Text

Proposed Tech Specs,Revising Administrative Control & Resolution of NRC Comments
	ML20117K958
Person / Time
Site:	Palisades
Issue date:	09/03/1996
From:	; CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	;
Shared Package
ML18065A908	List: ML20117K958;
References
	NUDOCS 9609120266
	Download: ML20117K958 (66)
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ATTACHMENT 2 CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 l l

TECHNICAL SPECIFICATION CHANGE REQUEST ADMINISTRATIVE CONTROLS ADDITIONAL CHANGES l

Replacement Proposed Pages i

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75 Pages 9609120266 960903 5 PDR ADOCK 0500

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, . PALISADES PLANT TECHNICAL SPECIFICATIONS

. . TABLE OF CONTENTS l

l SECTION DESCRIPTION PAGE NO

, 1.0 DEFINITIONS 1-1

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 e

, 2.1 SAFETY LIMITS - REACTOR CORE 2-1

2.2 SAFETY LIMITS - PRIMARY COOLANT SYSTEM PRESSURE 2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS - RPS 2-1

Table 2.3.1 Reactor Protective System Trip Setting Limits 2-2 B2.1 Basis - Reactor Core Safety Limit B 2-1 1

B2.2 Basis - Primary Coolant System Safety Limit B 2-2 B2.3 Basis - Limiting Safety System Settings B 2-3 3.0 LIMITING CONDITIONS FOR OPERATION 3-1 l 3.0 APPLICABILITY 3-1 3.1 PRIMARY COOLANT SYSTEM 3-lb

, 3.1.1 Operable Components 3-lb 3.1.2 Heatup and Cooldown Rates 3-4 1

Figure 3-1 Pressure - Temperature Limits for Heatup 3-5 Figure 3-2 Pressure - Temperature Limits for Cooldown 3-6 3.1.3 Minimum Conditions for Criticality 3-12 3.1.4 Maximum Primary Coolant Radioactivity 3-17 3.1.5 Primary Coolant System Leakage Limits 3-20 1 3.1.6 Maximum PCS 0xygen and Halogen Concentration 3-23 3.1.7 Primary and Secondary Safety Valves 3-24a 3.1.8 Over Pressure Protection Systems 3-25a Figure 3-4 LTOP Limit Curve 3-25c 3.1.9 Shutdown Cooling 3-25h l 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM 3-26 3.3 EMERGENCY CORE COOLING SYSTEM 3-29 3.4 CONTAINMENT COOLING 3-34 3.5 STEAM AND FEEDWATER SYSTEMS 3-38 3.6 CONTAINMENT SYSTEM 3-40 Table 3.6.1 Containment Penetrations and Valves 3-40b

, 3.7 ELECTRICAL SYSTEMS 3-41 3.8 REFUELING OPERATIONS 3-46 3.9 Deleted 3-49 3

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i Amendment No.

. PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTIDN DESCRIPTION PAGE N0 3.0 LIMITING CONDITIONS FOR OPERATION (continued) 3-1 3.10 CONTROL R0D AND POWER DISTRIBUTION LIMITS 3-50 3.10.1 Shutdown Margin Requirements 3-50 3.10.2 Deleted 3-51 3.10.3 Part-Length Control Rods 3-51 3.10.4 Misaligned or Inoperable Rod 3-52 3.10.5 Regulating Group Insertion Limits 3-52 3.10.6 Shutdown Rod Limits 3-53 3.10.7 Low Power Physics Testing 3-53 3.11 POWER DISTRIBUTION INSTRUMENTATION 3-56 3.11.1 Incore Detectors 3-56

-3.11.2 Excore Power Distribution Monitoring System 3-57 Figure 3.11-1 Axial Variation Bounding Condition 3-59 3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY 3-60 3.13 Deleted 3-60 3.14 CONTROL ROOM VENTILATION 3-61 3.15 Deleted 3-62 3.16 ESF SYSTEM INITIATION INSTRUMENTATION SETTINGS 3-63 Table 3.16.1 ESF System Initiation Instrument Setting Limits 3-63 B3.16 Basis - ESF System Instrumentation Settings B 3.16-1 3.17 INSTRUMENTATION AND CONTROL SYSTEMS 3-64 3.17.1 Reactor Protactive System Instruments 3-64 Table 3.17.1 Instrument Requirements for RPS 3-65 3.17.2 Engineered Safety Features Instruments 3-66 Table 3.17.2 Instrument Requirements for ESF Systems 3-67 '

