ML20003H428

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Amends 53 & 36 to Licenses DPR-53 & DPR-69,respectively, Adding Certain Lessons Learned Category a Requirements Re TMI Accident to Tech Specs in Response to 800702 Request
ML20003H428
Person / Time
Site: Calvert Cliffs  Constellation icon.png
Issue date: 04/21/1981
From: Clark R
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20003H429 List:
References
NUDOCS 8105050851
Download: ML20003H428 (37)
v • d • e


Text

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION o

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.I WASHINGTON, D. C. 20655

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BALTIMORE GAS AND ELECTRIC COMPANY DOCKET NO. 50-317 CALVERT CLIFFS NUCLEAR POWER PLANT UNIT NO. 1 AMEN 0 MENT TO FACILITY OPERATING LICENSE Amendment No. 53 License No. DPR-53 1.

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

The application for amendment by Baltimore Gas & Electric A.

Company (the licensee) dated November 10, 1980 and supplemented November 25, 1980 and January 23, 1981, 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; The facility will operate in confonnity with the application, B.

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 The issuance of this amendment is in accordance with 10 CFR E.

Part 51 of the Commission's regulations and all applicable requirements have been satisified.

t 810505973;,

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-53 is hereby amended to read as follows:

(2) Technical Specifications The Technical SDecifications Contained in Appendices A and B, as revised through Amendment No. 53, are hereby incorporated in the license. The licensee shall operate the facility in acc.Jance with the Technical Specifications.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

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lark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: April 21, 1981 l

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UNITED STATES

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i NUCLEAR REGULATORY COMMISSION g

g WASHINGTON, D. C. 20666 t

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

' AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 36 License No. DPR-69 1.

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

A.

The application for amendment by Baltimore Gas & Electric Company (the licensee) dated November 10, 1980 and supplemented November 25, 1980 and January 23, 1981, 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 will operate in confomity with the application, the provisions of the Act, and the rules and regtlations 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 satisified.

I

-2 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:

2 Technical Specifications T-he Technical Specifications contained in Appendices A and B, as revised through Amendment No. 36, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of ths date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION bert A. Clark, Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: April 21, 1981 t

i I

ATTACHMENT TO LICENSE AENDMENT NOS. 53 AND 36_

FACILITY OPERATING LICENSE NOS. DPR-53 AND DPR-69 DOCKET NOS. 50-317 AND 50-318 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain document completeness.

Page IV X

XVI 3/4 3-11 3/4 3-13

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3/4 3-17 3/4 3 3/4 3-22 3/4 3-23 3/4 3-41 3/4 3-42 3/4 4-3 3/4 4-4 3/4 4-5 B3/4 3-3 B3/4 4-1 B3/4 4-2 l

B3/4 4-2a (added) l 6-4 6-5 l

6-6 l

6-21 6-22 added

-,, - - -, - +. -,,

.-s.y

. _ - ~ -. - -. - -,, --

y

INDEX LIMITING CONDITIdNS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...........................................

3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 80 RATION CONTROL Shutdown Margin - T

> 200 F.......................

3/4 1-1 avg avg 200'F.......................

3/4 1-3 Shutdown Margin - T Boron Dilution...................................... 3/4 1-4 Modera tor Tempera ture Coefficient................... 3/4 1-5 Minimum Temperature for Cri ticality.................. 3/4 1-7 3/4.1.2 B0 RATION SYSTEMS Flow paths - Shutdown................................

3/4 1-8 Flow Paths - 0perating...............................

3/4 1-9 Cha rg i ng Pump - S hu tdown............................. 3/4 1-10 Charging Pumps - 0perating...........................

3/4 1-11 Boric Acid Pumps - Shutdown..........................

3/4 1-12 Boric Acid Pumps - 0perating.........................

3/4 1-13

  • Borated Water Sources - Shutdown.....................

3/4 1-14 Borated Water Sources - Operating....................

3/4 1-16 3/4.1.3 MOVABLE CONTROL ASSEMBLIES Full Length CEA Position.............................

3/4 1-17 Posi tion Indicator Channel s.......................... 3/4 1 -21 CEA Drop Time........................................

3/4 1-23 Shutdown CEA Insertion Limits........................

3/4 1-24 Regulating CEA Insertion Limits......................

3/4 1-25 CALVERT CLIFFS - UNIT 1 III Amendment No. 32 Amendment No. 18 CALVERT CLIFFS - UNIT 2

_ ~ _

.HDEX Z

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 L I N EA R H E AT R ATE.....................................

3/4 2-1 3/4.2.2 TOTAL PLANAR RADIAL PEAKING FACT 0R...................

3/4 2-6 3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR...............

3/4 2-9 3/4.2.4 AZIMUTHAL POWER TILT.................................

3/4 2-32 3/4.2.5 DELETED..............................................

3/4 2-13 3/4.2.6 DNB PA RA M ET ERS.......................................

