Text

Corrected Amends 76,71,72 & 77 to License DPR-58 & Amends 57,53 & 58 to License DPR-74,correcting Typographical & Processing Errors in Original Amends
	ML17320A893
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Site:	Cook
Issue date:	12/15/1983
From:	; NRC
To:	;
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ML17320A894	List: ML17320A893;
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TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS n

C)

'I FUNCTIONAL UNIT 1.

Manual Reactor Trip 2.

Power Range, Neutron Flux 3.

Power Range, Neutron Flux, High Positive Rate 4.

Powe'r Range,: Neutron Flux, High Negative Rate CHANNEL CHECK H.A.

N.A.

H.A.

CHANHEL CHAHNEL FUNCTIONAL CALIBRATION TEST H.A.

Siu(1)

D(2), M(3)

M and Q(6)

R(6)

NODES IN MHICH

'URVEILLANCE-

'E UIREO H,A, 1,

2 1,

2 5,

Intermediate

Range,

. '(eutron Flux 6.

Source

Range, Neutron Flux 7.

Overtemper ature bT 8.

Overpower

)I5T 9.

Pressurize~

Pressure Low 10.

Pressurizer Pressune--High 11.

Pressurizer Mater Level--High 12.

Loss of Flow - Single Loop R(6)-

R(e)

Slu(1}

M. and SIU(l) 3, 2and*

2(7), B(7),

4 and 5

~

1, 2

8312280157 83l2i 5

'<DR aDOCK 05OOOSZS P

-'PDR

TABLE 4.3-1 Continued NOTATI'ON

With the'reactor trip system breakers c'losed and the control rod drive system capable of rod withdrawal.

(1)

If not performed, in previous.? days.

(2)

Heat balance only, above 15K of RATED THERMAL POWER.

J (3)

Compare encore'o excore axial imbalance above 15% of RATED THERMAL POWER.

Recalibrate if absolute difference 3 percent percent.

I (4)

Manual ESF functional input check every 18 months.

(5)

Each train tested every other month.

(6)

Neutron detectors may be excluded from'HANNEL CALIBRATION.

(7)

Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.

D. C.

COOK-UNIT 1

3/4 3-14 Amendment No. 7~~-

n TABLE 3.3-4 Continued ENGINEERED SAFETY FEATURE ACTUATIOH SYSTEH INSTRUHENTATIOH TRIP SETPOINTS CD CD J

FUllCTIOHAL UNIT 6.

tlOTOR DRIVEN AUXILIARYFEEDWATER PUMPS a.

Steam Generator Water Level -- Low-Low TRIP SETPOINT 17K of narrow range instrument span each steam generator ALLOWADLE VALUES 16% of narrow range instrument span each steam generator b.

4 kv Dus Loss of Vol tage c.

Safety Injection d..

Loss of liain Feedwater Pumps 3196 volts with a 2-second delay.

tlot Appl i cab 1 e

- llot Applicable 3196; +10, -36 volts with a 2+.2 second delay Not Applicable Not Applicable O

7.

TURBINE DRIVEfl AUXILIARY FEEDWATER PUflPS a.

Steam Generator Water Level -- Lop-Low Reacto Coolant Pump Bus Und~rvolgpge D..

LPSS OF POWER a>

4 kv Bus Loss of Voltage b

4 kv Dus Degraded Voltage

> 17K of narrow range instrument span each steam generator

> 2750 Voltseach bus 3196 volts with a 2-second delay 3596 vol ts with a 2.0 inin. time delay

> 16$ or narrow range instrument span each steam generator

> 2725 Voltseach bus 3196, +10, -36 volts with a 2+.2 second delay 3596, +36, -18 volts with a 2.0 minute.+ 6 second time delay

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i6.0 ADMINISTRATIVE CONTROLS pl 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this responsi-bility during his absence.

6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organization for facility management and technical support. shall be as shown on Figure 6.2-1.

II FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b.

At least one licensed Operator shall be in the control room when fuel is in the reactor.

c.

