ML20217D269

From kanterella
Jump to navigation Jump to search
Proposed Tech Specs Re Change to Reactor Coolant Sys Flow Requirements to Allow Increased SG Tube Plugging
ML20217D269
Person / Time
Site: Calvert Cliffs  Constellation icon.png
Issue date: 04/21/1998
From:
BALTIMORE GAS & ELECTRIC CO.
To:
Shared Package
ML20217D232 List:
References
NUDOCS 9804240253
Download: ML20217D269 (26)
v • d • e


Text

.

ATTACHMENT (1) l h

UNIT 1 ]

MARKED-UP TECHNICAL SPECIFICATION PAGES l

2-3 2-5 3/42-8 3/4 7-4 l l B 3/4 7-3 I I i 9804240253 980421 PDR ADOCK 05000317 P pga Baltimore Gas and Electric Company License Amendment Request l April 21,1998 l

l

a

- $i3 r

- z*4M

- $Orymd2**k

=

4E r

mrQgO e

b4 Of 0

n eo i

NA w

gbM

_ 4 , 5 1

E 4 L

B A

fN 8 _

4 O

I 1 _

wT vA R CR E AE NP W UO 4 O P

4

_ 1 L -

A _

_ I E '

M R

LN E e BO AI 2 H

_ 53 l TT PA 1 T D

ER E -

&0 ( CE T _

CP A AO ,

0 R i

1 F o.$ { - O

- S N

- S O

- E I T

- L 8 C 6

- O OE ,

- E I TH 0 AR

- C L'

- N D A DT F N EF A U N - AR QE TO S E O -- WO S

_ I E 4R E MINV OL 6 T - P R1 a ,

I A - S TO L

LIT A 0 P AR AG O1 R

E

- A R HNN 2 B NAV P - L L OE NT SIA ouia A Arr I I L KH E TC O - OY R - CL NN IE S AATS R U R A , 4 i -

DN IE EYS T

I T

- AO A OMN XP T ALEF S G I

PB Y 0

/i- IN NdI OFR r- CO O iA C

nT RAI OD I

S N R*0 A '

EIT O8D ShsUi

_ e FARO RA I 5 N i

Aw- -

FR RE II T

M RU LDCO v DR OE I IS CNO AAC

, 2 0

N - OP AOA I EHE JOL VRFH U -- FO IFF RTS 0 0 0 0 0 0 4 2 0 E 8 5 W 5 5 3 8

4 g -@y hg gc 3o5m N,s "

%N oN nON 2 nDP g E m2yy 5a .

OGg7o= . cz H * + lj ,k e * ; ' haa$ {E 3c, M I.

o r.

i gM!U>Eghg y>

yy ,b=S i

. ~ i t

d n

_ n a d e

af D l

,o E o e

- T o f R%aA c f E0 R W3d r o ,

O nf o p Pfao t o c o L R% a t ate 0 e

_ MnW r w o

S_ RiO7 E EoP0 n l

_ U e Hp 1 . g e L l TtL R i b A b eA5E W s e a s V a esM a e S c v RfO d i s h T E i omEoP i I L l buH f p g s c M B p amTmL o i p n I A p i uA 0 s i W A % nD mM 0 p 5

_ L w4 8 0g

_ O 0iEiR T L t 1 mTxE o 2 4 6 1 n -

o AaH L l i N f 551 2r I A N 5aRmT O

P a T D .

_ E t S hEfR tToE n iA W a d P

I wRmO l o I- e e

Y R uP

,f m o f T Roil c E xA f N W%aM r o 1 O O0mR o

- I P3 E t p 2 T aH c o A Lf T a t 2 T Aod e

_ N M nD r w E E RtaE o L M En T n l B U e Hi ,A g e A R _

T T T l TORR i s

b -

S N b pE s

N I a etWf e a -

a e

_ I O P

T c veOo i osP b  %

d f

i s

p g s i h c

h(o E E l

p amL0 o i p n V S p uA 0 s i I 0 p 5 T P A %mM7 w 4 8 0g C 0iR0 E I t 1 nE1 2 4 6 1 n -

T R o iH l i O T N 5mT 5 >f 55t 1r _

R P _

O R N h h l T w g g - e C o i i v A L H H e e E r L R - - - u s r w e e s e p o r r e t i h l u u r a r g F s s P W T i s s H t er r o e r r o

r n o - a P P t t t l a a T c l o r t r r I a e o e n e e N e v C z e n n U R e i m e e L r r n G G L

A l

a r t o

c u i s a g m a

N u e s t a O n w a e

e n ew ew I a o R r o t o to P C Sl SL T M P C

N .

