Ae 44 sq. )W Vo'T b.c.. % >as u g.e t, e,. gg g,, e's ee rmosas M K.s t. a,a N w, p.t. i l l e I i C s

r 4 t ( RADI0 ACTIVE EFFLUENTS 3/4.11.3 50L: RA;;;a n;.i -,i.g INSERT-77 LIMITING CONDITION FOR OPERATION INSERT-78 3.11.3 The l,J.,. a en.,, n:.. shall be OPERABLE and used, sp:1!::t!: ' r. n: efth : ""CCE!! CCF'"0L "".CC"J.", for the SOLIDIFICATION and packaging of radioactive wastes to ensure meeting the requirements of 10 CFR Part 20 anc of 10 CFR Part 71 prior to shipment of radioactive wastes from the site. APPLICABILITY: At all times. ACTION: a. With the packaging requirements of 10 CFR Part 20 and/or 10 CFR Part 71 not satisfied, suspend shipments of defectively packaged soli.d radioactive wastes from the site. TNcroT_79 b. With the k::it r:d :: : :y;t;; inoperable for more than 31 days, in lieu of any other report required by Specification 6.9.1, prepare ans / submit to the Commission within 30 days pursuant to Specification 6.9.2 a Special Report which includes the following information: [ 1. Identification of the inoperable equipment or subsystems and the reason for inoperability, 2. Action (s) taken to restore the inoperable equipment to OPERABLE status, 3. A descripti'on of the alternative used for SOLIDIFCATION and I packaging of radioact.ive wastes, and 4. Summary oescription of action (s) taken to prevent a recurrence. c. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS INSERT-78 4.11.3.1 Theb:1!: :.:::: :,: e sna11 be demonstrated OPERABLE at least once per 92 cays by: a. Operating the solid radwaste system at least once in the previous 92 days 4a a:---d ne: t.: "RO ::: 00,t7n;; p,;;;;;;;, or i-- b. Verification of the existence of a valid contract for SOLICIFCATION to be performed by a contractor.'

rt;.= ui:r. : ?R0;;;; CONTECL T RO G RA".

2-0T:4 3/411-g e

t t PAGE 3/4 11-18 INSERT-77 " RAD'n'ASTE SOLIDIFICATION" INSERT-78 "radwaste solidification syste=" i l 1 1

t

7.,

4 RADI0 ACTIVE EFFLUENTS s. SURVEILLANCE REOUIREMENTS (Continued) i 4.11.3.2 THE FR^ 55 CCNTROL FR00RA!'. ; hall b; u;;d :: erify che SOLIDIFICA-

• ' ION of at least one representative test specimen from at least every tenth i

batch of each type of wet radioactive waste (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions, and sodium sulfate solutions). I a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICA-TION of the batch under test shall be suspended until such time as additional test specimens can be obtained alternative SOLIDIFICATION parameters can be determined '

rt:r.:: eith th: "ROCE!! CO PROL

~ PR0;R*,"., and a subsequent test verifies SOLIDIFICATION. SOLIDIFCATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters d:: rr.ir.:d by th: PROC 00 CONTROL PR00RA.".. b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the FRC C',; ^^S!E^ FR G L na;. e L. -.a s g.e. collection and testing of representative test specimens from each consecutive batch of the sa'me type of wet wastelunts) at least 3 consecutive initial test specimens demonstrate SOLIDIFCATION. -h e-PROCESS CONTROL PROCRA". sh:11 b; r. difi:d :: r:quir:d, :: pr:vic:d S;::i'ic: tier 5.'.2, t: :::ur: SOLICIFICATIC" Of : b;cque.-t batches-( w. -ww==. INSERT-79 s 9 F' ' T '. - ; - 3/411-g

..._u. t ,.i 4.. INSERT - 79 Page 3/4 11-19 "shall be conducted" l i l O b e h a 9 h, s k f 1 I i I t

RADIOLOGICAL TECHNICAL SPECIFICATIONS 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING l 3/4.12.1 MONITORING PROGRAM LIMITING CONDITION YOR OPERATION 3.12.1 The radiological environmental monitoring program shall I be conducted as specified in the Radiological Environmental Sampling Schedule in the ODCM. APPL"CABILITY: At all times. ACTION: a. With the radiological environmental monitoring program not being conducted as specified in the ODCM, pre-pare and submit to the Commission, in the Annual Radio-logical Operating Report, a description of the reasons i for not conducting the program as required and the plans for preventing a recurrence. b. With a confirmed measured concentration of radioactiv-l ity in an environmental sampling medium exceeding the reporting levels of Table 3.12-1 when averag'ed over any l calendar quarter, prepare and submit to the Commission within 30 days from the end of the affected calendar quarter a Report pursuant to 6.9.1.13. When more than one of the radionuclides in Table 3.12-1 are detected l l in the sampling medium this report shall be submitted. l if: I concentration (1) c'oncentration (2) +' +. . E 1.0 l. limit level (1) limit level (2) When radionuclides other than those in Table'3.12-1 l are detected and are confirmed as the result of plant effluents, this report shall be submitted if the poten-l tial annual dose to an individual is equal to or { greater than the calendar year limits of Specifications 3.11.1.2, 3.11.2.2, and 3.11.2.3. This report is not i required if the measured level of radioactivity was not l the result of plant effluents; however, in such an j event, the condition shall be reported and described in the Annual' Radiological Environmental Operating Report. c. With milk or fresh leafy vegetable samples unavailable from one or more of the sample locations, identify l -in the Annual Radiological Environmental Operating i Report the cause of the unavailability of samples and the locations for obtaining replacement samples. The locations from which samples were unavailable may J07-13-81 WCGS 3/4 12-1 { ~_-

m q r - y ( L RADIOLOGICAL ENVIRONMENTAL MONITORING LIMITING CONDITION FOR OPERATION (Continued) then be deleted from those in the sampling schedule in [ the ODCM, provided that the locations from Md.ch the t replacement samples were obtained are added to the A j environmental monitoring program as replacement locations. + This does not apply to locations from which samples are occasionally unavailable on a temporary basis. d. The provisions of Specifications '3.0.3 and 3.0.4 are not applicable., .m c s. \\ SURVEILLANCE REQUIREMENTS + 4.12.1 The radiological environmental monitoring samples shall be collected and analyzed in accordance with the Radiological' Environmental Sampling Schedule in the ODCM. N /,', '4 3 / { 6 i

f

./ g s ( .w s' l s 'a 'c y x J' i g \\ s<yg l ') [ } s i t t. / 1 i j \\ s j, t 'N if ' j' y < t e s 'q t, y,,, ( y ,i, / ,T t l O ' f, s 4 WCGS 3/4'12-2 07-13-81 i

I ~ 5

C s.O "Is-TABIE 3.12-1 l

's 1 N REPORTIfC IEVEIS FOR RADIGACf1VITY ODiGNfRATIOtB IN EtNIROtNENTAL SAMPLES 7 Reporting IcVels Water Airbome Particulgte Fish Milk Food Pro.lucts Analysis (pCi/1) __ or Gases (pCi/m ) (pCi/kg, wet) (pCi/1) (pCi/kg, wet) H-3 2 x 10 4 Mn-54 1 x 10 3 x 10 Fe-59 4 x 10 1 x 10 W 4 2 Co-58 1 x 10 3 x 10 e Co-60 3 x 10 1 x 10 w 8 2 4 W Zn-65 3 x 10 2 x 10 Zr + 95 4 x 10 l I-131 2 0.9 3 1 x 10 Cs-134 30 10 1 x 10 60 1 x 10 Cs-137 50 20 2 x 10 70 2 x 10 Ba-La-140 2 x 10 3 x 10 For drinking water samples. This is 40 CFR Part 141 value. o Y

• • Total for parent and daughter. t l

r W 2

[ i ~ o* TABIE 4.12-1 i MAXIMM VALUES EDR TIE IIXER LIMITS OF DETECTION (LLD)a,c 'I Water Airborne Particulgte Fish Milk Food Products Sedimant l Analysis (pCi/1) or Gas (pCi/m ) (pCi/kg, wet) (pCi/1) (pCi/kg, wet) (pCi/kg, dry) f -2 { gross beta 4 1 x 10 .l H 2000(1000 ) 54 Mn 15 130 59Fe 30 260 8,60Co 15 130 0 65 e Zn 30 260 w i E S

i.

l Zr 30' 95 s g g 131 ~ I 1 7 x 10 1 60 1 ~ 130 15 60 150 l Cs 15(10 ) 5 x 10 ~ Cs 18 6 x 10 150 18 80 180 140 Ba 60 60 140La 15 15 O a H .r- -.m.m., i==i

• P-6% er'+4 GN N'W *DPMT*P W*TWW+'W'**""**

f

TABLE 4.12-1 (Continued) TABLE NOTATION a. The LLD is the smallest concentration of radioactive material in a sample that will be detected with 95 percent probability with 5 percent probability of falsely concluding that a blank observation represents a "real" signal. For a particular me43urement system (which may include radio-chemical separation): 4.66s LD= (E) (V) (2.22) (Y [exp(- Aa t)] l Where: 4.66 is the standard deviation for a 95% confidence level. I LLD is the lower limit of detection as defined above (as l picocurie per unit mass or volume), s is the standard deviation of the background counting rNte or of the counting rate of a blank sample as appro-priate (as counts per minute). E is the counting efficiency (as counts per transformation), V is the sample size (in units of mass or volume), 2.22 is the number of transformations per minute per picocurie, Y is the fractional radiochemical yield (when applicable), A is the radioactive decay constant for the particular radionuclide, and A t,is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples). i The value of s used in the calculation of the LLD for a detection systNm shall be based on the actual observed variance of the background counting rate or of the counting rate of the blank samples (as appropriate) rather than on an unverified theoretically predicted variance. In calculating the LLD for a radionuclide determined by gamma-ray spectrometry, the background shall include the typical contributions of other radionuclides normally present in the samples (e.g., potassium-40 in milk samples). Typical values of E, V, Y, and A t shall be used in the calculations. l i WCGS 3/4 12-5 07-13-81 l