3.17.3 Isolation Functions Instruments 3-68 Table 3.17.3 Instrument Requirements Isolation Functions 3-69 3.17.4 Accident Monitoring Instruments 3-70 Table 3.17.4 Instrument Requirements for Accident Monitoring 3-71 3.17.5 Alternate Shutdown System Instruments 3-72 Table 3.17.5 Instruments for the Alternate Shutdown System 3-73 3.17.6 Other Safety Feature Instruments 3-74 Table 3.17.6 Instruments for Other Safety Features 3-77 B3.17 Basis - Instrumentation Systems B 3.17-1 3.18 Deleted 3-79 3.19 I0 DINE REMOVAL SYSTEM 3-79 3.20 SH0CK SUPPRESSORS (Snubbers) 3-80 3.21 CRANE OPERATIONS AND MOVEMENT HEAVY LOADS 3-81 3.22 Deleted 3-84 3.23 POWER DISTRIBUTION LIMITS 3-84 3.23.1 Linear Heat Rate 3-84 3.23.2 Radial Peaking Factors 3-86 3.23.3 Quadrant Power Tilt - Tq 3-87 l Table 3.23-3 Power Distribution Measurement Uncertainty 3-88 ii Amendment No.

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,' PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION DESCRIPTION PAGE N0 l l

4.0 SURVEILLANCE RE0VIREMENTS 41 I 4.1 OVER PRESSURE PROTECTION SYSTEM TESTS 4-6 4.2 EQUIPMENT AND SAMPLING TESTS 4-7 Table 4.2.1 Minimum Frequencies for Sampling Tests 4-9 )

Table 4.2.2 Minimum Frequencies for Equipment Tests 4-11 Table 4.2.3 Ventilation Systems Tests 4-14 4.3 SYSTEMS SURVEILLANCE 4-16 Table 4.3.1 Primary Coolant System Pressure Isolation Valves 4-18 4.4 Deleted 4-19 4.5 CONTAINMENT TESTS 4-19 4.5.1 Integrated Leakage Rate Tests 4-19 4.5.2 Local Leak Detection Tests 4-19 4.5.3 Containment Isolation Valves 4-21 4.5.4 Surveillance for Prestressing System 4-21 4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS 4-24 4.6.1 Safety Injection System 4-24 4.6.2 Containment Spray System 4-24 4.6.3 Pumps 4-24 4.6.4 Valves 4-24 4.6.5 Containment Air Cooling System 4-25 4.7 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-42 4.7.1 Diesel Generators 4-42 4.7.2 Station Batteries 4-42 4.7.3 Emergency Lighting 4-43 i

i' 4.8 MAIN STEAM STOP VALVES 4-44 l 4.9 AUXILIARY FEEDWATER SYSTEM 4-45 l 4.10 REACTIVITY AN0MALIES 4-46 4.11 Deleted- 4-46 4.12 AUGMENTED ISI PROGRAM FOR HIGH ENERGY LINES 4-60 4.13 Deleted 4-65

! 4.14 AUGMENTED ISI PROGRAM FOR STEAM GENERATORS 4-66 l 4.15 PRIMARY SYSTEM FLOW MEASUREMENT 4-70 j 4.16 ISI PROGRAM FOR SH0CK SUPPRESSORS (Snubbers) 4-71 4

4.17 INSTRUMENTATION SYSTEMS TESTS 4-75 Table 4.17.1 Surveillance for the RPS 4-76 i Table 4-17.2 Surveillance for ESF Functions 4-77 i Table 4-17.3 Surveillance for Isolation Functions 4-78 Table 4-17.4 Surveillance for Accident Monitoring 4-79

Table 4-17.5 Surveillance for Alternate Shutdown 4-80 j Table 4-17.6 Surveillance for Other Safety Functions 4-81 l

84.17 Basis - Instrumentation Systems Surveillance B 4.17-1 i

4.18 POWER DISTRIBUTION INSTRUMENTATION 4-83 4.18.1 Incore Detectors 4-83

4.18.2 Excore Monitoring System 4-83 i 4.19 POWER DISTRIBUTION LIMITS 4-84

4.19.1 Linear Heat Rate 4-84

4.19.2 Radial Peaking Factors 4-84

4.20 MODERATOR TEMPERATURE COEFFICIENT (MTC) 4-85 1

iii j Amendment No.