U . . . . . .

F 1 2 3 4 5 6 7

.l I4>c c #5 8 s ' i Z1, E l p*

T

!aE2E E. n c L

. ', l 3/4.2 _ POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown:

l

a. Cold Leg Temperature s 548"F l
b. Pressurize'r Pressure 2 2200 psia
  • 340,000 c.

/ -

l Reactor Coolant System Total Flow Rate 2 gpm l

d. AXIAL SHAPE INDEX, THERMAL POWER as specified in the COLR.

l APPLICABILITY: MODE 1.

ACTION: With any of the above parameters excee parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or. ding its limit, restore the reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS 4.2.5.1 Each of the parameters shall be verified to be within their limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. l-4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

i Limit not applicable during either a THERMAL POWER ramp increase in j

excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.

CALVERT CLIFFS - UNIT 1 3/4 2-8 Amendment No. 186 Me

3/4.7 PLANT SYSTEMS TABLE 4.7-1 STEAM LINE SAFETY VALVES PER LOOP VALVE LIFT SETTINGS

  • ALLOWABLE-ORIFICE SIZE
a. RV-3992/4000 935-995 psig R
b. RV-3993/4001 935-995 psig R  !
c. RV-3994/4002 935-1035 psig R
d. RV-3995/4003 935-1035 psig R
e. RV-3996/4004 M 935 g R
f. RV-3997/4005 935-R l g. RV-3998/4006 935 #

R

h. RV-3999/4007 935 R

l l

l l

l l

1050 peig - -

Lift settings for a given steam line are also acceptable if any 2 valves lift between 935 and 995 psig, any 2' other valves lift between and 1035'psig, and the 4 remaining valves lift between 935 and CALVERT CLIFFS - UNIT 1 3/4 7-4 taendment No. 186 I

\

L

e 3/4.7 PLANT SYSTEMS

. . BASES 3/4.7.1.4 Activity The limitations on Secondary System specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 %. primary to secondary tube

~

leak in the steam generator of the affecte steam line and concurrent loss of offsite electrical power. These values are consist nt with the assumptions used in the accident analyses.

1 IOO $a b s fe N'y l 3/4.7.1.5 Main Steam Line Isolation Valves The OPERABILITY of the main steam line isolation valves ensures that no I

more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the 1

event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. The main steam isolation valves are surveilled to close in less than 5.2 seconds to ensure that under reverse steam flow conditions, the valves will close in less than the 6.0 seconds assumed in the accident analysis.

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 80 F and 200 psig are based on steam generator secondary side limitations and are sufficient to prevent brittle fracture.

3/4.7.3 COMPONENT COOLING WATER SYSTEM The OPERABILITY of the Component Cooling Water System ensures that sufficient cooling capacity is available for continued operation of safety related equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

3/4.7.4 SERVICE WATER SYSTEM The OPERABILITY of the Service Water System ensures that sufficient cooling capacity is available for continued operation of equipment during normal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with the assumptions used in the accident analyses.

l CALVERT CLIFFS - UNIT 1 B 3/4 7-3 Amendment No. 186 l

ATTACHMENT (2)

UNIT 2 MARKED-UP TECHNICAL SPECIFICATION PAGES 2-1 2-3 2-3a 2-5 2-7 ,

I 3/42-8 3/4 7-4 B 3/4 7-3 1

Baltimore Gas and Electric Company License Amendment Request April 21,1998

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS REACTOR CORE 2.1.1 The combination of THERMAL POWER, pressurizer pressure, and highest operating loop cold le oolant temperature shall not exceed the limits shown in Figure 2.1-1 .

APPLICABILITY: MODES 1 and 2.