TABLE 4.12-1 (Continued) TABLE NOTATION b. LLD for drinking water. c. Other peaks which are measurable and identifiable, together with the radionuclides in Table 4.12-1, shall be identified and reported. O O l l l f WCGS 3/4 12-6 07-13-81

t 3 j I RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.2 LAND USE CENSUS LIMITING CONDITION FOR OPERATION 3.12.2 A land use census shall be conducted and shall identify the location of the nearest milk animal, the nearest residence, and the nearest garden

• of greater than 500 square feet producing fresh leafy vegetables in each of the 16 meteorological sectors within a distance of 5 miles.

l APPLICABILITY: At all times. ACTION: a. With a land use census identifying a location (s) which yields a calculated dose or dose commitment greater than the values currently being calculated in Specifica-tion 4.11.2.3, identify the new location (s) in the Annual Radiological Environmental Operating Report. b. With a land use census identifying a location (s) which yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from I which samples are currently being obtained in accordance l with Specification 3.12.1, identify the new location (s) in the Annual Radiological Environmental Operating Report. The new location (s) shall be added to the radiologi-cal environmental monitoring program within 30 days. The sampling location (s), excluding the control station location, having the lowest calculated dose or dose commitment (via the same exposure pathway) may be deleted from this monitoring program af ter (October 31) of the yea,r in which this land use census was conducted. c. The provisions of Specification 3.0.3 and 3.0.4 are not applicable. r l

• Broad leaf vegetation sampling may be performed at the site boundary in the sector with the highest D/Q in lieu of the garden census.

WCGS 3/4 12-7 07-13-81 l l L

_\\ ? r 3/4.12.2 LAND USE CENSUS i LIMITING CONDITION FOR OPERATION i SURVEILLANCE REQUIREMENTS 4.12.2 The land use census shall be conducted at least once per year between the dates of June 1 and October 1 using a method which provides the best results, such as a door-to-door survey, aerial survey, or by consulting local agricultural authorities. / I l WCGS 3/4 12-8 07-13-81 i

a [ RADIOLOGICAL ENVIRONMENTAL MONITORING f 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM LIMITING CONDITION FOR OPERATION 1 3.12.3 Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program which has been approved by the Commission. APPLICABILITY: At all times. ACTION: a. With analyses not being performed as required above, report the corrective actions taken to prevent a recur-rence to the Commission in the Annual Radiological Environmental Operating Report. b. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.12.3 A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report. O h I l WCGS 3/4 12-9 07-13-81 - * ~ ~ ~ ~

INSTRUMENTATION BASES b 3/4.3.3.7 CHLORINE DETECTION SYSTEMS s The OPERABILITY of the chlorine detection system ensures that sufficient capability is available to promptly detect and initiate protecti'e action in v 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, " Protection of Nuclear Powe Plant Control Room Operators Against an Accidental Chlorine Release," February (97.5. 1 3/4.3.3. FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability.is required in order to detect and locate fires in their early stages. Prompt detection of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program. h In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas ( is required to provi,de detection capability until the inoperable instrumentation I is restored to OPERABILITY. f/ OOSE-PART DETECTION INSTRUMENTATION The OPERABILITY loose par tection instrumentation ensures that sufficient capability is ava o detect loose metallic parts in the primary system and avoid or gat mage to primary system components. The allowable out-of-service mes and survei e requirements are consistent with the recommendat s of Regulatory Guide 1. " Loose-Part Detection i Program for the mary System of Light-Water-Cooled rs," May 1981. 3 TURBINE OVERSPEED PROTECTION This specificat s provid o ensure that the turbine overspeed protection instrumentation e turbine speed control valves are OPERABLE and will protect the tur from ssive overspeed. Protection from turbine excessive overspeed equired since ex ve overspeed of the turbine could generate potenti damaging missiles which cot act and damage safety related comp nts, equipment, or structures. w e.GS - ' STC - B 3/4 3-4 d .-u .~i.

7 _. l i I f^- REACTOR COOLANT SYSTEM %/ i BASES (c.e,W b OPERATIONAL LEAKAGE (,C _ im m ) [ f PRESSURE BOUNDARY LEAKAGE of any magnitude is unacceptable since it may [ be indicative of an impending gross fai. lure of the pressure bound.ary. Therefore, the presence of any PRESSURE BOUNDARY LEAKAGE requires the unit to be promptly placed in COLD SHUTOOWN. 3/4.4.7 CHEMISTRY ~ The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits', for the specified limited time intervals without having a significant effect on p the structural integrity of the Reactor Coolant System. The time interval ri., permittin'g continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concen-trations to within the Steady State Limits. The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action. 3/4.4.8 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting.2-hour doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that a specific site parameters of the ( ) +&te., such as site boundary location and meteorological conditions, were ot considered in this evaluation. S O on wp c. W. Cmev c.b3 r= w e.u ' :T3 8 3/4 4-5 %CY 2 0 C ~ i

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i 4 3/a.11.2.2 DOSE. HOBLE GASES This specification is provided to implement the requirements of Sections II.S. I*I.A and IV.A of Appendix I, 10 CFR Part 50..The Limiting Condition for Oseration implements the guides set forth in Section II.B of Appendix 1. The ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix I to assure tnat the releases of radioactive material in gaseous effluents will be kept- - "as low as is reasonably achievable". The Surveill,ance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be shown by calculational procedures based on models and cata such that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated. The dose calculations established in the 00CM for calculating the doses due to the actual release rates of [*) radioactive noble gases in gaseous effluents are consistent with the methodology provided in f.egulatory Guide 1.109, " Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory, f Guide 1.111. " Methods for Estimating Atmospheric Transport and Dispersion o Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors," Revision 1, July 1977. ~ '. ; 0"CM :;_;t':n: pr:vid:d f:r ::t:r ' 'm; th: ci-

et tne ;ite b;.nd;ry er; b;;;d -pen in; ri;;;ri;;l..;r;;; ;;.;;;pn;ci; 0

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2/4.11.2.3 DOSE. RADIO!ODINES. RADIOACTIVE MATERIALS IN PARTICULATE FORM AND RADIONUCtIDES OTHER THAN NOSLE GASES This specification is provided to imolement the recuirements of Sections II.C. .II.A and IV.A of Appenoix I, 10 CFR Part 50. The Lim.iting Condition for Opera-tien are the guides set forth in Section II.C of Appendix 1. The ACTION state-cents provice the required operating flexibility and at the same time implement tne guides set forth in Section IV.A of Appendix I to assure that the releases of radioactive materials in gaseous effluents will be kept "as low as is easonably acnievable". The 00CM calculational methocs specified in the Surveillance Requirements implement the requirements in Section III.A of Appendix I that conformance with the guides of Appendix I be snown by calcula-tional procedures based on models and data, such snat the actual exposure of an individual tnrough appropriate pathways is unlikely to be substantially underestimated. The 00CM calculational methods for calculating the doses due to the actual release rates of tne subject materials are consistent witn the ([') methodology provided in Regulatory Guide 1.109, " Calculation of Annual Doses u C45 w R - 3 5 - r-5 3/4 11-3

t l RADICACTIVE EFFLUENTS i BASES l to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Ccmpliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Of spersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," Revision 1, July 1977. Tt :: : :ti: :.:1:: pr:vid: f:r :::r ' 'n;

n: :::::1 ::::: b:::: ;;;r th

'f:::rf::1 : :r:;: t ::ph:r': ::nditi:n:. T,e release rate specifications for radioiodines, radioactive materials in L"ticulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways .ni.:n were examined in the development of these calculations were:

1) individual innalation of airborne radionuclides, 2) deposition of radionuclides onto

- ein leafy vegetation with subsequent consumption by man, 3) deposition onto s gesssy areas where milk animals and meat producing animals graze with consump-tion of the milk and meat by man, and 4) deposition on the ground with subsequent exposure of man. 1/4.11.2.4 CASE 005 RADVASTE TREATMENT O The OPERABILITY of the gaseous racwaste treatment system and the ventila-I tion exhaust treatment system ensures that the systems'will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement tnat the appropriate portions of tnese systems be used, when specified, provides reasonable assurance that the releases of radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable". This specification implements the requirements of 10 CFR Part 50.36a, General Cesign Criterion 60 of Appendix A to 10 CFR Part 50, and the casign objectives given in Section II.0 of Appendix ! to 10 CFR Part 50. The specified limits governing the use of appropriate portions of tne systems were 1,)ecified as a suitable fraction of the dose design objectives set forth in Sections II.B anc II.C of Appendix I,10 CFR Part 50, for gaseous effluents.

• /a.11.2.5 EXPLOSIVE GAS MIXTURE This specification is provided to ensure that the concentration of cotentially explosive gas mixtures contained in the waste gas holduo system is,

saintained below the flammacility limits of in,;..;s.. ,3so. sa.... 4.. n iINSERT-86

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N: :- : '.t ';:.) Maintaining the concentration ,f nydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with g the requirements of General Design Criterion 60 of Appancix A to 10 CFR Part 50. r 1 vo c.65 '0 '!. L B 3/4 11-4 i l

t 2 6 O INSERT - 86 i f Page B 3/4 11-4, 3/4.'11.2.5 "a hydrogen and oxygen gas mixture" t G 1 l t I I I i e _.1

P 2 i I ) Ra0!OACTIVE EFFLUENTS j t BASES + h 3/4.11.2.6 GAS STORAGE TANKS I Restricting the quantity of radioactivity contained in each gas storage tsnk provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to an individual at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan.13.7.1, "Wasta Gas System Failure". s. 3/4.11.3 SOLID RADI0 ACTIVE WASTE

• The OPERABILITY of the solid radwaste system ensures that the system will be available for use whenever solid radwastes require processing and packaging p.-ior to being shipped offsite.