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PALISADES PLANT TECHNICAL SPECIFICATIONS *

. TABLE OF CONTENTS SECTIDN DESCRIPTION PAGE N0 5.0 DESIGN FEATURES 5-1 5.1 SITE 5-1 5.2 CONTAINMENT DESIGN FEATURES 5-1 5.2.1 Containment Structures 5-1 i

5.2.2 - Penetrations 5-2 5.2.3 Containment Structure Cooling Systems 5-2 :

5.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) 5-2 i 5.3.1 Primary Coolant System 5-2 ,

5.3.2 Reactor Core and Control 5-3

5.3.3 Emergency Core Cooling System 5-3 5.4 FUEL STORAGE 5-4 3

5.4.1 New Fuel Storage 5-4 5.4.2 Spent Fuel Storage 5-4a i Figure 5-1 Site Environment TLD Stations 5-5 6.0 ADMINISTRATIVE CONTR011 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 l 6.2.1 Onsite and Offsite Organizations 6-1 ;

6.2.2 Plant Staff 6-2 6.3 PLANT STAFF QUALIFICATIONS 6-3 6.4 PROCEDURES 6-4 6.5 PROGRAMS AND MANUALS 6-5 6.5.1 Offsite Dose Calculation Manual 6-5 6.5.2 Primary Coolant Sources Outside Containment 6-6 6.5.3 Post Accident Sampling Program 6-6 1 6.5.4 Radioactive Effluent Controls Program 6-7 6.5.5 Containment Structural Integrity Sury. Program 6-8 6.5.6 Primary Coolant Pump Flywheel Sury. Program 6-8 i 6.5.7 Inservice Inspection and Testing Program 6-8 l 6.5.8 Steam Generator Tube Surveillance Program 6-9 6.5.9 Secondary Water Chemistry Program 6-14 6.5.10 Ventilation Filter Testing Program 6-15 6.5.11 Reserved 6-16 6.5.12 Technical Specification Bases Control Program 6-16 6.5.13 Reserved 6-17 6.5.14 Containment Leak Rate Testing Program 6-17 6.5.15 Process Control Program 6-18 6.6 REPORTING REQUIREMENTS 6-19 6.6.1 Occupational Radiation Exposure Report 6-19 6.6.2 Radiological Environmental Operating Report 6-19 6.6.3 Radioactive Effluent Release Report 6-19 6.6.4 Monthly Operating Report 6-19 6.6.5 Core Operating Limits Report 6-20 6.6.6 Reserved 6-21 6.6.7 Accident Monitoring Instrument Report 6-21 6.6.8 Containment Structural Integrity Surveillance Report 6-22 l 6.6.9 Steam Generator Tube Surveillance Report 6-22 6.7 HIGH RADIATION AREA 6-23 ,

iv Amendment No.

TECHNICAL SPECIFICATIONS I

! 1.0 DEFINITIONS

! The following terms are defined for uniform interpretation of these Technical Specifications.

I l ASSEMBLY RADIAL PEAKING FACTOR - F,^

ASSEMBLY RADIAL PEAXING FACTOR shall be the maximum ratio of the power generated in any fuel assembly, to the average fuel assembly )ower.

l (Each of these power terms shall be integrated over core heigit and j shall include tilt.)

AVERAGE DISINTEGRATION ENERGY - E AVERAGE DISINTEGRATION ENERGY shall be the average (weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling) of the sum of the average beta and

. gamma energies per disintegration (in MEV) for isoto)es, other than i iodines, with half lives greater than 15 minutes, ma(ing up at least i 95% of the total noniodine activity in the coolant.

AXIAL OFFSET or AXIAL SHAPE INDEX - A0 or ASI AXIAL OFFSET or AXIAL SHAPE INDEX shall be the ratio of the power generated in the lower half of the core minus the power generated in the upper half of the core, to the sum of those powers.

CHANNEL CALIBRATION A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors.

The CHANNEL CALIBRATION shall encompass the entire channel including the sensor, alarm, interlock, and trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is calibrated. Neutron detectors may be excluded I from CHANNEL CALIBRATIONS. !

CHANNEL CHECK A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and status with other-indications and status derived from independent instrument channels measuring the same parameter. A CHANNEL CHECK shall include verification that the monitored parameter is within limits imposed by the Technical Specifications.

1-1 Amendment No. M, 43, 64, H, 68, Me, M4, M8, 4N, M2,

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l l 1.0 DEFINITIONS (continued)' i 1

CHANNEL FUNCTIONAL TEST A CHANNEL FUNCTIONAL TEST shall be the injection of a simulated signal into the channel to verify that it is OPERABLE, including any alarm and trip initiating function.

COLD SHUTDOWN The COLD SHUTDOWN condition shall be when the arimary coolant is at SHUTDOWN B0RON CONCENTRATION and T., is less tian 210*F.