3/4 2-14 3/4.3 INSTRUMENTATION '

3/4.3.1 R EA CTOR PROTECTIVE INSTRUMENTATION................... 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION....................................

3/4 3-10 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.................

3/4 3-25 I

Incore Detectors.....................................

3/4 3-29 Sei smi c Instrumentati on..............................

3/4 3-31 Meteorological Instrumentation....................... 3/4 3-34 Remote Shutdown Instrumentation......................

3/4 3-37 Post-Accident Instrumentation........................

3/4 3-40 Fire Detection Instrumentation....................... 3/4 3-43 3 /4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS................................

3/4 4-1 3/4.4.2 S A F E T Y V A L'J E 3.................,...................... 3/4 4-3 f

3/4.4.3 RELIEF VALVES.................,......................

3/4 4-4 CALVERT CLIFFS - UNIT 1 IV Amendment No. 77, 75,/}9, 53 CALVERT CLIFFS - UNIT 2 Amendment No. ), JJ, JS, 36 l

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INDEX BASES 4

SECTION PAGE 3/4.0 APPLICABILITY..........................................

B 3/4 0-1 3/4.1 RE' ACTIVITY CONTROL SYSTEMS 3/4.1.1 B0 RATION CONTR0L...,..................................

B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS.....................................

B 3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES...........................

B 3/4 1-3 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 LINEAR HEAT RATE.....................................

B 3/4 2-1 3/4.2.2 TOTAL PLANAR RADIAL PEAKING FACTOR...................

B 3/4 2-1 4

3/4.2.3 TOTAL INTEGRATED RADIAL PEAKING FACTOR...............

B 3/4 2-1

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3/4.2.4 AZIMUTHAL POWER TILT.................................

B 3/4 2-1 3/4.2.5 FCEL RESIDENCE TIME..................................

B 3/4 2-2 3/4.2.6 DNB PARAMETERS.......................................

B 3/4 2-2 3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION...........................

B 3/4 3-1 3/4.3.2 ENGINEERED S AFETY FEATURE INSTRUMENTATION............ B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION........................... B 3/4 3-1 CALVERT CLIFFS - UNIT 1 Amendment No. 21

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CALVERT CLIFFS-UNIT 2 IX Amendment No. 9

INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS...............................

B 3/4 4-1 3/4.4.2 SAFETY VALVES........................................

B 3/4 4-1 3/4.4.3 RELIEF VI.LVES................'....................... B 3/4'4-2 3/4.4.4 PRESSURIZER.........................................

B 3/4 4-2 3/4.4.5 STEAM GENERATORS....................................

B 3/4 4-2 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE......................

B 3/4 4-3 3/4.4.7 CHEMISTRY...........................................

B 3/4 4-4 B 3/4 4-4 3/4.4.8 SPECIFIC ACTIVITY...................................

3/4.4.9 PRESSURE / TEMPERATURE LIMITS.........................

B 3/4 4-5 3/4.4.10 STRUCTURAL INTEGRITY................................

B 3/4 4-12 3/4.4.11 CORE BARREL MOVEMENT............._...................

B 3/4 4-12 3/4.4.12 LETDOWN LINE EXCESS FLOW............................

B 3/4 4-12 3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3/4.5.1 SAFETY INJECTION TANKS..............................

B 3/4 5-1 t

3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS........................

B 3/4 5-1 l

3/4.5.4 REFUELING WATER TANK (RWT)..........................

B 3/4 5-2 l

CALVERT CLIFFS - UNIT 1 Amendment No. 53 CALVERT CLIFFS - UNIT 2 X

Amendment No. 6, 36 l

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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY............................................

6-1 6.2 ORGANIZATION r

0 f f s i t e...................................................

6-1 Fa c i.1 1 ty S t a f f............................................

6-1 6.3 FACIL ITY STAFF QUALIFICATIONS............................. 6-6 6.4 TRAINING..................................................

6-6 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS AND SAFETY REVIEW COMMITTEE (POSRC)

Fu n c t i o n................................................

6-6 Composition.............

............................ 6-6 Al t e rn a t e s..............................................

6-6 Meeting Frequency.......................................

6-7 l

Quorum'..................................................

6-7 Responsibi11 ties........................................

6-7 A u t h o r i ty............................................... 6 - 8 R e c o r d s.... '............................................. 6 - 8 6.5.2 0FF SITE SAFETY REVIEW COMMITTEE (OSSRC)

Fu n c t i o n................................................

6 - 8 Composition........................................

.... 6-9 Al t e r n a t e s..............................................

6-9 Co n s u l t a n t s.............................................

6-9 Meeting Frequency.......................................

6-9 Quorum..................................................6-9 Review..................................................6-10 Audits..................................................6-11 A u t h o r i ty...............................................

6-1 1 R e c o r d s.................................................