This specification implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters 'n:! d:d '

j ':r'n; th: "20:55! CCP~'C'.

'70C17". may include, but are not limited to waste type, waste pH, waste / liquid / solidification ageng/ catalyst ratios, waste oil content, waste principal gag enemical constituents, mixing and curing times. 3/A.11.4 TOTAL 00SE This specification is provided to meet the dose limitations of An CFR 190. Tne specification requires the preparation and submittal of a Special Report -nenever the calculated doses from plant radioactive ef fluents exceed twice the cesign objective doses of Appendix.I. For sites containing uo to a reactors. it is n'ghly unlikely that the resultant dose tol: n;; :; c. will exceeo 40 CFR 190 if the individual reactors remain within the reporting requirement level. The Special Report will cescribe a course of action wnich should result l in the limitation of cose tol; r;;. nci.ic.;'. for 12 consecutive montns to within the 40 CFR 190 limits. For the so-noses of the Soecial Reoort. it mav i l te assumed that the dose commitment to kr.;.....~... .; from other uranium i. fuel cycle sources is negligible, with the exception that cose contributions ( from other nuclear fuel cycle facilities at the same site o'r within a radius cf 5 miles must be considerea. INSERT-88 e INSERT-87 l i u c (.s. l '"' 0 7 B 3/ 11-5 _.._.,...)

t PACE B 3/4 11-5 INSERT-8 7 "a member of the public" INSERT-88 "A member of the public is considered an individual that works and/or resides outside of the plant restricted area. An individual is not considered a member of the public during any period in which he/she is engaged in any nuclear fuel cycle activities." b i lO i i ..,_.%4 .w,.w.- -.***ew""'Nar*'

.-~ t 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES 3/4.12.1 MONITORING PROGRAM The radiological monitoring program required by this Specification provides measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides which lead to I the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby sup-plements the radiological effluent monitoring program by verifiying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of the environmental expo-sure pathways. The initially specified monitoring program will be effective for at least the first 3 years of commercial operation. Following this period, program changes may be initiated based on operational experience. The detection capabilities required by Table 4.12-1 are state-of-the-art for routine environmental measurements in industrial laboratories. The LLDs for drinking water meet the requirements of 40 CFR 141. 3/4.12.2 LAND USE CENSUS This Specification is provided to ensure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are made if required by the results of this census. The best survey information from door-to-door surveys, aerial surveys, consulting with local agricultural authorities, or other methods shall be used. This census satisfies the I requirements of Section IV.B.3 of Appendix I to 10 CFR Part 50. Restricting the census to gardens of greater than 500 square feet 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 consump-tion by a child. To determine this minimum garden size, the l following assumptions were used, 1) that 20 percent of the garden l was used for growing broad leaf vegetation (i.e., similar to lettuce and cabbage), and 2) a vegetation yield of 2 kg/ square meter. WCGS B 3/4 12-1 07-13-81 __J

l f 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING BASES (Continued) 3/4.12.3 INTERLABORATORY COMPARISON PROGRAM The requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid. 9 F 9 WCGS B 3/4 12-2 13-81 i

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..m__ -.-. - - = - _. - MSE u) C GS ~l 2 PI%E 5 -(, Olloc.g b5 Q40 e, N a h DESIGN FEATURES iL c;. 79, Tp

5. 3 REACTOR CORE

'S' I i FUEL ASSEMBLIES ~ ~ 5.3.1 The reactor core shall contain 193 fuel assemblies with each fuel k assembly containing A 4 fuel rods clad with (Zircaloy -4). Each fuel rod shall have a nominal active fuel length of 143.7 inches and contain a maximum l total weight of v14% grams uranium. The initial core loading shall have a maximum enrichment of Lt weight percent U-235. Reload fuel shall be similar. in physical design to the initial core loading and shall have a maximum enrichment ~of 3.5 weight percent U-235. CONTR0t pnn qsst."3LI:; .3.2 The reactor core shall contain 53 full length aM _ pa"t t..gth rol rod assemblies. hThe iuli len grr control rod ' emb!400-sha contain co a no. nal 142 inches of absorber material. The art S$3*9N3 t ra? accomh iac shall enntain a n0:-inal 30 -inch 0; Of ;b-n"ber e teria [r^at their t.!:r erh. The nominei voiues of absuiLc. m&tuiici shdii [bc Y[ne"ca 2 cilue", e-cent %dier and 5 percent cadmiur. All cent-s-sha44-be-clad m u. se \\rh -.ic== steei i.ub i us. The Lolauwe of the void ivuwih a ine parc o le ;th *Mr 11 c mt:in aluminar cyide. o n 5.4 REACTOR C00LAN SYSTEM DESIGN PRESSURE AND T RATURE 5.4.1 The reactor coolant ' stem is designe and shall be maintained: a. In accordance with th code re irements specified in Section (5.2) of the FSAR, with allow nce f r normal degradation pursuant to the applicable Surveillance q rements, b. For a pressure of p ig, nd c. For a temperature of F, ex pt for. the pressurizer which is

• F.

VOLUME 5.4.2 The total water nd steam volume of the reac r coolant system is + cubic feet at nominal T,yg of (525) F. j 5.5 METEOROLOGI AL TOWER LOCATION 5.5.1 The m eorological tower shal' be located as shown on 'gure (5.1-1). i wcus l b I / l J A_ 5- ,7 _ S^N -

.~ ~ -... l r DESIGN FEATURES I i l 5.3 REACTOR CORE PQEL ASSEMBLIES 4 5.3.1 reactor core shall contain fuel assemblies each '^ comp d of a 17 X 17 s array with 264 fuel rods clad in Zircaloy The r position in the fuel assembly is reserved fo ore instrumentation. The remaining 24 positi in the. assembly have guide thimbles for the r uster control as lies (RCCA's). Each fuel rod shall have a nominal active fuel noth of 144 inches. CONTROL ROD ASSEMBLIES 5.3.2 The reactor core shall contain 53 full length control rod assemblies. A control rod assembly shall contain 24 control rods.each with a nominal 142 inches of absorber material. Each control rod consists of hafnium clad in stainless steel. 5.4 REACTOR COOLANT SYSTEM PARAMETERS PRESSURE AND TEMPERATURE 1 i 5.4.1 The reactor coolant system is designed and shall be maintained: In accordance with the code requirements specified a. in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance l Requirements, b. For a pressure of 2485_ psig, and For a temperature of 6_50F, except for the pressurizer c. which is 680F. VOLUME 5.4.2 a. The total system liquid volume including pressurizer water at maximum guaranteed power is 11,393 cubic ft. i b. The total system volume, including pressurizer and surge line is 12,135 cubic ft. g ~ 5.5 METEOROLOGICAL TOWER LOCATION ,("}5.5.1 The meteorological tower shall be located as shown on Figure 5.1-1. i WCGS 5-6 u

t 1-INDEX t ADMINISTRATIVE CONTROLS I I SECTION PAGE 6.1 RESPONSIBILITY 6-1 i 6.2 ORGANIZATION 6.2.1 OFFSITE 6-1 6.2.2 UNIT STAFF 6-1 6.2.3 SHIFT TECHNICAL ADVISOR. 6-5 6.3 UNIT STAFF OUALIFICATIONS 6-5 l 6.4 TRAINING. 6-5 6.5 REVIEW AND AUDIT 6.5.1 PLANT SAFETY REVIEW COMMITTEE Function 6-6 Composition. 6-6 Alternates 6-6 Meeting Frequency. 6-7 Quorum. 6-7 Responsibilities. 6-7 Authority. 6-8 Records. 6-8 6.5.2 NUCLEAR SAFETY REVIEW COMMITTEE Function. 6-8 Composition. 6-9 Alternates 6-9 Consultants. 6-9 Meeting Frequency. 6-9 Quorum. 6-10 Review. 6-10 Audits 6-11 Authority. 6-12 Records. 6-12 6.6 REPORTABLE OCCURRENCE ACTION. 6-12 6.7 SAFETY LIMIT VIOLATION. 6-12 E e WCGS I 4-

