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D.C. Cook, Technical Specifications, Steam Generator Integrity
	ML070810133
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Issue date:	03/14/2007
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UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Page 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation ................................................................... 3.3.1-1 Table 3.3.1-1, Reactor Trip System Instrumentation .................................................... 3.3.1-11 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation ........................ 3.3.2-1 Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation ............. 3.3.2-7 3.3.3 Post Accident Monitoring (PAM) Instrumentation ........................................................... 3.3.3-1 Table 3.3.3-1, Post Accident Monitoring Instrumentation ............................................. 3.3.3-4 3.3.4 Remote Shutdown Monitoring Instrumentation ............................ 3.3.4-1 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ................................ 3.3.5-1 3.3.6 Containment Purge Supply and Exhaust System Isolation Instrumentation ......... 3.3.6-'1 Table 3.3.6-1, Containment Purge Supply and Exhaust System Isolation Instrum entation ......................................................................................................... 3 .3 .6-4 3.3.7 Control Room Emergency Ventilation (CREV) System Actuation Instrumentation .......... 3.3.7-1 Table 3.3.7-1, CREV System Actuation Instrumentation .............................................. 3.3.7-3 3.3.8 Boron Dilution Monitoring Instrumentation (BDMI) .......................................................... 3.3.8-1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB)

Lim its .................................................. ........................................................................ 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 ............... I.............................................. 3.4.3-1 Figure 3.4.3-1, Reactor Coolant System Pressure versus Temperature Limits -

Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 E FP Y ) .......................................................................................................... 3 .4 .3-3 Figure 3.4.3-2, Reactor Coolant System Pressure versus Temperature Limits -

Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY) ........ 3.4.3-4 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 Pressure Isolation Valve (PIV) Leakage ................................................................. 3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation ........................................................................ 3.4.15-1 3.4.16 R C S Specific Activity ...................................................................................................... 3.4.16-1 Figure 3.4.16-1, Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER ........................................................ 3.4.16-3 3.4.17 Steam Generator (SG) Tube Integrity ............................................................................. 3.4.17-1 Cook Nuclear Plant Unit 1 Page 2 of 5 Amendment No. 287, 298

UNIT 1 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Page 5.0 ADMINISTRATIVE CONTROLS 5 .1 R e spo nsibility .................................................................................................................... 5 .1-1 5.2 Organization .................................................................................................................... 5.2-1 5.2.1 Onsite and Offsite Organizations ................................................................................... 5.2-1 5 2 .2 U nit S taff ........................................................................................................................ 5 .2-1 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 (ODCM) ..................................................................... 5.5-1 5.5.2 Leakage Monitoring Program ........................................................................................ 5.5-2 5.5.3 Radioactive Effluent Controls Program ............ ...................... ................................. 5.5-2 5.5.4 Component Cyclic or Transient Limits ........................ ................................................... 5.5-3 5.5.5 Reactor Coolant Pump Flywheel Inspection Program ........................ 5.5-4 5.5.6 Inservice Testing Program .............................................................................................. 5.5-4 5.5.7 Steam Generator (SG) Program ........................................... I......................................... 5.5-5 5.5.8 Secondary W ater Chemistry Program ........................................................................... 5.5-7 5.5.97 Ventilation Filter Testing Program (VFTP) ................................... ............................... 5.5-7 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program .................. 5.5-10 5.5.11 Diesel Fuel Oil Testing Program ...................................... 5.5-10 5.5.12 Technical Specifications (TS) Bases Control Program ..... .......................I 5.5-11 5.5.13 Safety Function Determination Program (SFDP) .......................................................... 5.5-12 5.5.14 Containment Leakage Rate Testing Program..............,........................... 5.5-13 5.5.15 Battery Monitoring and Maintenance Program, .................... ........................................5.5-14 5.6 Reporting Requirements ............................................. 5.6-1 5.6.1 Deleted......................................................................................................... ......... 5.6-1 5.6.2 Annual Radiological Environmental Operating Report .................................................... 5.6-1 5.6.3 Radioactive Effluent Release Report .............................................................................. 5.6-2 5 .6 .4 D eleted .......................................................................................................................... 5 .6-22 5.6.5 CORE OPERATING LIMITS REPORT (COLR) .............................................................. 5.6-2 5.6.6 Post Accident Monitoring Report .................................................................................... 5.6-4 5.6.7 Steam Generator Tube Inspection Report ...................................................................... 5.6-4 5.7 High Radiation Area ....................................................................... ................................... 5.7-1 Cook Nuclear Plant Unit 1 Page 5 of 5 Amendment No. 2-87, 298