6-12 l

CALVE 7.T CLIFFS - UNIT 1 XV Amendment No. 26 CALVERT CLIFFS - UNIT 2 Amendment No. 11 l

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INDEX ADMINISTRATIVE CONTROLS

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SECTION PAGE 6.6 REPORTABLE OCCURRENCE ACTI0N..............................

6-12 6.7 SAFETY LIMIT VIOLATION....................................

6-13 6.8 PROCEDURES................................................

6.

6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS AND REPORTABLE OCCURRENCES..............

6-14 6.9.2 SPECIAL REP 0RTS.........................................

6-18 6.10 RECORD RETENTION.........................................

6-19 6.11 RADIATION PROTECTION PR0 GRAM.............................

6-20 6.12 HIGH RADIATION AREA......................................

6-20 6.13 ENVIRONMENTAL QUAL IFICATI UN.............................

6-21 6.14 SYSTEM INTEGRITY........................................

6-21 6.lE IODINE MONITORING.......................................

6-22 l

l CALVERT CLIFFS - UNIT 1 Amendment No. 19, 53 CALVERT CLIFFS - UNIT 2 XVI Amendment F3. 79, 36

52 g '

TABLE 3.3-3 GG 53 g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 4 MINIMUM Pp TOTAL NO.

CHANNELS CHANNELS APPLICABLE Jig; FUNCTIONAL UNIT.

OF CHANNELS TO TRIP OPERABLE MODES ACTION io SAFETY INJECTION (SIAS) )

1.

c: e a.

Manual (Trip Buttons 2

1 2

1, 2, 3, 4 6

5555

, 4 b.

Containment Pressure -

"3 "

High 4

2 3

1.2,3 7*

c.

Pressurizer Pressure -

Low 4

2 3

1,2,3(a) 7*

2.

CONTAINMENT SPRAY (CSAS) a.

Manual (Trip Buttons) 2,

1 2

1,2,3,4 6

i 4d b.

Containment Pressure --

High 4

2 3

1,2,3 11 w

3.

CONTAINMENT ISOLATION (CIS)#

l a.

Manual CIS (Trip Buttons) 2 1

2 1,2,3,4 6

3E R b.

Containment Pressure -

9g High 4

2 3

1,2,3 7*

EE 99

-i -4

  1. Containment isolation of non-essential penetrations is also initiated by SIAS (functional units 55 1.a and 1.c).

MO 9

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.- w TABLE 3.3-3 (Continued)

,,,3 hkhk ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 54

(($

MINIMUM TOTAL NO.

CHANNELS CHANNELS APPLICABLE i 'i

- UIUI FUNCTIONAL UNIT 0F CHANNELS TO TRIP OPERABLE MODES ACTION gg 4.

MAIN STEAM LINE ISOLATION a.

Manual (MSIV n, _,

Hand Switches and Feed Head Isolation Hand Switches) 1/ valve 1/ valve 1/ valve 1, 2, 3, 4 6

l b.

Steam Generator 4/ steam 2/ steam 3/ steam 1,2,3(c) 7*

Pressure - Low generator generator generator 5.

CONTAINMENT SUMP

g RECIRCULATION (RAS) a.

Manual RAS (Trip Buttons) 2 1

2 1,2,3,4 6

i n3 b.

Refueling Water Tank - Low 4

2 3

1,2,3 7*

t 4

4

TABLE 3.3-3 (Continued) hh ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTUMENTATION mm

@Q MINIMUM nn CC TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPFRABLE MODES ACTION 6. CONTAINMENT PURGE N EE VALVES IS0Lt. TION qq , 19 - a. Manual (Purge Valve Control Switches) 2/ Penetration 1/ Penetration 2/ Penetration 1, 2, 3, 4 6 b. Containment Radiation - High Area Monitor 4 2 3 6 8 7. LOSS OF POWER w 1 a. 4.16 kv Emergency Bus Undervoltage (Loss of w Voltage) 4/ Bus, 2/ Bus 3/ Bus 1, 2, 3 7* i b. 4.16 kv Emergency Bus 3 Undervoltage (Degraded Voltage) 4/ Bus 2/ Bus 3/ Bus 1,2,3 7* F.F .h

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

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l 99 TABLE 3.3-3 (Continued)_ mm i 3333 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION c, c, E E: l '3l3l MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS _TO TRIP OPERABLE M00ES ACTION c c. 55 4 1 4 8. CVCS ISOLATION m. i a. Hanual (CVCS . 1/ Valve' 1/ Valve 1/ Valve 1, 2, 3, 4 6 Isolation Valve Control Switches) i b. West Penetration 4 2 3 1,2,3,4 7 Room / Letdown Heat ]; Exchanger Room Pressure - High h 1 li g h 1 g i i 4

