Text

Proposed STS Re Safety Sys Settings,Reactor Protective Instrumentation & ESF Actuation Sys Instrumentation.Proposal for one-time Exemption to Tech Spec 3.7.1.2 Allowing Thorough Test of Auxiliary Feedwater Actuation Sys Encl
	ML20066F973
Person / Time
Site:	Calvert Cliffs
Issue date:	11/17/1982
From:	; BALTIMORE GAS & ELECTRIC CO.
To:	;
Shared Package
ML20066F961	List: ML20066F959; ML20066F973;
References
	NUDOCS 8211190249
	Download: ML20066F973 (24)
	v • d • e

_ ._ _ _ _

s t

on 9

r JARLE 2.2-1 (Cent'd) 23; Tf i REACTOR PROTECTIVE INSTRUNENTATION TRIP SETPOINT LIMIIS -

! j;3 Ei l 0 0 ra 9% C filNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES o* :;l .

l 88 4'. Pressurizer Pressure - liigh 1 400 2 psia 1 2400 psia o-uO" E 5. Containment Pressure - litch -< 4 psig ~< 4 psig

. @7 El 685 6tf

- 6. SteamGeneratorPressure-Low (2) 1-H6 psia ; 1 H e psia l t

7. Steam Generator Water Level - Low > 10 inches below top > 10 inches below top of feed ring. Ff feed ring.

D. Axial flux offset (3) Trip setpoint adjusted to Trip setpoint adjusted to

not exceed the limit lines not exceed the Ilmit lines of Figure 2.2-1. of Figure 2.2-1.

9. Thermal Hargin/ Low Pressure (1)

a. Four Reactor Coolant Pumps Trip setpoint adjusted to Trip setpoint adjusted to 1 Operating not exceed the limit lines not exceed the limit lines of Figures 2.2-2 and 2.2-3. of Figures 2.2-2 and 2.2-3.

h. Steam Generator Pressure ~

< 135 psid < 135 psid Dif ference - !!!gh (1) ~

10. Loss of Turbine -- Hydraulic 1 1100 psig > 1100 psig -

@ Fluid Pressure - low (3) n -

Rate of Change of Power - liigh (4) y 11. 1 2.6 decades per minute 1 2.6 decades per minute

) g: .

1 TABLE NOTATION

-4 (1) Tripmaybehypassedbejow10 1 of RMED THERIM POWER; bypass shall be' automatically removed when M Ill[lulAL POWElt is >_ 10 % of RAILD TilFRHAl. PollE!!.

i

9 .

G TABLE 2.2-1 (Con t,* d ) -

El

~'

TABLE NOTATIONS (Cont'd) .

U

j. (2) Trip may be manually bypassed below 6M psla; bypass shall be autoisatically7removed 72 5 p at v.

. (3) Trip may be bypassed below 15% of RATED TilERHAL P0llER; bypass shall be automatically removed when g IllEltlML P0llER is > 15% of RATED TilERilAL POWER. -

(4) Trip may be hyp.4ssed below 10 1 -4 and above 12% of RATED TilERHAL POWER.

!. z*.

~

L 6ie o

R

, e il 8

2 e

e

%g

i b

/

, f. .N. T *.u

'l* '. A P =.~. V. t.'K. .~T_.1 t_ _~i -?

i . ".. *e5 ,

t' EAsis

~

- /. 2 3 c;eration of the react:r at reduced ; wer if cne or we react:r c: clan:

pu=ps are taken out of service. The icw-ficw M; en cic s anc A11 cwa:le

','alue f:r :ne vaMcus react:r c:olan: ;u== c=bir ti:ns have been dedved in c:nsiceratien of ins:-=en: err:rs and res ense tim equic ent involvec :: =ain ain =e U.'iER above h u:nceralner:eration es of I and ex;ec ad transients. ror react:r = era-ion win enly rso or =ree reac: r c: lan: c:s :: era:ing, the Peac=r C :lan: .:lew-L:w tri: se:-

points, the Pcwer Lavel-High tric se=:ints, anc =e Ther al Margin /L:w Pressure tri; sa=:ints are au:=atically enanged wnen the :=; =nciti:n select:r swich is =ar.ually sa: :: :he desired wo- er =ree-.;c: g.7,3 position. Changing nase :M: se=oints curing :,,c anc :nree :u :

operaticn =revents the =ini=um value of DNER fr:m ceing beicw 1.195 during i normal :: era:icnal transients and anticipa:ad transients wnan Only wo r three reac= r cc lan: pu=;s are c;erating.