Definitions 1.1 1.1 Definitions ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its actuation setpoint TIME at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. 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 verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or 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 water injection or leakoff) that is not identified LEAKAGE; and

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

Cook Nuclear Plant Unit 1 1.1-3 Amendment No. 287, 298

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. 0.8 gpm unidentified LEAKAGE;

c. 10 gpm identified LEAKAGE; and

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Unidentified LEAKAGE A.1 Verify source of unidentified 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

> 0.8 gpm. LEAKAGE is not the pressurizer surge line.

OR A.2 Reduce unidentified 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months LEAKAGE to within limit.

B. Unidentified LEAKAGE B.1 Reduce unidentified 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

> 1.0 gpm. LEAKAGE to < 1.0 gpm.

Cook Nuclear Plant Unit 1 3.4.13-1 Amendment No. 287, 298

RCS Operational LEAKAGE 3.4.13 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Identified LEAKAGE not C.1 Reduce LEAKAGE to within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months within limits, limits.

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, AND or C not met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE.

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

Cook Nuclear Plant Unit 1 3.4.13-2 Amendment No. 24W, 298

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS (continued)

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

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

Verify primary to secondary LEAKAGE is - 150 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gallons per day through any one SG.

Cook Nuclear Plant Unit 1 3.4.13-3 Amendment No. 28-7, 298

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 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 NOTE= -----------------------------------------------------------

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

Program.

AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program. next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR SG tube integrity not maintained.

Cook Nuclear Plant Unit 1 3.4.17-1 AmendmentNo. 298

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program. a SG tube inspection Cook Nuclear Plant Unit I 3.4.17 -2 AmendmentNo- 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 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:

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. SG 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.

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 1 gpm for all SGs.

AmendmentNo.28-A 298 5.5-5 Cook Nuclear Plant Cook Nuclear Unit 1 Plant Unit 1 5.5-5 . Amendment No.2-97, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued)

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 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.

Amendment No. 2~8-7~ 298 5.5-6 Nuclear Plant Cook Nuclear Unit 1 Plant Unit 1 5.5-6 Amendment No. 243-7, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Secondary Water Chemistry Pro-gram This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. 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;

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 is required to initiate corrective action.

5.5.9 Ventilation Filter Testing Program (VFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.9.a and 5.5.9.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.

Tests described in Specification 5.5.9.c shall be performed once per 24 months; after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of adsorber operation; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the charcoal adsorber capability.

Tests described in Specification 5.5.9.d shall be performed once per 24 months.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.

Amendment No. 28-A 298 5.5-7 Nuclear Plant Cook Nuclear Unit 1 Plant Unit 1 5.5-7 Amendment No. 297, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testingq Proqram (VFTP) (continued)

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a removal efficiency of > 99%

of the dioctyl phthalate (DOP) when tested in accordance with the standard and at the system flowrate specified below:

ESF Ventilation System ANSI Standard Flowrate (cfm)

CREV System N510-1975 > 5,400 and 5 6,600 ESF Ventilation System N510-1980 > 22,500 and < 27,500 FHAEV System N510-1980 > 27,000 and < 33,000

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a removal efficiency of > 99% of a halogenated hydrocarbon refrigerant test gas when tested in accordance with the standard and at the system flowrate specified below:

ESF Ventilation System ANSI Standard Flowrate (cfm)

CREV System N510-1975 5,400 and 5 6,600 ESF Ventilation System N510-1980 > 22,500 and S 27,500 FHAEV System N510-1980 > 27,000 and 5 33,000

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers, shows the methyl iodide penetration less than or equal to the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30°C (86°F) and the relative humidity (RH) specified below:

Cook Nuclear Plant Unit 1 5.5-8 Amendment No. 28-7, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)

ESF Ventilation System Face Velocity (fpm) Penetration (%) RH CREV System NA 1 95 ESF Ventilation System 45.5 5 95 FHAEV System 46.8 5 95 In addition, the carbon samples not obtained from test canisters shall be prepared by either:

1. Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed; or

2. Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below:

Delta P ESF Ventilation System (inches water gaugqe) Flowrate (cfm)

CREV System 6 >5,400 and !6,600 ESF Ventilation System 6 > 22,500 and < 27,500 FHAEV System 6 > 27,000 and < 33,000 Amendment No. 28-A 298 Unit 1 5.5-9 Nuclear Plant Cook Nuclear Plant Unit 1 5.5-9 Amendment No. 247-, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor temporary liquid storage tanks.