_ ~.. - - - - - = .- -.,.=,. . s' t INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.8 PROCEDURES AND PROGRAMS 6-13 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6-16 Startup Report 6-16 Annual Reports. 6-17 Annual Radiological Environmental Operating Report 6-17 Semiannual Radioactive Effluent Release Report 6-18 Monthly Reactor Operating Report 6-20 Reportable Occurrences 6-20 Prompt Notification With Written Followup 6-20 Thirty Day Written Reports. 6-22 Radial Peaking Factor Limit Report 6-23 6.9.2 SPECIAL REPORTS 6-23 6.10 RECORD RETENTION 6-24 6.11 RADIATION PROTECTION PROGRAM 6-25 6.12 HIGH RADIATION AREA 6-25 6.13 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6-26 6.14 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-27 I i h l t 9 l WCGS II

i ', s' ADMINISTRATIVE CONTROLS l i 6.1 RESPONSIBILITY l 6.1.1 The Plant Superintendent shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence. 6.1.2 The Supervising Operator, under the Shift Supervisor, l shall be responsible for the Control Room command function. A management directive to this effect, signed by the Vice President ( Nuclear shall be reissued to all station personnel on an annual basis. 6.2 ORGANIZATION i OFFSITE 6.2.1 The offsite organization for unit management and technical support shall be as shown in Figure 6.2-1. UNIT STAFF i 6.2.2 The Unit 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 Reactor 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 Reactor Operator shall be in the Control Room. i r c. ALL CORE ALTERATIONS shall 'be observed and directly super-vised by either a licensed Senior Reactor Operator or Senior f Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation. d. A site Fire Brigade of at least 5 members shall be maintained i onsite at all times.# The Fire Brigade shall not include i 3 members of the minimum shift crew necessary for safe l I shutdown of the unit and any personnel required for other i essential functions during a fire emergency. l The Fire Brigade composition may be less than the minimum l requirements for a period of time not to exceed 2 hours in order to accommodate absence provided immediate action is taken to fill the required positions. t j WCGS 6-1

b o KG'&E GENERAL OFFICE ORGANIZATION tn PRESIDENT S CHIEF EXEC OFFICER 1 VICE PRESIDENT NUCLEAR m oa NSRC NUCLEAR SAFETY e 8 REVIEW m COMMITTEE I MANAGER MANAGER MANAGER DIRECTOR SUPERVISOR NUCLEAR QUALITY NUCLEAR PLANT NUCLEAR PLA N 8 SERVICES ASSURANCE ENGINEERING OPER ATIONS CONTROLS TECHNICAL & ADMIN. ASSISTANCE i i m

Plant Supt. ~- -Maint. Supv. -Operations Supv. (SIO) -Tech. Supp. Supv. -Plant Supp. Supv. -Admin. Supv. -Maint. Engr. -Oper. Coord. (SJO). -Ibactor Ehgr. Supv. -Fire Prot. Spec. -Admin. Asst. -Maint. Coord. -Surveillanc Coord. l-Ehgr. .l-Clerk -Ebc. Qxit. Supv. -Mech. Supv. -Shift Supv. (SRO) -I&C Supv. -01ief of Security

• l-Doc. Cont.

-Nuc Mech-Mach -Supv. Oper. (SIO) -Eng./ Specialist -Security Ops Supv Spec. -Nuc Mech-Weld -Reactor Oper. (IO) -I&C Techs -Lieutenant -Nuc Mech Elec -Nuc. Station Oper. -Sergeants -Utility Helper -Utility Helper -Iead (bmp. Ehgr. -Guards -Bldg Serviceman -Engr./ Specialist -Conputer Tedi. -Qual. (bnt. Supv.* 1 l-Qual. Cont. Spec. -Elec. Supv. l-Nuclear Elect -Chemist -Olem. Supv. -Training Supv. cn -Olem. Tedi. -Training Spec. 1 -Utility Helper -Training Eng -Warehouse Supv. l-Warehouse Attendant -Health Physicist * -Ibsults Ehgineering Supv. -Health Phy. Supv. -Results Eng. (M) -Health Phy. Tech. -Results Eng. (E) -Utility Helper Figure 6.2-2 NOGS NORMAL OPERATItG OIGANIZATION

• For technical matters of an innediate nature, the respective individual reports directly to the Plant Superintendent.

i 'l

v / Tcblo 6.2-1 MINIMUM SHIFT CREW COMPOSITION l \\ J! \\ r> POSITION NUMBER OF INDIVIDUALS REQUIRED TO FILL POSITION I MODES 5& 6 MODES 1, 2, 3& 4' 1 SS 1 SRO 1 , /. /None RO 2 li SO 4 L 't HP 1 24cne i None II CHM 1 None YI STA 1** SS - Shif t Supervisor with a Senior Reactor Operator's License SRO - Individual with a Senior Reactor Operator's License' ' RO - Individual with a Reactor Operator's License e g I, h SO - Station Operator s / ,' / e l, 'O '.[ h STA - Shift Technical Advisor t HP - Health Physics Personnel 5 ? t CHM - Chemistry Personnel j/ I i The Shif t Crew Composition may be one less than the minimum requi.rtments of Table 6.2-1 for a period of time not to exceed 2.hourofin order to accommodate unexpected absence of on-duty; shif t crew', members provided immediate action is taken to restore.the Shift Crew'Composit.fon to with-l i in the minimum requirements of Table 6.2-1. This provisija does not permit any shift crew position to be unmanned upon shift' change due to an oncoming shift crewman being late or absent.( j During any absence of the~ Shift Supervisor frob the Control Room whil the unit is in MODE 1, 2, 3 or 4, an individual (other than the Shif t Technical Advisor) with a valid SRO 1.,icensk shall be designdted to ~ assume the Control Room command functionDuring any abgencelof the Shift Supervisor from the Control' Room while the unit is~in MODE 5,or y, an individual with a valid RO license 'i(other than the Shif1t Technical) Advisor) shall be designated to assums)the Control Room admmand function. ~j j r Personnel performing safety related furctions shall not work more than:*' l. 16 hours straight, 2. 24 hours in any forty-eight hour period, 3. 72 hours in any 7-day period, /, ~ 'f 4. There should be a break of 8 hours.(which can include

1. /

shift turnover time)'between al1 work periods,. j/ ?

s f J 1

.s /,/ 4 Deviation from these requiremen'ds mr.y be authorized by/ the Plant., Superintendent in accordance withJestablished procedures'and with documentation of the cause. Overt,ime.. limits do not include shift turnover time. 'li. .\\ / s

• • Shif t Technical Advisor is requiredrabytime the Shift Superdisce on duty does not meet the educational ' equirement of a STA as defined r

in the October 31, 1980 letter from Darrell G. Eisenhut, Divis'J.wi of. Licensing, to all licensees ar.d applicants. ') j' WCGS 6-4, v

s / l t ADMINISTRATIVE CONTROLS i \\ N 3 \\ 6.'2.3 SHIFT TECHNICAL ADVISOR l 1 The Shif t Technical Advisor when required as specified in Table 6.2-1 shall serve in an advisory capacity to the Shif t Supervisor on i v' matters pertaining to the engineering aspects assuring safe operation I i, /', of, the unit. 8 6.3 UNIT STAFF QUALIFICATIONS l / i:. [' 6.3.1 Each member of the unit staff shall meet or exceed the f minimuy~ qualifications of ANSI 3.1, 1978 for comparable positions and the supplemental requirements specified in Section A and C of Enclosure 1 of the March 28, 1980 NRC Letter to all licensees, except for the Health Physicist who shall meet or exceed the qualifications of Regolatory Guide 1.8, September 1975, pertaining to Radiation Protection Manager. 6.4 TRAINING 6.4.1 A retraining and replacement training program for the unit otaff shall he maintained under the administration of the Training Supervisor and shall mcet or exceed the requirements and recommendations of ANSI 3.1, 1978 and Appendix "A" of 10 CFR Part 55 and the supplemental requirements specified in Section A and C of Enclosure 1 of the March 28, 1980 NRC letter to all licensees, and shall include familiarization with relevant industry operational experience. I \\ I. g 1 0 h9 ,Gf

• l

,/. + f [ WCGSI 6-5 /> L - fl/

a..... -..... t 't ADMINISTRATIVE CONTROLS l 6.5 REVIEW AND AUDIT l This section discusses the safety review groups for Wolf Creek. 'I Other licensing documents describe other groups within the nuclear l organization with their respective review and/or audit l responsibilities. l l 6.5.1 PLANT SAFETY REVIEW COMMITTEE l FUNCTION l 6.5.1.1 The Plant Safety Review Committee (PSRC) shall function to l Edvise the Plant Superintendent on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The Plant Safety Review Committee shall be composed of the: l Chairman: Plant Superintendent Member: Operations Supervisor Member: Technical Support Supervisor l Member: Maintenance Supervisor i Member: Plant Support Supervisor Member: Instrument and Control Supervisor Member: Reactor Engineering Supervisor Member: Health Physicist Member: Chemist Member: Results Engineering Supervisor 1 ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the i PSRC Chairman to serve on a temporary basis; however, no more than 1 two alternates shall participate as voting members in PSRC activities i et any one time. } i WCGS 6-6

s. f ADMINISTRATIVE CONTROLS t MEETING PREOUENCY 6.5.1.4 The PSRC shall meet at least once per calendar month and as l convened by the PSRC Chairman or his designated alternate. QUORUM 6.5.1.5 The minimum quorum of the PSRC necessary for the performance of the PSRC responsibility and authority provisions of these Technical Specifications shall consist of the Chairman or his designated alternate and four members including alternates. RESPONSIBILITIES 6.5.1.6 The Plant Safety Review Committee shall be responsible for: l a. Review of 1) all procedures required by Specification 6.8 and changes thereto, 2) all programs required by Specification 6.8 and changes thereto, 3) any other proposed procedures or changes thereto as determined by the Plant Superintendent to affect nuclear safety. b. Review of all proposed tests and experiments that affect nuclear safety. c. Review of all proposed changes to Appendix "A" Technical Specifications. d. Review of all proposed changes or modifications to unit systems or equipment that affect nuclear safety. Investigation of all violations of the Technical Specifications e. including the preparation and forwarding of reports covering evaluation and recommendations to prevent recurrence to the Director Nuclear Operations and to the Nuclear Safety Review; Committee (NSRC). f. Review of events requiring 24-hour written notification to the Commission. l g. Review of unit operations to detect potential nuclear safety hazards. h. Performance of special reviews, investigations or analyses and reports thereon as requested by the Plant Superintendent or the NSRC. i. Review of the Security Plan and shall submit recommended changes to the NSRC. WCGS 6-7 h u_ 7----