Pressurizer Pressure-Hieh The Pressuri:e'r Pressure-Hign tric, ba xed u: by the :ressuri:er i= e safety valves anc main sten line safety valves. :r: vices reacer c:::an: .

system pretactica agains: OvertressuM :ation in th2 aven: Of loss Of loac witacu: reac=r :M :. This trip's se:=in is ICO :si telew ce nemual lift se::ing (2500 psia) Of the pressuri:er ::ce safety valves and its c:ncurren: c: era:icn wita he ;cwer-o:erated relief vcives avoics :ne undesiracle cpera:ica of the pressuri:er =ce safety valves.

Centairrent : essure Wieh The 0:ntainment Pressure-Hign tri; revices assurance that a rea:= r tri: is ini-iatac ::ncurrently 41 = 1 safety injecti:n. he sen:1n:

f:r :nis trip is icentical to the safety injecti:n set; int.

Stea. Generan t 2 essure-Lew The Steam Genera::r Pressure-L w M ; revices :re:ectic. agains l

an excessive subsecuent c: ole:wn rate Of we Of reac=r hea; :=lan:.

extracti:n fr:ning he se . =ef staam n ;sia genera =rs jgg is sufficiently belcw :ne full-lcac c:erating : ia.: Of 550 :sia s:

as no: = inter See with ner-:al =eratica, but still nign en ugn =

pr: vide :ne recuirec pre:ectica in ce even: cf excessively Sign staa fl ew.

Sg l ~11s se :ing was used wit: anuncertain:yfac=r:f_ psi g l in the ac:ican analyses.which wa4 ba4ed m At % gg yge ww6

(

CAL.t...

u . . .,,

. . .-:. -Ut. u 1 3 2-2

,.cenc en: ic. ..u , -c

LIMITING SAFETY SYSTEM SETTINGS 1

BASES ,

l Steam Generator Water Level .

l The Steam Generator Water Level-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity and assures that the pressure of the reactor coolant system will not exceed its SEE 3afety Limit.f 9; :pnified ntp;i-t ; :';id;; ells....ca that the. . wl11 msar be ruf'm e"* water !4 ::::r;> in -h; ::::r ; re-?t:r: t += the M t..-ip ;c p.e.,J . m ,3;.. gf m;,. ..m..jS .ainutu uciw.. c "i:ry ivaar M G E

-f::i;;t;r is re wir;d.

Axial Flux Offset f,2 3 The axial flux offset trip is provide ' to ensure that excessive axial ceaking will not cause fuel damage. The axial flux offset is determined from the axial;y split excore tectors. The trip setpoints ensure that neither a DNBR of less than 1.195 nor a peak linear heat rate which corresponds to the temperature for .ue centerline melting will exist as a consequence of axial power ma1 distributions. These trip set-

. points were derived frorr an analysis of many axial power shaces with

, allowances for instrumentation inaccuracies and the uncertainty associated

. [ with the excore to incore axial flux offset relationship.

i .

!TherralMarcin/LowPressure g3 il ure trip is provided to prevent operation The Themal Margin / Low Pr when the DNBR is less than 195 .

/87f "

The trip is initiated en er the reactor coolant system pressure

' signal drops below either 750 psia or a ecmputed value as described below, whichever is higher. ihe computed value is a function of the

highee of ai power or neutron power, reactorThe inlet temperature, and the .

minimum value of reactor I number of reactor coolant pumps operating.

P coolant flow rate, the maximum AZIMUTHAL 70WER TILT and the maximum CEA deviation gemitted for continuous operation are assumed in the genera-tion of this trip function. In addition CIA group secuencing in accor-Finally, the dance with Specifications 3.1.3.5 and 3.1.3.5 is assumed.

, maximum insertion of CIA banks which can occur during any anticipated l operational occurrence prior to a Power Level-High trip is assumed. '

I l

3 2-6 Amendment No. 73. 3I OAlt.'Er CLIFFS - UNIT 2 t

l l

Tce specified setpoint in ccebination with the auxiliary feedwater actuation system ensures that sufficient water inventory exists in both steam generators to remove decay heat following a loss of main feedwater h

flow event. ;

I

, e

t e

TABLE 3.3-1 (Continued)

~

,e TABLE NOTATION J

' *With the protective system trip breakers in the closed position and .

the CEA drive system capable of CEA withdrawal.