TABLE 3.3-4 99 GG ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES 99 wH pp ALLOWABLE qq FUNCTIONAL UNIT TRIP SETPOINT VALUES ?? 1. SAFETY INJECTION (SIAS) cc a. Manual (Trip Buttons) Not Applicable Not Applicable 53 b. Containment Pressure - High 1 4.75 psig 1 4.75 psig N".. c. Pressurizer Pressure - Low > 1578 psia > 1578 psia 2 2. CONTAINMENT SPRAY (CSAS) a. Manual (Trip Buttons) Not Applicable Not Applicable b. Containment Pressure -- High 1 4.75 psig 1 4.75 psig 3. CONTAINMENT ISOLATION (CIS) # { a. Manual CIS (Trip Buttons) Not Applicable Not Applicable b. Containment' Pressure - High 1 4.75 psig 1 4.75 psig 4. MAIN STEAM LINE ISOLATION a. Manual (MSIV Hand Switches and Feed Head Isolation gy Hand Switches) Not Applicable Not Applicable y& b. Steam Generator Pressure - Low > 570 psia > 570 psia nn 2z??

  1. Containment isolation of non-essential penetrations is also initiated by SIAS (functional gp units 1.a and 1.c).

0

TABLE 3.3-4 (Continued) %,:o ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP VALUES ALLOWABLE (( FUNCTIONAL UNIT TRIP VALUE VALUES 33 5. CONTAINMENTSUMPRECIRCULATION(RAS) N a. Manual RAS (Trip Buttons) - Not Applicable Not Applicable 33 b. Refueling Water Tank - Low > 24 inches above > 24 inches above tank bottom tank bottom 6. CONTAINMENT PURGE VALVES ISOLATION ## g a. Manual (Purge Valve Control Switches) Not Applicable Not Applicable b. Containment Radiation - High Area Monitor 1 220 mr/hr 1 220 mr/hr y 7. LOSS OF POWER a. 4.16 kv Emergency Bus Under-2450+105 volts with a 2450+105 volts with a voltage (Loss of Voltage) 210.2 second time delay 2l0.2 second time delay- ~ b. 4.16 kv Emergency Bus Under-3628+25 volts with a 3628+25 volts with a voltage (Degraded Voltage) 8+0.4 second time delay 810.4 second time delay FF (( ff Containment purge valve isolation is also initiated by SIAS (functional units 1.a. 1.b, and 1.c). A5 .O O. i R*l l l r

TABLE 3.3-5 (Continued) -ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 6. Steam Generator Pressure-Low a. Main Steam Isolation 5, 6. 9 b. Feedwater Isolation 5, 80 7. Refueling Water Tank-Low a. Containment Sump Recirculation < 80 8. Reactor Trip a. Feedwater Flow Reduction to 5% 5, 20 9. Los of Power a. 4.16 kv Emergency Rus --< 2.2 Undervoltage (Loss of Voltage) b. 4.16 kv Emergency Bus 5, 8. 4 Undervoltage (Degraded Voltage) TABLE NOTATION

  • Diesel generator starting and sequence loading delays included.
    • Diesel generator starting and sequence loading delays not included.

Offsite power available.

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

CALVERT CLIFFS - UNIT 1 Amendment No. 40 CALVEdT CLIFFS - UNIT 2 3/4 3-21 Amendment No. 22

TABLE 4.3-2 hh ENGINEEREDSAFETYFEATUREACTUATION5YSTEMINSTRUMENTAIONSURVEILLANCEREQUIREMENTS MM CHANNEL MODES IN WHICH $g CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE n n C; FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED n5h 1. SAFETY INJECTION (SIAS) '- i a. Manual (Trip Buttons) N.A. N.A. R N.A. E c= b. Containment Pressure - High S R M 1,2,3 c. Pressurizer Pressure - Low S R M 1,2,3 p5 d. Automatic Actuation Logic N.A. N.A. M(1)(3) 1, 2, 3 m] 2. CONTAINMENT SPRAY.(CSAS) a. Manual (Trip Buttons) N.A. N.A. R N.A. b. Containment Pressure -- High S R M 1,2,3 c. Automatic Actuation Logic H.A. N.A. M(1) 1, 2, 3 R 3. CONTAINMENT ISOLATION (CIS) # l a. Manual CIS (Trip Buttons) N.A. N.A. R N.A. b. Containment Pressure - High S R M 1,2,3 c. Automatic Actuation Logic N.A. M.A. M(1)(4) . 1, 2, 3 4. MAIN STEAM LINE ISOLATION (SGIS) a. Hanual SGIS (MSIV Hand EE Switches and Feed Head E8 Isolation Hand Switches) N.A. N.A. R N.A. b. Steam Generator Pressure - Low S R M 1, 2, 3 Automatic Act,uation Logic N.A. N.A. M(1)(5) 1, 2, 3 c. g 5 F