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CONTAINMENT INTEGRITY CONTAINMENT INTEGRITY is defined to exist when all the following are l true:

a. All nonautomatic containment isolation valves and blind flanges are closed (OPERABLE) except as noted in Table 3.6.1.

b. The equipment hatch is properly closed and sealed.

c. At least one door in each personnel air lock is properly closed and sealed.

d. All automatic containment isolation v;.hes are OPERABLE i (as demonstrated by satisfying isolation times specified in l Table 3.6.1 and leakage criterion in Specification 4.5.2) or are i

locked closed. i

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e. The uncontrolled containment leakage satisfies Specification 4.5.

j CONTROL RODS CONTROL RODS shall be all full-length shutdown and regulating rods.

CORE OPERATING LIMITS REPORT (COLR)

The COLR is the document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with l Specification 6.6.5. Plant operation within these limits is addressed in individual Specifications.

I DOSE EQUIVALENT I-131 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (gCi/gm) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134 and 1-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."

l-2 Amendment No. H , 43, 64, 9 , 68, M B, M 4, 444, 4 W , M a,

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l 1.0 DEFINITIONS (continued) l HOT SHUTDOWN The HOT SHUTDOWN condition shall be when the reactor is subcritical by an amount greater than or equal to the margin as specified in Technical Specification 3.10 and T.,, is greater than 525'F.

I HOT STANDBY l is greater than 525'F and I The any ofHOT STANDBY the CONTROL RODS condition shalland are withdrawn bet when T~ he neutron flux power ra instrumentation indicates less than 2% of RATED POWER.

LOW POWER PHYSICS TESTING LOW POWER PHYSICS TESTING shall be testing performed under approved i written procedures to determine CONTROL R00 worths and other core I nuclear properties. Reactor power during these tests shall not exceed '

2% of RATED POWER, not including decay heat and PCS T.,, and PCS pressure .

shall be in the range of 371*F to 538'F and 415 psia to 2150 psia, !

respectively. Certain deviations from normal operating practice which l are necessary to enable performing some of these tests are permitted in '

accordance with the specific provisions in these Technical ;

Specifications.

l OPERABLE - OPERABILITY ;

1 A system, subsystem, train, component, or device shall be OPERABLE, or i have OPERABILITY, when it is capable of performing its specified i functions, and when all necessary attendant instrumentation, controls, electrical )ower, cooling or seal water, lubrication, or other auxiliary equipment t1at are required for the system, subsystem, train, component, or device to perform its specified functions are also capable of performing their related support functions.

EQ9ER OPERATION The POWER OPERATION condition shall be when the reactor is critical and the neutron flux power range instrumentation indicates greater than 2% of RATED POWER.

fdjaQRANT POWER TILT - T, QUADRANT POWER TILT shall be the algebraic ratio of quadrant power minus average quadrant power, to average quadrant power.

RATED POWER RATED POWER shall be a steady state reactor core output of 2530 MW,.

1-3 Amendment No. M, 43, 64, U, 68, M8, M4, M8, M7, M2,

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, l 1.0 DEFINITIONS (continued)

REACTOR CRITICAL I

The reactor is considered critical for pur)oses of administrative control when the neutron flux wide range ciannel instrumentation indicates greater than 10 % of RATED POWER.

REFUELING BORON CONCENTRATION REFUELING BORON CONCENTRATION shall be a Primary Coolant System boron I

concentration of at least 1720 ppm AND sufficient to assure the reactor l is subcritical by 2 5% Ap with all CONTROL RODS withdrawn.

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i REFUELING OPERATION A REFUELING OPERATION shall be any operation involving movement of core

, components (except for incore detectors) when the reactor vessel head is untensioned or removed with fuel in the reactor vessel.

1 REFUELING SHUTDOWN 4

The REFUELING SHUTDOWN condition shall be when the primary coolant is at REFUELING BORON CONCENTRATION and T . is less than 210*F.

SHUTDOWN BORON CONCENTRATION

. SHUTDOWN BORON CONCENTRATION shall be a Primary Coolant System boron concentration sufficient to assure the reactor is subcritical by 2 2% Ap with all CONTROL RODS in the core and the highest worth CONTROL R00 fully withdrawn.

SHUTDOWN MARGIN i

SHUTDOWN MARGIN shall be the instantaneous amount of reactivity by which 4 the reactor is subcritical or would be subcritical from its present condition assuming that all CONTROL RODS are fully inserted except for the single highest worth CONTROL R00 which is assumed to be withdrawn.

TOTAL RADlg PEAKING FACTOR - F,'

The TOTAL RADIAL PEAXING FACTOR shall be the maximum product of the ratio of individual assembly power to core average assembly power, times the highest local peaking factor integrated over the total core height, including tilt. Local peaking factor is defined as the maximum ratio of an individual fuel rod power to the assembly average rod power.

