Text

Proposed Tech Specs Re Requirements to Allow Increased SG Tube Plugging
	ML20134D304
Person / Time
Site:	Calvert Cliffs
Issue date:	01/31/1997
From:	; BALTIMORE GAS & ELECTRIC CO.
To:	;
Shared Package
ML20134D290	List: ML20134D287; ML20134D304;
References
	NUDOCS 9702050075
	Download: ML20134D304 (25)
	v • d • e

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ATTACHMENT Q) 4 i

UNIT I MARKED-UP TECHNICAL SPECIFICATION i

PAGES 2-1 2-3 2-3a 2-5 2-7 3/4 2-8 3/4 7-4 B 3/4 7-3 j[820$00ck obob j7'.

P Baltimore Gas and Electric Company License Amendment Request January 31,1997

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 1 coolant temperature shall not exceed the limits ~

shown in Figure 2.1-1 ~

Aoo .nsterisk APPLICABILITY: MODES I and 2.

ACTION:

l

a. Whenever the point defined by the combination of the highest I operating loop cold leg temperature and THERMAL POWER has exceeded the appropriate pressurizer pressure line, be in H0T STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l

d. A Safety Limit Viola +. ion Report shall be prepared and submitted !

to the Comission, the offsite review function and the Vice President - 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 Whenever the Reactor Coolant System pressure has exceeded l a.

2750 psia, be in H0T STANDBY with the Reactor Coolant System pressure within its limit within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

d. A Safety Limit Violation Report shall be prepared and submitted to the Comission, the offsite review function and the Vice President - Nuclear Energy within 14 days of the violation.

2- - ---

Amendment No._216_

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TABLE 2.2-1 [*

n REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS E TRIP SETPOINT ALLOWA8LE VALUES h bn FUNCTIONAL UNIT Manual Reactor Trip Not Applicable Not Applicable g a 1. E

< 10% above THERMAL POWER, with a < 10% above THERMAL POWER, and

2. Power Level - High minimum setpoint of 30% of RATED a minimum setpoint of 30% of M E THERMAL POWER, and a maximum of RATED THERMAL POWER and a g Z < 107.0% of RATED THERMAL POWER. maximum of 5 107.0% of RATED o

J. THERMAL POWER. g W ign reactor coolant of design reactor coolant

3. Reactor Coolant Flow - Low Pressurizer Pressure - High 5 2400 psia (2% 5 2400 psia y,

4. ;;i Containment Pressure - High 5 4 psig 5 4 psig m 5.

1 685 psia 3 685 psia

6. SteagGeneratorPressure- g Low 210 inches below top of feed 2 10 inches below top of feed y

7. Steam Generator Water Level - ring ,

Low ring m i

m

! j 2

p .-

a n

e,

i 4

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS TABLE 2.2-1(Continued)

TABLE NOTATION See Specification 3.2.5, "DNB Parameters," for the designJeactor coolant flow. -

PfII) Trip may be bypassed below 10"% OF RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is 110 "% of RATED THERMAL POWER.

(2) Trip may be manually bypassed below 785 psia; bypass shall be automatically removed at or above 785 psia.

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

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

- %- v

Tre %he C.JaJ Row -1on by sdpad anot allowasle value shall be h 95'A of desip reacfor coolad flow flrou)

Unit 1, Cycle 13 / ^

I CALVERT CLIFFS -UNIT 2 -- -- - - - Amendment- No.- 216- - !

3/4.2 _P_0WER 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:

I

a. Cold Leg Temperature s; 548 F l

b. Pressurizer Pressure 2 2200 psia

3fD,000

/ l ;

c. 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 exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS , 4.2.5.1 Each of the parameters shall be verified to be within their limits I

at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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.

- 7Ce Lek C . lad S ykm Al /% Rde lim;f eAall be 370,000 y fdrougA lhil 1 C cle y 13.