ADMINISTRATIVE CONTROLS j. Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the NSRC. k. Review of every unplanned onsite release of radioactive material to the environs, including the preparation and forwarding of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence to the Plant Superintendent and to the Nuclear Safety Review Committee. 1. Review of major changes to radwaste systems. h i AUTHORITY 6.5.1.7 The Plant Safety Review Committee shall: l a. Recommend in writing to the Plant Superintendent approval or disapproval of items considered under 6.5.1.6 (a) through (d) above. b. Render determinations in writing with regard to whether i or not each item considered under 6.5.1.6 (a) through (e) above constitutes an unreviewed safety question. Provide writt'n notification within 24 hours to the c. e Director Nuclear Operations and the-NSRC of disagreement between the PSRC and the Plant Superintendent; however, the Plant Superintendent shall have responsibility for resolution of such disagreements pursuant to 6.1.1 above. l RECORDS 6.5.1.8 The Plant Safety Review Committee shall maintain written minutes of each PSRC meeting that, at a minimum, document the results of all PSRC activ.ities performed under the responsibility and authority provisions of these technical specifications. Copies shall be provided to the Director Nuclear Operations and the NSRC. [ 6.5.2 NUCLEAR SAFETY REVIEW COMMITTEE F FUNCTION I 6.5.2.1 The Nuclear Safety Review Committee (NSRC) shall function l to provide independent review and audit of designated activities in the areas of: a. nuclear power plant operations t WCGS 6-8 L

ADMINISTRATIVE CONTROLS f b. nuclear engineering c. chemistry and radiochemistry d. metallurgy e. instrumentation and control k f. radiological safety g. mechanical and electrical engineering h. quality assurance practices l COMPOSITION 6.5.2.2 The NSRC shall be composed of the: Chairman - Manager Nuclear Services, KG&E Vice Chairman - Manager Nuclear Plant Engineering, KG&E Member - Quality Assurance Coordinator, KG&E Member - Director Nuclear Operations, KG&E Member - Manager Licensing, KG&E Member - Vice President-Engineering, KG&E Member - Manager Safety Engineering ALTERNATES 6.5.2.3 All alternate members shall be appointed in writing by the NSRC Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in NSRC activities at any one time. CONSULTANTS 6.5.2.4 Consultants shall be utilized as determined by the NSRC Chairman to provide expert advice to the NSRC. MEETING FREQUENCY 6.5.2.5 The NSRC shall meet at least once per calendar quarter during the initial year of unit operations following fuel loading cnd at least once per six months thereaf ter. WCGS 6-9 e

i ADMINISTRATIVE CONTROLS i QUORUM 6.5.2.6 The minimum quorum of the NSRC necessary for the performance y of the NSRC review and audit functions of these Technical Specifications i shall consist of the Chairman or the Vice Chairman and greater than f one-half of the total number of NSRC members including alternates. ) No more than a minority of the quorum shall have line responsibility for operation of the unit. l REVIEW 6.5.2.7.1 The NSRC shall review: a. Written safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of 10 CFR 50.59, to verify that l such actions did not constitute an unreviewed safety question. b. Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in 10 CFR 50.59. I c. Proposed tests or experiments which involve an unreviewed safety question as defined in 10 CFR 50.59. l d. Proposed changes to Technical Specifications or this Operating License. e. Violations of applicable codes, regulations, orders, l Technical Specifications, license requirements, or of internal procedures or instructions.having nuclear i safety significance. l 1 f. Significant operating abnormalities or deviations from normal and expected performance of unit equipment that affect nuclear safety. g. Events requiring 24 hour written notification to the Commission. h. All recognized indications of an unanticipated deficiency in some aspect of design or operation of structures, systems, or components that could affect nuclear safety. i. Reports and meeting minutes of the PSRC. l WCGS 6-10 I

't i ADMINISTRATIVE CONTROLS I AUDITS 6.5.2.8 Audits of unit activities shall be performed under the cognizance of the NSRC. These audits shall encompass: I a. The conformance of unit operation to provisions contained j within the Technical Specifications and applicable license conditions at least once per 12 months. b. The performance, training and qualifications of the I entire unit staff at least once per 12 months. c. The results of actions taken to correct deficiencies occurring in unit equipment, structures, systems or method of operation that affect nuclear safety at least once per 6 months. d. The performance of activities required by the Operational Quality Assurance Program to meet the criteria of Appendix "B", 10 CFR 50, at least once per 24 months. e. The Emergency Plan and implementing procedures at least once per 24 months. f. The Security Plan and implementing procedures at least once per 24 months. g. Any other area of unit operation considered appropriate by the NSRC or the Vice President-Nuclear. l h. The Fire Protection Program and implementing procedures at least once per 24 months. i. An independent fire protection and loss prevention inspection and audit to be performed annually utilizing either qualified offsite licensee personnel or an outside fire protection firm. j. An inspection and audit of the fire protection and loss prevention program to be performed by an outside qualified fire consultant at intervals no greater than 3 years. WCGS 6-11

"M1P 't ADMINISTRATIVE CONTROLS i AUTHORITY 6.5.2.9 The NSRC shall report to and advise the Vice President-Nuclear on those areas of responsibility specified in Sections 6.5.2.7 and 6.5.2.8. RECORDS 6.5.2.10 Records of NSRC activities shall be prepared, approved and l 6 distributed as indicated below: a. Minutes of each NSRC meeting shall be prepared, reviewed by participating members and forwarded to the Vice President-Nuclear within 14 days following each meeting. b. Reports of reviews encompassed by Section 6.5.2.7 above, shall be prepared, reviewed by participating members and forwarded to the Vice President-Nuclear within 14 days following completion of the review. c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the Vice President-Nuclear and to I the management positions responsible for the areas (odited within 30 days after completion of the audit by the auditing organization. 6.6 REPORTABLE OCCURRENCE ACTION 6.6.1 The following actions shall be taken for REPORTABLE OCCURRENCES: a. The Commission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9. b. Each REPORTABLE OCCURRENCE requiring 24 hour notification to the Commission shall be reviewed by the PSRC and submitted to the NSRC,and the Director Nuclear Operations. 6.7 SAFETY LIMIT VIOLATION 6.7.1 The following actions shall be taken in the event a Safety Limit is violated: a. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour. The Director of Nuclear Operations and the NSRC shall be notified within 24 hours. WCGS 6-12

4 t ADMINISTRATIVE CONTROLS r' I b. A Safety Limit Violation Report shall be prepared. The [ report shall be reviewed by the PSRC. This report shall l ii describe (1) applicable circumstances preceding the L violation, (2) effects of the violation upon facility components, systems or structures, and (3) corrective action taken to prevent recurrence. c. The Safety Limit Violation Report shall be submitted to the Commission, the NSRC, and the Director Nuclear l Operations within 14 days of the violation. d. Critical operation of the unit shall not be resumed until authorized by the Commission. 6.8 PROCEDURES AND PROGRAMS 6.8.1 The plant shall be operated and maintained in accordance with approved procedures. Major procedures, supported by appropriate minor procedures (such as checkof f lists, operating instructions, data sheets, alarm responses, etc.) shall be provided for the following operations where these operations involve nuclear safety of the plant: a. The applicable procedures recommended in Appendix "A" of Regulation 1.33, Rev 2, Feb 1978 b. Refueling Operation c. Emergency Plan Implementation d. Surveillance and Testing of safety related equipment. l e. Fire Protection Program Implementation f. Radioactive - Waste Processing Implementation. g. OFFSITE DOSE CALCULATION MANUAL Implementation. 6.8.2 Approval of Procedures a. All major procedures of the categories listed in 6.8.1 and modifications to the intent thereof shall be reviewed by the Plant Safety Review Committee and approved by the Plant Superintendent prior to implementation and reviewed periodically as set forth in Administrative Procedures. WCGS 6-13 L

m a h ADMINISTRATIVE CONTROLS I.. b. Minor procedures (checkof f lists, operating instructions, data sheets, alarm responses, chemistry analytical procedures, technical instructions, special and routine maintenance procedures, laboratory manuals, etc.) shall, prior to initial use, be approved by the PSRC. 6.8.3 Changes to Procedures a. Temporary changes to major procedures, of the categories i listed in 6.8.1 which do not change the intent of the original or subsequent approved procedure, may be made provided such changes to operating procedures are approved by the Shift Supervisor (SRO licensed) and one of the Call Superintendents. For temporary changes to major procedures under the jurisdiction of Maintenance, Instrumentation and Control, Reactor Engineering, Chemistry, or Health Physics which do not change the e intent, changes may be made upon approval of the cognizant group leader and a Call Superintendent. All Temporary changes to major procedures (made by a Call Superintendent and either a cognizant group head or the Shift Supervisor) shall subsequently be reviewed by the Plant Safety Review Committee and appro*Jed by the Plant Superintendent within 2 weeks; except that temporary changes to major procedures made during a refueling outage may be reviewed and approved at any time prior to initial criticality of the reload core. All permanent changes to major procedures shall be made in accordance with Step 6.8.2.a. b. All temporary or permanent changes to minor operating procedures (checkof f lists, alarm responses, data sheets, operating instructions, etc.) shall be approved by the Shift Supervisor, and shall be subsequently reviewed and approved by the Operations PSRC Subcommittee. All temporary or permanent changes to other minor procedures under the jurisdiction of Maintenance, Instrumentation and Control, Reactor Engineering, Chemistry, or Health Physics, shall be approved by a supervisor of the cognizant group and shall be l subsequently reviewed and approved by the appropriate PSRC Subcommittee. WCGS 6-14

e t ADMINISTRATIVE CONTROLS 6.8.4 The following programs shall be established, implemented, and maintained: a. Primary Coolant Sources Outside Containment A program to reduce leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. The systems include the recirculation spray and safety injection. The program l shall include the following: (i) Preventive maintenance and periodic visual inspection requirements, and (ii) Integrated leak test requirements for each system at refueling cycle intervals or less. b. In-Plant Radiation Monitoring A program which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following: (i) Training of personnel, (ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analysis equipment. WCGS 6-15 L