1-4 Amendment No. M, 43, 64, U, 68, 4M, M4, MS, 4W, Ma, MB,

l 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit - Reactor Core i

The Minimum DNBR of the reactor core shall be maintained greater than or i equal to the DNB correlation safety limit. 1 Correlation Safety Limit i XNB 1.17 !

ANFP 1.154 i HTP 1.141 i Acolicability Safety Limit 2.1 is applicable in HOT STANDBY and POWER OPERATION.

Action 2.1.1 If a Safety Limit is exceeded, the reactor shall be shut down immediately and not restarted until the Commission authorizes resumption of operation in accordance with 10 CFR 50.36(c)(1)(i)(A).

2.2 Safety Limit - Primary Coolant System Pressure (PCS) l The PCS Pressure shall not exceed 2750 psia.

J Anolicability Safety Limit 2.2 is applicable when there is fuel in the reactor.

Action 1 2.2.1 If a Safety Limit is exceeded, the reactor shall be shut down immediately and not restarted until the Commission authorizes resumption of operation in accordance with 10 CFR 50.36(c)(1)(i)(A).

2.3 Limitino Safety System Settinas - Reactor Protective System (RPS)

The RPS trip setting limits shall be as stated in Table 2.3.1. I Anolicability Limiting Safety System Settings of Table 2.3.1 are applicable when the associated RPS channels are required to be OPERABLE by Specification 3.17.1.

Action 2.3.1 If an RPS instrument setting is not within the allowable settings of Table 2.3.1, immediately declare the instrument inoperable and complete corrective action as directed by Specification 3.17.1.

Amendment No. M, M, 43, MB, M7, MO, MB, 2-1

3.17 INSTRUMENTATION SYSTEMS Snecification 3.17.4 The Accident Monitoring Instruments listed in Table 3.17.4 shall be OPERABLE. (Specifications 3.0.3, 3.0.4, and 4.0.4 do not apply.)

Acolicability Specification 3.17.4 applies when the PCS temperature is > 300*F.

Action 3.17.4.1 With one required channel of functions 1 through 14 inoperable for one or more functions:

a. Restore channel to OPERABLE status within 7 days.

3.17.4.2 With two required channels of functions 1 through 14 inoperable for one or more functions:

a. Restore one channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3.17.4.3 With position indication inoperable for one or more Containment Isolation Valves:

a. Restore the indication to OPERABLE status or lock the associated valves in the closed position within 7 days.

3.17.4.4 If any action required by 3.17.4.1 throu associated completion time has expired, gh 3.17.4.3 is not met AND the

a. The reactor shall be placed in HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , and

b. The reactor shall be placed in a condition where the affected equipment is not required, within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3.17.4.5 With one channel of functions 16 through 21 inoperable for one or more functions:

a. Restore the channel to OPERABLE status within 7 days.

3.17.4.6 With two required channels of functions 16 through 21 inoperable for one or more functions:

a. Restore one channel to OPERABLE status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

3.17.4.7 If any action required by 3.17.4.5 or 3.17.4.6 is not met AND the associated completion time has expired:

a. With two CETs in any one quadrant ino erable, complete Action 3.17.4.4 in lieu of Action 3.17.4.7 c ,

b. With two RVWL channels inoperable, initiate alternate monitoring within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ,

c. Submit a report to the NRC in accordance with Specification 6.6.7. l

d. Restore the channels to OPERABLE status prior to startup from the next refueling.

Amendment No. M6, M7, M3, 3-70

, ,- 4.0 SURVEILLANCE REQUIREMENTS

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4.0.1 Surveillance requirements shall be applicable during the reactor

operating conditions associated with individual Limiting Conditions for Operation unless ot%rwise stated in an individual surveillance requirement.

4.0.2 Each Surveillance Requirement shall be performed within the specified i surveillance interval with a maximum allowable extension not to exceed

i 25 percent of the surveillance interval.

! 4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute noncompliance with the operability requirements for a Limiting Condition

for Operation. The time limits of the action requirements are

applicable at the time it is identified that a Surveillance Requirement ,

has not been performed. The action requirements may be delayed for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the allowable outage time limits of the action requirements are less than

. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance Requirements do not have to be performed on

! inoperable equipment.

4.0.4 Entry into a reactor operating condition or other specified condition shall not be made unless the Surveillance Requirements associated with a Limiting Condition of Operation has been performed within the stated surveillance interval or as otherwise specified. This provision shall not prevent passage through or to plant conditions as required to comply with action requirements.

