ML060690055

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Units, 1 and 2, Tech Spec Pages for Amendments 278 and 255 Steam Generator Tube Integrity
ML060690055
Person / Time
Site: Calvert Cliffs  
Issue date: 03/09/2006
From:
Plant Licensing Branch III-2
To:
References
TAC MC8067, TAC MC8068
Download: ML060690055 (25)
v • d • e


Text

TABLE OF CONTENTS 3.3.4 Engineered Safety Features Actuation System (ESFAS) Instrumentation.........

3.3.4-1 3.3.5 Engineered Safety Features Actuation System (ESFAS) Logic and Manual Actuation.......

3.3.5-1 3.3.6 Diesel Generator (DG)-Loss of Voltage Start (LOVS).3.3.6-1 3.3.7 Containment Radiation Signal (CRS).3.3.7-1 3.3.8 Control Room Recirculation Signal (CRRS).3.3.8-1 3.3.9 Chemical and Volume Control System (CVCS)

Isolation Signal.3.3.9-1 3.3.10 Post-Accident Monitoring (PAM) Instrumentation.

3.3.10-1 3.3.11 Remote Shutdown Instrumentation.3.3.11-1 3.3.12 Wide Range Logarithmic Neutron Flux Monitor Channels.3.3.12-1 3.4 REACTOR COOLANT SYSTEM (RCS).3.4.1-1 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits......

3.4.1-1 3.4.2 RCS Minimum Temperature for Criticality............

3.4.2-1 3.4.3 RCS Pressure and Temperature (P/T) Limits..........

3.4.3-1 3.4.4 RCS Loops -

MODES 1 and 2........

3.4.4-1 3.4.5 RCS Loops -

MODE 3.............

3.4.5-1 3.4.6 RCS Loops -

MODE 4.............

................... 3.4.6-1 3.4.7 RCS Loops -

MODE 5, Loops Filled......

3.4.7-1 3.4.8 RCS Loops -

MODE 5, Loops Not Filled..............

3.4.8-1 3.4.9 Pressurizer........................................ 3.4.9-1 3.4.10 Pressurizer Safety Valves..........................

3.4.10-1 3.4.11 Pressurizer Power-Operated Relief Valves (PORVs)... 3.4.11-1 3.4.12 Low Temperature Overpressure Protection (LTOP)

System......................................... 3.4.12-1 3.4.13 RCS Operational LEAKAGE............................

3.4.13-1 3.4.14 RCS Leakage Detection Instrumentation..............

3.4.14-1 3.4.15 RCS Specific Activity.............................. 3.4.15-1 3.4.16 Special Test Exception (STE) RCS Loops - MODE 2.... 3.4.16-1 3.4.17 Special Test Exception (STE) RCS Loops - MODES 4 and 5...........................................

3.4.17-1 3.4.18 Steam Generator (SG) Tube Integrity................

3.4.18-1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS).3.5.1-1 3.5.1 Safety Injection Tanks (SITs).3.5.1-1 3.5.2 ECCS-Operating.3.5.2-1 3.5.3 ECCS-Shutdown.3.5.3-1 CALVERT CLIFFS - UNIT 1 ii Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

TABLE OF CONTENTS 3.5.4 Refueling Water Tank (RWT).........................

3.5.4-1 3.5.5 Trisodium Phosphate (TSP)..........................

3.5.5-1 3.6 CONTAINMENT SYSTEMS....................................

3.6.1-1 3.6.1 Containment........................................

3.6.1-1 3.6.2 Containment Air Locks..............................

3.6.2-1 3.6.3 Containment Isolation Valves.......................

3.6.3-1 3.6.4 Containment Pressure...............................

3.6.4-1 3.6.5 Containment Air Temperature........................

3.6.5-1 3.6.6 Containment Spray and Cooling Systems..............

3.6.6-1 3.6.7 Deleted 3.6.8 Iodine Removal System (IRS)........................

3.6.8-1 3.7 PLANT SYSTEMS.......................................... 3.7.1-1 3.7.1 Main Steam Safety Valves (MSSVs)...................

3.7.1-1 3.7.2 Main Steam Isolation Valves (MSIVs)................