ACTION:

l a. Whenever the point defined by the combination of the highest l I

o)erating loop cold leg temperature and THERMAL POWER has exceeded t1e appropriate pressurizer pressure line, be in H0T STANDBY within j 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. '

b. The NRC Operations Center shall be notified by telephone as soon as possible and in all cases within one hour.
c. The Vice President-Nuclear Energy and the offsite review function  ;

shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

d. A Safety Limit Violation Report shall be prepared and submitted to  !

the Connission, the offsite review function and the Vice President - i Nuclear Energy within 14 days of the violation.

REACTOR COOLANT SYSTEM PRESSURE 2.1.2 The Reactor Coolant System pressure shall not exceed 2750 psia.

APPLICABILITY: MODES 1, 2, 3, 4 and 5.

ACTION:

MODES 1 and 2

a. Whenever the Reactor Coolant System pressure has exceeded 2750 psia, l be in HOT STANDBY with the Reactor Coolant System pressure within its -

limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

l b. The NRC Operations Center shall be notified by telephone as soon as l possible and in all cases within one hour.

t

c. The Vice President-Nuclear Energy and the offsite review function shall be notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
d. A Safety Limit Violation Report shall be prepared and submitted to the Connission, the offsite review function and the Vice President .

Nuclear Energy within 14 da s of the violation. i f flGure O-l~~IA SY 1 NUY CALVERT CLIFFS - UNIT 2 ___

2-1 _

Amendment No. 193 _

n e>' n

~. yEM D,O ry 2* m MgHQ ,rQg=* .

50

' @0 a3 4

1 a

M f

_ \

._ B A'

i N 6 O

I 1

wTA 4R B

_ E AE NP W 4 OF

, UO ,

$ 1 L

A

_ 4 M R R

%"4 T R O N E

$ AI 2

1 H

T

- rT

  • O4 , wR A D E
E T dP A Non ' AO ,

0 R 1 F

- . O

- S N

$d -

S E

O T-

- I L 8 C

- O OE ,

- E I TH 0 AR

- - C L' I N D A D F N E 'F N -

O --

AR WO A U QE IOSE I'

I T - E S

DR IO S

E R-1 MNV I OL , 6 A - L P LIA T 0 R - P A AR R A G O12 NAV E -

P -

L L OE HNND SIA OU I I n O -- OY NT L KH E TC R - CL NN A AT R U A m D N IE A S IE RAA 4 SSI G EYS f , 0 n -- AO TMN O XP ALEF PBY A

r - L NdI OFR RAL N C.NO O uT I 0A _

C tA ODS F A RO u

EIT sU R'8 O 5D _

c RA ShI DROE I N _

c-A PR RE T IC I

I LDCO IS CNO AA C

, 2 0

N -

OP M IOL RU AOA I EHE U -- FO LFF VRFD RTS 0 0 0 0 Q

~ 8 5 @ 6 4 2 6

0 6 a ,

g- g$ g8 m $D% &. 3 Q b Q*hb xb y$mm N,.a. e

!g m#n$ N n m 42 - $>r g m e z m> ? y _, 5i-2Gm? p~m gaW E8gg" = Gw l' 5HII N ,'

iIl 1 ,

Ill r

lllIl

ao m43 G " NO rE dE* >9 mN4E - " g"m 0

, , 3

%, S L

E  %

- M N

, 6 MO I L

ET CA CR S

E AE NP W

_ UO , 4 O

. L PL A

M R

R EN E h B AO

, 2 I 5 TD L D PA D ER CE E

T CP A AO ,

O R h-- L F

.O

- S N

- S O E

I T

I

- O OE

, E O AC

- E I T H R

- - C L N D A DT F F

AR N U E IOS N - WO A Q E O -

E DR S B rNV 4

, 6 D-A - S UO E

P R.1 ROL IA 0 R - P A ARR A G O12 ITV NA E -

P - L OE H NNB SIA OUY I IT O --

I OY NT KIE TCE E - CL NN IE S LADR A U A

R AF A , 4 L DN IES XP SG EYS

- A 0 B - AO N TMO NUI N

ALRF I

PBY CO O RT RAIN OFR M -

C SUTA I

ODS O R'0 A E

C -

EI RA T

SN P AR O8D 8 N

- PR I I r

i DROE IS I

CNO , 2 a

A - RE MRw ILDCO AOA f AAO EHB W- OP FO LFM I O VRFU RTB

- - ~ - - - _

0 04 0 0 0

=a 8

5 6 2

5 0

5 8

4 .Q g *-

3gRsh 9sy b s n

D%bn% L a

3EmM N pb

,m o8 S=m -4%E>F g=0z m> mM _,E::