  1. Containment isolation of non-essential penetrations is also initiated by SIAS (functional units w (n 1.a and 1.c).

1 l l

99 1ABLE 4.3-2 (Continued) GG Gg ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 7 9M CHANNEL MODES IN WHICH GG CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED i. C C 55 5. CONTAINMENT SUMP RECIRCULATION (RAS) a. Manual RAS (Trip Buttons) N.A. N.A. R N.A. b. Refueling Water 1,2,3 Tank - Low N.A. R M c. Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3 II l 6. CONTAINt1ENT PURGE VALVES ISOLATION w a. Manual (Purge Valve Control 2 Switches) N.A. N.A. R N.A. b. Containment Radiation - High w k Area Monitor S R M 6 w 7. LOSS OF POWER a. 4.16 kv Emergency Bus Undervoltage (Loss of Voltage) N.A. R M 1,2,3 b. 4.16 kv Emergency Bus gp Undervoltage (Degraded .m {$ Voltage) N.A. R M 1,2,3 m a R5 8. CVCS ISOLATION N.A. R M 1,2,3,4 2 2 West Penetration Room / P P Letdown Heat Exchanger yg Room Pressure - High Mm

    1. Containment purge valve isolatio:: is also initiated by SIAS (funct.ional units 1.a. 1.b and 1.c).

l m 4

TABLE 4.3-2 (Continued) TABLE NOTATION (1) The logic circuits shall be tested nanually at least once per 31 days. (3) SIAS logic circuits A-5, B-5, A-10 and B-10 may be exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown. (4) CIS lo'gic circuits A-5 and B-5 may be exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown. (5) SGIS logic circuits A-1 and B-1 may be exempted from testing during operation; however, these logic circuits shall be tested at least once per 18 months during shutdown. CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3/4 3-24 Amendment No. 3

52 r;52r; TABLE 3.3-10 S259 ' 4 POST-ACCIDENT MONITORING INSTRUMENTATION PP MM gy MINIMUM CHANNELS INSTRUMENT OPERABLE

c. c 23 5 1 1.

Power Range Nuclear Flux 2 n>. 2. Containment Pressure 2 i 3. Wide Range Logarithmic Neutron Flux Monitor 2 4. Reactor Coolant Outlet Temperature 2 l 5. Reactor Coolant Total Flow 2 w 1 6. Pressurizer Pressure 2 3 7. Pressurizer Level 2 8. Steam Generator Pressure 2/ steam generator 9. Steam Generator Level 2/ steam ;;terator l[j[ 10. Feedwater Flow 2 ER l g;g; 11. Auxiliary feedwater Flow Rate 1/ steam generator 4 ff 12. RCS Subcooled Margin Monitor 1 S$El 13. PORV/ Safety Valve Acoustic Flow Monitoring 1/ valve l 14. PORY Solenoid Power Indication 1/ valve

s TABLE 4.3-10 POST-ACCIDENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS nn i ?? 1

  • M i

4Q CHANNEL CHANNEL INSTRUMENT CHECK CALIBRATION e, n CC

q q 1.

Power Range Nuclear Flux M Q mm 2. Containment Pressure M R ~ '@ E i pp 3. Wide Range Logarithmic Neutron Flux Monitor M N.A. 3 m-4, Reactor Coolant Outlet Temperature M R i 5. Reactor Coolant Total Flow M R R ~6. Pressurizer Pressure M R Y 7. Pressurizer Level M R 8. Steam Generator Pressure M R 9. Steam Generator Level M R [ 10. Feedwater Flow M R

11. Auxiliary Feedwater Flow Rate M

R i 22 12. kCS Subcooled Margin Monitor M R i ER 22 13. PORV/ Safety Valve Acoustic Monitor N. A.. R 55 j 14. PORY Solenoid Power Indication N.A. N.A. 2x b

  1. D i

REACTOR COOLANT SYSTEM SAFETY VALVES i LIMITING CONDITION FOR OPERATION 3.4.2.1 The following pressurizer code safety valves shall be OPERABLE: Valve lift Settings ( 1%) RC-200 2500 psia RC-201 2565 psia APPLICABILITY: MODES 1, 2 and 3. ACTION: With one pressurizer code safety valve inoperable, either restore the inoperable valve to OPERABLE status within 15 minutes or be in HOT SHUTOOWN within 12 hours..' 3.4.2.2 At least one of the above pressurizer code safety valves shall be OPERABLE:* APPLICABILIT,Y : MODES 4 and 5. ACTION: With no pressurizer code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE shutdown cooling loop into operation. SURVEILLANCE REQUIREMENTS 4.4.2 No additional Surveillance Requirements other than those required by Specification 4.0.5.

  • Both valves may be removed in MODE 5 provided at least one valve is replaced by a spool piece which allows the pressurizer to relieve directly to the quench tank.

