Text

Amends 228 & 202 to Licenses DPR-53 & DPR-69,respectively, Revising TS for Reduction of Total Reactor Coolant Sys Flow Limit from 370,000 Gpm to 340,000 Gpm in Support of Increased SG Tube Plugging
	ML20248B013
Person / Time
Site:	Calvert Cliffs
Issue date:	05/23/1998
From:	Bajwa S; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20248B019	List: ML20248B013; ML20248B024;
References
	NUDOCS 9806010143
	Download: ML20248B013 (37)
	v • d • e

p2 REog g

t UNITED STATES

)

g j

NUCLEAR REGULATORY COMMISSION

]

2 WASHINGTON, D.C. 20666 4 001 49.....,o BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.228 License No. DPR-53 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for arqendment by Baltimore Gas and Electric Company (the licensee) dated January 31,1997, as supplemented February 13, February 28, Marcli 25, Apnl 16, August 19, and September 29,1997, January 22, March 17, April 8, and April 21, and May 22,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility wil! operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2. of Facility Operating License No. DPR-53 is hereby amended to read as follows:

i 9806010143 980523 i

PDR ADOCK 05000317 P

PDR l

.. 2.C.2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 228, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

2.C.6 Additional Conditions The Additional Conditions contained in Appendix C, as revised through Amendment No. 228, are hereby incorporated into this license. Baltimore Gas and Electric Company shall operate the facility in accordance with the Additional Conditions.

3.

This License Amendment No. 228 is effective as of the date of its issuance and shall be I

implemented within 6 months after restart from the spring 1998 refueling outage.

Implementation of this amendment shallinclude the changes to the Technical Specifications and certain changes in the Updated Final Safety Analysis Report (USFAR) i regarding Main Steam Line Break, Steam Generator Tube Rupture, Seized Rotor, and Boron Dilution Analyses.

FOR THE NUCLEAR REGULATORY COMMISSION W/

A S. Singh Bajwa, Director 1

Project Directorate 1-1

]

Division of Reactor Projects - 1/Il Office of Nuclear Reactor Regulation Attachments: 1. Pages 3 and 4 to License

No. DPR-53 i

and page 1 of Appendix C of License

No. DPR-53.

2. Changes to the Technical Specifications Date of issuance:

May 23, 1998

Pages 3 and 4 of the license and page 1 of Appendix C are attached, for convenience, for the composite license to reflect this change, i

i I

3.

Pursuant to the Act and 10 CFR Parts 30,40, and 70, to receive, possess, and use any byproduct, source or special nuclear material without l

restriction to chemical or physical form, for sample analysis or instrument I

calibration or associated with radioactive apparatus or components; 4.

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C.

This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70;is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafterin effect; and is subject to the additional conditions specified or incorporated below:

1.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2700 megawatts (thermal).

2.

Technical Sp_ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 228, are hereby incorporated to the license. The licensee shall operate the facility in accordance with the Technical Speci'ications.

(a)

For Surveillance Requirements (SRs) that are new, in Ameridment to Final Operating License DPR-53, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 227. For SRs that existed prior to Amendment 227, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first per'ormance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 227.

3.

The licensee is required to implement and maintain the administrative controls identified in Section 6 of the NRC's Fire Protection Safety Evaluation on the facility dated September 14,1979.

4.

Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR Amendment No. 3,41,45,189,220,227.228

73.21, are entitled : "Calvert Cliffs Nuclear Power Plant Physical Security Plan," with revisions submitted through FebruW 17,1988; "Calvert Cliffs Nuclear Power Plant Guard Training and Qualification Plan," with revisions submitted through November 1,1985; and *Calvert Cliffs Nuclear Power Plant Safeguards Contingency Plan," with revisions submitted February 9, 1988. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.

5.

Secondary Water Chemistry Monitorino Prooram The licensee shat! implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. This program shall include:

a.

Identification of a sampling schedule for the critical parameters and control points for these parameters.

b.

Identification of the procedures used to quantify parameters that are critical to control points.

c.

Identification of process sampling points.

d.

Procedure for recording and management of data.

e.

Procedures defining corrective actions for off control point chemistry conditions; and f.

A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate corrective action.

6.

Additional Conditions The Additional Conditions contained in Appendix C as revised through Amendment 228 are hereby incorporated into this license. Baltimore Gas and Electric Company shall operate the facility in accordance with the Additional Conditions.

D.

This license is effective as of the date of issuance and shall expire at midnight July 31, 2014.

FOR THE ATOMIC ENERGY COMMISSION Original signed by:

Rogers. Boyd A. Giambusso, Deputy Director for Reactor Projects Directorate of Licensing Attaciment:

Appendices A & B -

Technical Specifications Amendment No. 41,45,59,13,189,220,227,228

Appendix C Additional Conditions Facility Operatino License No. DPR-53 Baltimore Gas and Electric Company (BGE, the licensee) shall comply with the following condition on the schedule noted below:

l f j Amendment Number Additional Condition implementation Date i

227 The licensee is authorized This amendment is to relocate certain effective immediately and Technical Specification shall be implemented by requirements to licensee-August 31,1998.

controlled documents.