I Amendment No. 30, M, HQ, MB, 4M, 4-1 1

, 4.0 SURVEILLANCE RE0VIREMENT (Continued) l 4.0.5 Deleted l

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i Amendment No. 430, MB, 4

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! , 4.0 RASIS Specifications 4.0.1 through 4.0.4 establish the general requirements- l applica.ble to Surveillance Requirements. These requirements are based on the Surveillance requirements stated in the code of Federal Regulations, 10 CFR 50.36(c)(3):

" Surveillance requirements are requirements relating to test, !

calibration, or inspection to ensure that the necessary !

quality of systems and components is maintained, that facility i operation will be within safety limits, and that the limiting !

conditions of operation will be met."-

]

Specification 4.0.1 establishes the requirement that surveillances must be performed during reactor operating conditions or other conditions for which j the requirements of the Limiting Conditions for Operation apply, unless i otherwise stated in an individual Surveillance Requirement. The purpose of l this specification is to ensure that surveillances are performed to verify

! the operational status of systems and components and that parameters are ,

within specified limits to ensure safe operation of the facility when the l plant is in a reactor operating condition or other specified condition for which the associated Limiting Conditions for Operation are applicable.

Surveillance Requirements do not have to be performed when the facility is in I an operational condition for which the requirements of the associated l Limiting Condition for Operation do not apply, unless otherwise specified. )

The Surveillance Requirements associated with a Special Test Exception are only applicable when the Special Test Exception is used as an allowable exception the requirements of a specification. -

Specification 4.0.2 establishes the limit for which the specified time interval for Surveillance Requirements may be extended. It permits an allowable extension of the normal surveillance interval to facilitate surveillance scheduling and consideration of plant operating conditions that may not be suitable for conducting the surveillance; e.g., transient conditions or other ongoing surveillance or maintenance activities. It also provides flexibility to accommodate the length of a fuel cycle for surveillances'that are performed at each refueling outage and are specified with an 18-month surveillance intr. val. It is not intended that this provision be used repeatedly at a convenience to extend the surveillance intervals beyond that specified for surveillances that are not performed during refueling outager,. The limitation of Specification 4.0.2 is based on engineering judgment ar.d the recognition that the most probable result of any particular surveillance being performed is the verification of conformance with the Surveillance Requirements. This provision is sufficient to ensure that the reliability ensured through surveillance activities is not significantly degraded beyond that obtained from the specified surveillance interval.

Specification 4.0.3 establishes the failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by the provisions of Specification 4.0.2, as a condition that constitutes a failure to meet the operability requirements for a Limiting Condition for Operation, i Under the provisions of this specification, systems and components are assumed to be operable when Surveillance Requirements have 1

! Amendment No. +30, MB, +7+,

4-3

i 4.0 RASIS (Continued) j i

Specification 4.0.4 establishes the requirement that all applicable surveillances must be met before entry into a reactor operating condition or

, other condition of operation specified in the Applicability statement. The i

purpose of this specification is to ensure that system and component l operability requirements or parameter limits are met before entry into an i operational condition for which these systems and components ensure safe operation of the facility. This provision applies to changes in reactor

, operating conditions or other specified conditions associated with plant l

shutdown as well as startup.

Under the provisions of this specification, the applicable Surveillance l Requirements must be performed within the surveillance interval to ensure that the Limiting Conditions for Operation are met during initial plant startup or following a plant outage.

When a shutdown is' required to comply with action requirements, the provisions of Specification 4.0.4 do not apply because this would delay placing the facility in a-lower operational condition.

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I Amendment No. 430, 44B, 4-5 i

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. 4.1 OVERPRESSURE PROTECTION SYSTEM TESTS l =

l Surveillance Reauirements In addition to the requirements of The Inservice Inspection and Testing Program, Specification 6.5.7, each PORV flow path shall be demonstrated OPERABLE by:

1. Testing the PORVs in accordance with the inservice inspection requirements for ASME Boiler and Pressure Vessel Code,Section XI, Section IWV, Category B valves.

2. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.

3. When the PORV flow path is required to be OPERABLE by Specification 3.1.8.1:

(a. Performing a complete cycle of the PORV with the plant above COLD SHUTDOWN at least once per 18 months.

l (b. Performing a complete cycle of the block valve prior to heatup from COLD SHUTDOWN, if not cycled within 92 days.

4. When the PORY flow path is required to be OPERABLE by Specification 3.1.8.2:

(a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per ;

31 days.

(b. Verifying the associated block valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

5. Both High Pressure Safety Injection pumps shall be verified incapable of injection into the PCS at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , unless the reactor head is removed, when either PCS cold leg temperature is < 300*F, or when both shutdown cooling suction valves, M0-3015 and M0-3016, are open.