3.7.2-1 3.7.3 Auxiliary Feedwater (AFW) System...................

3.7.3-1 3.7.4 Condensate Storage Tank (CST)......................

3.7.4-1 3.7.5 Component Cooling (CC) System...................... 3.7.5-1 3.7.6 Service Water (SRW) System.........................

3.7.6-1 3.7.7 Saltwater (SW) System..............................

3.7.7-1 3.7.8 Control Room Emergency Ventilation System (CREVS).. 3.7.8-1 3.7.9 Control Room Emergency Temperature System (CRETS).. 3.7.9-1 3.7.10 Emergency Core Cooling System (ECCS) Pump Room Exhaust Filtration System (PREFS).3.7.10-1 3.7.11 Spent Fuel Pool Exhaust Ventilation System (SFPEVS).3.7.11-1 3.7.12 Penetration Room Exhaust Ventilation System (PREVS).

....................................... 3.7.12-1 3.7.13 Spent Fuel Pool (SFP) Water Level..................

3.7.13-1 3.7.14 Secondary Specific Activity........................

3.7.14-1 3.7.15 Main Feedwater Isolation Valves (MFIVs)............

3.7.15-1 3.7.16 Spent Fuel Pool (SFP) Boron Concentration..........

3.7.16-1 3.7.17 Spent Fuel Pool (SFP) Storage......................

3.7.17-1 3.8 ELECTRICAL POWER SYSTEMS...............................

3.8.1-1 3.8.1 AC Sources-Operating.........

3.8.1-1 3.8.2 AC Sources-Shutdown.........

3.8.2-1 3.8.3 Diesel Fuel Oil.............

3.8.3-1 3.8.4 DC Sources-Operating.........

3.8.4-1 3.8.5 DC Sources-Shutdown.........

3.8.5-1 3.8.6 Battery Cell Parameters............................

3.8.6-1 CALVERT CLIFFS - UNIT 1 iii Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

TABLE OF CONTENTS 3.8.7 Inverters-Operating.........

3.8.7-1 3.8.8 Inverters-Shutdown.........

3.8.8-1 3.8.9 Distribution Systems-Operating......

3.8.9-1 3.8.10 Distribution Systems-Shutdown......

3.8.10-1 3.9 REFUELING OPERATIONS...................................

3.9.1-1 3.9.1 Boron Concentration................................

3.9.1-1 3.9.2 Nuclear Instrumentation............................

3.9.2-1 3.9.3 Containment Penetrations...........................

3.9.3-1 3.9.4 Shutdown Cooling (SDC) and Coolant Circulation-High Water Level....

3.9.4-1 3.9.5 Shutdown Cooling (SDC) and Coolant Circulation-Low Water Level..........

3.9.5-1 3.9.6 Refueling Pool Water Level.........................

3.9.6-1 4.0 DESIGN FEATURES............................................

4.0-1 4.1 Site Location..........................................

4.0-1 4.2 Reactor Core...........................................

4.0-1 4.3 Fuel Storage........................................... 4.0-2 5.0 ADMINISTRATIVE CONTROLS....................................

5.1-1 5.1 Responsibility......................................... 5.1-1 5.2 Organization...........................................

5.2-1 5.2.1 Onsite and Offsite Organizations...................

5.2-1 5.2.2 Unit Staff.........................................

5.2-2 5.3 Unit Staff Qualifications..............................

5.3-1 5.4 Procedures.............................................

5.4-1 5.5 Programs and Manuals...................................

5.5-1 5.5.1 Offsite Dose Calculation Manual....................

5.5-1 5.5.2 Primary Coolant Sources Outside Containment........

5.5-2 5.5.3 Not Used...........................................

5.5-2 5.5.4 Radioactive Effluent Controls Program...............

5.5-2 5.5.5 Component Cyclic or Transient Limit................

5.5-5 5.5.6 Concrete Containment Tendon Surveillance program... 5.5-6 5.5.7 Reactor Coolant Pump Flywheel Inspection Program

... 5.5-6 5.5.8 Inservice Testing Program..........................

5.5-6 5.5.9 Steam Generator (SG) Program.......................

5.5-7 5.5.10 Secondary Water Chemistry Program..................