The provisions of Epecification 3.0.4 are not applicable. ,

(a) Trip may be bypassed below 10~4 of RATED THEVAL POWER; bypass shall be automatically removed when THERMAL POWER is~ > 10-' of RATED THERMAL POWER.

(b) Trip may be manually bypassed below automatically removed at or above 465 psia.

psia; bypass shall .be -[

~45 (c) Trip may be b'ypassed'below 15% of RATED THEFAL POWER; bypass shall be automatically removed when THEPAL POWER is ~

> lE of RATED THEFAL POWER.

(d) Trip may be bypassed below 10 % and above 12% of MTED THE??AL POWER.

(e) Trip may be bypassed during testing pursuant to Special Test Excep- i tion 3.10.3. J (f). There shall be at least two decades of overlap between the Wjde -

Range Logarithmic Neutron Flux Monitoring Channels and the Power

. Range Neutron Flux Monitoring Channels. -

ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Cnannels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and/or open the protective systam trip breakers.

ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in either tha bypassed or tripped condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . For the purpcses of testing and maintenance, the inoparable channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of initial loss of OPERABILITY; hcwever, the inocerable channel shall then be either restored to OPERABLE status or placad in the tripped condition.

1 CALVERT CLIFFS - UNIT 1 3/4 2 4 Acencmen: No . 1-5 l

o ATTACHMENT 2

% TABLE 3.3-3 (Continued) 63 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION kg:

' MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE

,9 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION t1 ro 8. CVCS ISOLATION

a. Manual (CVCS isolation Valve Control Switches) 1/ Valve 1/ Valve 1/ Valve 1, 2, 3, 4 6

b. West Penetration Room / Letdown Heat Exchanger Room

} Pressure - High 4 2 3 1, 2, 3, 4 7*

'f 9. AUXILIARY FEEDWATER

$ Actuation System

a. Manual (Trip Buttons) 2 sets of 2 I set of 2 2 sets of 2 1,2,3 6 per S/G per S/G per S/G

b. Steam Generator Level - Low 4/SG 2/SG 3/SG 1,2,3 7 4

c. Steam Generator

A P High 4/SG 2/SG 3/SG 1,2,3 7 if a

a

.b

[r .' a" --. . : ww. .

i

- TABLE 3.3-3'(Continued) l TABLE NOTATION f

/705 I

- (a) Trip function may be bypassed in this MODE when pressuri:er * ':

pressure is < psia; bypass shall be automatically removed when pressurizar pressure is 1 W98 psia.

igee (c) Trip function may be bypassed in this MODE below kpsia; bypass r h

shall be. automatically removed at or above ces psia.

W -

The provisions of Epecification 3.0.4 are not applicable.

ACTION STATEMENTS .

ACTION 6 - With the number of OPERABLE channels one less than the Totil Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least

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

ACTION 7 - With the number of OPERABLE channels one less than the Total Number of Channels, operation may proceed provided the following conditions are satisfied:

a. The inoperable' channel is placed in either the bypassed or tripped. condition within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . For the purposes of testing and maintenance, the inoperable channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from time of initial loss of OPERASILITY; however, the inoperable channel shall then be either restored to OPERABLE status or placed in the tripped condition.

b. Within one hour, all functional units receiving an input from the inoperable channel are also placed in the same condition (either bypassed or tripped, as applicable) as that required by a. above for .the inoperable channel.

c. The Minimum Channels OPERABLE requirement is met; however, one additional channel may be bypassed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months while performing tests and maintenanca on that channel provided the othat inoperable channel is placed in the tripped condition.

3/4 's-15 k:endment No.48 CALVERT CLIFFS - UNIT 1

= __ _

__ )

i .

TABLE 3.3-4 99 GG ENGlHEERE0 SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES .,

gm .

HH pP ALLOWABLE mm Ip4CTIONALUNIT TRIP SETPOINT VALUES g

' ?AN

, i 1. SAFETY INJECTION (SIAS) cc a. Hanual(TripButtons) Not Applicable Not Appitcable .

I i35 HH i

b. Containment Pressure - liigh 1 4.75 psig 1 4.75 psig .