't i ADMINISTRATIVE CONTROLS I 6.9 REPORTING REQUIREMENTS ROUTINE REPORTS AND REPORTABLE OCCURRENCES 6.9.1 In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Director of the Regional Office of Inspection and Enforcement unless otherwise noted. STARTUP REPORT 6.9.1.1 A summary report of plant startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant. 6.9.1.2 The Startup report shall address each of the tests identified in the FSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report. 6.9.1.3 Startup reports shall be submitted within (1) 90 days following completion of the startup test program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i. e., initial criticality, completion of startup test program, and resumption or commencement of commercial operation) supplementary reports shall be submitted at least every three months until all three events have been completed. WCGS 6-16 1

ADMINISTRATIVE CONTROLS ANNUAL REPORTS 6.9.1.4 Annual reports covering the activities of the unit as described below for the previous calendar year shall be submitted prior to March 1 of each year. The initial report shall be submitted l , prior to March 1 of the year following initial criticality. 3 i 6.9.1.5 Reports required on an annual basis shall include: i A tabulation on an annual basis of the number of station, utility, l } cnd other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated manrem exposure j cccording to work and job functions,l/ e.g., reactor operations i and surveillance, inservice inspection, routine maintenance, special j maintenance (describe maintenance), waste processing, and refueling. The dose assignments to various duty functions may be estimated based on pocket dosimeter, TLD, or film badge measurements. Small exposures totalling less than 20 percent of the individual total do not need be accounted for. In the aggregate, at least 80 percent of the total whole body dose received from external sources should be assigned to specific major work functions. l ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6.9.1.6 Routine radiological environmental operating reports covering the operation of the unit 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. 6.9.1.7 The annual radiological environmental operating reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and pr evious environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required i by Specification 3.12.2. If harmful effects or evidence of irreversible demage are detected by the monitoring, the report shall provide an cnalysis of the problem and a planned course of action to alleviate the problem. 1/This tabulation supplements the requirements of paragraph l 20.407 of 10 CFR Part 20. WCGS 6-17

MW ; t ADMINISTRATIVE CONTROLS i i The annual radiological environmental operating reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period. In the event that some results are l not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as possible in a supplementary report. The reports shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and the results of licensee participation in the Interlsboratory Comparison Program, required by Specification 3.12.3. SEMIANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6.9.1.8 Routine radioactive effluent release reports covering the operation of the unit during the previous 6 months of operation shall be submitted within 60 days af ter January 1 and July 1 of each year. The period of the first report shall begin with the date of initial criticality. 6.9.1.9 The 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 summarized on a quarterly basic following the format of Appendix B thereof. l The radioactive effluent release report to be submitted 60 days l after January 1 of each year and shall include an annual summary of ~ hourly meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing i of wind speed, wind direction, atmospheric stability, and precipitation (if measured) on magnetic tape, or in the form of joint frequency i l distributions of wind speed, wind direction, and atmospheric stability. This same report shall include an assessment of the radiation doses i due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. l l l WCGS 6-18

. - ~ - - - 6 ADMINISTRATIVE CONTROLS t All assumptions used in making these assessments (i. e., specific activity, exposure time, and location) shall be included in these g reports. The meteorological conditions concurrent with the time of j release of radioactive materials in gaseous effluents ~ (as determined i by sampling frequency and measurement) shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual (ODCM). The radioactive effluent release report to be submitted 60 days af ter January 1 of each year shall also include an assessment of radiation doses to the likely most exposed real individual from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and direct radiation) for the previous 12 consecutive months to show conformance with 40 CFR 190, Environmental Radiation Protection Standards for Nuclear Power Operation. Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev. 1. The radioactive effluent release reports shall include the following information for each type of solid waste shipped offsite during the report period: a. Container volume, b. Total curie quantity (specify whether determined by measurement or estimate), c. Principal radionuclides (specify whether determined by measurment or estimate), d. Type of waste (e.g., spent resin, compacted dry waste, evaporator bottoms), e. Type of container (e.g., LSA, Type A, Type B, Large Quantity), and f. Solidification agent (e.g., cement, urea formaldehyde). The radioactive effluent release reports shall include unplanned releases from the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis. WCGS 6-19 L

t ADMINISTRATIVE CONTROLS t i MONTHLY OPERATING REPORT e 6.9.1.10 Routine reports of operating statistics and shutdown I L experience shall be submitted on a monthly basis to the Director, l Office of Management and Program Analysis, U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, with a copy to the Regional Office of Inspection and Enforcement, no later than the 15th of each month following the calendar month covered by the report. Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days of when the change (s) became effective. In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by the PSRC. REPORTABLE OCCURRENCES 6.9.1.11 The REPORTABLE OCCURRENCES of Specifications 6.9.1.12 and 6.9.1.13 below, including corrective actions and measures to prevent recurrence, shall be reported to the NRC. Supplemental reports may be required to fully describe final resolution of occurrence. In case of corrected or supplemental reports, a licensee event report shall be completed and reference shall be made to the original r.eport date. PROMPT NOTIFICATION WITH WRITTEN FOLLOWUP 6.9.1.12 The type of events listed below shall be reported within l 24 hours by telephone and confirmed by telegraph, mailgram, or facsimile transmission to the Director of the Regional Office, or his designate no later than the first working day following the event, with a written followup report within 14 days. The written followup report shall include, asfa minimum, a completed copy of a licensee event report form. Information provided on the licensee t event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event. a. Failure of the reactor protection system or other systems subject to limiting safety system settings to initiate the required protective function by the time a monitored parameter reaches the setpoint specified as the limiting safety system setting in the technical specifications or failure to complete the required protective function. WCGS 6-20

r ADMINISTRATIVE CONTROLS b. Operation of the unit or affected systems when any parameter or operation subject to a limiting condition for operation is less conservative than the least conservative aspect of the Limiting Condition for Operation established in the Technical Specifications. t c. Abnormal degradation discovered in fuel cladding, reactor coolant pressure boundary, or primary containment. d. Reactivity anomalies involving disagreement with the predicted value of reactivity balance under steady state conditions during power operation greater than or equal to 1% A k/k; a calculated reactivity balance indicating a SHUTDOWN MARGIN less conservative than specified in the Technical Specifications; short-term reactivity increases that correspond to a reactor period of less than 5 seconds or, if suberitical, an unplanned reactivity insertion of more than 0.5% a k/k; or occurrence of any unplanned criticality. e. Failure or malfunction of one or more components which prevents or could prevent, by itself, the fulfillment of the functional requirements of system (s) used to cope with accidents analyzed in the SAR. f. Personnel error or procedural inadequacy which prevents or could prevent, by itself, the fulfillment of the functional requirements of systems required to cope with accidents analyzed in the SAR. g. Conditions arising from natural or man-made events that, as a direct result of the event require unit shutdown, operation of safety systems, or other protective measures required by Technical Specifications. h. Errors discovered in the transient or accident analyses or in the methods used for such analyses as described in the safety analysis report or in the bases for the Technical Specifications that have or could have permitted l l reactor operation in a manner less conservative than assumed in the analyses. t WCGS 6-21 1

NDMINISTRATIVE CONTROLS i 1. Performance of structures, systems, or components that requires remedial action or corrective measures to prevent operation in a manner less conservative than assumed in the accident analyses in the safety analysis report or Technical Specifications bases; or discovery during unit life of conditions not specifically considered in the safety analysis report or Technical Specifications that require remedial action or corrective measures to prevent the existence or development of an unsafe j condition. j. Offsite releases of radioactive materials in liquid and gaseous ef fluents which exceed the limits of Specification 3.11.1.1 or 3.11.2.1. k. Exceeding the limits in Specification 3.11.1.4 or 3.11.2.6 for the storage of radic active materials in the listed tanks. The written follow-up report shall include a schedule and a description of activities planned and/or taken to reduce the contents to within the specified limits. THIRTY DAY WRITTEN REPORTS 6.9.1.13 The types of events listed below shall be the subject of written reports to the Director of the Regional Office within thirty days of occurrence of the event. The written report shall include, as a minimum, a completed copy of a licensee event report form. Information provided on the licensee event report form shall be supplemented, as needed, by additional narrative material to provide complete explanation of the circumstances surrounding the event. s a.* Reactor protection system or engineered safety. feature l instrument settings which are found to be less conservative than those established by the Technical Specifications but which do not prevent the fulfillment of the functional requirements of affected systems. b.* Conditions leading to operation in a degraded mode l permitted by a Limiting Condition for Operation or plant shutdown required by a Limiting condition for Operation.