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a Surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b. A Surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of > 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and

c. A Surveillance program to ensure that the quantity of radioactivity contained in all outdoor temporary liquid storage tanks that are not surrounded by liners, dikes, or walls capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

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.

5.5.11 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 applicable 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 API gravity, an absolute specific gravity, or a specific gravity within limits; Cook Nuclear Plant Unit 1 5.5-10 Amendment No. 28-7, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Diesel Fuel Oil Testing Program (continued)

2. A flash point within limits and, if the gravity was not determined by comparison with the supplier's certification, a kinematic or saybolt viscosity within limits; and

3. A clear and bright appearance with proper color;

b. Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in Specification 5.5.11 .a above, are within limits; and

c. Total particulate concentration of the fuel oil is < 10 mg/I when tested every 31 days in accordance with ASTM D-2276, Method A.

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

5.5.12 Technical Specifications (TS) 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 TS 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 TS incorporated in the license; or

2. A change to the updated FSAR 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 5.5.12.b 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).

Amendment No. ~ 298 5.5-11 Nuclear Plant Cook Nuclear Unit 1 Plant Unit 1 5.5-11 Amendment No. 28-7, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.13 Safety Function Determination Program (SFDP)

This program ensures loss of safety function is detected and appropriate actions taken. Upon entry into LCO 3.0.6, an evaluation shall be made to determine if loss of safety function exists. Additionally, other appropriate actions may be taken as a result of the support system inoperability and corresponding exception to entering supported system Condition and Required Actions. This program implements the requirements of LCO 3.0.6.

a. The SFDP shall contain the following:

1. 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;

2. Provisions for ensuring the unit is maintained in a safe condition if a loss of function condition exists;

3. Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

4. Other appropriate limitations and remedial or compensatory actions.

b. A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), 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:

1. A required system redundant to the system(s) supported by the inoperable support system is also inoperable;

2. A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

3. A required system redundant to the support system(s) for the supported systems described in Specifications 5.5.13.b.1 and 5.5.13.b.2 above is also inoperable.

c. 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. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

Cook Nuclear Plant Unit 1 5.5-12 Amendment No. 28-7, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 Containment Leakage Rate Testing Program

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September, 1995, as modified by the following exceptions:

1. The Type A testing Frequency specified in NEI 94-01, Revision 0, Paragraph 9.2.3, as "at least once per 10 years based on acceptable performance history" is modified to be "at least once per 15 years based on acceptable performance history." This change applies only to the interval following the Type A test performed in October 1992.

2. A one-time exception to the requirement to perform post-modification Type A testing is allowed for the steam generators and associated piping, as components of the containment barrier. For this case, ASME Section XI leak testing will be used to verify the leak tightness of the repaired or modified portions of the containment barrier. Entry into MODES 3 and 4 following the extended outage that commenced in 1997 may be made to perform this testing.

b. The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 12 psig.

c. The maximum allowable containment leakage rate, La, at Pa, shall be 0.25%

of containment air weight per day.

d. Leakage rate acceptance criteria are:

1. Containment leakage rate acceptance criterion is 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and < 0.75 La for Type A tests.

2. Air lock testing acceptance criterion is overall air lock leakage rate is

< 0.05 La when tested at > Pa.

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

5.5-13 Amendment No. 28w, 298 Cook Nuclear Plant Unit I1 Plant Unit 5.5-13 Amendment No. 2-97, 298

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.15 Battery Monitoring and Maintenance Program This program provides for battery restoration and maintenance, based on the recommendations of IEEE Standard 450-1995, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer including the following:

a. Actions to restore battery cells with float voltage < 2.13 V; and

b. Actions to equalize and test battery cells that had been discovered with electrolyte level below the minimum established design limit.

Cook Nuclear Plant Unit 1 5.5-14 Amendment No. 2-97-, 298

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring Report When a report is required by Condition B or H of LCO 3.3.3, "Post Accident Monitoring (PAM) 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.

5.6.7 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 5.5.7, 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,

f. Total number and percentage of tubes plugged to date, and

g. The results of condition monitoring, including the results of tube pulls and in-situ testing.