Irr.plementation of this amendment shallinclude the relocation of these requirements to the appropriate documents as described in the licensee's application dated December 4,1996, as supplemented by letters dated March 27, June 9, June 18, July 21, August 14, August 19, September 10, October 6, October 20, October 23, November 5,1997, and January 12, January 28, and March 16,1998, evaluated in the NRC staff's Safety Evaluation enclosed with this amendment.

228 The licensee is authorized The updated UFSAR shall I

to incorporate in the be implemented within 6 USFAR certain changes months after restart from regarding Main Steam Line the spring 1998 refueling Break, Steam Generator outage.

Tube Rupture, Seized Rotor, and Boron Dilution Analyses.

l

F.

m

,a *

4 UNITED STATES g

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. enmaa annt 4,.....

BALTIMORE GAS AND FI FCTRIC COMPANY DOCKET NO. 50-318 CALVERT CLIFFS NUCLEAR POWER PLANT. UNIT NO. 2

)

AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 202 License No. DPR-69 J

1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Baltimore Gas and Electric Company (the licensee) dated January 31,1997, as supplemented February 13, February 28, March 25, April 16, August 19, and September 29,1997, January 22, March 17, I

April 8, April 21, and May 22,1998, complies with the standards and requirements j-of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR ChapterI; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this l

amendment can be conducted without endangering the health and safety l

of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; l

D.

The issu ance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

l l

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.2. of Facility Operating License No. DPR-69 is hereby amended to read as follows:

l

\\ 2.C.2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 202, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

1 2.C.6 Additional Conditions l

The Additional Conditions contained in Appendix C, as revised through Amendment No. 202 are hereby incorporated into this license. Baltimore Gas and 3

Electric Company shall operate the facility in accordance with the Additional Conditions.

l 3.

This license amendment is effective as of the date of its issuance and shall be implemented within 6 months after restart from the spring 1999 refueling outage.

Implementation of this amendment shallinclude the changes to the Technical Specifications and certain changes in the UFSAR regarding Main Steam Line Break, Steam Generator Tube Rupture, Siezed Rotor, and Boron Dilution Analyses.

FOR THE NUCLEAR REGULATORY COMMISSION D

S. Singh Bajwa, Director Project Directorate 1-1 Division of Reactor Projects - 1/11 Office of Nuclear Reactor Regulation Attachments: 1. Pages 3,4, and 5 to License

No. DPR-69 l

and page 1 of Appendix C of License

No. DPR-69.

2. Changes to the Technical Specifications Date of issuance: May 23, 1998

Pages 3,4, and 5 of the license and page 1 of Appendix C are attached, for convenience, for the composite license to reflect this change.

I i

.A

3.

Pursuant to the Act and 10 CFR Parts 30,40, and 70, to receive, possess, and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed

{

sources forinstrumentation and radiation monitoring equipment calibration, and as fusion detectors in amounts as required; 4.

Pursuant to the Act and 10 CFR Parts 30,40, and 70, to receive, possess, and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument calibration or associated with radioactive apparatus or components; 5.

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the @eration of the facility.

C.

This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below-1.

Maximum Power Level The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 2700 megawatts (thermal).

2.

Technical Specifications l

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 202, are hereby incorporated to the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(a)

For Surveillance Requirements (SRs) that are new, in Amendment 201 to Final Operating License DPR-53, the first performance is due at the end of the first surveillance interval that begins at implementation of Amendment 201. For SRs that existed prior to Amendment 201, including SRs with modified acceptance criteria and SRs whose frequency of performance is being extended, the first performance is due at the end of the first surveillance interval that begins on the date the Surveillance was last performed prior to implementation of Amendment 3.

Less Than Four Pumo Operation The licensee shall not operate the reactor at power levels in excess of five (5} percent of rated thermal power with less than four (4) reactor coolant pumps in operation. This condition shall remain in effect until the licensnee has submitted safety analyses for less than four pump operation, and approval for such operation has been granted by the Commission by amendment of this license. Amendment No. 9,25,166,202

4.

The licensee is required to implement and maintain the administrative controls identified in Section 6 of the NRC's Fire Protection Safety Evaluation on the facility dated September 14,1979.

5.

Secondarv Water Chemistry Monitorina Prooram The licensee shall implement a secondary water chemistry monitoring program to inhibit steam generator tube degradation. This program shallinclude:

a.

Identification of a sampling schedule for the critical parameters and control points for these parameters.

b.

Identification of the procedures used to quantify parameters that ti.te critical to control points.

~

c.

Identification of process sampling points, d.

Procedure for recording and management of data.

e.

Proceduras defining corrective actions for off control point chemistry conditions; cnd f.