Limit not applicable during either a 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 1 3/4 2-8 Amendment No. 186

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

ALLOWABLE ORIFICE SIZE

a. I RV-3992/4000 935-995 psig R i

b. RV-3993/4001 935-995 psig R

c. RV-3994/4002 935-1035 psig R l l

d. RV-3995/4003 935-1035 psig R

e. RV-3996/4004 935 R

f. RV-3997/4005 935- R

~ ~ ~

g. RV-3998/4006 935 R l

h. RV-3999/4007 935 R

7Ee maxianum aHowable /#f seh%g b lAe highest set valves shall be 1065 psig firoupA thif 1, C cle y 13.

A w y -

1050 psig (befween 935 and 106S sig y $brougb Un;f 1, Cycle 13).

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

1 and 1035 psig, and the 4 remaining valves lift between 935 and 1

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

^

b i

l- 3/4.7 PLANT SYSTEMS

\

j BASES 3/4.7.1.4 Activity l _

k_ dag._

l f

The limitations on Secondary System specific activity ensure that the resultant off-site radiation dose will be limited to

!. 10 CFR Part 100 limits in the event of a s".eam line rupture. T11s dose i i

also includes the effects of a coincidentC .0 l 'O primary to secondary tube l

! leak in the steam generator of the affectecEsteam ine and concurrent loss !

! of offsite electrical power. These values are con istent with the assumptions used in the accident analyses.

g 3/4.7.1.5 Main Steam Line Isolation Valves The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event'of a steam line l

rupture. This restriction is required to 1) minimize the positive i reactivity effects of the Reactor Coolant System cooldown associated with I 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.

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. 1 l

3/4.7.3 COMPONENT COOLING WATER SYSTEM li 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 nomal and accident conditions. The redundant cooling capacity of this system, assuming a single failure, is consistent with' the assumptions used in the i accident analyses.

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

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ATTACHMENT (4)

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UNIT 2 MARKED-UP TECHNICAL SPECIFICATION i

PAGES i

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3/47-4

B 3/4 7-3 i

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Baltimore Gas and Electric Company

License Amendment Request January 31,1997 1

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS .

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d,'EllALVEikDTIL DEIT100 MnMDtVM FIGURE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT CALVERT CLIFFS - UNIT 2 2-3 Amendment No. 193 l

,m g TABLE 2.2-1 G

g REACTOR PROTECTIVE INSTRUMENTATION TRIP SETPOINT LIMITS y ,

TRIP SETPOINT ALLOWABLE VALUES b FUNCTIONAL UNIT Not Applicable Not Applicable g h 1. Manual Reactor Trip i5

' Power Level - High < 10% above THERMAL POWER, with a < 10% above THERMAL POWER, and 2.

iiiinimum setpoint of 30% of RATED a minimum setpoint of 30% of d ,

E THERMAL POWER, and a maximum of RATED THERMAL POWER and a g iZ o N $ 107.0% of RATED THERMAL POWER. maximum of 5 107.0% of RATED -

THERMAL POWER.

3. Reactor Coolant Flow - Low W g of design reactor coolant thof design reactor coolant flow flow g Pressurizer Pressure - High 5 2400 psia 5 2400 psia

4. '

5. Containment Pressure - High 5 4 ps? 5 4 psig :

m

>.685 psia h -

6. > 685 .ia ,

SteagGeneratorPressure-Low g

Steam Ger.erator Water Level - 3 10 inches below top of feed 210 inches below top of feed y

7. ring , ;

Low ring b

=

S iv la a

n . .

w r

3/4.2 POWER DISTRIBUTION LIMITS

3/4.2.5 DNB PARAMETERS i

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.

300,000 Pressurizer Pressure 2 2200 psia, j

c. Reactor Coolant System Total Flow Rate 2 370,000 gpm

, l ,

d. AXIAL SHAPE INDEX, THERMAL POWER as specified in the COLR l 4

APPLICABILITY: MODE 1.

l i ACTION: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to

. less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

1

\

SURVEILLANCE REQUIREMENTS 1

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

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 4

l 1

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

3/4.7 PLANT SYSTEMS TABLE 4.7-1 STEAM LINE SAFETY VALVES PER LOOP VALVE NUMBER 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 935 psig R j RV-3997/4005 f.