5.5-10 5.5.11 Ventilation Filter Testing Program.................

5.5-11 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program..........................

5.5-13 5.5.13 Diesel Fuel Oil Testing Program....................

5.5-14 CALVERT CLIFFS - UNIT 1 iv Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No.255

TABLE OF CONTENTS 5.5.14 5.5.15 5.5.16 5.6 5.6.1 5.6.2 5.6.3 5.6.4 5.6.5 5.6.6 5.6.7 5.6.8 5.6.9 Technical Specifications Bases Control Program.....

Safety Function Determination Program (SFDP).......

Containment Leakage Rate Testing Program...........

Reporting Requirements.................................

Deleted............................................

Annual Radiological Environmental Operating report Radioactive Effluent Release Report................

Deleted............................................

CORE OPERATING LIMITS REPORT (COLR)................

Not Used...........................................

Post-Accident Monitoring Report....................

Tendon Surveillance Report.........................

Steam Generator Tube Inspection Report.............

5.5-15 5.5-15 5.5-17 5.6-1 5.6-1 5.6-1 5.6-2 5.6-2 5.6-2 5.6-8 5.6-9 5.6-9 5.6-9 CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 v

Amendment No. 278 Amendment No. 255

Definitions 1.1 1.1 Definitions LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE
1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal leakoff), that is captured and conducted to collection systems or a sump or collecting tank;
2. LEAKAGE into the containment atmosphere from sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE; or
3. Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE).
b. Unidentified LEAKAGE All LEAKAGE (except RCP seal leakoff) that is not identified LEAKAGE;
c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) l through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

MODE A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolts specified in Table 1.1-1 with fuel in the reactor vessel.

CALVERT CLIFFS - UNIT 1 1.1-4 Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

Definitions 1.1 1.1 Definitions OPERABLE-OPERABILITY PHYSICS TESTS A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation.

These tests are:

a. Described in Chapter 13, Initial Tests and Operation of the Updated Final Safety Analysis Report;
b. Authorized under the provisions of 10 CFR 50.59; or
c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER (RTP)

REACTOR PROTECTIVE SYSTEM (RPS) RESPONSE TIME CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 RTP shall be a total reactor core heat transfer rate to the reactor coolant of 2700 MWt.

The RPS RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its RPS trip setpoint at the channel sensor until electrical power to the CEAs drive mechanism is interrupted. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

In lieu of measurement, response time may be verified for selected components provided that the components and methodology for 1.1-5 Amendment No. 278 Amendment No. 255

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a. No pressure boundary LEAKAGE;
b. 1 gpm unidentified LEAKAGE;
c.

10 gpm identified LEAKAGE; and I

d.

100 gallons per day primary to any one steam generator (SG).

secondary LEAKAGE through I

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS operational A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LEAKAGE not within within limits.

limits for reasons other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3.4.13-1 Amendment No. 278 Amendment No. 255

RCS Operational LEAKAGE 3.4.13 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1

NOTES ------------------

1. Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.
2. Not applicable to primary to secondary LEAKAGE.

Verify RCS Operational LEAKAGE is within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> limits by performance of RCS water inventory balance.

CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3.4.13-2 Amendment No. 278 Amendment No. 255

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.4.13.2 -------------

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

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation.

Verify primary to secondary LEAKAGE is 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

< 100 gallons per day through any one SG.

CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3.4.13-3 Amendment No. 278 Amendment No. 255

SG Tube Integrity 3.4.18 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.18 Steam Generator (SG) Tube Integrity LCO 3.4.18 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

APPLICABILITY:

MODES 1, 2, 3, and 4.

ACTIONS

NOTES------------------------------------

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more SG tubes A.1 Verify tube integrity 7 days satisfying the tube of the affected repair criteria and tube(s) is maintained not plugged in until the next accordance with the refueling outage or SG Steam Generator tube inspection.

Program.

AND A.2 Plug the affected Prior to tube(s) in accordance entering MODE 4 with the Steam following the Generator Program.

next refueling outage or SG tube inspection CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3.4. 18-1 Amendment No. 278 Amendment No. 255

SG Tube Integrity 3.4.18 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3.