A procedure identifying the authority responsible for the interpretation of the data and the sequence and timing of administrative events required to initiate corrective action.

6.

Additional Conditions The Additional Conditions contained in Appendix C as revised through Amendment 202 are hereby incorporated into this license. Baltimore Gas and Electric Company shall operate the facility in accordance with the AdditionalConditions.

D.

Physical Protection l

The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards coritingency plans including amendments made pursuant to provisions of the l

Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain Safeguards Information protected under 10 CFR 73.21, are entitled:

"Calvert Cliffs Nuclear Power Plant Physical Security Plan," with revisions submitted through February 17,1988; "Calvert Cliffs Nuclear Power Plant Guard Training sind Qualification Plan," with revisions submitted through November 1,1985; and "Calvert Cliffs Nuclear Power Plant Safeguards Contingency Plan," with revisions submitted February 9,1988. Changes made in accordance with 10 CFR 73.55 shall be l

implemented in accordance with the schedule set forth therein.

) Amendment No. 23,25,41,166,202 I

t

E.

If harmful effects or uvidence of irreversible damage are detected by the biological monitoring program, hydrological moni oring program, and the radiological monitoring program specified in the Appendix B Technical Specifications, the licensee will provide to the staff a detailed analysis of the problem and a program of remedial action to be taken promptly to eliminate or significantly reduce ther detrimental effects or damage.

F.

This license is effective as of the date of issuance and shall expire at midnight August 13, 2016.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by-Roger S. Boyd, Director Division of Project Management Office of Nuclear Reactor Regulation

Attachment:

Changes to Technical Specifications, Appendix B Date of Issuance: November 30,1976 Amendment No. 28,84,113,116,201,202

i..

Anoendix C Additional Conditions Facility Ooerstina License No. DPR-69 i

Baltimore Gas and Electric Company (BGE, the licensee) shall comply with the following condition on the schedule noted below:

Amendment Number Additional Condition implementation Date 201 The licensee is authorized This amendment is to relocate certain effective immediately and Technical Specification shall be implemented by requirements to licensee-August 31,1998.

controlled documents.

I implementation of this amendment shallinclude the relocation of these requirements to the appropriate documents as described in the licensee's l

application dated l

December 4,1996, as supplemented by letters dated March 27, June 9, June 18, July 21, August 14, August 19, September 10, October 6, October 20, October 23, November 5,1997, and January 12, January 28, and March 16,1998, evaluated in the NRC staffs Safety Evaluation enclosed with this amendment.

202 The licensee is authorized The updated UFSAR shall to incorporate certain be implemented within 6 changes in the USFAR months after restart from 1

Regarding Main Steam the spring 1999 refueling Line Break, Steam Generator outage Tube Rupture, Seized Rctor and Boron Dilution Analyses.

l

1 l

i ATTACHMENTTO LICENSE AMENDMENTS i

AMENDMENT NO. 228 FACILITY OPERATING LICENSE NO. DPR-53 AMENCMENT NO. 202 FACILITY OPERATING LICENSE NO. DPR-69 DOCKET NOS. 50-317 AND 50-318 i

l Revise Appendix A Technical Specifications, including the issued but not yet

]

implemented improved Technical Specifications, by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by amendment number and l

contain marginal lines indicating the areas of change.

Current Technical Specifications - Unit No.1 Remove Pgte,s insert Paaes 2-3 2-3 2-5 2-5 3/4 2-8 3/4 2-8 3/4 7-4 3/4 7-4 8 3/4 7-3 8 3/4 7-3 Current Technical Specifications - Uriit No. 2 Remove Paaes insert Paaes 2-1 2-1 l

2-3 through 2-7 2-3 through 2-8 3/4 2-8 3/4 2-8 3/4 7-4 3/4 7-4 B 3/4 7-3 8 3/4 7-3 j

l 1

Imoroved Technical Specifications - Unit Nos.1 and 2 Remove Paaes insert Paaes 2.0-1 2.0-1 2.0-2 2.0 2 and 2.0-3 3.3.1-9 3.3.1-9 3.3.1-11 3.3.1-11 3.4.1-1 3.4.1-1 3.4.1-2 3.4.1-2 3.7.1-4 3.7.1-4 8 3.7.1-3 8 3.7.1-3 B 3.7.14-1 B 3.7.14-1

_________________u

O

.O

(

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

\\

600 h

UNACCEPTABLE OPERATION 580 0.93 1.02 1.09 1.13 FOR PRE-CLAD COLLAPSE

\\

OPERATION ONLY UNACCEPTABLE F

OPERAT!0N N

LIMITS CONTAIN NO ALLOWANCE FOR INSTRUMENT ERROR OR k

FLUCTUATIONS go c

g s,

g VAllD FOR AXIAL SHAPES AND 4 I6 I 49 4 ROD PADIAL PEAKING 520 FACTORS LESS THAN OR EQUAL TO THOSE ON FIGURE B2.1-1 b

ACCEFTABLE g 500 REACTOR OPERATION LIMITED TO LESS OPERATION 2-THAN 580F BY ACTUATION OF THE SECONDARY SAFETY VALVES 480 433 1.67 1.74 1.80 1 E2 0