At least two licensed Operators shall. be present in the control room during reactor start-up,. scheduled reactor shutdown and during recovery from reactor trips.

d.

An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

e.

ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior'eactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f.

A site Fire Brigade of at least 5 members shall be maintained onsite at all t~s.

The fire Brigade shall not include 3

members of the minimum shift crew necessary for safe shutdown of the unit or any personnel required for other essential functions during a fire emergency.

g.

The amount of overtime worked by plant staff members performing safety-related functions must be limited in accordance with NRC Policy Statement on working hours (Generic Letter No. 82-12).

D. C.

COOK - UNIT 1 6-1 Amendment No. 77:

i' I

n.

AUTHOR.TY T;",= rrGi. 5 Cu 't,",u,qr0=r:)

a 1G 1 pie,.

n lAg proceGUl es ol (solidifiCetiori Of radioactive wastes at least once'per.24'-

sssors ihs

~

The performance o

activi-.ies required by the gual jty Assurance Program to meet the c. iteria of Regulatory Guide 1.21, Rev.

June 1974 and Regulatory Guide 4.1, Rev.

1,! April 1975 at lees; once per 12 months.

6.5.2.9 The NSDRC shall report to arC advise the Vice Chairman, Engineering and Cons.ruction, A:PSC, on those areas o.- resoonsibili iy specified in Sections 6.5.?.7 and 6.5.2.8.

. 5. 2. 10 Recorcs o-,

HSDRC a tivities shall be prepa. ed, approved and oistribvteC as irdicated below:.-

b.

tiirsutes 0 i e-

'AEPSC)

'Wi tslil'i ch NSDRC l 'eeti nc she 1:1 be the Vic=- Chairman, Encineo 14 C vs foiiqwin each Repor.S'f rev j ewS esCa-,paSSe" by ceC

prepared, ap=rov=r and forwardod

.o

-..".=-

anC Cnnstructior, AEPSC, wi;h'n 14 days

.he review.

prepal od approvod rlrc and Constru

-.lors) eting.

or. 6.5.2.7 abov, shall b'ollcwinc c

>>-lie-icr. of r

s'sis Wc

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7" e o"..1 sslQA rig r << ~

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0.

C.

COOK UNIT 1

6-12 Amendment Nos ?2

ADMINISTRATIVE CONTROLS 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 follow-ing 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 conmencement of comnercial power operation),

supplementary reports shall be submitt d at least every three months until all three events have been completed.

Imam ~oars

~

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 prior to March 1 of the year following initial criticality.

6.9.1;5 Reports required on an annual basis shall include:

a.

A tabulation on an annual basis of the number of station, utility and other personnel

'(including contractors) receiving exposures greater than 100 mrem/yr and their assocSged man rem exposure according to work.and

)ob functions, e.g.,

reactor operations an8 surveillance, inservice inspection, routine maintenance, special maintenance "

(describe maintenance),

waste processing, and refueling.

The dose assignment to various duty functions may be estim'ates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20X of the individual total dose need not be accounted for.

In the aggregate, at least 80X of the total whole body dose received from externa1 sources shall be-assigned to specific ma)or work functions.

b.

The complete results of steam generator'"

tube inservice inspections performed during the report period (reference Specification 4.4.5.5.b).

c.

Documentation of all challenges to the pressurizer power operated relief valves (PORVs) or safety valves..

1 A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common to all units at the station.

2 This tabulation supplements the requirements of 20.407 of 10 CPR Part 20.

'D. C.

COOK - UNIT 1

6-15 c

Amandmeat No. 77

TABLE 4,3-]

REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE RE UIREMENTS FUNCTIONAL UNIT 4

1.

Manual Reactor Trip 2.

Power Range, Heutron Flux

//

3.

,Power Range, Neutron Flux,

)(jgh Positive Rate ).

.rr

/

4.

Pov)er Range, Neutron Flux, High Negative Rate 5.

Intermediate

Range, Neutron Flux CHANNEL CHECK N.A.