CALVERT CLIFFS - UNIT 1 3/4 4-3 Amendment No.34, 53 CALVERT CLIFFS - UNIT 2 Amendment No. 76, 36

t REACTOR COOLANT SYSTEM RELIEF VALVES LIMITING CONDITION FOR OPEPATION 3.4.3 Two power operated relief valves (PORVs) and their associated block ' 1 valves shall be OPERABLE. 3 APPLICABILITY: MODES 1, 2 and 3. i ACTIO4: With one or more PORV(s) inoperable, within'l hour eithr.- restore a. the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN within the following 30 hours. b. With one or more block valve (s) inoperable, within 1 hour either restore the block valve (s) to OPERABLE status or close the block j valve (s) and remove power from the block valve (s); otherwise, be i . in at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN I within the following 30 hours. SURVEILLANCE REQUIREMENTS

4. 4. 3.1 Each PORY shall be demonstrated OPERABLE:

1 At least once per 31 days by performance of a CHANNEL FUNCTIONAL a. TEST, in accordance with Table 4.2-1, Item 4. b. At least once per 18 months by performance of a CHANNEL CALIBRATION. i 4.4.3.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel, i i o i CALVERT CLIFFS - UNIT 1 3/4 4-4 Amendment No. 53 CALVERT CLIFFS - UNIT 2 Amendment No. 36

REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.4 The pressurizer shall be OPERABLE with a steam bubble and with at least 150 kw of pressurizer heater capacity capable of being supplied by emergency power. The pressurizer level shall be within 5 percent of its programmed value. APPLICABILITY: MODES 1 and 2. ACTION: a. With the pressurizer inoperable due to an inoperable emergency power supply to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 12 hours. b. With the pressurizer otherwise inoperable, be in at least HOT STANDBY with the reactor trip breakers open within 6 hours and in HOT SHUTDOWN within the following 6 hours. SURVEILLANCE REQUIREMENTS 4.4.4 The pressurizer water level shall be determined to be within 5 per-cent of its programmed value at least once per 12 hours. CALVERT CLIFFS - UNIT 1 3/4 4-5 Amendment No. 53 CALVERT CLIFFS - UNIT 2 Amendment No. 36

REACTOR COOLANT SYSTEM STEAM GENERATORS, LIMITING CONDITION FOR OPERATION 3.4.5 Each steam generator shall be OPERABLE. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With one or more steam generators inoperable, restore the inoperable generator (s) to OPERABLE status prior to increasing T,yg above 200*F. SURVEILLANCE REQUIREMENTS 4.4.5.0 Each steam generator shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program and the require-ments of Specification 4.0 5. 4.4.5.1 Steam Generator Sample Selection and Inspection - Each steam generator shall be determined OPERABLE during shutdown by selecting ar.d inspecting at least the minimum number of steam generators specified in Table 4.4-1. 4.4.5.2 Steam Generator Tube Sample Selection and Inspection - The steam generator tube minimum sample size, inspection result classification, and the corresponding action required shall be as specified in Table 4.4-2. The inservice inspection of steam generator tubes shall be performed at the frequencies specified in Specification 4.4.5.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Speci-fication 4.4.5.4. The tubes selected for each inservice inst -tion shall include at least 3% of the total number of tubes in all steam generators; the tubes selected for these inspections shall be selected on a random basis except: a. Where experience in similar plants with similar water chemistry l indicates critical areas to be inspected, thcn at least 50% of the tubes inspected shall be from these critical areas. b. The first inservice inspection (subsequent to the preservice inspection) of each steam generator shall include: 1. All nonplugged tubes that previously had detectable wall penetrations (>20%),and CALVERT CLIFFS - UNIT 1 3/4 4-6 CALVERT CLIFFS - UNIT 2

1 i INSTRUMENTATION BASES 3/4.3.3.6 POST-ACCIDENT INSTRUMENTATION The OPERABILITY of the post-accident instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess these variables following an accident. This capability is consistent with the recommendations of Regulatory Guide 1.97 "Instrumenta-tion for Light-Water-Cooled Nuclear Plants to Assess Plant Conditions During and Following an Accident". December 1975, and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations". 3/4.3.3.7 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires. This capability is required in order to detect and locate fires in their early stages., Prompt detection of fires will reduce the potential for damage to safety related equipment and is an integral element in the overall facility fire protection program. In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABILITY. i l CALVERT CLIFFS - UNIT 1 B 3/4 3-3 Amendment No. 25,53 CALVERT CLIFFS - UNIT 2 Amendment No.11,36