0.2 0.4 0.6 0.8 1

1.2 1.4 1.6 1.8 2

FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT CALVERT CLIFFS - UNIT 1 2-3 Amendment No. 228

t d

n n

a d

af D

l e

,o E

o e

T o

f R%aA c

E0 R

f W3d r

o O

nf o

Pfao t

p o

c o

L R%

a t

ate 0 e

Mni f

r w

S Ri07 o

E E oP0 n

l U e Hp 1

g e

L l

TtL R i b

esM<W A b eA E

s V a e

a s

S c v RfO d i

a e

T E i omEoP s

i h

L l

buH f

p g s c

IM B

p amTmL o i p n

A p

i uA 0

s i

IL W A %nDmM )

0 p 5 O

0iEiR 2w 4 8

0g T

L t 1 mTxE 9o 2 4 6 1 n L

o AaH l

i N

O 5 5 3 2r A N < aRmT 2f I

PT a

E D

hEfR t

S tToE n

P iA W

a d

wRmO l

e I

R uP o

e

,f m o

f T

Roil c

E xA f

N 1

O W%aM r

o O0mR o

I 2

T P3 E

t p

A aH c

o Lf T

a t

2 TN Aod e

E E

M nD r

w Rt aE o

L M

E n T

n l

B U

A i

e Hi

,A g

e t

T T

T l

TORE i

b S

N b pE s

I a etWf e

a s

N O c veOo d

i a

e I

P i osP s

i h

E T

l b

f p g s c

E p amL0 o

i p n

V S

p uA 0 s i

I T

A %mM7 0

p 5 C

P 0iR0 2w 4 8

0g I

t 1 nE1 9o 2 4 6 1 n ET R o iH l

i O

T N < mT5 2f $ $ t 3r RP W

ROT h h l

C w

g g -

e A

o i i v

E L

H H e e

r L

R us 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 e e r r

r n

r r o o

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 l

o u i g

m A a r t

s a N u e c

s t a

a O n w a

e n ew ew I

a o e

r o t o t e T M P R

P C Sl SL CN li F

1 2

3 4 5 6 7

9rh~ bm5 ' EZ -

o GE&en i

r

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:

a.

Cold Leg Temperature 5 548 F b.

Pressurizer Pressure 2 2200 psia

i c.

Reactor Coolant System Total Flow Rate 2 340,000 gpm l

d.

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

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 J

1ess 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 .

{

l SURVEILLANCE REQUIREMENTS l

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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The Reactor Coolant System total flow rate shall be detemined to i

be within its limit by measurement at least once per 18 months.

J 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. 228

(

1 1

3/4.7 PLANT SYSTEMS TABLE 4.7-1 l

STEAM LINE SAFETY VALVES PER LOOP VALVE LIFT SETTINGS" ALLOWABLE ORIFICE SIZE l

a.

RV-3992/4000 935-995 psig R

l 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 935-1050 psig R

l f.

RV-3997/4005 935-1050 psig R

l g.

RV-3998/4006 935-1050 psig R

l h.

RV-3999/4007 935-1050 psig R

l 1

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 935 and 1035 psig, and the 4 remaining valves lift between 935 and 1050 psig.

l CALVERT CLIFFS - UNIT 1 3/4 7-4 Amendment No. 228 J

]

i 3/4.7 PLANT SYSTEMS BASES j

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 100 gallons per day primary to l

secondary tube leak in the steam generator of the affected steam line and concurrent loss of offsite electrical power. These values are consistent with the assumptions used in the accident analyses.

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 stes'.a line rupture. This restriction is required to 1) minimize the positive

{

reactivity effects of the Reactor Coolant System cooldown associated with 1

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.

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 I

with the assumptions used in the accident analyses.

)

3/4.7.4 SERVICE WATER SYSTEM J

The OPERABILITY of the Service Water System ensures that sufficient cooling capacity is available for continued operation of equipment during normal j

and accident conditions. The redundant cooling capacity of this system,

)

assuming a single failure, is consistent with the assumptions used in the 1

accident analyses.

{

CALVERT CLIFFS - UNIT 1 B 3/4 7-3 Ameadment No. 228

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 o)erating loop cold leg coolant temperature shall not exceed the limits s 10wn in Figure 2.1-1 l

APPLICABILITY: MODES 1 and 2.

ACTION:

a.

Whenever the point defined by the combination of the highest 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 .

l 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.

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 I and 2 a.

Whenever the Reactor Coolant System pressure has exceeded 2750 psia, be in NOT 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 preparped and submitted to the Commission, the offsite review function and the Vice President -

Nuclear Energy within 14 days of the violation.