935-h R

g. RV-3998/4006 935-h R l h. RV-3999/4007 935-hpsig R

.1050 Lift settings for a given steam line are also acceptable if any 2 v

k alves lift 035 between psig, 935 and4995 and the psig, any remaining valves 2 other valves lift lift between 935 between and CALVERT CLIFFS - UNIT 2 3/4 7-4 Amendment No. 163 l~

i

{ 3/4.7 PLANT SYSTEMS j BASES i

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 j- NOT STANDBY conditions for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months 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 j ;

The limitations on Secondary System specific activity enturn that the resultant off-site radiation dose will be limited to; nr -

'st;;; :',

j. 10 CFR also Part 100 includes the limits effects in of theaevent of a s"a coincidenti.gn_,]Jperupture. This dosetube

. . m primary to secondary i

leak in the steam generator of the affected steamMine and a concurrent i

! loss of offsite electrical power. These values ar e consistent with the J j assumptions used in the accident analyses. ~ -

l 3/4.7.1.5 Main Steam Line Isolation Valves 100 yks y olay l l

' The OPERABILITY of the main steam line isolation valves ensures that no more than one steam generator will blowdown in the event of a steam line 4

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 !

i event the steam line rupture occurs within containment. The PPERABILITY of i

the main steam isolation valves within the closure times of tue surveillance requirements are consistent with the assumptions used in the j accident analyses. The main steam isolation valves are surveilled to close in less than 5.2 seconds to ensure that under reverse steam flow i conditions, the valves will close in less than the 6.0 seconds assumed in !

t the accident analysis,

, q i

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 90 F and .

l 200 psig are based on steam generator secondary side limitations and are '

!. sufficient to prevent brittle fracture.

l 4

i j

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

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ATTACHMENT (5)

MARKED-UP IMPROVED TECHNICAL SPECIFICATION PAGES i

l 2.0-1 i

2.0-2 2.0-2a 3.3.1-9 3 3.3.1-11 3.4.1-1 3.4.1-2 B 3.7.14-1 ,

B 3.7.14-2 Baltimore Gas and Electric Company License Amendment Request January 31,1997

SLs

2.0 l 2.0 SAFETY LIMITS (SLs) i 2.1 SLs

'2.1.1 Reactor Core SLs 2.1.1.1 In MODES 1 and 2, the combination of THERMAL POWER, ,

pressurizer pressure, and the highest operating loop i cold leg coolant temperature shall not exceed the limits shown in Figure 2.1.1-1. I 2.1.1.2 In MODES 1 and 2, the peak linear heat rate (LHR) shall bes21.0kw/ft.

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

2.2 SL Violations 4

~

2.2.1 If SL 2.1.1 is violated, restore compliance and be in MODE 3 ;

within I hour.

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 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . i l

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

- - - - - - - - - - NO TE - - - - - - - - - - - -

( For Unif f only, Rpre 2.1.1-1a sAall apply $broagb Cy ele 13.

. - ~ - A CALVERT CLIFFS - UNITS 1 & 2 2.0-1 Revision 0

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SLs 2.0 l

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p .%S .A i l- 4 Il g N'-

a' l l/

, , , i

,k k t , , , i UNACCEPTABLE OPERATION UNACCEPTABLE 58C- OPERATION -

FOR PRE-CLAD COLLAPSE

OPERATION ONLY LIMITS CONTAIN NO ALLOWANCE

'% *g%, *g I

% FOR INSTRUMENT ERROR OR *g 9 g 54C- FLUCTUATIONS _

l VAltD FOR AXtAL SHAPES AND i ROD RADIAL PEAKING 52C- FACTORS WITHIN LIMITS -

50C_ REACTOR OPERATION LIMITED TO LESS E THAN 580*F BY ACTUATION OF THE SECONDARY SAFETY VALVES ACCEPTABLE 4,g, OPERATION 46r e i i i i i i i i t.