EAE nEm* E$ N mL 9 g g g.g# o. w 6 in>Gm Ct l ,:

m sE4 gEd >E r - g m yy_0=8 u

t d n d n a e _

af D l e

o E o f T o R%aA c f E0 R o W3d r O nf o p Pfao t c o o a t -

L R% e ate 0 w MnW r o S RiO7 E EoP0 n g

l e

U e Hp 1 .

b L l TtL R i A

V b

a esM eA$E W s e a s c v RfO d i a e S E i omEoP s i h T L l buH f p g s c ,

I B p amTmL o ip n M A p i uA 0 s i I 0 p S L W A %nDmM w 4 8 0g O

T N

I L

L A

t N o 0i E 1mTxE 5aRmT AaHi R hl tf 553 o 2 4 6 .

1n 2r i

O P

T a E D .

S hEfR t tToE n P iA W a 4 d e

I wRmO uP l

o 4 e R ,f m o - f T Roil c N E xA r = f o

1 O W%aM o

- I O0mR p 2 T P3 E t o

A aH c t 2 T Lf T a N Aod e w E E M nD r o L M RtaE B U En ,T n g

l e

A R e Hi A b T T T l TORR pE i

s S N b a s N I a etWf e a e I O c veOo d i P i osP s i h _

T l b  % f p g s c _

E E p amL0 o i p n V _

S p uA 0 s i I p 5 T A %mM7 w 4 0

8 0g C P 0iR0 1 n E I t 1 nE1 o 2 4 6 i R o iH l _

T O T N 5mT 5 2f 5 5t tr .

R P

R O

N h h l .

T C w g g - e o i i v A

E 'L H H e e r L R - - - u s r w e e s e p o r r e t i h l u s s P u r W a

r g F T i s s H t e e r r r n r r o o o - a P P t t t l a ar T c l o r t r e

I a e o e n e n N e v C z e n U R e r

i r n G m e G e

L L

A l

a r o

t u i s a g m a

N u e c s t a O n w a e n ew ew I a o e r o t o to T M P R P C SL SL '.g C

N . . . .

U . . .

F 1 2 3 4 5 6 7 g m " bm? ' EG m m*  ?.

( .

O F

E. _

i

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

, . TABLE 2.2-1 (Continued)

. TABLE _ NOTATION See Specification 3.2.5, ONB Parameters," for the design reactor coolant flow, f(1) Trip may be bypassed below 10-'% OF RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is t 10 '% of RATED THERMAL POWER.

(2) 1 rip may be mariaally bypassed below 785 psia; bypass shall be automatically removed at or above 785 psia.

(3) Trip may be bypassed below 15% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 115% of RATED THERMAL POWER.

(4) Trip may be bypassed below 10-'% and above 12% of RATED THERMAL POWER.

& j)]g. [gpf,hr hC)41t YW ~ BOW allowable valut M k 5 95$ 'l A"f'T Coolantflon) Nro")h U U k dyA N-l I

CALVERT CLIFFS - UNIT 2 2-7_

Amendment No. 193 __

l

e 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB related parameters shall be maintained within the limits shown: l

a. Cold Leg Temperature s 548 F l 3+0,000
b. Pressurizer Pressure 2 2200 psia. l
c. Reactor Coolant System Total Flow Rate 2 gpm l
d. AXIAL SHAPE INDEX, THERMAL POWER as specified in the COLR l j i

APPLICABILITY: MODE 1. '

ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

l SURVEILLANCE REQUIREMENTS l

4.2.5.1 Each of the parameters shall be verified to be within their limits l l at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

4.2.5.2 The Reactor Coolant System total flow rate shall be detennined to be within its limit by measurement at least once per 18 months.

l l

A M The kador (csjad S shin 1 70AY El** '

sh d) be a. 370,000 f?") & 0 4 ; y ,' y ; 4 c je p .