Figure 2.1-la shall apply through Unit 2 Cycle 12.

l CALVERT CLIFFS - UNIT 2 2-1 Amendment No. 202

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 600 UNACCEPTABLE OPERATION 580 0.93 1.02 1.09 1.13 FOR PRE-CLAD COLLAPSE OPERATION ONLY UNACCEPTABLE p

OPERATION J 560

!s LIMITS CONTAIN NO ALLOWANCE 3

FOR INSTRUMENT ERROR OR M

FLUCTUATIONS g 640

% %,#g %,

g VALID FOR AXIAL SHAPES AND y

ROD RADIAL PEAKING

$ 520 FACTORS LESS THAN OR EQUAL T0 THOSE ON FIGURE B2.1-1 500 REACTOR OPERATION LIMITED TO LESS I

THAN 580F BY ACTUATION OF THE SECONDARY SAFETY VALVES 480 460 1.67 1.74 1.80 1.82 0

0.2 0.4 0.6 0.8 1

1.2 1.4 1.6 1.8 2

FRACTION OF RATED THERMAL POWER FIGURE 2.1-1 REACTOR CORE THERMAL MARGIN SAFETY LIMIT EFFECTIVE AFTER UNIT 2 CYCLE 12 CALVERT CLIFFS - UNIT 2 2-3 Amendment No. 202

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2

i e

i i

p

$- 3 l

- l5

_ 3

/

3l t$

/

5

- D p

ort 00 1,

_ Sg i

g l

O S

0 h lll

- Eg 5!

$5 E!

h2 5g

- 8 5

I ol

! Igl l ll R

8<E5 5

lnilliIl'dg

- =

ei i

e i

i e

d.'ratuvasanstramoonnnIxyn FIGURE 2.1-A REACTOR CORE THERMAL MARGIN SAFETY LIMIT EFFECTIVE THROUGH UNIT 2 CYCLE 12 CALVERT CLIFFS - UNIT 2 2-4 Amendment No. 202

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFET_Y,.,1YSTEM SE TINGS

, REACTOR TRIP SETPOI'4TS 2.2.1 The reactor protective instrumentation setpoints shall be set t.onsistent with tne Trip Setpoint values shown in Table 2.2-1.

APPLICA1IJLITY: As shown for each cilannel in Table.3.31.

ACTION: With a reactor protectiva instrumentation setpoint less i

conservative than the value shown in the Allo ~< table Values column of Table 2.2-1, declare the channel inoperab e and apply the applicab:e ACTION

)

statement requirement of Specification 3.3.1.3 until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoin: value.

(

\\

l l

1 1

l l

i i

l l

l CALVERT CLIFFS - UNIT 2 2-S Amendment No. 202 l

l u

n d

af D

e

,o E

e T

f R%aA E0 R

r f

W3d o

o O

nf t

Pfao c

p o

a o

L R%

e t

ate 0 r

MnW w

S iiO7 n

o t

E EoP0 g

l U e Hp 1

. i e

L l

TtL R s b

A b eA<E e,w V a esM W do a s

c v RfO l

i a

e omE oP f f s

i h

E i L

l bun o

p g s c

S B p amYmL t

i p n

T A

p i

uA " n 0 s i

I W A %nDmM 4a 0 p 5 M

O 0iEiR 2l 4

8 0g I

L t 1 mTxE 9o 2 4 6 1 n L

L o

Aa!

o i

A N <aRmT 2c 5 5 2 3r TN I

a O

D t

P hEfR n

T tToE a

E iA W

l d

S wRmO o

e P_

uP o

e

,f m c

f I

Roil R

E xA r

f T

W%aM o

o O0mR t

N P3 E

c p

1 O

aH a

o I

Lf T

e t

2 T

Aod r

A M

nD w

2 T

RtaE n

o N

En

,T g

l E

E e

Hi A

e i

L M

T ToRP s

b l

B U

N b pE e

A R

I a etWf d

a s

T T

O c veOo i

a e

S P i osP f

s i

h N

T l

b o

p g s c

I E

p amL0 i p n

S p

uA 0 s i

E A %mM7 0 p 5 V

P 0i i0 2w 4 8

0g t

I I t 1 nE1 9o 2 4 6 1 n T

R o iH l

i

<ii T5 2f 5 5 t 2r C

T N ETO R

W P

h h l

R w

g g e

O o

i i v

T L

H H e e

C r

L A

u E

s r

R w

e e s e

p o

r r e t

i h l

u u r a

r g s s P W

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

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

m e e

U R e

i L

r r n G G

L l

o u i g

m A

a r t

s a N

u e c

s t a

a O

n w a

e n ew ew I

a o e

r o t o to I M P R

P C Sl SL C!