Basis With the reactor vessel head installed when the PCS cold leg temperature is less than 300*F, or if the shutdown cooling system isolation valves M0-3015 and M0-3016 are open, the start of one HPSI pump could cause the Appendix G or the shutdown cooling system pressure limits to be exceeded; therefore, both pumps are rendered inoperable.

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Amendment No. MG, 449, MG, MB, M3, M4, 4M, 4-6 i

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4.2 EQUIPMENT SAMPLING AND TESTS l

Table 4.2.3 VENTILATION SYSTEM TESTE The Control Room Ventilation and Isolation System and the Fuel Storage Area HEPA/ Charcoal Exhaust System shall be demonstrated to be OPERABLE by the following tests:

1

l. Performing required Control Room Ventilation and Fuel Storage Area l filter testing in accordance with the Ventilation Filter Testing Program.

l l 2. At least once per refueling cycle by:

i a. Verifying that on a containment high-pressure and high-radiation test signal, the Control Room Ventilation system

automatically switches into the emergency mode of operation with flow through the HEPA filter and charcoal adsorber bank.

! b. Verifying that the Control Room Ventilation system maintains

the Control Room at a positive pressure ;t 1/8 inch WG relative to the outside atmosphere during system emergency mode operation.

, c. Verifying that the Fuel Pool Ventilation System is OPERABLE by

! initiating flow through the HEPA filter and charcoal adsorbers j from the control room.

! l 3. Verifying that the Control Room temperature is s 90*F; once per i 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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l 4-14 Amendment No. 81, M2, 1

., Basis - Table 4.2.2 Item 12 - Trisodium Phosohate (TSP) Tests i Item 12.a - TSP quantity verification i

Verification of the quantity of TSP in the baskets ensures that neither leakage nor other water sources in the containment reduce the basket content below the required minimum. This requirement ensures that there is an adequate quantity of TSP to adjust the pH of the post LOCA sump solution to a i

value between 7.0 and 8.0.

! Item 12.b - TSP quality verification l Periodic testing is performed to ensure the solubility and buffering ability i of the TSP after exposure to the containment environment. Satisfactory l completion of this test assures that the TSP in the baskets is " active" as required by Specification 3.19.

Adequate solubility is verified by submerging a representative sample of TSP i

from one of the baskets in containment in un-agitated borated water heated to a temperature representing post-LOCA conditions; the TSP must completely dissolve within a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. The test time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is specified to

allow time for the dissolved TSP to naturally diffuse through the un-agitated '

4 test solution. Agitation of the test solution during the solubility

. verification is prohibited, since an adequate standard for the agitation intensity (other than no agitation) cannot be specified. The flow and

)l turbulence in the containment sump during recirculation would significantly '

i decrease the time required for the TSP to dissolve.

Adequate buffering capability is verified by a measured pH of the sample 1 solution, following the solubility verification, between 7 and 8. The sample l 1s cooled and thoroughly mixed prior to measuring pH.

1 The quantity of the TSP sample, and quantity and boron concentration of the j water are chosen to be representative of post-LOCA conditions.

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i 4-15 l Amendment No. ME,

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,' _4.3 SYSTEMS SURVEILLANCE

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APPLICABILITY Applies to preoperational and inservice structural surveillance of the reactor vessel and other Class 1, Class 2 and Class 3 system components.

OBJECTIVE To insure the integrity of the Class 1, Class 2 and Class 3 piping systems and components.

SPECIFICATIONS a,b,c,d,e,f - Deleted j 1

g. A surveillance program to monitor radiation induced changes in the l l mechanical and impact properties of the reactor vessel materials shall be maintained as described in Section 4.5.3 of the FSAR.

h. Periodic leakage testing'd* on each check valve listed in l l Table 4.3.1 shall be accomplished prior to returning to the Power l 0)eration Condition after every time the plant has been placed in tie Refueling Shutdown Condition, or the Cold Shutdown Condition for more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months if such testing has not been accomplished within the previous 9 months, and prior to returning the check valves to service after maintenance, repair or replacement work is performed on the valves.

i. Whenever integrity of a pressure isolation valve listed in l Table 4.3.1 cannot be demonstrated and credit is being taken for compliance with Specification 3.3.3.b, the integrity of the remaining check valve in each high pressure line having a leaking valve shall be determined and recorded daily and the position of the other closed valve located in that pressure line shall be recorded daily.

j. Following each use of the LPSI system for shutdown cooling, the l reactor shall not be made critical until the LPSI check valves (CK-3103, CK-3118, CK-3133 and CK-3148) have been verified closed.