O 0.? O.4 0.0 0.8 1.0 1.2 1.4 1.6 1.8

}2.0 FRACTION OF RATED THERMAL POWER 5SS M

% 0 *4

=:. R %

Figure 2.1.1-1 Reactor Core Thermal Margin Safety Limit CALVERT CLIFFS - UNITS 1 & 2 2.0-2 Revision 0

SLs 2.0 3

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i i l i i i i i i UNACCEPTABLE OPERATION - -

UNACCEPTABLE 58C- OPERATION - l FOR PRE-CLAD COLLAPSE

=

OPERATION ONLY I

u. SSC .

E

% LIMITS CONTAIN NO ALLOWANCE %4 %%4 %4 ,

FOR INSTRUMENT ERROR OR *g *g *g *g 54C- FLUCTUATIONS l 3 i VALID FOR AXIAL SHAPES AND ROD RADIAL PEAKING

{52C- FACTORS W. THIN LIMITS _

REACTOR OPERATION LIMITED TO LESS 50C_ . THAN 580*F BY ACTUATION OF THE _

SECONDARY SAFETY VALVES ACCEPTABLE 48C. OPENION

**o g 4 $

46C ' ' ' ' i ' ' ' '

O O.2 0.4 0.6 0.8 1.0 1.2 1.4 1.6 1.8 2.0 FRACTION OF RATED THERMAL POWER

-_ - - go7g - .

His $ $ute only App lies bo UnN $ tom 9h Cy cle 13.

4 f, A

. Figure 2.1.1-lh Reactor Core Thermal Margin Safety Limit k

CD CALVERT CLIFFS - UNITS 1 5 2 Revision 0 2.0-2h 4

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

Reactor Protective System Instrumentation SURVEILLANCE l FUNCTION MODES REQUIREMENTS ALLOWABLE VALUE I l

1. Power Level-High 1, 2 SR 3.3.1.2 s 10% RTP above j 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 l SR 3.3.1.9

2. Rate of Change of 1, 2 SR 3.3.1.1(') s 2.6 dpm l Power-Higb(*) SR 3.3.1.6 SR 3.3.1.7 9p .g SR 3.3.1.8 Ap,

3. Reactor C,oolant 1, 2 SR 3.3.1.1 2 of Design Flow Flow-Low () SR 3.3.1.4 1 SR 3.3.1.7 SR 3.3.1.8 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

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. SteamGenerItbr. 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

h RPS Instrumentation-Operating 3.3.1 :

Table 3.3.1-1 (page 3 of 3)

Reactor Protective System Instrumentation i

Id Bistable trip unit may be bypassed when THERMAL POWER is < 1E 4% RTP or '

> 12% RTP. Bypass shall be automatically removed when THERMAL POWER is ;

2 IE-4% RTP and s 12% RTP.

Id Bistable trip unit may be bypassed when THERMAL POWER is < 1E-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 l bypassed below 5% RTP. I l

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.

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

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

Id Trip is only applicable in MODE 1215% RTP.

I') CHANNEL CHECK only applies to Wide Range Logarithmic Neutron Flux Monitor.

(g) 7h %k Cooluf Flow -Low a& aalle value s),all Le h 95 */o kr j Ur,;fion!,lbroa ;

I I

sq l

l

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

=_ - . _ .- _ - _ - . - - _ _ _ - - - _ - - - - .

i RCS Pressure Temperature, and Flow DNB Limits !

3.4.1 >

3.4 REACTORCOOLANTSYSTEM(RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling i (DNB) Limits LC0 3.4.1 RCS DNB parameters for pressurizer pressure, cold leg temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure 2 2200 psia; ;

i

b. RCS cold leg temperature (Tc ) s 548'F; and

c. RCS total flow rate 2 .

& MO,000 gpm.

i APPLICABILITY: MODE 1.