• NOTE:

Routine surveillance testing, instrument calibration, or preventative maintenance which requires system configurations as described in a and b above need not be reported except when test results themselves reveal a degraded mode as described above. l WCGS 6-22 l

r IDMINISTRATIVE CONTROLS I i c. Observed inadequacies in the implementation of administrative or procedural controls which threaten to cause reduction of degree of redundancy provided in reactor protection j systems or engineered safety feature systems. d. Abnormal degradation of systems other than those specified in 6.9.1.12.c above designed to contain radioactive material resulting from the fission process. e. An unplanned offsite release of 1) more than 1 curie of [ radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous ef fluents, or 3) more than 0.05 curies of radioiodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information: 1. A description of the event and equipment involved. 2. Cause(s) for the unplanned release. 3. Actions taken to prevent recurrence. 4. Consequences of the unplanned release. f. Measured levels of radioactivity in an environmental sampling medium determined to exceed the reporting level values of Table 3.12-l when averaged over any calendar l quarter sampling period. RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.14 The F limit for RATED THERMAL POWER (FE P) challbeprovidebtotheDirector'oftheRegionalO{fficeofInspection cnd Enforcement, with a copy to the Director, Nuclear Reactor Regulation, Attention Chief of the Core Performance Branch, U.S. Nuclear Regulatory Commission, Wa'shington, D.C. 20555 for all core planes containing bank "D" control rods and all unrodded core planes ct least 60 days prior to cycle initial criticality. In the event that the limit would be submitted at some other time during core life, it will be submitted 60 days prior to the date the limit would l become effective unless otherwise exempted by the Commission. l l AnyinformationneededtosupportF@[Pwillbebyrequest l from the NRC and need not be included in this report. l i SPECIAL REPORTS l' 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. ( WCGS 6-23

q r ,goMYNTSTRATIVECONTROLS 6.10 RECORD RETENTION I In addition to the applicable record retention requirements of Title i 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated. I 6.10.1 The following records shall be retained for at least five l years: l 6 a. Records and logs of unit operation covering time interval l at each power level. l b. Records and logs of principal maintenance activities, inspe ct ions, repair and replacement of principal items of equipment related to nuclear safety. c. ALL REPORTABLE OCCURRENCES submitted to the Commission. d. Records of surveillance activities; inspections and calibrations required by these Technical Specifications. e. Records of changes made to the procedures required by Specification 6.8.1. f. Records of radioactive shipments. g. Records of sealed source and fission detector leak tests and results. h. Records of annual physical inventory of all sealed source material of record. 6.10.2 The following records shall be retained for the duration of 'the Unit Operating License: Records and drawing c'hanges reflecting unit design a. modifications made to, systems and equipment described in the Final Safety Analysis Report. b. Records of new and irradiated fuel inventory, fuel transfers, and assembly burnup histories, c. Records of radiation exposure for all individuals entering radiation control areas. d. Records of gaseous and liquid radioactive material released to the environs. e. Records of transient or operational cycles for those unit components identified in Table 5.7-1. f. Records of reactor tests and experiments. WCGS 6-24

... ~ _.. - - - 1 i l

• ADMINISTRATIVE CONTROLS l

g. Records of training and qualification for current members of the unit staff. h. Records of in-service inspections performed pursuant to these Technical Specifications. t i. Records of Quality Assurance activities required by the QA Manual. r j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to 10 CFR 50.59. i k. Records of meetings of the PSRC and the NSRC. l } t 1. Records of the service lives of all snubbers listed in Tables 3.7-4a and 3.7-4b including the date at which the service life commences and associated installation and maintenance records. Records of secondary water sampling and water quality. m. n. Records of analyses required by the radiological environmental monitoring program. 6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure. 6.12 HIGH RADIATION AREA 6.12.1 In lieu of the " control device" or " alarm signal" required by paragraph 20.203(c) (2) of 10 CFR 20, each high radiation area in which the intensity of radiation is greater than 100 mrem /hr but less that 1000 mrem /hr 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).

• Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:
• Health Physics personnel or personnel escorted by Health Physics personnel may be exempt from the RWP issuance i

requirement during the performance of their assigned radiation protection duties, provided they are otherwise following plant radiation protection procedures for entry into high radiation areas. t I WCGS 6-25

.. + ADMINISTRATIVE CONTROLS a. A radiation monitoring device which continuously indicates the radiation dose rate in the area. I b. A radiation monitoring device which continuously integrates I 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 level in the area has been established and 3 personnel have been made knowledgeable of them. c. An individual qualified in radiation protection procedures l who is equipped with a radiation dose rate monitoring device. This individual shall be responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the facility Health Physicist in the Radiation Work Permit. 6.12.2 The requirements of 6.12.1, above, shall also apply to each high radiation area in which the intensity of radiation is greater than 1000 mrem /hr. In addition, locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor and/or the Site Health Physicist. 6.13 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6.13.1 The ODCM shall be approved by the Commission prior to implemen-tation. 6.13.2 Licensee initiated changes'to the ODCM: 1. Shall be submitted to'the Commission in the Monthly Operating Report within 90 days of the date the change (s) was/were made effective. This submittal shall contain: l a. Sufficiently detailed information to totally support the rationale for the change without benefit of additional or supplemental information. Information submitted should consist of a package of those pages of the ODCM to be changed with each page numbered and provided with an approval and i date box, together with appropriate analyses or evaluations justifying the change (s) ; l WCGS 6-26

_ = _, _e a w a e

• ADMtNISTRATIVE CONTROLS b.

A determination that the change will not reduce the accuracy or reliability of dose calculations or setpoint determinations; and c. Documentation of the fact that the change has been reviewed and found acceptable by the PSRC. l 2. Shall become effective upon review and acceptance by the PSRC. l 6.14 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (Liquid, Gaseous, and Solid) 6.14.1 Licensee initiated major changes to the radioactive waste systems (liquid, gaseous, and solid) : 1. Shall be reported to the Commission in the Monthly Operating Report for the period in which the evaluation was reviewed by the Plant Safety Review Committee. The l discussion of each change shall contain: a. A summary of the evaluation that led to the determin-ation that the change could be made in accordance with 10 CFR 50.59. b. Sufficent detailed information to totally support the reason for the change without benefit of additional or supplemental information. c. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems. d. An evaluation of.the change which shows the predicted releases of radioactive material in liquid and gaseous effluents and/or quantity of solid waste that differ from those previously predicted in the license application and amendments thereto. e. An evaluation of the change which shows the expected maximum exposures to individual in the unrestricted area and to the general population that differ from those previously estimated in the license application and amendments thereto. WCGS 6-27

I I ~IDMINISTRATIVECONTROLS f. A comparison of the predicted releases cf radioactive materials, in liquid and gaseous effluente and in solid waste, to the actual releases for the period prior to when the changes are to be made. g. An estimate of the exposure to plant operating personnel as a result of the change. i h. Documentation of the fact that the change was reviewed and found acceptable by the PSRC. l 2. Shall become effective upon review and acceptance by the PSRC. I WCGS 6-28

.. _... _ _ _ _ _ _ ~ i l' t 8 WOLF CREEK GENERATING STATION UNIT NO. 1 ENVIRONMENTAL PROTECTION PLAN (NON-RADIOLOGICAL) l TABLE OF CONTENTS Section Page f 1.0 objectives of the Environmental Protection 1-1 Plan 3 l 2.0 Environmental Protection Issues 2-1 2.1 Water Quality Issues 2-1 2.2 Aquatic Issues 2-2 i 2.3 Terrestrial Issues 2-2 3.0 Consistency Requirements 3-1 L 3.1 Plant Design and Operation 3-1 3.2 Reporting Related to the NPDES Permit 3-3 3.3 Changes Required for Compliance with Other 3-3 Environmental Regulations 4.0 Environmental Conditions 4-1 4.1 Unusual or Important Environmental Events 4-1 t 4.2 Environmental Monitoring and Management 4-1 4.2.1 Fog Monitoring 4-1 4.2.2 Waterfowl Impaction 4-2 4.2.3 Land Management 4-2 t l l i I ll

n: i e TABLE OF CONTENTS (continued) Section Page 5.0 Administrative Procedures 5-1 l 5.1 Review and Audit 5-1 l I 5.2 Retention of Program Documentation 5-1 5.3 Changes to the Environmental Protection Plan 5-1 5.4 Plan Reporting Requirements 5-2 5.4.1 Routine Reports 5-2 5.4.2 Nonroutine Reports 5-3 ii l ~

f i e 1.0 Objectives of the Environmental Protection Plan The Environmental Protection Plan (EPP) provides the framework for all environmental activities initiated at Wolf Creek Generating Station (WCGS) to monitor and protect the quality of the environment in areas surrounding the plant site. It serves to verify KG&E's commitment to operate WCGS in a safe and efficient manner while maintaining controls to minimize the environmental impact of the station operation. The EPP provides as its chief objectives the means for: (a) Demonstrating regulatory compliance for those environmental criteria specified by the NRC, State and local regulatory agencies and applied to WCGS. 'l (b) Tracking, KG&E's environmental commitments, as specified in the ER(OLS), FE S', SAR and ensuring that they are suitably fulfilled. l (c) Ensuring that the primary intent is to operate the plant in an environmeAt511y accep$able manner and that, should an event occur whicli, significantly impacts the environment, timely l actions are tr. ken to evaluate, mitigate and document the l j causative problem. l l i 1-1 [ l

y.p 3 '1 / .) i 't V f: ,;)# ( / e t / ~~ j 1 A ', .i{ .) f~ f >~ , (- w 2.0 Environmental Protection.Issuris M ~,.s .,4

A';

i + In the FES-OL dated June, 1982, the staff considered the environ- ., !t "c mental impacts associatedLwith the ophration of WCGy. Certain I // ,r, environmental issues were idenyified whichJrequired monitoring,, ., r - j.. studyorlicenseconditbnstoresolveenvironme,ntalconcernsahO' jf ! to assure adequate protection, of the environment. ' / .c.4 / + ?t; G <i

/

j i / -j s 2.1 Water Quality Issues i, 7. 'I f 7 t J '., [~ l l i 4 a (a) That chlorine dosin,g is controlled within th'o'se frequencien' ^/ m 3<. i' and discharge cencentrations. predicted; and e, valuated.byN. he t /. p, s taf f.'- d.Z " ' I N ( 'I .~ , '[j l t 7 ,.,I ~ Section 4.2.6.1; Scalirig and Biocide Tre'at.ner.it), j (FES-OL: f .o J s. '[/ Wastes) 7 ,o .s 'k?, ? ir, f_ \\ ) v' -*j [.. (b) That dischargas.from Wolf Creek Cooling Lakef(WCCL) be regulated sE.that Total Dissolved Solids (TDS)', sulfite.[nd C. chloride; concentrations in the Neosho River after complelc /,' / 'r .i i .d ' mixing are at most 500, 250 and 250 mg/1, respectively, #/ 1 e according to Kansas Water Quality Criteria and 40 CFR 143. 755 s ,t i{ ./ z (FES-OL: Se t on 5.3.2.1; Neosho River) , t s,- 7 u (c) Additional water quality issues shall conform to the effec-l/, ' IG tive NPDES permit implemented by the Kansas Department of i' Hee.lth and Environment. yt, l 2-1 / -r ,e l