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops and associated reactor coolant pumps in operation, and maintain DNBR above 1.30 during all normal operatjons and anticipated transients. STARTUP and POWER OPERATION may be initiated and may proceed with one or two reactor coolant pumps not in operation after the setpoints for the Power Level-High, Reactor Coolant Flow-Low, Thermal Margin / Low Pressure and Axial Flux Offset trips have been reduced to their specified values. Reducing these trip setpoints ensures that the DNBR will be maintained above 1.30 during three pump opera-tion and that during two pump operation the core void fraction will be limited to ensure parallel channel flow stability within the core and thereby prevent premature DNB. A single reactor coolant loop with its steam generator filled above the low level trip setpoint provides sufficient heat removal capability for core cooling while in MODES 2 and 3; however, single failure considerations require plant cooldown if component repairs and/or corrective actions cannot be made within the allowable out-of-service time. The restrictions on starting a Regctor Coolant Pump during MODES 4 and 5 with one or more RCS cold legs < 275 F are provided to prevent RCS pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into or (2) by restrict-ing starting of the RCPs to wheg the gecondary water temperature of each steam generator is less than 46 F (34 F when measured by a surface contact instrument)'above the coolant temperature in the reactor vessel. 3/4.4.2 SAFETY VALVES l The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psia. Each safety valve is designed to relieve 7.6 x 105 lbs per hour of saturated steam at the valve setpoint. The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating shutdown cooling loop. connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psia. The combined relief capacity of these valves is sufficient to CALVERT CLIFFS - UNIT 1 B 3/4 4-1 Amendment No. #, 53 CALVERT CLIFFS - UNIT 2 Aniendment No. 75, 78, 36 m

REACTOR COOLANT SYSTEM BASES limit the Reactor Coolant System pressure to within its Safety Limit of 2750 psia following a complete loss of turbine generator load while operat-ing at RATED THERMAL POWER and assuming no reactor trip until the first Reactor Protective System trip setpoint (Pressurizer Pressure-High) is reached (i.e., no credit is taken for a direct reactor trip on the loss of turbine) and also assuming no operation of the pressurizer power operateG relief valve or steam dump valves.. Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Vessel Code. 3/4.4.3 RELIEF VALVES The power operated relief valves (PORVs) operate to relieve RCS pressure below the setting of the pressurizer code safety valves. These relief valves have remotely operated block valves to provide a positive shutoff capability should a relief valve become inoperable. The electrical power for both the relief valves and the block valves is capable of being supplied from an emergency power source to ensure the ability to seal this possible RCS leak-age path. 3/4.4.4 PRESSURIZER A steam bubble in the pressurizer with the level as programmed ensures that the RCS is not a hydraulically. solid system and is capable of accommo-dating pressure surges during operation. The programmed level also protects the pressurizer code safety valves ind power operated relief valve against water relief. The power operated relief valves function to relieve RCS l pressure during all design transients. Operation of the power operated relief valve in conjunction with a reactor trip on a Pressurizer--Pressure-High signal, minimizes the undesirable opening of the spring-loaded pressurizer code safety valves,. The requirement.that 150 kw of pressurizer heaters.and their associated controls be capable of being supplied electrical power from an emergency bus provides assurance that these heaters can be energized during a loss of off-site power condition to maintain natural circulation at HOT STANDBY. 3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tubes ensure that the structural integrity of this portion of the RCS will be maintained. The program for inservice inspection of steam generator tubes is based on a modifiution of Regulatory Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to CALVERT CLIFFS - UNIT 1 B 3/4 4-2 Amendment No. M, 53 CALVERT CLIFFS - UNIT 2 Amendment No. M, 36

REACTOR COOLANT SYSTEM BASES maintain surveillance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. In-service inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that correc-tive measures can be taken. i The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. 4 The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the primary coolant system and the secondary coolant system (primary-to-secondary leakage = 1 gallon per minute, total). Cracks having a primary-to-secondary leakage t CALVERT CLIFFS - UNIT 1 B 3/4 4-2a Amendment No. 53 Amendment No. 36 CALVERT, CLIFFS - UNIT 2 pg

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TABLE 6.2-1 MINIMUM SHIFT CREW COMPOSITION # Condition of Unit 2 - No Fuel in Unit 1 LICENSE APPLICABLE MODES CATEGORY 1, 2, 3 & 4 5&6 SOL i 1* OL 2 1 Non-Licensed 2 1 Shift Technical Advisor l## 0 Condition'of Unit 2 - Unit 1 in MODES 1, 2, 3 or 4 APPLICABLE MODES LICENSE CATEGORY 1, 2, 3 & 4 5&6 S0L** 2 2* OL** 3 2 Non-Licensed 3 3 Shift Technical Advisor l## l## i Condition of Unit 2 - Unit 1 in MODES 5 or 6 LICENSE APPLICABLE MODES CATEGORY 1, 2, 3 & 4 5&6 50L** 2 1* OL** 2 2 Non-Licensed 3 '3 Shift Technical Advisor l## 0 CALVERT CLIFFS - UNIT 1 Amendment No. 53 CALVERT CLIFFS - UNIT 2-6-4 Amendment No. 36

TABLE 6.2-1 (Continued)