NOTE----------.--.------.--.-

1 Pressurizer pressure limit does not apply during:

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

b. THERMAL POWER step > 10% RTP.

ACTIONS f CONDITION REQUIRED ACTION COMPLETION TIME 1

A. Pressurizer pressure A.1 Restoreparameter(s) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months t

or RCS flow rate not to within limit.

within limits.

x m m_______________)

76 ACS Mal flow rak hmil slu!i be h 370,000 b Unil 1 '

___2. y peonI , drougf C l 13.

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

i l

, 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months !

~

associated Completion Time of Condition A

not met.

C. RCS cold leg C.1 Restore cold leg 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months temperature not temperature to within l 4

within limits. limits.

D. Required Action and D.1 Reduce THERMAL POWER 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months i associated Completion to s 305; RTP.

Time of Condition C not met.

]

SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure 2 2200 psia. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.2 Verify RCS cold leg temperature s 548*F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

NOTE------------------- 4 SR 3.4.1.3 @0nly required to be met in MODE 1.

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

2. & ud j al , glAe y ??CS /dal fim uk sball be h 370,000 flrough C y s -

le 13.

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

I Secondary SpGeific Activity B 3.7.14 )

B 3.7 PLANT SYSTEMS B 3.7.14 Secondary Specific Activity i

BASES BACKGROUND Activity in the secondary coolant results from steam generator tube outleakage from the Reactor Coolant System (RCS). Under steady state conditions, the activity is primarily iodines with relatively short half lives, and thus )

is indication of current conditions. During transients, !

I-131 spikes have been observed as well as increased l releases of some noble cases. Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary coolant.

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

This limit is lo an the activity value that might be ;

expected from a ' --- tube leak (LC0 3.4.13. "RCS Operational LEAKA of primary coolant at the limit of i 1.0 Ci/gm (LC0 3.4.15, "RCS Specific Activity"). The main ;

steam line break (MSLB) is assumed to result in the release of the noble gas and iodine activity contained in the steam i generator inventory, the feedwater, and reactor coolant !

LEAKAGE. Most of the iodine isotopes have short half lives (i .e. , < 20 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months ) .

1 APPLICABLE The accident analysis of the MSLB, as discussed in the SAFETY ANALYSES Updated Final Safety Analysis Report (UFSAR), Chapter 14 (Ref. 2), assumes the initial secondary coolant specific activity to have a radioactive isotopo concentration of 0.10pCi/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

Secondary Specific Activity B 3.7.14 BASES theradiologicalconsequencesofanMSLBdonotexceed@

Te '$ctfsCc')the unit exclusion area boundary limits for whole body and thyroid dose rates.

With the loss of offsite power, the remaining steam generator is available for core decay heat dissipation by venting steam to the atmosphere through main steam safety valves (MSSVs) and atmospheric dump valves (ADVs). The i Auxiliary Feedwater System supplies the necessary makeup to ,

the steam generator. Venting continues until the reactor l i coolant temperature and pressure have decreased sufficiently ,

for the Shutdown Cooling System to complete the cooldown. I In the evaluation of the radiological consequences of this accident, the activity released from the steam generator connected to the failed steam line is assumed to be released directly to the environment. The unaffected steam generator j is assumed to discharge steam and any entrained activity l through MSSys and ADvs during the event. !

l Secondary specific activity limits satisfy l l

10 CFR 50.36(c)(2)(ii), Criterion 2. l

, LC0 As indicated in the Applicable Safety Analyses, the specific ,

1' activity limit in the secondary coolant system of I s 0.10 Ci/gmDOSEEQUIVALENTI-131tolimitthe !

radiological consequences of a Design Basis Accident (DBA) to a small fraction of the required limit (Ref.1). j Monitoring the specific activity of the secondary coolant ensures that when secondary specific activity limits are exceeded, appropriate actions are taken in a timely manner to place the unit in an operational MODE that would minimize the radiological consequences of a DBA.

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

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ML20134D304

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