N.A.

CHANNEL CAL'IBRATION N.A.

O(2), M(3) and q(6)

R(8}

R(6}

R(6)

CHANNEL FUNCTIONAL TEST S/U(1,)

S/U(1)

MODES IN WHICH SURVEILLANCE RE UIRED

,RA; 1

/2 1

2 1 j 2 a

1, 2 and

6.

Source

Range, Neutron Flux 7.

Overtemper ature hT 8,

Overpower hT 9.

Pressurizer Pressure Low 10.

Pressurizer Pressure--High 11; Pressurizer Hater Level--High 12.

Loss of Flow - Single Loop S

R(6)

R

)4 and S/U(l) 2(7) 3(.7).

4 I))

M 1,

2

l. 2 1 j 2

1, 2

lj 2

TABLE 4.3-1 Continued NOTATION (1)

(2)

(3)

M'ith the reactor trip system breakers closed and the control rod dr~ve system'capable of rod withdrawal.

If not performed in previous 7 days.

He t balance only, above 154 of RATED THERMAL POWER.

Adjust channel 'if absolute difference

> 2 percent.

Compare incore to excore axial offset above 15% of RATED THERMAL POWER.

Recalibrate if absolute difference

> 3 percent.

Manual ESF functional input check every 18 months.

(5)

Each train tested every other month.

(6)

Neutron detectors may be excluded from CHANNEL CALIBRATION.

(7)

Below P-6 (BLOCK OF SOURCE RANGE REACTOR TRIP) setpoint.

D.

C.

COOK - UNIT 2 3/4 3-13 Amendment No.; '7

CDn Pl CD CD TABLE 3. 3-4 Continued ENGINEERED SAFElY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT 6.

MOTOR DRIVEN AUXILIARYFEEDMATER PUMPS a.

Steam Generator Mater Level -- Low-Low TRIP SETPOINT '

21X of narrow range instrument span each

.steam generator ALLOWABLE VALUES R 20X of. narrow range instrument span each steam generator GJ 7

b.

4 kv Bus Loss of Voltage c.

Safety Injection d.

Loss of Hain Feedwater Pum s

p TURBINE DRIVEN AUXILIARYFEEDMATER PUMPS a.

Steam Generator Water Level -- Low-Low b.

Reactor Coolant Pump Bus Undervoltage 3196 volts with,a 2 second delay Not Applicable Hoi Applicable R 21X of narrow range instrument span each steam generator R 2750.Volts--each bus 3196, +10

-36 volts with a 2 i 0..2 second delay

. Hot Applicable Hot Applicable R 20X of narrow range instrument span each steam generator R 2725 Volts--each bus O

8r 'OSS OF POWER a.

4 kv Bus Loss of Voltage b.

4 kv Bus Degraded Vol tage 3196 volts with a 2 second delay 3596 volts with a 2.0'minute time delay 3196~ +16

-36 volts with a 2..i:0.2 second de'lay

+36 ~ -1 8 v~] ts with 2.0 minute t 6 second time delay

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Bc S:-S 3/4.7.6

~MP ~ILATIONSYS~~

CP The OPZRABIEITY of the ESP ventilation sys em ensures that radio-active materials lea3cing from the ECCS equipment within the pep oom following a TEA aze filtered p ioz to reaching the environmen The ope=ation of 8's systmn and the resulwt effect on offsite aosage cal-culations was assumed in the acciaent analyses.

3/4. 7. 7 EKBAKIC SRHBH5

~l sos a

e ecu'd OP~V. to ~e tha" the stw~~al teg ity of the reactor coolant system ana aQ other safety related systems is main~ed du='ng and following a seimc or othez event iwtia~g dynamic loads.

Snubhezs excluded from Ms inspection prog am are those insalled on nonsafety-related systems and then on'y 'f ~Heir fa'luze;or failu=e of the system on which Bey are iwta>>ed, would have no adve

'se ef ect on any safety-related system.