N'.

c. Pressurizer Pressure - Low >Ja psia > M78 psia ;

~ esas

<nf

' 2. CONTAINHENT SPRAY (CSAS)

a. Manual (Trip Buttons) Not Applicable Not Applicable

't

!I b. Containment Pressure -- High 1 4.75 psig 1 4 75 psig i

3. CONTAlHMENT ISOLATION (CIS) g i ,!

{ a. Manual CIS (Trip Buttons.) fiat Appilcable Not Appilcable I

~

{ b. Containment Pressure - liigh 1,4.75 psig 1 4 75 psig

~ .

4 HAIN STEAM LINE ISOLATION

a. Manual (HSly lland Switches , !-

and feed licad Isolation l.

FN llandSwltches). Not Appilcable Not Applicabis

88 Q b. Steam Generator Pressure - Low > M psia > 998 psia .;

h,.Ra,. ~6TS 6e6 .

2z ?

PP # Containment isolation of non-essential penetrations is also initiated by SIAS (functional -

@ units 1.a and 1.c). ,

p -

l;

O ATTACHMENT 3 1

, TABLE 3.3-4 (Continued) p ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES 4 ,

E

, ALLOWABLE g FUNCTIONAL UNIT TRIP VALUE VALUES

8. CVCS ISOLATION to West Penetration Room / Letdown Heat f 0.5 psig <0.5 psig Exchanger Room Pressure - High

9. AUXILIARY FEEDWATER ACTUATION SYSTEM

a. Manual (trip buttons) Not Applicable Not Applicable y" b. Steam Generator (A or B) Level- Low -194" to -149" -194" to -149" w (inclusive) (inclusive) h c. Steam Generator 4 P-High f 130.0 psid f 130.0 psid #

(SG-A > SG-B)

d. Steam Generatord P-High $ 130.0 psid $ 130.0 psid (SG-B > SG-A) e n

EF O

., . wp y- _' ^ ~' ' ~

s --

TABLE 3:3-5 y ENGINEERED ~SAFETf FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS~ l 1

1. Manual

a. ' SIAS Safety Injection. (ECCS) Not pplicable

b. CSAS Containment Spray Not Applicable

~

c. CIS -

containcent. Isolation Not Applicable

d. - RAS Contain=ent Sump Recirculation Not Applicable

_ [_ , ku##.seag FrewAffg houps ' ~ [h [phhg3,gg

2. Pressur4:er F essure-Lcw

a. Safety Injection (ECCS) < 30*/.30",

a Containment ?* essure-Hich 3 .-

a. Safety Injectien (ECCS) 1 30*/30 "

b. Containment Iselation 1 30

c. Contain=ent Fan Coolers 1 35*/10 "

4 Containment Pressure--Hich

a. Containment Spray 1 50*/60 "

S. Containment Radiatien-Mich

a. Containment Furg~e Valves Isolation 15 t

v l

CALVERT CLIFFS - UNIT 1 3/i 3-E0 Amendmen- No. 48

ATTACHMENT 4 TABLE 3.3-5 (Continued)

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

6. Steam Generator Pressure-Low I

a. Main Steam isolation f 6.9

b. Feedwater Isolation 6 80

7. Refueling Water Tank-Low

a. Containment Sump Recirculation f30

8. Reactor Trip

a. Feedwater Flow Reduction to 5% 6 20

9. Loss of Power

a. 4.16 kv Emergency Bus Undervoltage i 2.2***

(Loss of Voltage)

b. 4.16 kv Emergency Bus Undervoltage f 8.4***

(Degraded Voltage)

10. Steam Generator Level- Low

a. Steam Driven AFW Pump _$ 54.5

b. Motor Driven AFW Pump I 54.5* / 14.5**

11. Steam Generator AP-High

a. Auxiliary Feedwater Isolation f 20.0 TABLE NOTATION

Diesel generator starting and sequence loading delays included.

- Diesel generator starting and sequence loading delays not included.

Offsite power available.

- - Response time measured from the incidence of the undervoltage condition to the diesel generator start signal.

CALVERT CLIFFS - UNIT 2 3/4 3-21 Amendment No.

J

k .