• 'f e #

6 l 1

/ r U l' s'. IP ~ s t 2.2 Aquatic Issues l The need for aquatic monitoring programs to confirm that effects on aquatic biota due to plant operations are no greater than predicted. Y (FES-OLi Section 5.1) j r f e <2.3 Terrestrial Issues (a) That dhe composition and structure of vegetation in the / h-453-ha (li20-acre) exclusion zone will be selectively con-trolled to be compatible with the function and security of station facilities. (FES-OL: Section 5.5.1.1; Station Site) t (b) That the vegetation within a buffer zone surrounding the cooling lake will be retained in, or allowed to develop toward a natural state, i.e., naturally occurring biotic communitics. t i'- (FES-OL: Section 5.5.1.1; Station Site) i 1

• 'i )

2-2 1

t 4 I a-1 (c) That herbicides used for the maintenance of transmission-l line corridors will be limited to herbicides approved by the U.S. EPA and the Statc of Kansas at the time of such use. i i (FES-OL: Section 5.5.1.2; Energy-Transmission System) I i (d) That in the event of a serious disease problem involving waterfowl attributable to station operation occurs the actions specified in the reference will be initiated follow-ing technical evaluation if deemed necessary. (FES-OL: Section 5.5.1.1; Station Site) (e) The need for vegetational and wildlife monitoring programs and that a general survey program for waterfowl collision events be accomplished. r i* (FES-OL: Section 5.5.1.2, Energy-Transmission System) (f) The need for a fog monitoring program to document any potential increase in fogging due to the operation of the l cooling lake heat-dissipation system. 4 l (FES-OL: Section 5.4.1; Fog and Ice) l l [ 1 I 2-3

li I e 9 3.0 Consistency Requirements 3.1 Plant Design and Operation Changes in station design, operation, performance testing, experimental analysis, etc. that do not impact the environment are exempt from the EPP requirements and may be initiated according to procedure or operator judgment. Changes in these [ parameters that do impact the environment are not exempt and I are subject to EPP control as follows. j Prior to engaging in activities which have the potential to significantly impact the environment, the licensee shall prepare and record an environmental evaluation of such activity. If this evaluation indicates nominal environmental impact, which is [ considered within the general envelope of the WCGS ER(OLS), FES

1. and SAR, the evaluation will be retained on file and the activity or changes shall commence.

If the evaluation indicates that such i an activity involves an unreviewed environmental question, the l licensee shall provide a written evaluation of such activities and obtain prior approval from the Director, Office of Nuclear Reactor I Regulation. When such activity involves a change in the EPP, such activity and change to the EPP may be implemented only in accord-ance with an appropriate license amendment as set forth in Section 5.3. f

• This provision does not relieve the licensee of the requirements of 10 CFR 50.59.

t-3-1 i

m,. i i b A proposed change, test or experiment shall be deemed to involve an unreviewed environmental question if it concerns (1) a matter which may result in a significant increase in any adverse envi-ronmental impact previously evaluated in the final environmental statement (FES) as modified by staff s testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or in any decisions of the Atomic Safety and Licensing Board; or (2) a significant change in effluents or power level [in accordance with 10 CFR Part 51.5(b)(2)] or (3) a matter not previously reviewed and evaluated in the documents specified in (1) of this Subsection, which may have a significant adverse environmental impact. l The licensee shall maintain documentation of changes in facility design or operation and of tests and experiments carried out pursuant to this Subsection. This documentation shall include a written evaluation which provide bases for the determination that the change, test, or experiment does not involve an unreviewed environmental question nor constitute a decrease in the effectiveness of this EPP to meet the objectives specified in Section 1.0. The licensee shall include as part of his Annual Environmental Operating Report (per Subsection 5.4.1) brief descriptions, analyses, interpretations, and evaluations of such changes, tests and experiments. l l 3-2

'I

l s'

D 3.2 Reporting Related to the NPDES Permit Violations of the NPDES Permit shall be reported to the NRC by submittal of copies of the reports required by the NPDES Permit. Changes and additions to the NPDES Permit shall be reported to the NRC within 30 days following the date the change is approved. [ If a permit, in part or in its entirety, is appealed and stayed, the NRC shall be notified within 30 days following the date the stay is granted. The NRC shall be notified of changes to the effective NPDES Per-mit proposed by the licensee by providing NRC with a copy of the proposed change at the same time it is submitted to the per-mitting agency. The notification of a licensee-initiated change shall include a copy of the requested revision submitted to the permitting agency. The licensee shall provide the NRC a copy of the application for renewal of the NPDES permit at the same time the application is submitted to the permitting agency. 3.3 Changes Required for Compliance with Other Environmental Regulations Changes in plant design or operation and performance of tests or experiments which are required to achieve compliance with other Federal, State, or local environmental regulations are not sub-ject to the requirements of Section 3.1. 3-3

4 a. 4.0 Environmental Conditions 4.1 Unusual or Important Environmental Events Any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related f to plant operation shall be recorded and promptly reported to the NRC within 24 hours by telephone, telegraph, or facsimile trans-l missions followed by a written report per Subsection 5.4.2. The following are examples: excessive bird impaction events, onsite i plant or animal disease outbreaks, mortality or unusual occurrence of any species protected by the Endangered Species Act of 1973, fish kills, increase in nuisance organisms or conditions and unanticipated or emergency discharge of waste water or chemical substances. I No routine monitoring programs other than as described in the EPP are required to implement this condition. i i 4.2 Environmental Monitoring and Management i Environmental monitoring and management activities shall be i undertaken as outlined in Section 2 and as described in the following. j l 4.2.1 Fog Monitoring j i 4 A fog monitoring program shall be accomplished to document the I frequency of occurrence of natural fog and future cooling lake l 4-1 l t . -. ~,....

) E operation induced fog. A visiometer and continuous recorder l shall be utilized in a conservative location throughout the program. 4.2.2 Waterfowl Impaction A general survey program shall be accomplished to document sig-nificant waterfowl collision events and determine if mitigation is warranted. 4.2.3 Land Management There shall be a land management program instituted at WCGS to provide for revegetation, maintenance and restoration of the WCGS site. This program shall attempt to achieve a balance between production and conservation values on site property through the implementation of conservation and wildlife management techniques. There shall be no reporting requirements associated with this condition. l l l 4-2 .~

a 5.0 Administrative Procedures 5.1 Review and Audit The licensee shall provide for review and audit of compliance with the EPP. A description of the organization structure utilized to achieve the review and audit function and results of the audit j activities shall be maintained and made available for inspection. 5.2 Retention of Program Documentation Program documentation relative to the environmental aspects of plant operation shall be made and retained in a manner convenient { i for review and inspection. Program documentation shall be made available to NRC on request. I t l Documentation of modifications to plant structures, systems and components determined to potentially affect the continued pro-f tection of the environment shall be retained for the life of the plant. All other information, data and finalized reports relating to this EPP shall be retained for five years or, where applicable, I in accordance with the requirements of other agencies. I 5.3 Changes in Environmental Protection Plan I Request for change in the EPP shall include an assessment of the environmental impact of the proposed change and a supporting i justification. Implementation of such changes in the EPP shall l l l 5-1 i

4 I b' not commence prior to NRC approval of the proposed changes in the form of a license amendment incorporating the appropriate revision l i' to the EPP. i 5.4 Plant Reporting Requirements 5.4.1 Routine Reports An Annual Environmental Operating Report describing implementation of this EPP for the previous year shall be submitted to the NRC prior to May 1 of each year. The initial report shall be submitted prior to May 1 of the year following issuance of the operating license. The period of the first report shall begin with the date of issuance of the operating license. The report shall include summaries and analyses of the results of the environmental protection activities required by Subsection 4.2 of this EPP for the report period, including a comparison with preoperational studies, operational controls (as appropriate), previous non-radiological environmental monitoring reports, and an assessment of the observed impacts of the plant operation on the environment. If harmful effects or evidence of trends towards irreversible damage to the environment are observed, the licensee shall provide a detailed analysis of the data and a proposed course of action to alleviate the problem. l 5-2

i } i s' The Annual Environmental Operating Report shall also include: I l l (a) A list of EPP noncompliances and the corrective actions taken to remedy them. 1 (b) A list of all changes in station design or operation, tests, j and experiments made in accordance with subsection 3.1 which involved a potentially significant unreviewed environmental issue. (c) A list of nonroutine reports submitted in accordance with Subsection 5.4.2. In the event that some results are not available ey the report due date, the report shall be submitted noting and explaining the missing results. The missing data shall be submitted as soon as possible in a supplementary report. l 5.4.2 Nonroutine Reports A written report shall be submitted to the NRC within 30 days of occurrence of nonroutine event. The report shall (a) describe, analyze, and evaluate the event, including extent and magnitude of the impact and plant operating characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar i 1 5-3

occurrences involving similar components or systems, and (e) indicate the agencies notified and their preliminary responses. i L l Events reportable under this Subsection which also require reports to other Federal, State or local agencies shall be reported in i accordance with those reporting requirements in lieu of the j requirements of this Subsection. The NRC shall be provided a copy of such report at the same time it is submitted to the other agency. I 1 i 5-4 __I ,}}