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> associated Completion Time of Condition A AND not met.

B.2 Be in MODE 5.

36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> OR SG tube integrity not maintained.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.18.1 Verify SG tube integrity in accordance with In accordance the Steam Generator Program.

with the Steam Generator Program SR 3.4.18.2 Verify that each inspected SG tube that Prior to satisfies the tube repair criteria is entering MODE 4 plugged in accordance with the Steam following a SG Generator Program.

tube inspection Amendment No. 278 Amendment No. 255 CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 3.4.18-2

Programs and Manuals 5.5 5.5 Programs and Manuals

a. Testing frequencies specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as follows:

ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice testing activities Required Frequencies for performing inservice testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Biennially or every 2 years At At least least once per once per 7

31 days days At least once per 92 days At At At least least least once once once per per per 184 276 366 days days days At least once per 731 days

b. The provisions of SR required Frequencies activities; 3.0.2 are applicable to the above for performing inservice testing
c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and
d. Nothing in the ASME Boiler and Pressure Vessel Code shall construed to supersede the requirements of any Technical Specification.

be 5.5.9 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 5.5-7 Amendment No. 278 Amendment No. 255

Programs and Manuals 5.5 5.5 Programs and Manuals

a. Provisions for condition monitoring assessments. Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage.

The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b. Performance criteria for SG tube integrity.

Steam generator tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady-state full power operation primary to secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

CALVERT CLIFFS - UNIT 1 5.5-8 Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

Programs and Manuals 5.5 5.5 Programs and Manuals

2. Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 100 gpd per SG.
3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c. Provisions for SG tube repair criteria.

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

d. Provisions for SG tube inspections.

Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial, and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria.

The tube-to-tubesheet weld is not part of the tube.

In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.
2. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power CALVERT CLIFFS - UNIT 1 5.5-9 Amendment No.278 CALVERT CLIFFS - UNIT 2 Amendment No.255

Programs and Manuals 5.5 5.5 Programs and Manuals months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period.

No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected.

3. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less).

If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary LEAKAGE.

5.5.10 Secondary Water Chemistry Program This program provides controls for monitoring secondary water chemistry to inhibit steam generator tube degradation and low pressure turbine disc stress corrosion cracking.

The program shall include:

a.

Identification of a sampling schedule for the critical variables and control points for these variables;

b. Identification of the procedures used to measure the values of the critical variables;
c. Identification of process sampling points which shall include monitoring the discharge of the condensate pumps for evidence of condenser in leakage; CALVERT CLIFFS - UNIT 1 5.5-10 Amendment No.

278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

Programs and Manuals 5.5 5.5 Programs and Manuals

d. Procedures for the recording and management of data;
e. Procedures defining corrective actions for all 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, which are required to initiate corrective action.

5.5.11 Ventilation Filter Testing Program A program shall be established to implement the following required testing of engineered safety feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.11.a and 5.5.11.b shall be performed once per 18 months for ventilation systems other than the Iodine Removal System (IRS) and 24 months for the IRS; after each complete or partial replacement of the high efficiency particulate air (HEPA) filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter or charcoal adsorber housing; and following painting, fire, or chemical release in any ventilation zone communicating with the system.

Tests described in Specification 5.5.11.c shall be performed once per 18 months for ventilation systems other than the IRS and 24 months for the IRS; after 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of system operation; after any structural maintenance on the HEPA filter or charcoal adsorber housing; and following painting, fire, or chemical release in any ventilation zone communicating with the system.

Tests described in Specification 5.5.11.d shall be performed once per 18 months for ventilation systems other than the IRS and 24 months for the IRS.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Ventilation Filter Testing Program test frequencies.