j Fj 1 2

3 4 5 6 7

g Gg*

G'EZm m

EaRan

tn oi e

n bf R

d o

L oe o

eO td t rhC e

i et t

dv dg e

u eo er)h n

t r ta2t i

sp sl(

m u

u n

js jeri r

S dt dho e

E ai at

,d p

U m

e L

ti t nad s

A nl naii e

V i ihsv d

oe otpo d a

E ph p

r i c

L ttR t s5p s e

S B

e L

es7 p

d T

A sdO se8s W

eC l 1t 5

6 IM O pe p

i 3 L ice i t) m 1

2 IL L

rxh ro1i A

A Tet Tn(l 5

N 5 TN I

OP tn tn T

oi oi E

n n

S d

d oe oe

)

P td td e

i i

t d

I e

R dv dv u

u T

eo eo n

t r tr i

n i

N sp sp m

t O

u u

js js r

n I

o T

dt dt e

C A

ai ai p

(

T m

m T

ti ti s

N 1

E N

nl nl e

I i

i d

M 2

U O oe oe d

a P

ph ph i

c R

2 T

T ttR ttR s

e E

e L

p d

S S

sdO sdO E

N eC eC 5

6 L

I P pe pe 3

B I ice i ce 1

2 A

E R

rxh rxh A

T V

T Tet Tet 5

N 5 ITCET I

O U

RP

)

s h

U p

g RO m

e i

e u r

H T

r P u

C u

s A

s t g

E r

s n R

e a r

e r

l Ph w

)

P o

g o

U o

ri P

w C oH t

e t

f e

L r

a-o s

/ o r

f n t ee e

f i cg nc g

o g an en d n

T r

ei Ge a a I

x a Rt r o h N

u M

a me L C U

l rr af

_ f l

ue ef f f L

a op ti o o SD m FO A

l N a s e O i e

s t

I x

h o a T A T a b

L R C

N U

0 1

F 8 9

1 1

gGg~bnaigZ" ma n

an

2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS j

TABLE 2.2-1(Continued)

TABLE NOTATION See Specification 3.2.5, "DNB Parameters," for the design reactor coolant flow.

The Reactor Coolant Flow - Low trip setpoint and allowable value shall be t 95% of design reactor coolant flow through Unit 2 Cycle 12.

(1)

Trip may be bypassed below 10-*% OF RATED THERMAL POWER; Lypass shall be automatically removed when THERMAL POWER is 110-'% of RATED THERMAL PGWER.

(2)

Trip may be manually bypassed below 785 psia; bypass shall be automatically removed et or above 785 psia.

(3)

Trip may be bypassed below 15% of RATED Tf D MAL POWER; bypass shall be automatically removed when THERMAL POWER is E 15% of RATED THERMAL POWER.

(4)

Trip may be bypassed below 10-'% and above 12% of RATED THERMf4L POWER.

i

)

CALVERT CLIFFS - UNIT 2 2-8 Amendment No. 202

r --

~-

t 3/4.2, POWER DISTRIBUTION LIUT.S_

l 3/4.2.5 DNB PARffiETERS l LIMITING CONDITIGN FOR OPEr3TICH 3.2.5 The following DNB related p6rameters shall be taaintained withir, the i

limits shown:

I a.

Cold Leg Temperature 5; 546'F b.

Pressurizer Pressure 2 2200 psia" c.

k actor Coulant System Total Flow Rate 2 340,000 ppm" l

l d.

AXIAL SHAPE.7DEX, THERMAL POWER u sp&ified in the COLR APPLICABILITY: MODE 1.

ACTIOJ{: With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hwrs 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 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.

j I

i l

l l

I Limit not applicable during either THERMAL POWER ramp increase in l

I excess of 5% of RATED THERMAL POWER per minute or a THERMAL POWER step I

increase of greater than 10% of RATED THERMAL POWER.

The Reactor Coolant System Total Flow P. ate limit shall be 1370,000 gpm I

througn Unit 2 Cycle 12.

CALVERT CLIFFS - UNIT 2 3/4 2-8 Amendment No. 202

i 3/4.7 PLANT SYSTEMS TABLE 4.7-1 STEAM LINE SAFETY VALVES PER LOOP l

VALVE NUMBER LIFT SETTINGS

ALLOWABLE ORIFICE SIZE a.

RV-3992/4000 935-995 psig R

i

(

b.

RV-3993/4001 935-995 psig R

c.

RV-3994/4002 935-1035 psig R

d.

RV-3995/4003 935-1035 psig R

t e.

RV-3996/4004 935-1050** psig R

l f.

RV-3997/4005 935-1050" psig R

l g.

RY-3998/4006 935-1050" psig R

l h.

RV-3999/4007 935-1050" psig R

l l

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

1050 psig (between 935 and 1065 psig through Unit 2, Cycle 12).

The maximum allowable lift setting for the highest set valves shall be 1065 psig through Unit 2, Cycle 12.