ATTAC"tBE.'r 5 -

O TABLE 4.3-2 (Continued) h '

j I ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL MODES IN WHICH E CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

, FUNCTIONAL UNIT CHECK CALIBRATION TEST _ REQUIRED

5. CONTAINMENT SUMP RECIRCULATION (RAS)

N

a. Manual RAS (Trip Buttons) N.A. N.A. R N.A.

b. Refueling Water Tank - Low N.A. R M 1, 2, 3

c. Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3

6. CONTAINMENT PURGE VALVES ISOLATION ##

a. Manual (Purge Valve Control Switches N.A. N.A. R N.A.

w b. Containment Radiation - High

) Area Monitor S R M 6 W

L

7. LOSS OF POWER

" a. 4.16 kv Emergency Bus Undervoltage (Loss of Voltage N.A. R M 1, 2, 3

b. 4.16 kv Emergency Bus Undervoltage (Degraded Voltage N.A. R M 1, 2, 3

8. CVCS ISOLATION N.A. R M 1, 2, 3, 4 '

West Penetration Room /

Letdown Heat Exchange Room Pressure - High p

d 9. AUXILIARY FEEDWATER

" a. Manual (Trip Buttons) q N.A. N.A. R N.A.

b b. Steam Generator Level-Low S R M 1, 2, 3 B c.

d.

Steam Generator AP-High 5 R M 1, 2, 3 U

Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3 P

1. Containment purge valve isolation is also initiated by SIAS (functional units 1.a,1.b and 1.c).

.1

h o ATTACHMENT 6 h TABLE 3.3-9 a

REMOTE SHUTDOWN MONITORING INSTRUMENTATION d '

' MINIMUM READOUT MEASUREMENT CHANNELS C INSTRUMENT LOCATION RANGE OPERABLE b

rv 1. Reactor Trip Breaker Indication Cable Spreading Room OPEN-CLOSE 1/ trip breaker N

2. Reactor Coolant Cold Leg Temperature 2C43 212-7050 p g

3. Pressurizer Pressure 2C43 0-4000 psia 1

4. Pressurizer Level 2C43 0-360 inches 1 m 5. Steam Generator Pressure 2C43 0-1200 psig 1/ steam generator N

[ 6. Steam Generator Level 2C43 -401 to +63.5 inches 1/ steam generator M.

b il a

(t

9 ATTACHMENT 7

g. TABLE 4.3-6 P

, REMOTE SHUTOOWN MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS o

CHANNEL E INSTRUMENT CHECK CALIBRATION i

g; 1. Reactor Trip Breaker Indication M N.A. 4

$ 2. Reactor Coolant Cold Leg Temperature M R

3. Pressurizer Pressure M g

4. Pressurizer Level M R

5. Steam Generator Pressure M R

. w 6. Steam Generator Level M R

b Y

E a

EF O

d l

O ATTACHMENT 8 l h '

{; TAug 3.3-10 9 POST-ACCIDENT MONif0 RING INSTRUMENTATION b

a I

q MINIMUM h

H CHANNELS INSTRUMENT OPERABLE ro

l. Containment Pressure 2 N

2. Wide Range Logarithmic Neutron Flux Monitor 2

3. Reactor Coolant Outlet Temperature 2 g l 4. Pressurizer Pressure 2 Wa 5. Pressurizer Level 2

'f 6. Steam Generator Pressure 2/ steam generator

7. Steam Generator Level (Wide Range) 2/ steam generator M

8. Auxiliary Feedwater Flow Rate 2/ steam generator

9. RCS Subcooled Margin Monitor 1

10. PORV/ Safety Valve Acoustic Flow Monitoring 1/ valve ,

11. PORV Solenoid Power Indication 1/ valve u

lb n

if O

o ATTACHMENT 9 D,

te TABLE 4.3-10 o POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS V<

h CHANNEL CHANNt:L INSTRUMENT CHECK CALIBRATION E2 1. Containment Pressure M R 2 M

ru 2. Wide Range Logarithmic Neutron Flux Monitor M N.A.

3. Reactor Coolant Outlet Temperature h4 R

4. Pressurizer Pressure M R

5. Pressurizer Level M R

6. Steam Generator Pressure M R Y= 7.

1 Steam Generator Level (Wide Range) M R [

8. Auxiliary Feedwater Flow Rate M K

9. RCS Subcooled Margin Monitor M R

10. PORV/ Safety Valve Acoustic Monitor N.A. R

11. PORV Solenoid Power Indication N.A. N.A.

N* ,

n

.EF

J .