Limit not applicable during either THERMAL POWER ramp increase in excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step increase of greater than 10% of RATED THERMAL POWER.

CALVERT CLIFFS - UNIT 2 3/4 2-8 Amendment No. 163

I .

3/4.7 PLANT SYSTEMS TABLE 4.7-1 STEAM LINE SAFETY VALVES PER LOOP J VALVE NUMBER LIFT SETTINGS

  • ALLOWABLE ORIFICE SIZE l a. RV-3992/4000 935-995 psig R  !

1

b. RV-3993/4001 935-995 psig R l c. RV-3994/4002 935-1035 psig R
d. RV-3995/4003 935-1035 psig R
e. RV-3996/4004 935 R j f. RV-3997/4005 935- h D fO $ d N R
g. RV-3998/4006 935- R
h. RV-3999/4007 935-hpsig R i

l l

$$ f}g daghwwt b8l W VAlM$ 4)14ll Alltu%U bt fpbf (6i]kl69hCfr&

$ff"1l' ll)M'b kt}b 2,[y& d $-

l 10 50pQ (%" "'""'* *'"'"'"' Y ^ 4 l

l Lift settings for a given steam line are also acceptable if any 2 valves lift between 935 and 995 psig, any 2 other valves lift between nd 1035 psig, and the 4 remaining valves lift between .935.and psig.

CALVERT CLIFFS - UNIT 2 3/4 7-4 Amendment No. 163 l'

3/4.7 PLANT SYSTEMS BASES 3/4.7.1.3 Condensate Storace Tank The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at NOT STAND 5Y conditions for 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> with steam discharge to atmosphere with concurrent and total loss of offsite power. The contained water volume limit includes an allowance for water not usable because of tank discharge line location or other physical characteristics.

3/4.7.1.4 Activity The limitations on Secondary System specific activity ensure that the resultant off-site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam _line rupture. This dose also includes the effects of a coincidentWF.0W primary to secondary tube leak in the steam generator of the affected steam line and a concurrent i loss of offsite electrical power. These values re consistent with the

! assumptions used in the accident analyses.

3/4.7.1.5 Main Steam Line Isolation Valves /00 don l0tN5 /F /

The OPERABILITY of the main steam line isolation valves ensures that no l more than one steam generator will blowdown in the event of a steam line rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and 2) limit the pressure rise within containment in the event the steam line rupture occurs within containment. The OPERABILITY of the main steam isolation valves within the closure times of the surveillance requirements are consistent with the assumptions used in the accident analyses. The main steam isolation valves are surveilled to close in less than 5.2 seconds to ensure that under reverse steam flow conditions, the valves will close in less than the 6.0 seconds assumed in the accident analysis.

l 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION The limitation on steam generator pressure and temperature ensures that the

, pressure induced stresses in the steam generators do not exceed the maximum l allowable fracture toughness stress limits. The limitations of 90 F and 200 psig are based on steam generator secondary side limitations and are sufficient to prevent brittle fracture.

CALVERT CLIFFS - UNIT 2 B 3/4 7-3 Amendment No. 163 l

ATTACHMENT (3)

MARKED-UP IMPROVED TECHNICAL SPECIFICATION PAGES 2.0-1 2.0-2 2.0-2a 1

i 3.3.1-9

3.3.1-11 3.4.1-1 3.4.1-2 1

3.7.1-4 B 3.7.1-3 B 3.7.14-1 Baltimore Gas and Electric Company License Amendment Request April 21,1998

, v SLs ,

2.0 l l

2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs l

, 2.1.1.1 In MODES 1 and 2, the combination of THERMAL POWER, pressurizer pressure, and the highest operating loop cold leg coolant temperature shall not exceed the t limits shown in Figure 2.1.1-1.

l 2.1.1.2 In MODES 1 and 2, the peak linear heat rate (LHR) shall i

bes21.0kw/ft.

l 2.1.2 Reactor Coolant System (RCS) Pressure SL In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained j s 2750 psia.

l 2.2 SL Violations ~ ~

2.2.1 If SL 2.1.1 is violat'ed, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

2.2.2 If SL 2.1.2 is violated:

2.2.2.1 In MODE 1 or 2, restore compliance and be in MODE 3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

L L 2.2.2.2 In MODE 3, 4, or 5, restore compliance within 5 minutes.