Cook Nuclear Plant Unit 1 5.6-4 Amendment No. 247, 2-, 298

UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapter/Specification Paqe 3.3 INSTRUMENTATION 3.3.1 Reactor Trip System (RTS) Instrumentation ................................................................... 3.3.1-1 Table 3.3.1-1, Reactor Trip System Instrumentation ................................................... 3.3.1-11 3.3.2 Engineered Safety Feature Actuation System (ESFAS) Instrumentation ........................ 3.3.2-1 Table 3.3.2-1, Engineered Safety Feature Actuation System Instrumentation ............. 3.3.2-7 3.3.3 Post Accident Monitoring (PAM) Instrumentation .......................................................... 3.3.3-1 Table 3.3.3-1, Post Accident Monitoring Instrumentation ............................................. 3.3.3-4 3.3.4 Remote Shutdown Monitoring Instrumentation ............................................................... 3.3.4-1 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation ................................ 3.3.5-1 3.3.6 Containment Purge Supply and Exhaust System Isolation Instrumentation ................ 3.3.6-1 Table 3.3.6-1, Containment Purge Supply and Exhaust System Isolation Instrum entation ......................................................................................................... 3.3.6-4 3.3.7 Control Room Emergency Ventilation (CREV) System Actuation Instrumentation .......... 3.3.7-1 Table 3.3.7-1, CREV System Actuation Instrumentation .............................................. 3.3.7-3 3.3.8 Boron Dilution Monitoring Instrumentation (BDMI) .......................................................... 3.3.8-1 3.4 REACTOR COOLANT SYSTEM (RCS) 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 ................. I.................................................... 3.4.2-1 3.4.3 RCS Pressure and Temperature (P/T) Limits ............................. 3.4.3-1 Figure 3.4.3-1, Reactor Coolant System Pressure versus Temperature Limits -

Heatup Limit, Criticality Limit, and Leak Test Limit (Applicable for service period up to 32 ,E FP Y) ....................................................... ............................................... 3.4 .3-3 Figure 3.4.3-2, Reactor Coolant System Pressure versus Temperature Limits -

Various Cooldown Rates Limits (Applicable for service period up to 32 EFPY) ........ 3.4.3-4 3.4.4 RCS Loops - MODES 1 and 2 ........................................................................................ 3.4.4-1 3.4.5 RC S Loops - MO D E 3 .................................................................................................... 3.4.5-1 3.4.6 RC S Loops - MO D E 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 P ressurizer ..................................................................................................................... 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 Pressure Isolation Valve (PIV) Leakage ................................................................. 3.4.14-1 3.4.15 RCS Leakage Detection Instrumentation...................................................................... 3.4.15-1 3.4.16 R C S Specific Activity ...................................................................................................... 3.4.16-1 Figure 3.4.16-1, Reactor Coolant DOSE EQUIVALENT 1-131 Specific Activity Limit Versus Percent of RATED THERMAL POWER ........................................................ 3.4.16-3 3.4.17 Steam Generator (SG) Tube Integrity ............................................................................ 3.4.17-1 Cook Nuclear Plant Unit 2 Page 2 of 5 Amendment No. 2-69, 279

UNIT 2 APPENDIX A TECHNICAL SPECIFICATIONS TABLE OF CONTENTS Chapte r/Specification Page 5.0 ADMINISTRATIVE CONTROLS 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 ........................................................................................................................ 5.2-1 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 (ODCM) ...................................................................... 5.5-1 5.5.2 Leakage Monitoring Program ......................................................................................... 5.5-2 5.5.3 Radioactive Effluent Controls Program ........................................................................... 5.5-2 5.5.4 Component Cyclic or Transient Limits ........................................................................... 5.5-3 5.5.5 Reactor Coolant Pump Flywheel Inspection Program ..................................................... 5.5-4 5.5.6 Inservice Testing Program ................................................................................. ......... 5.5-4 5.5.7 Steam Generator (SG) Program .................................................................................... 5.5-5 5.5.8 Secondary W ater Chemistry Program ........................................................................... 5.5-7 5.5.9 Ventilation Filter Testing Program (VFTP) ............................... 5.5-7 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Program .............. 5.5-10 5.5.11 Diesel Fuel Oil Testing Program ....................................................................................... 5.5-10 5.5.12 Technical Specifications (TS) Bases Control Program ..... .: ...................................... 5.5-11 5.5.13 Safety Function Determination Program (SFDP) ...................... :....................................... 5.5-12 5.5.14 Containment Leakage Rate Testing Program ................................................................. 5.5-13 5.5.15 Battery Monitoring and Maintenance Program .... ...................................................... 5.5-13 5.6 Reporting Requirem ents ................................................................................................. 5.6-1 5.6.1 Deleted ........................................................................................................................... 5.6-1 5.6.2 Annual Radiological Environmental Operating Report .................................................... 5.6-1 5.6.3 Radioactive Effluent Release Report ............................................................................. 5.6-2 5.6.4 Deleted ........................................................................................................................... 5.6-2 5.6.5 CORE OPERATING LIM ITS REPORT (COLR) .............................................................. 5.6-2 5.6.6 Post Accident Monitoring Report .................................................................................... 5.6-4 5.6.7 Steam Generator Tube Inspection Report ...................................................................... 5.6-4 5.7 High Radiation Area ....................................................................................................... 5.7-1 Cook Nuclear Plant Unit 2 Page 5 of 5 Amendment No. 2-9, 279