PLANT SYSTEMS ATTACHMENT 10 AUXILIARY FEEDWATER SYSTEM LIMITING CONDITION FOR OPERATION 3.7.1.2 Two auxiliary feedwater trains consisting of one steam driven and one motor driven pump and associated flow paths capable of automatically initiating flow shall be d OPERABLE. (An OPERABLE Steam driven train shall egnsist of one pump aligned for automatic flow initiation and one pump aligned in standby .)

APPLICABILITY: MODES 1,2, and 3.

ACTION:

a. With one motor driven pump inoperable:

1. Restore the inoperable motor driven pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b. With one steam driven pump inoperable:

Y

1. Align the standby steam driven pump to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

2. Restore the inoperable steam driven pump to standby status (or OPERABLE status if the other steam driven pump is to be placed in standby) within the next 30 days or be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. Whenever a subsystem (consisting of one pump, piping, valves and controls in the direct flow path) required for operability is inoperable for the performance of periodic testing (e.g. manual discharge valve closed for pump Total Dynamic Head test) a dedicated operator will be statione,d at the local station with direct communication to the Control Room. Upon completion of any testing, the subsystem required for operability will be returned to its proper status and verified in its proper status by an independent operator check.

With the requirements of 3.7.1.2.a. or 3.7.1.2.b. met, the provisions of specification g 3.0.4 are not applicable.

IA standby pump shall be available for operation but aligned so that automatic flow 4 initiation is defeated upon AFAS actuation.

CALVERT CLIFFS-UNIT 2 3/47-5 Amendment No.

1

ATTACHMENT 11 PLANT SYSTEMS SURVEILLANCE REQUIREMENTS 4.7.1.2 Each auxiliary feedwater flowpath shall be demonstrated OPERABLE:

a. At least once per 31 days by:

1. Verifying that each steam driven pump develops a Total Dynamic Head of 2 2800 ft. on recirculation flow. (If verification must tm @

demonstrated during startup, surveillance tgting shall be performed upon achieving an RCS temperature 2 300 F and prior to entering MODE 1).

2. Verifying that the motor driven pump develops a Total Dynamic Head of 2 3100 ft. on recirculation flow.

3. Cycling each testable, remote operated valve that is not in its operating position through at least one complete cycle.

4. Verifying that each valve (manual, power operated or automatic)in the direct flow path is in its correct position. The AFW flow control valves may be verified by observing a 160 gpm setpoint on the flow indicator controller in Control Room.

b. Before entering MODE 3 after a COLD SHUTDOWN of at least 14 days by completing a flow test that verifies the flow path from the condensate storage tank to the steam generators.

c. At least once per 18 months by verifying that each automatic valve in the flow path actuates to its correct position and each auxiliary feedwater pump automatically starts and delivers a modulated flow of 160 gpm + 10 gpm to each flow leg upon receipt of each auxiliary feedwater actuation system (AFAS) test signal.

CALVERT CLIFFS-UNIT 2 3/4 7-Sa Amendment No.

ATTACHMENT 1 PLANT SYSTEMS BASES U = Maximum number of inoperable safety valves per operating steam line 106.5 = Power Level-High Trip Setpoint for two loop operation 46.8 = Power Level-High Trip Setpoint for single loop operation with two reactor coolant pumps operating in the same loop X = Total relieving capacity of all safety valves per steam line in Ibs/ hour Y = Maximum relieving capacity of any one safety valve in Ibs/ hour 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 300 F from normal operating conditions in the event of a total loss of offsite power. A capacity of 400 gpm is sufficient to ensure that adequate feedwater flow is available to rengve decay heat and reduce the Reactor Coolant System temperature to less than 300 F when the shutdown cooling system may be placed into operation.

Flow control valves, installed in each leg supplying the steam generators, maintain a nominal flow setpoint of 160 gpm plus or minus 10 gpm for operator setting band. The p

nominal flow setpoint of 160 gpm incorporates a total instrument loop error band of plus 47 gpm (217 gpm total flow per leg) and minus 60 gpm (90 gpm total flow per leg).

In the spectrum of events analyzed in which automatic initiati$n of auxiliary feedwater occurs the nominal setting of 160 gpm allows a minimum of 10 minutes before operator action is required. At 10 minutes after automatic initiation of flow the operator is assumed to be available to increase or decrease auxiliary feedwater flow to that required for existing plant conditions.