.- - _ _ _ - - -. - go 72 - - - - - - - -- - - - -

Fo~ r llisll'1 onlp Rg ute 2.1.I~la sball 2PPly $b'0"1 h Cyele 12.

. A A e A ~

' CALVERT CLIFFS - UNITS 1 & 2 2.0-1 Revision 0

.- 1

\

l

. I SLs 2.0

'l 1

Oh l ." .A i 60r i i L<

l t

/

i i i ll I i i UNACCEPTABLE OPERATION UNACCEPTABLE SBC- OPERATION -

FOR PRE-CLAD COLLAPSE OPERATION ONLY i

E S LIMITS CONTAIN NO ALLOWANCE '\ 4*%4%* .% j g 54C-FOR INSTRUMENT ERROR OR FLUCTUATIONS

  • g *N4*g 4*g n

VALID FOR AXIAL SHAPES AND -

ROD RADIAL PEAKING g 52C- FACTORS WITHIN LIMITS _

d SOC- REACTOR OPERATION LIMITED TO LESS _

THAN 580*F BY ACTUATION OF THE k SECONDARY SAFETY VALVES 2

ACCEPTABLE j 48C.. OPERATION 46r e i i i i i i i ik k

\ >

O 0.2 0.4 0.0 0.8 1.0 1.2 1.4 1.6 .1.8 'j2.0 FRACTION OF RATED THERMAL POWER 5 g b M l o> o>

4 D N o J$

l e

Figure 2.1.1-1 Reactor Core Thermal Margin Safety Limit l

{

CALVERT CLIFFS - UNITS 1 & 2 2.0-2 Revision 0

o SLs 2.0 60r I i i i i I i i UNACCEPTABLE OPERATION 8 - -

UNACCEPTABLE

} SBC ----

OPERATION _

l FOR PRE-CLAD COLLAPSE OPERATION ONLY

, tL SGC. ~

l t g LIMITS CONTAIN NO ALLOWANCE  % %%

4 4 4

\4

$ FOR INSTRUMENT ERROR OR *g oY*g *g i

l } 54C- FLUCTUATIONS _

g I VALID FOR AXtAL GHAPES AND l

ROD RADIAL PEAKING

{ 52C-l FACTORS WITHIN LIMITS _

b o

50C_ REACTOR OPERATION LIMITED TO LESS _

THAN 580*F BY ACTUATION OF THE SECONDARY SAFETY VALVES ACCEPTABLE 4BC. OPERATION

\ > > > >

I 46r e i i t t i t *

'oo oA c, i

l O O.2 0.4 0.6 0.0 1.0 1.2 1.4 1.6 1,8 2.0 l

FRACTION OF RATED THERMAL POWER

- - ko rE - ~~~ ~

Eis Sgure only app lies bo Unda f rough Cy cle Lt. 1

, , A- O Figure 2.1.1-lh Reactor Core Thermal Margin Safety Limit A

h0D CALVERT CLIFFS - UNITS 1 & 2 2.0-2h Revision 0

~

c RPS Instrumentation-Operating 3.3.1 Table 3.3.1-1 (page 1 of 3)

Reactor Protective System Instrumentation SURVEILLANCE FUNCTION MODES REQUIREMENTS ALLOWABLE VALUE

1. Power Level-High 1, 2 SR 3.3.1.2 s 10% RTP above SR 3.3.1.3 current THERMAL POWER SR 3.3.1.4 but not < 30% RTP nor '

SR 3.3.1.5 > 107% RTP SR 3.3.1.8 SR 3.3.1.9

2. Rate of Change of 1, 2 SR 3.3.1.1") s 2.6 dpm Power-Higb(*) SR 3.3.1.6 SR 3.3.1.7 gg .g SR 3.3.1.8 Ap,
3. Reactor Coolant 1, 2 SR 3.3.1.1 2 of Design Flow I Flow-Low ) SR 3.3.1.4 SR 3.3.1.7 SR 3.3.1.8 l SR 3.3.1.9
4. Pressurizer 1, 2 SR 3.3.1.1 s 2400 psia Pressure-High SR 3.3.1.4 SR 3.3.1.8 SR 3.3.1.9 1
5. Containment 1, 2 SR 3.3.1.1 s 4.0 psig Pressure-High SR 3.3.1.4 SR 3.3.1.8 SR 3.3.1.9
6. SteamGener'atbr- 1, 2 SR 3.3.1.1 2 685 psia Pressure-Low (') SR 3.3.1.4 SR 3.3.1.7 SR 3.3.1.8 SR 3.3.1.9 CALVERT CLIFFS - UNITS 1 & 2 3.3.1-9 Revision 0 e

o ,

i .