The visual iwpecaon f agency is basea upon maint~~g a consmt level of snuhbe protec 'on to system.

The efoze, the zec~ed inspec"'on inte~ va-ies inve sely w'th KD ohse~ed snubber failu=.es and. 's dete~'ned....

by the ne'er of inope able snuhhe s founa d-ing an inspection.

Inspections performed before tMt interval has elapsed may be used as a new refe ence point to dete~e the next inspec"ion.

Boweve,

~Me results of such ea=ly pections pe~armed before the original ecuized we internal has elapsed (nominal ~e less 25%)

may not be used to 3.eng Men the r~ed inspection intewal.

Any inspec""'on whose results remi'"ed a shorter inspection inte~ wi'l ove~ide the prev'ous

scheaule, When the cause of the rejection of a snubbe" is clea ly esMlished and remedied. for that.snubbe~

ana for any other snMaezs "m" ~y ee gene=ical'y susceptible, ana vezi ied by insezvice funct'onal tes~g, ~t snubber may be exempted frcm heing counted as inoperable.

Generically susceptible snuhbe"s aze those which aze of a soecific make c moael and have the same design features di=ectly related to zejection of the snubhe" by v'sual inspection, or aze s~larlv located or exposed to the same environmental condit'ons such as temperature, radiation, and vibration.

When a snubber is found inoperable, an engineering evaluat'on is pe~armed, in addition to the detestation o

the snubber mode of

ailure, in o der to dete~ne if any safety-related component or sys-em has been advezsly a fected by the inoperability o the snubbe The enginee ing evaluation shall dete~e whether or not the snubbe" mode of failuze has impar ed a significant effect or degradation on the supported component or system.

D. C.

COOK - UNIT 2 B 3/4 7-5 Amendment No',

53

i,

~

BASM To -~de ass~ance of snuhbe func='onaL relia~i=J',

a reoresentat've sample of the insured sn~~em wiLL be un@ 'onally tested du~kg plan shu~ns at 18'onth inta~i, Observed failures o<< these sawle snuhhers sha3 1 ecui e ~c on2Ll te~g of additional unL Me s~me 3'

of a snnhher is evaluated Wm manufacturer's input and-',:

czm"'on "~~gh c=ns&~eration of the sw&1e semce cond=tions

and,

'"'ssoci2Lted i>>>>~~

~ on 2L1d ~tenanc8 eco<<~M (ncwl'v ~~<<aLLed snubber g

seal repLLced, s- ~g ~pla~, in high ~Ha 'on a ea,, "gh t~e~t=e a=ca es"..}.

Me recu -erne to mon'>>~

Me snye= se~ce

~s'~e 's c'uded N 4;=e M= >>Ae snu~'~

oe "od"ca'~y undem~

a pe~once evalw~ in v.'ev o" Ze'- age and opening c=M"='ow.

Nese rods vM~ -~M st~ ~meal m~es c

fn

-e c=ns'de=a='cn o

snuhhe= serv'e

~'ce.

he ~'-emena. for De ~twice of reco ds and Ae ma&er se~ce

e re."'.e~ a=e ~ ~tended>>m a

ec'plant ope~~.

3/4. 7. S S~D 'CUB~ C2H~~TZQH cn Bmovahle ~~rLina~

or so'~~>>

<<~<<<<i <<g 'g 1 ooA 4 <<c z1 aha ymca ~<<s is hased cn LQ ~

7Q g9 (c) ~g<<ts go>>

en~e ~t Leakage ~ hyurodnc", st c

and soec~

nnclea>> w-e=M souce>> sviLL not exceed. a~'avahle w<<~ vaLLles.

3/4. 7. '9 ~~ SQPKUSSZON'YP~)

r 4 QP~

of Ae '-e sL~ess~

sos-ms ensu=es ~ ad~

ce <<a s wa<<es~aw camah+4<<y is avai@ll 4

<<to c ~~+4 e and ext~i~h r<<<<es oc==-g in any po='on o

the fatty vhe e s2Lfety ela~ ~pnL'ent Loca ed.