CALVERT CLIFFS UNIT-2 B 3/4 7-2 Amendment No.

5.0 Initial Startup Test Program for Auxiliary Feedwater Actuation Systern 5.1 Background Discussion Cycle 5 and subsequent cycles will incorporate long term Auxiliary Feedwater System upgrade requirements discussed in NUREG-0737 consisting of a new electric driven' train and associated controls. During the startup of Unit 1, Cycle 6 difficulty was experienced in maintaining adequate ster.m generator pressures during testing of the auxiliary feedwater automatic initiation logic. This was due to the lack of an adequate heat source while feeding the generators in MODE 3 following refueling. In order to perform an adequate test nuclear heat is required to maintain steam generator pressure constant during the the short period of time necessary to complete testing and adjustment of the feedwater system.

5.2 Technical Specification Exemption In order to use nuclear heat a one-time exception to Technical Specification 3.7.1.2 is required to allow entry into MODE 1 with a partially surveilled Auxiliary Feedwater System. Prior to entry into MODE 2 the following measures will be taken to insure operability of the Auxiliary Feedwater System: 1) system cleanliness will be verified; 2) system hydrostatic tests (within the scope requested in Reference 1) will be performed; 3) electric pump and control valve checks will be performed; and 4) actuation logic i

control checks will be made. Additionally, surveillance on the steam driven Auxiliary Feedwater Pumps (AFP) will be performed to prove the pumps are i available for manual steam generator level control prior to entry into MODE

2. The remainder of testing to be performed during MODE 1 will consist of steam and electric driven pumps automatic flow initiation performance and equipment response time testing at normal operating temperature and pressure.

I

Once MODE 1 is entered for testing any unrelated problem which results in an exit from MODE 1 will not preclude reentering MODE 1 to complete testing.

.5.3 Justification The Auxiliary Feedwater System is designed to provide feedwater to the steam generator for the removal of heat from the Reactor Coolant System and to cool the Reactor Coolant System to 3000 in the event of the loss of normal feedwater flow. The Auxiliary Feedwater Automatic Initiation Logic provides an adequate primary heat sink in the unlikely event that no operator action is taken for up to 10 minutes following a low steam generator water level transient.

The system will be hydrostatically tested prior to entering MODE 3 and all system flushes will be completed. Prior to entry into MODE 2 the Auxillary Feedwater System will be verified operational from the standpoint that the steam driven AFPs are available for manual feed addition. In addition, the ficw control valve will be stroke tested to ensure the operator has manual control capability. Upon entering MODE 2, experienced operations personnel will be available to manually operate the system using the steam driven AFPs to maintain adequate steam generator inventory. Should an event occur which requires the mitigating effect of auxiliary feedwater, decay heat will be minimal. The ambient heat losses at the end of refueling nearly equal the production of heat from the reactor plus the heat from the Reactor Coolant Pumps. Decay heat alone is not enough to overcome ambient heat losses.

Since the syatem will be functionally capable of maintaining steam generator inventory and operations personnel will be available to operate the system, the intent of Technical Specification 3.7.1.2 will be met.

1. A. E. Lundvall to R. A. Clark Itr. dated 8/30/82, " Inservice Inspection Program for relief from ASME Code Section II Requirements Determined to be Impracticable."

I s

t

s .

6.0 References For Supplement

1. R. E. Henry, H. K. Fauske, "The Two Phase Critical Flow of One-Component Mixtures in Nozzles, Orifices, and Short Tubes," Journal of Heat Transfer, Transactions of the ASME, May,1971.

2. CENPD-188-A, "HERNITE Space-Time Kinetics," July, 1976.

3 CENPD-161-P, TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core," July,1975

4. R. V. MacBeth, "An Appraisal of Forced Convection Burn-Out Data," Proc.

Instn. Mech. Engrs., 1965-66, Vol. 180, Pt. 3C, pp. 37-50.

5 D. M. Lee, "An Experimental Investigation of Fcrced Convection Burnout in High Pressure Water; Part IV, Large Diameter Tubes at About 1600 Psia,"

AEEW-R, November, 1966.

6. CENPD-206-P, " TORC Code, Verification and Simplified Modeling Methods,"

January, 1977

.

ML20066F973

Text

Navigation menu

Search