RPS Instrumentation-Operating i 3.3.1 Table 3.3.1-1 (page 3 of 3)

Reactor Protective System Instrumentation I')

Bistable trip unit may be bypassed when THERMAL POWER is < IE-4% RTP or

> 12% RTP. Bypass shall be automatically removed when THERMAL POWER is h IE-4% RTP and s 12% RTP. ,

i N

Bistable trip unit may be bypassed when THERMAL POWER is < IE-4%.

Bypass shall be automatically removed when THERMAL POWER is 2 1E-4% RTP. During testing pursuant to LC0 3.4.16, trips may be bypassed below 5% RTP.

I'I Bistable trip unit may be bypassed when steam generator pressure is <

785 psig. Bypass shall be automatically removed when steam generator pressure is 2 785 psig.

M Bistable trip unit may be bypassed when THERMAL POWER is < 15% RTP.

Bypass shall be automatically removed when THERMAL POWER is 215% RTP.

M Trip is only applicable in MODE 1215% RTP.

N CH.ANNEL CHECK only applies to Wide Range Logarithmic Neutron Flux Monitor.

(g) JXe %k Coalad Row -Low a&4/e value c),all Le h 95 0/, &r

^

lAnifof onl }brou g.

4-CALVERT CLIFFS - UNITS 1 & 2 3.3.1-11 Revision 0 l

n.

e l

L

. ", \

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3,4.1 RCS Pressure. Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LC0 3.4.1 RCS DNB parameters for press'urizer pressure, cold leg temperature, and RCS total flow rate shall be within the l limits specified below:

1

! a. Pressurizer pressure 2 2200 psia;

b. RCS cold leg temperature (T c) <; 548*F; and A^ 40,0 0
c. RCS total flow rate 2 , O gpm.

APPLICABILITY: MODE 1.

.....______.---------------NOTE---------------_-_----_---_.

Pressurizer pressure limit does not apply during:

, a. THERMAL POWER ramp > 5% RTP per minute; or *

b. TilERMAL POWER step > 10% RTP.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l A. Pressurizer pressure A.1 Restoreparameter(s) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or RCS flow rate not to within limit.

within limits.

~

___'__1_

76 ??CS Mal flow nk h.,,;f shIl be & 370,000 foc Un;/d only , drouf C p le 1.1.

CALVERT CLIFFS - UNITS 1 & 2 3.4.1-1 Revision 0

V RCS Pressure. Temperature, and Flow DNB Limits 3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ~

associated Completion Time of Condition A not met.

C. RCS cold leg C.1 Restore cold leg 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> tt -arature not temperature to within within limits. limits.

D. Required Action and D.1 Reduce THERMAL POWER 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion to s 30% RTP.

Time of Condition C not met.

SURVEILLANCE REQUIREMENTS  !

i SURVEILLANCE FREQUENCY l

SR 3.4.1.1 Verify pressurizer pressure 2 2200 psia. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.1.2 Verify RCS cold leg temperature s 548'F. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.1.3 -------------------NOTE-------------------

h0nly required to be met in MODE 1.

---------------------------------- M0,ooo Verify RCS total flow rate 2 gpm. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.4.1.4 Verify measured RCS total flow rate is 24 months within limits.

2. Foe Un;} ) sn , fAe ??CS hhl flm rde sball be W370,000 lbough Cyle D.

CALVERT CLIFFS - UNITS 1 & 2 3.4.1-2 Revision 0

V f

MSSVs 3.7.1 Table 3.7.1-2 Main Steam Safety Valve Lift Settings VALVE NUMBER ~

l LIFT SETTING"'

_ Steam Generator #1 Steam Generator #2 (psig)

RV-3992 RV-4000 935-995 .