The c'~e supp ession syst2sn consis

~ of.>>>e ~ter ~~

s-=ay ~/cc. sp=~e s, M2, Salon and

~-e hose s~tions.

Me co13.ec-

"='ve c2Lpah&> of the <<ire suppression sys m is 2Ld2acaate to ~'Lm' po en>>~ hmge w safety r2LlaM ~anent

'and is a major elenent in Ae a~ y ~~-e detection grog 2LnL Ea the event thar erne or mora of tha required Lav prsasura CO systems a

4 isolated or personnel protacsionf to peMtebi 4

ry fc routine tours, n~"exes, construction or survailm~ce tas~~,

the f~ e deca "cn systs (s} ra~ad by spec'~ica "cn 3,3.3.8 s~ ba ve '<<iad to ba oparabla 2Lnd a M~ Para Patch Pat ol essabXMad

~a M affected a"aas. not occlqiad by vor'hers..

Tha Rov~ 7'w Watch Pa@ o'(s) shall cons'st of ona or nLoro parsons Mevladgaabla of tha cation. ~ oparat'cn o<< fira fight~ equipaans and good

'=a

~

pzotactionl pcrsowel safety p ac 'cas such as ~tecanca of access and eg=.ass rou Rs and pe scnnal accounsabil'~

maasu as.

The f~c~ns of the M~ Eira Viasch Patwl can ba fulXQLad by psrsonnaL ~~lvad ~~

D. C.

COOK UN' B 3/4 7>>6 Amendment NO.

53

I I

6.0 ADMINISTRATIVECONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the succession to this respon<<

sf.bility during his absence.

6 '

ORGANIZATION OFZS ITE 6.2.1 The offsite organization for facility management and technical support shall be as shown in Figure 6.2-1.

FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:

a.

Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.

b.

At least one licensed Operator shall be in the control room when fuel is in the.reactor.

c.

At least two licensed Operators shall be present in the control room during reactor start-up, scheduled -reactor shutdown and during recovery from reactor trips.

d.

. M individual qualified in radiation protection procedures shall be on site when fuel is in the reactor.

e.

ALL CORE ALTERATIONS shall be directly sdpervised by either a licensed Senior Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.

f.

A site Fire Brigade of at, least 5 members shall be maintained onsite at all times.

The Fire Brigade shall not include 3 members of the minimum shift crew necessary for safe shut-down of the unit or any personnel required for other essential functions during a fire emergency.

g.

The amount of overtime worked by plant staff members performing safety-related functions must be limited iri accordance with the NRC Policy Statement on working hours (Generic Letter No. 82-12).

D.

C.'NCOOK - UNIT 2 6-1 Amendment No.~5S

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.ADMINISTRATIVE CONTROL'S I ~

, power operation)

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supplementary reports shall be submitted at least every three months until "all three events have been completed.

'ANNUAL REPORTS

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~r 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 prior to March 1 of the year following initial criticality.

6.9.1.5 Reports required on an annual basis shall include:

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem/yr and their assocfged man rem exposure according to work and gob functions, e.g.,

reactor operations anh surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance),

waste processing, and refueling.

The dose assignment to various duty functions may be estimates based on pocket dosimeter, TLD, or film badge measurements.

Small exposures totalling less than 20K of the individual total dose need not be accounted for.

In the aggregate, at least 80X of the total whole body dose received from external sources shall be. assigned to specific major work functions.

b'.

The complete results of steam generator tube inservice inspections performed during the report period (reference Specification 4.4.5.5.b).

C ~

Documentation of all challenges to the pressurizer power operated relief valves (PORVs) or safety valves.

1 A single submittal may be made for a multiple unit station.

The submittal should combine those sections that are common to all units at the station.

2 This tabulation supplements the requirements of 20.407 of 10 CFR Part 20.

D. C.

COOK - UNIT 2 6-15 Amendment Ho. f 88-e r rg

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ML17320A893

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