  • Does not include the licensed Senior-Reactor Operator or Senior Reactor Operator Limited to Fuel Handling, supervising CORE ALTERATIONS during fuel reloading.
    • Assumes each individual is licensed on each unit.
  1. Shift crew composition may be less than the minimum requirements for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2.1.
    1. With one unit in MODE 5 or 6, and the other unit in MODE 1, 2, 3 or 4,,

the 50L holder other than the Shift Supervisor may serve as STA. With one unit defueled and the other unit in MODE 1, 2, 3 or 4, the STA must be an S0L holder in addition to the one 50L required. With both units in MODE 1, 2, 3 or 4, the STA must be an S0L holder in addition to the two SOL's required. 3 ( ( l l CALVERT CLIFFS - UNIT 1 Amendment No. 53 CALVERT CLIFFS - UNIT 2 6-5 Amendment No. 36 I i .~.

ADMINISTRATIVE CONTROLS 6.3 FACILITY STAFF QUALIFICATIONS 6.3.1 Each member of the facility staff shall meet or exceed the minimum-qualifications of ANSI N18.1-1971 for comparable positions, except for the Radiation Safety Engineer who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, and (2) the Shift Technical Advisor who shall have a Bachelor's Degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents. 2 6.4 TRAINING 6.4.1 ' A retraining and replacement training program for the facility staff shall be maintained under the direction of the General Supervisor - Training and Technical Services and shall meet or exceed the requirements and recom-mendations of Section 5.5 of ANSI N18.1-1971 and Appendix "A" of 10 CFR Part 55. 1 6.4.2 A training program for the Fire Brigade shall be maintained under the direction of the General Supervisor - Training and Technical Services and shall meet or exceed the requirements of Section 27 of the NFPA Code-1975. 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS AND SAFETY REVIEW COMMITTEE (POSRC) FUNCTION 6.5.1.1 The POSRC shall function to advise the Plant Superintendent on all matters related to nuclear safety. COMPOSITION 6.5.1.2 The POSRC shall be composed of the: Chairman: Plant Superintendent Member: General SQpervisor'- Operations. Member: General Supervisor; ' Electrical and Controls Member: General Supervisor - Chemistry Member: Principal Engineer - Plant Engineering Nuclear Member: General Foreman - Maintenance and Modifications Member: Supervisor - Nuclear Fuel Management Member: General Supervisor - Radiation Safety l Member: General Supervisor - Training and Technical Services l ALTERNATES 6.5.1.3 All alternate members shall be appointed in writing by the POSRC g Chairman to serve on a temporary basis; however, no more than two alternates shall participate as voting members in POSRC activities at any one tina. CALVERT CLIFFS - UNIT 1 Amendment No. 26, O, 53 CALVERT CLIFFS - UNIT 2 6-6 Amendment No. JJ, 26, 36 l ..~._.,..._...,,,_.%

ADMINISTRATIVE CONTROLS b. A high raqiation. area in which the intensity of radiation is greater than 1000 mrem /hr shall be subject to the provisions of o.12.1.a. above, and in addition locked barricades shall be provided to prevent unauthorized entry into such areas and the keys shall be maintained under the administrative control of the Shift Supervisor on duty. 6.13 ENVIRONMENTAL QUALIFICATION 6.13.1 By no later than June 30, 1982 all safety-related electrical equipment in the facility shall be qualified in accordance with the provisions of: Division o.f Operating Reactors " Guidelines for Evaluating Environmental Qualification of Class IE Electrical Equipment in Operating Reactors" (DOR Guidelines); or NUREG-0588 " Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" December 1979. Copies of these documents are attached to Order for Modification of Licenses DPR-53 and DPR-69 dated October 24,1980. 6.13.2 By no later than December 1,1980, complete and auditable records must be available and maintained at a central location which describe the environmental qualification method used for all safety-related electrical equipment in sufficient detail to document the degree of compliance with the D0R Guidelines or NUREG-0588. Thereafter, such records should be updated and maintained current as equipment is replaced, further tested, or otherwise further qualified. ~ 6.14 SYSTEM INTEGRITY The licensee shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as low as practical levels. This program shall include the following: 1. Provisions establishing preventive maintenance and periodic visual inspection requirements, and 2. Leak test requirements for each system at a frequency not to exceed refueling cycle intervals. l 2 CALVERT CLIFFS - UNIT.1 6-21 9/ddf ditd4 9ttdEdf AA4 7980 Amendment No. 53 CALVERT CLIFFS - UNIT 2 prppypppppppppppyppjJppp j Amendment No. 36

ADMINISTRATIVE CONTROLS 6.15 IODINE MONITORING The licensee shall implement a program which will-ensure the capability to l accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following: 1. Training of personnel, 2. Procedures for monitoring, and 3. Provisions for maintenance of sampling and analysis equipment. l l CALVERT CLIFFS - UNIT 1 Amendment No. 53 CALVERT CLIFFS - UNIT 2 6-22 Amendment No. 36 . ~. ...}}

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