CALVERT CLIFFS - UNIT 2 3/4 7-4 Amendment No. 202

3/4.7 PLANT SYSTEMS BASES 3/4.7.1.3 Condensate Storaae Tank The OPERABILITY of the condensate storage tank with the minimu'n water volume ensures that sufficient water is available to maintain the RCS at NOT STAN05Y 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 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 100 gallons per day primary to l

secondary tube leak in the steam generator of the affected steam line and a concurrent loss of offsite electrical power. These values are consistent with the assumptions used in the accident analyses.

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 rupture. This restriction is required to 1) minimize the positive reactivity effects of the Reactor Coolant System cooldown associated with the blowdown, and ?.) 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 90 F and 200 psig are based on steam gen m tor secondary side limitations and are sufficient to prevent brittle fracture.

j l

l I

l i

CALVERT CLIFFS - UNIT 2 B 3/4 7-3 Amendment No. 202 i

l SLs l

2.0 1

l 2.0 SAFETY LIMITS (SLs) i 2.1 SLs i

2.1.1 Reactor Core SLi

{

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 l

limits shown in Figure 2.1.1-1.

............................ NOTE......-....-...'...............

For Unit 2 only Figure 2.1.1-la shall apply through Cycle 12.

l l

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 f

In MODES 1, 2, 3, 4, and 5, the RCS pressure shall be maintained s 2750 psia.

l 2.2 SL Violations 2.2.1 If SL 2.1.1 is violated, 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 .

)

2.2.2 If SL 2.1.2 is violated:

1 2.2.2.1 In MODE.1 ur 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 .

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

l CALVERT CLIFFS - UNIT 1 2.0-1 Amendment No. 228.

CALVERT CLIFFS - UNIT 2 Amendment No. 202

{

s.

SLs 2.0 -

600

'hACCEPTABLE OPERATION g

0.93 1.02 1.09 1.13 FOR PRE CLAD COLLAPSE OPERAT10 ONLy -

UNACCEPTABLE

{f..g OPERATION LlHITS CONTAIN NO ALLOWANCF FOR INSTRUMENT ERROR OR g '

FLUCTUATIONS

,j c

g g

h' VAllD FOR AXIAL SHAPES AND y # %, %,

4 a

R00 RADIAL PEAKING

@ 620 FACTORS LESS THAN OR EQUAL TO THOSE ON FIGURE B2-.1-1 l

i b

W ACCEPTABLE

~

REACTOR OPERATION LIMITED TO LESS OPERATION THAN 500F BY ACTUATION OF 'RiE SECONDARY SAFE!Y VALVES 450 i

460 1.67 1.74 1.,80 1.82 0

0.2 0.4 0.6 3.8 1'

~

.2 1.4 1.6 1.8 2

1 FRACTION OF RATED THERMAL POWER l

Figure 2.1.1-1 (page 1 of IF Unit 1 and Unit 2 Reactor Core Thermal Margin Safety Limit CALVERT CLIFFS - UNIT 1 2.0-2 Amendment No. 228 CALVERT CLIFFS - UNIT 2 Amendment No. 202

O 6e SLs 2.0 2.0 SAFETY LIMITS (SLs) l 60r' I

I i

i i

UNACCEPTABLE OPERATION UNACCEPTABLE 5BC-OPERATION FOR PRE-CLAD COLLAPSE OPERATION ONLY u.

560-

v 99 g

LIMITS CONTA!N NO ALLOWANCE FOR INSTRUMENT ERROR OR g 54C-FLUCTUATIONS VALID FOR AXIAL SHAPES AND

{ 52C.

FACTORS WITHIN LIMITS ROD RADIAL PEAK:NG REACTOR OPERATION LIMITED TO LESS soc _

THAN 580 F BY ACTUATION OF THE SECONDARY SAFETY VALVES ACCEPTABLE

4gg, OPERATION o ogg q o

46C O

o.2 o.4 o.6 o.8 1.o 1.2 1.4 1.6 1.8 2.o FRACTION OF RATED THERMAL POWER

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

For Unit 2 only, Figure 2.1.1-la shall apply through Cycle 12.

Figure 2.1.1-la (page 1 of 1) l Unit 1 and Unit 2 Reactor Core Thermal Margin Safety Limit CALVERT CLIFFS - UNIT 1 2.0-3 Amendment No. 228 CALVERT CLIFFS - UNIT 2 Amendment No. 202

E 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 j

1.

Power Level-High 1, 2 SR 3.3.1.1 s 10% RTP above SR 3.3.1.2 current THERMAL POWER i

SR 3.3.1.3 but not < 30% RTP nor SR 3.3.1.4

> 107% RTP SR 3.3.1.5 SR 3.3.1.8 i

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 SR 3.3.1.8 3.