Definitions 1.1 1.1 Definitions ENGINEERED SAFETY The ESF RESPONSE TIME shall be that time interval from FEATURE (ESF) RESPONSE when the monitored parameter exceeds its actuation setpoint TIME at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.). Times shall include diesel generator starting and sequence loading delays, where applicable. 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 verification have been previously reviewed and approved by the NRC.

LEAKAGE LEAKAGE shall be:

a. Identified LEAKAGE

1. LEAKAGE, such as that from pump seals or valve packing (except reactor coolant pump (RCP) seal water injection or 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 water injection or leakoff) that is not identified LEAKAGE; and

c. Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

Cook Nuclear Plant Unit 2 1.1 -3 Amendment No. 2-64, 279

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

d. 150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

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

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

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

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Pressure boundary LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

Cook Nuclear Plant Unit 2 3.4.13-1 Amendment No. 269, 279

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ---------------------- NOTES ---------------

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

2. Not applicable to primary to secondary LEAKAGE.

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

SR 3.4.13.2 ---------------------- NOTES----- --------

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

Verify primary to secondary LEAKAGE is - 150 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gallons per day through any one SG.

Cook Nuclear Plant Unit 2 3.4_13-2 Amendment No. 26-9, 279

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 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

NO r! ----------------------------------------------------------

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 of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged maintained until the next in accordance with the refueling outage or SG Steam Generator tube inspection.

Program.

AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program. next refueling outage or SG tube inspection B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR SG tube integrity not maintained.

Cook Nuclear Plant Unit 2 3.4.17-1 AmendmentNo- 279

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the In accordance Steam Generator Program. with the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program. a SG tube inspection Cook Nuclear Plant Unit 2 3.4.17-2 AmendmentNO. 279

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Proqram 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:

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. SG 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.

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 I gpm for all SGs.

Cook Nuclear Plant Unit 2 5.5-5 Amendment No. 264, 279

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.7 Steam Generator (SG) Program (continued)

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 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.

Amendment No. 26~, 279 Unit 22 5.5-6 Nuclear Plant Cook Nuclear Plant Unit 5.5-6 Amendment No. 2499, 279

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.8 Secondary Water Chemistry Progqram This program provides controls for monitoring secondary water chemistry to inhibit SG tube degradation. 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;

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 is required to initiate corrective action.

5.5.9 Ventilation Filter Testing Program (VFTP)

The VFTP shall establish the required testing of Engineered Safety Feature (ESF) filter ventilation systems. Tests described in Specifications 5.5.9.a and 5.5.9.b shall be performed once per 24 months; after each complete or partial replacement of the HEPA filter bank or charcoal adsorber bank; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the filter bank or charcoal adsorber capability.

Tests described in Specification 5.5.9.c shall be performed once per 24 months; after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of adsorber operation; after any structural maintenance on the HEPA filter bank or charcoal adsorber bank housing; and, following painting, fire, or chemical release in any ventilation zone communicating with the subsystem while it is in operation that could adversely affect the charcoal adsorber capability.

Tests described in Specification 5.5.9.d shall be performed once per 24 months.

The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the VFTP test Frequencies.