CALVERT CLIFFS - UNIT 1 5.5-11 Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

Programs and Manuals 5.5 5.5 Programs and Manuals

a. Demonstrate for each of the ESF systems that an inplace test of the HEPA filters shows a penetration and system bypass

< 1.0% when tested in accordance with Regulatory Positions C.5.a and C.5.c of Regulatory Guide 1.52, Revision 2, and ANSI N510-1975, at the system flowrate specified as follows +/- 10%:

ESF Ventilation System Flowrate Control Room Emergency Ventilation System 2,000 cfm (CREVS)

Emergency Core Cooling System (ECCS) Pump 3,000 cfm Room Exhaust Filtration System (PREFS)

Penetration Room Exhaust Ventilation 2,000 cfm System (PREYS)

Spent Fuel Pool Exhaust Ventilation 32,000 cfm System (SFPEVS)

IRS 20,000 cfm

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a penetration and system bypass < 1.0% when tested in accordance with Regulatory Positions C.5.a and C.5.d of Regulatory Guide 1.52, Revision 2, and ANSI N510-1975, at the system flowrate specified as follows +/- 10%:

ESF Ventilation System Flowrate CREVS 2,000 cfm ECCS PREFS 3,000 cfm PREVS 2,000 cfm SFP Ventilation System 32,000 cfm IRS 20,000 cfm

c. Demonstrate for each of the ESF systems within 31 days after removal that a laboratory test of a sample of the charcoal adsorber, when obtained as described in Regulatory Position C.6.b of Regulatory Guide 1.52, Revision 2, shows the methyl iodide penetration less than the value specified below when tested in accordance with ASTM D3803-1989 at a CALVERT CLIFFS - UNIT 1 5.5-12 Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

Programs and Manuals 5.5 5.5 Programs and Manuals temperature of 30'C and greater than or equal to the relative humidity specified as follows:

ESF Ventilation System Penetrations RH CREVS 5%

70%

ECCS PREFS 50%

95%

PREVS 35%

95%

SFP Ventilation System 15%

95%

IRS 35%

95%

d. For each of the ESF systems, demonstrate the pressure drop across the combined HEPA filters, the prefilters, and the charcoal adsorbers is less than the value specified below when tested in accordance with Regulatory Guide 1.52, Revision 2, and ANSI N510-1975 at the system flowrate specified as follows +/- 10%:

ESF Ventilation System Delta P Flowrate CREVS 4 inwg 2,000 cfm ECCS PREFS 4 inwg 3,000 cfm PREVS 6 inwg 2,000 cfm SFP Ventilation System 4 inwg 32,000 cfm IRS 6 inwg 20,000 cfm 5.5.12 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides control for potentially explosive gas mixtures contained in the Waste Gas Holdup System and the quantity of radioactivity contained in gas storage tanks.

The gaseous radioactivity quantities shall be determined following the methodology in the ODCM.

The program shall include:

a. The limits for concentrations of oxygen in the Waste Gas Holdup System and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the CALVERT CLIFFS - UNIT 1 5.5-13 Amendment No.278 CALVERT CLIFFS - UNIT 2 Amendment No.255

Programs and Manuals 5.5 5.5 Programs and Manuals system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion); and

b. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than or equal to 58,500 curies noble gases (considered as Xe-133).

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance Frequencies.

5.5.13 Diesel Fuel Oil Testing Program A Diesel Fuel Oil Testing Program to implement required testing of both new fuel oil and stored fuel oil shall be established.

The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with ASTM Standards.

The purpose of the program is to establish the following:

a. Acceptability of new fuel oil for use prior to addition to storage tanks by determining that the fuel oil has:
1.

An American Petroleum Institute gravity or an absolute specific gravity within limits,

2.

A flash point and kinematic viscosity within limits for ASTM 2D fuel oil, and

3. Water and sediment < 0.05%.
b. Within 31 days following addition of new fuel oil to the storage tanks, verify that the properties of the new fuel oil, other than those addressed in a., above, are within limits for ASTM 2D fuel oil; and
c. Total particulate concentration of the fuel oil, when determined by gravimetric analysis based on ASTM D2276-1989, is < 10 mg/l when tested every 92 days.

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Programs and Manuals 5.5 5.5 Programs and Manuals

d. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Diesel Fuel Oil Testing Frequencies.

5.5.14 Technical Specifications Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

a. Changes to the Bases of the Technical Specifications shall be made under appropriate administrative controls and reviews.
b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
1. A change in the Technical Specifications incorporated in the license; or
2. A change to the UFSAR or Bases that requires NRC approval pursuant to 10 CFR 50.59.
c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
d. Proposed changes that meet the criteria of Specification 5.5.14b above shall be reviewed and approved by the NRC prior to implementation.

Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).

5.5.15 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken.

Upon entry into Limiting Condition for Operation (LCO) 3.0.6, -an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate limitations and remedial or compensatory actions may be identified to be taken as a result of the support system inoperability and CALVERT CLIFFS - UNIT 1 5.5-15 Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

Programs and Manuals 5.5 5.5 Programs and Manuals corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.

The SFDP shall contain the following:

a. Provisions for cross-train checks to ensure a loss of the capability to perform the safety function assumed in the accident analysis does not go undetected;
b. Provisions for ensuring the plant is maintained in a safe condition if a loss of function condition exists;
c. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and
d. Other appropriate limitations and remedial or compensatory actions.

A loss of safety function exists when, assuming no concurrent single failure, a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable, and:

a. A required system redundant to system(s) supported by the inoperable support system is also inoperable; or
b. A required system redundant to system(s) in turn supported by the inoperable supported system is also inoperable; or
c. A required system redundant to support system(s) for the supported systems (a) and (b) above is also inoperable.

The SFDP identifies where a loss of safety function exists.

If a loss of safety function is determined to exist by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety function exists are required to be entered.

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Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.16 Containment Leakage Rate Testing Program A program shall be established to implement the leakage testing of the containment as required by 10 CFR 50.54(o) and 10 CFR Part 50, Appendix J, Option B. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, including errata, as modified by the following exceptions:

a. Nuclear Energy Institute (NEI) 94 1995, Section 9.2.3:

The first Unit 1 Type A test performed after the June 15, 1992 Type A test shall be performed no later than June 14, 2007.

b. Unit 1 is excepted from post-modification integrated leakage rate testing requirements associated with steam generator replacement.
c. Unit 2 is excepted from post-modification integrated leakage rate testing requirements associated with steam generator replacement.

The peak calculated containment internal pressure for the design basis loss-of-coolant accident, Pa, is 49.4 psig.

The containment design pressure is 50 psig.

The maximum allowable containment leakage rate, Las shall be 0.20 percent of containment air weight per day at P^a Leakage rate acceptance criteria are:

a. Containment leakage rate acceptance criterion is S 1.0 La.

During the first unit startup following testing, in accordance with this program, the leakage rate acceptance criterion are

  • 0.60 La for Types B and C tests and < 0.75 La for Type A tests.

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Programs and Manuals 5.5 5.5 Programs and Manuals

b. Air lock testing acceptance criteria are:
1. Overall air lock leakage rate is < 0.05 La when tested at 2 Pa.
2. For each door, leakage rate is < 0.0002 La when pressurized to 2 15 psig.

The provisions of SR 3.0.2 do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program.

The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

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Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.7 Post-Accident Monitoring Report When a report is required by Condition B or F of LCO 3.3.10, "Post Accident Monitoring Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.8 Tendon Surveillance Report Any abnormal degradation of the containment structure detected during the tests required by the Pre-Stressed Concrete Containment Tendon Surveillance Program shall be reported to the NRC within 30 days.

The report shall include a description of the tendon condition, the condition of the concrete (especially at tendon anchorages), the inspection procedures, the tolerances on cracking, and the corrective action taken.

5.6.9 Steam Generator Tube Inspection Report A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG)

Program.

The report shall include:

a. The scope of inspections performed on each SG,
b. Active degradation mechanisms found,
c. Nondestructive examination techniques utilized for each degradation mechanism,
d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
e. Number of tubes plugged during the inspection outage for each active degradation mechanism, CALVERT CLIFFS - UNIT 1 5.6-9 Amendment No. 278 CALVERT CLIFFS - UNIT 2 Amendment No. 255

Reporting Requirements 5.6 5.6 Reporting Requirements

f. Total number and percentage of tubes plugged to date,
g. The results of condition monitoring, including the results of tube pulls and in-situ testing,
h. The effective plugging percentage for all plugging in each SG.

CALVERT CLIFFS - UNIT 1 CALVERT CLIFFS - UNIT 2 5.6-10 Amendment Nc. 278 Amendment No-255

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