Reactor Coolant 1, 2 SR 3.3.1.1 2 92% of Design Flow (8) l Flow-Low SR 3.3.1.4 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

{

CALVERT CLIFFS - UNIT 1 3.3.1-9 Amendment No. 228 CALVERT CLIFFS - UNIT 2 Amendment No. 202

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

Reactor Protective System Instrumentation SURVEILLANCE FUNCTION N0 DES REQUIREMENTS ALLOWABLE VALUE 9b. Asymmetric Steam 1, 2 SR 3.3.1.1 5; 135 psid Generator Transient SR 3.3.1.4 (ASGT)")

SR 3.3.1.7 SR 3.3.1.8 SR 3.3.1.9

10. Loss of Load")

1)

SR 3.3.1.6 NA 4

SR 3.3.1.7 "I

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

> I2% RTP.

Bypass shall be automatically removed when THERMAL POWER is 2 1E-4% RTP and < 12% RTP.

)

Bistable trip unit may be bypassed when THERMAL POWER is < 1E-4%.

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

"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.

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

Bypass shall be automatically removed when THERMAL POWER is 2 15% RTP.

"I Trip is only applicable in MODE 1215% RTP.

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

(8)

The Reactor Coolant Flow-Low allowable value shall be 2 95% for Unit 2 only, through Cycle 12.

CALVERT CLIFFS - UNIT 1 3.3.1-11 Amendment No. 228 CALVEP; CLIFFS - UNIT 2 Amendment No. 202

I o

l u

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 l

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; b.

RCS cold leg temperature (T ) s 548 F; and c

c.

RCS total flow rate 2 340,000 gpm.

l

NOTE----------------------------

The RCS total flow rate limit shall be 2 370,000 for Unit 2 only, through Cycle 12.

APPLICABILITY:

MODE 1.

NOTE-------------------------_--

Pressurizer pressure limit does not apply during:

a.

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

THERMAL POWER step > 10% RTP.

l 1

CONDITION REQUIRED ACTION COMPLETION TIME A.

RCS DNB parameter (s)

A.1 Restoreparameter(s) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months not within limits.

to within limit.

CALVERT CLIFFS - UNIT 1 3.4.1-1 Amendment No. 228 CALVERT CLIFFS - UNIT 2 Amendment No. 202

f U

- ' - ~ ~

~ ~ ~ - - - - - - ~ ~ ~ - - - -

C 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 not met.

ZRVEILLANCE REQUIREMENTS 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 Veriff RCS cold leg temperature s 548 F.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.4.1.3 NOTE--------_----------

For Unit 2 only, the RCS total flow rate shall be 2 370,000 through Cycle 12.

Verify RCS total flow rate 2 340,000 gpm.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months l

l l

SR 3.4.1.4 Verify measured RCS total flow rate is 24 months

{

within limits.

l CALVERT CLIFFS - UNIT 1 3.4.1-2 Amendment 7:3. 228 CALVERT CLIFFS - UNIT 2 Amendment No. 202

o d

MSSVs 3.7.1 Table 3.7.1-2 Main Steam Safety Valve lift Settings VALVE NUMBER LIFT SETTING )

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

RV-3992 RV-4000 935-995 RV-3993 RV-4001 935-995 RV-3994 RV-4002 935-1035 RV-3995 RV-4003 935-1035 RV-3996 RV-4004 935-1050 J

RV-3997 RV-4005 935-1050 j

RV-3998 RV-4006 935-1050 RV-3999 RV-4007 935-1050 UI 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 1050 psig.

l

NOTE-------------------------------------

For Unit 2 only, the maximum allowable lift setting for the four highest set valves for a given steam line shall be 1065 psig through Cycle 12.

CALVERT CLIFFS - UNIT 1 3.7.1-4 Amendment No. 228 CALVERT CLIFFS - UNIT 2 Amendment No. 202 l

r-u

]

)

MSSVs B 3.7.1 l

BASES lift between 935 and 1050** psig. Thus, the MSSVs still l

perform that design basis function properly.

l t

This LC0 provides assurance that the MSSVs will perform their designed safety function to mitigate the consequences 1

of accidents that could result in a challenge to the RCPB.

- The four remaining valves lift between 935 and 1065 psig for Unit 2 only, through Cycle 12.

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

)

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.

l In MODES 4 and 5, there 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.

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

4 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 power level. The number of inoperable MSSVs will determine i

the necessary level of reduction in secondary system steam flow and THERMAL POWER required by the reduced reactor trip i

l CALVERT CLIFFS - UNITS 1 & 2 B 3.7.1-3 Revision 1

________A

a e

Secondary Specific 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, 1-131 :. pikes have been observed as well as increased releases of some noble gases. Other fission product isotopes, as well as activated corrosion products in lesser amounts, may also be found in the secondary ca snt.

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 lower than the activity value that might be expected from a 100 gallons per day tube leak (LCO 3.4.13, l

"RCS Operational LEAKAGE") of primary coolant at the limit of 1.0 Ci/gm (LCO 3.4.15, "RCS Specific Activity"). The I

mainsteamlinebreak(MSLB)isassumedtoresultinthe release of the noble gas and iod%e >ctivity contained in the steam senerator 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 ).

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

ML20248B013

Text

Attachment:

Navigation menu

Search