Cook Nuclear Plant Unit 2 5.5-7 Amendment No. 26W, 279

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)

a. Demonstrate for each of the ESF systems that an inplace test of the high efficiency particulate air (HEPA) filters shows a removal efficiency of > 99%

of the dioctyl phthalate (DOP) when tested in accordance with the standard and at the system flowrate specified below:

ESF Ventilation System ANSI Standard Flowrate (cfm)

CREV System N510-1975 Ž 5,400 and 5 6,600 ESF Ventilation System N510-1980 Ž 22,500 and 5 27,500 FHAEV System N510-1980 > 27,000 and 5 33,000

b. Demonstrate for each of the ESF systems that an inplace test of the charcoal adsorber shows a removal efficiency of > 99% of a halogenated hydrocarbon refrigerant test gas when tested in accordance with the standard and at the system flowrate specified below:

ESF Ventilation System ANSI Standard Flowrate (cfm)

CREV System N510-1975 Ž 5,400 and 5 6,600 ESF Ventilation System N510-1980 22,500 and

27,500 FHAEV System N510-1980 > 27,000 and

33,000

c. Demonstrate for each of the ESF systems that a laboratory test of a sample of the charcoal adsorber, when obtained from either at least one test canister or at least two carbon samples removed from one of the charcoal adsorbers, shows the methyl iodide penetration less than or equal to the value specified below when tested in accordance with ASTM D3803-1989 at a temperature of 30'C (86°F) and the relative humidity (RH) specified below:

Cook Nuclear Plant Unit 2 5.5-8 Amendment No. 2-6-, 279

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Ventilation Filter Testing Program (VFTP) (continued)

ESF Ventilation System Face Velocity (fpm) Penetration (%) RH (%)

CREV System NA 1 95 ESF Ventilation System 45.5 5 95 FHAEV System 46.8 5 95 In addition, the carbon samples not obtained from test canisters shall be prepared by either:

1. Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed; or

2. Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed.

d. Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters and the charcoal adsorbers is less than the value specified below when tested at the system flowrate specified below:

Delta P ESF Ventilation System (inches water gauge) Flowrate (cfm)

CREV System 6 > 5,400 and __6,600 ESF Ventilation System 6 > 22,500 and < 27,500 FHAEV System 6 > 27,000 and < 33,000 Amendment No. 26~, 279 5.5-9 Cook Nuclear Cook Nuclear Plant Unit 2 Plant Unit 2 5.5-9 Amendment No. 2-W 279

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.10 Explosive Gas and Storage Tank Radioactivity Monitoring Pro-gram This program provides controls for potentially explosive gas mixtures contained in the Waste Gas Holdup System, the quantity of radioactivity contained in gas storage tanks and the quantity of radioactivity contained in unprotected outdoor temporary liquid storage tanks.

The program shall include:

a. The limits for concentrations of hydrogen and oxygen in the Waste Gas Holdup System and a Surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria (i.e., whether or not the system is designed to withstand a hydrogen explosion);

b. A Surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than the amount that would result in a whole body exposure of > 0.5 rem to any individual in an unrestricted area, in the event of an uncontrolled release of the tanks' contents; and

c. A Surveillance program to ensure that the quantity of radioactivity contained in all outdoor temporary liquid storage tanks that are not surrounded by liners, dikes, or walls capable of holding the tanks' contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System is less than the amount that would result in concentrations less than the limits of 10 CFR 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area, in the event of an uncontrolled release of the tanks' contents.

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.

5.5.11 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 applicable 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 API gravity, an absolute specific gravity, or a specific gravity within limits; 5.5-10 Amendment No. 2~, 279 Cook Nuclear Unit 2 Plant Unit Nuclear Plant 2 5.5-10 Amendment No. 2-99, 279

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.11 Diesel Fuel Oil Testing Program (continued)

2. A flash point within limits and, if the gravity was not determined by comparison with the supplier's certification, a kinematic or saybolt viscosity within limits; and

3. A clear and bright appearance with proper color;

b. Within 31 days following addition of the new fuel oil to storage tanks, verify that the properties of the new fuel oil, other than those addressed in Specification 5.5.11 .a above, are within limits; and

c. Total particulate concentration of the fuel oil is < 10 mg/I when tested every 31 days in accordance with ASTM D-2276, Method A.

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

5.5.12 Technical Specifications (TS) 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 TS 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 TS incorporated in the license; or

2. A change to the updated FSAR 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 5.5.12.b 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).

Amendment No. 2&~, 279 5.5-11 Cook Nuclear Unit 22 Plant Unit Nuclear Plant 5.5-11 Amendment No. 249, 279