RV-3993 RV-4001 935-995 l RV-3994 RV-4002 935-1035 RV-3995 RV-4003 935-1035 1 RV-3996 RV-4004 935- 065 l RV-3997 RV-4005 935 5*-fogo i RV-3998 RV-4006 I 935 RV-3999 RV-4007 935 5 l

(1) Lift settings for a given steam line are also acceptable if any two valves lift between 935 and 995 psig, any two other valves lift between 935 and 1035 psig, and the four remaining valves lift between 935 and 1065 psig.

~ v- N v

- - NOTE- - - - -

For

} un;f of only, tAe maximum alloazble liH seffing for ibe four highest se t valves for A given steam line shall be 106Sysig droug h Cycle 1,8.

--~

--- A  % ~ s i Ava J 1

i CALVERT CLIFFS - UNITS 1 & 2 3.7.1-4 Revision K t

v

. i 4

+ MSSVs B 3.7.1 i

. . BASES lift between 935 andhpsig. Thus, the MSSVs still n l perform that design basis function properly.

l l

This LC0 provides assurance that the MSSVs will perform their designed safety function to mitigate the consequences of accidents that could result in a challenge to the RCPB.

APPLICABILITY In MODES 1, 2, and 3, a minimum of five MSSVs per steam l

generator are required to be OPERABLE, according to Table 3.7.1-1 in the accompanying LCO, which is limiting and bounds all lower MODES.

1 In MODES 4 and 5, trere are no credible transients requiring the MSSVs.

The steam generators are not normally used for heat removal in MODES 5 and 6, and thus cannot be overpressurized; there is no requirement for the MSSVs to be OPERABLE in these MODES.

l ACTIONS The ACTIONS table is modified by a Note indicating that separate Condition entry is allowed for each MSSV.

f l

A.1 and A.2 An alternative to restoring the inoperable MSSV(s) to OPERABLE status is to reduce power so that the available MSSV relieving capacity meets Code requirements for the l l power level. The number of inoperable MSSVs will determine  !

( the necessary level of reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip f)fe py yypriaintiny VAAVt3 AIff N"'"Y WAY ?U

.fw $lt,1 ortf , NN9A RA A*

ll CALVERT CLIFFS - UNITS 1 & 2 8 3.7.1-3 Revisionp' i

.. v

, l 9

Secondary Specific Activity B 3.7.14 B 3.7 PLANT SYSTEMS B 3.7.14 Secondary Specific Activity BASES i

BACKGROUND Activity in the secondary coolant results from steam ]

generator tube outleakage from the Reactor Coolant System I (RCS). Under steady state conditions, the activity is primarily iodines with relatively short half lives, and thus is indication of current conditions. During transients, 1-131 spikes have been observed as well as increased releases of some noble gases. Other fission product isotopes, as well as activated corrosion products in lesser l

amounts, may also be found in the secondary coolant.

1 l

A limit on secondary coolant specific activity during power operation minimizes releas.es to the environment because of normal operation, anticipated operational occurrences, and I accidents.

l This limit is lo an the activ'ity value that might be expected from a - tube leak (LC0 3.4.13. "RCS Operational LEAK of primary coolant at the limit of

, 1.0 Ci/gm (LC0 3.4.15, "RCS Specific Activity"). The main l steam line break (MSLB) is assumed to result in the release

of the noble gas and iodine activity contained in the steam  ;

l generator inventory, the feedwater, and reactor coolant i LEAKAGE. Most of the iodine isotopes have short half lives (i .e. , < 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />) .

I l APPLICABLE The accident analysis of the MSLB, as discussed in the l SAFETY ANALYSES Updated Final Safety Analysis Report (UFSAR), Chapter 14

! (Ref. 2), assumes the initial secondary coolant specific l

activity to have a radioactive isotope concentration of 0.10 Ci/gmDOSEEQUIVALENTI-131. This assumption is used in the analysis for determining the radiological consequences of the postulated accident. The accident analysis, based on this and other assumptions, shows that

=

CALVERT CLIFFS - UNITS 1 & 2 B 3.7.14-1 Revision 0 l 1

. I

Retrieved from "https://kanterella.com/w/index.php?title=ML20217D269&oldid=3203194"