'd To satisfy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

" Reduced pressure testing is acceptable (see footnote 5, Table 4.3.1).

Minimum test differential pressure shall not be less than 150 psid.

4-16 l Amendment No. 63, M, MG, MB,

,- 4.3 SYSTEMS SURVEILLANCE (Cont'd)

Basis The inspection program specified places major emphasis on the areas of highest stress concentration as determined by general design evaluation and experience with similar systems."' In addition, that portion of the reactor vessel shell welds which will be subjected to a fast neutron dose sufficient to change ductility properties will be inspected. The inspections will rely primarily on ultrasonic methods utilizing up-to-date analyzing equipment and trained personnel. To the extent applicable, based upon the existing design and construction of the plant, the requirements of Section XI of the Code ;

shall be complied with. Significant exceptions are detailed in the requests for relief which have received NRC approval and are contained in the Class 1, Class 2 and Class 3 Long-Term Inspection Plans.

Valve Testina To ensure the continued integrity of selected check valves which are relied upon to preclude a potential LOCA outside containment, special requirements for periodic leak tests are specified. In addition a valve disk position "

check for the LPSI check valves is specified following each use of the LPSI system for shutdown cooling. This position check ensures that the four LPSI check valves have reclosed upon cessation of shutdown cooling flow.

References I

(1) FSAR, Section 4.5.6 (2) Deleted (3) Systematic Evaluation Program Topic V-II.A, NRC letter to the licensee transmitting the final topic evaluation dated November 9,1981.

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l 4-17 Amendment No. 53, 73, 430, 442,

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TABLE 4.3.1 PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES Maximum W

System Valve No. Allowable Leakaae High Pressure Safety Injection Loop 1A, Cold Leg 3101 5.0 gpm l 3104 5.0 gpm l Loop 1B, Cold Leg 3116 5.0 gpm 3119 5.0 gpm l

l Loop 2A, Cold Leg 3131 5.0 gpm 3134 5.0 gpm I Loop 2B, Cold leg 3146 5.0 gpm 3149 5.0 gpm Low Pressure Safety Injection loop 1A, Cold Leg 3103 5.0gpm Loop 1B, Cold Leg 3118 5.0gpm Loop 2A, Cold Leg 3133 5.0gpm l Loop 28, Cold Leg 3148 5.0gpm l

Footnote:

w Leakage rates less than or equal to 1.0 gpm are considered acceptable.

1.

2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are i considered acceptable if the latest measured rate has not exceeded the rate determined by the previous test by an amount that reduces the i margin between measured leakage rate and the maximum permissible rate of l 5.0 gpm by 50% or greater.

l 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater.

4. Leakage rates greater than 5.0 gpm are considered unacceptable.

5. Measured leakage rates must be adjusted for test pressures less than the l maximum potential pressure differential across the valve by assuming

! leakage to be directly proportional to the pressure differential to the l one-half power.

j 4-18 l l NRC Order Dated Amendment No. &3, April 20, 1981

,- l 4. 4 Deleted 4.5 CONTAINMENT TESTS 4.5.1 Intearated Leakaoe Rate Tests The containment integrated leak rate testing shall be performed in I accordance with the Containment Leak Rate Testing Program.

4.5.2 Local Leak Detection Tests

a. Iltsi (1) local leak rate tests shall be performed at a pressure of not less than 55 psig.

(2) Local leak rate tests for checking air lock door seals within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of each door opening shall be performed at a pressure of not less than 10 psig.

(3) Acceptable methods of testing are halogen gas detection, soap bubble, pressure decay, or equivalent.

(4) The local leak rate shall be measured for each of the following components:

(a) Containment penetrations that employ resilient seal gaskets, sealant compounds, or bellows.

(b) Air lock and equipment door seals.

(c) Fuel transfer tube.

(d) Isolation valves on the testable fluid systems' lines penetrating the containment.

(e) Other containment components which require leak repair in order to meet the acceptance criterion for any integrated leak rate test.

b. Acceptance Criteria (1) The total leakage from all penetrations and isolation valves shall not exceed 0.60 L,.

(2) The leakage for an air lock door seal test shall not exceed 0.023 L,.

l 4-19 Amendment No. H , M 6, H 5,

4.5 CONTAINMENT TESTS

4.5.2 Local Leak Detection Tests (continued)

c. Corrective Action l l (1) If at any time it is determined that 0.60 L, is exceeded, repairs shall be initiated immediately. If repairs are not '

completed and conformance to the acceptance criterion of 4.5.2.b(1) is not demonstrated with 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , the Plant shall be placed in at least hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

(2) If at any time it is determined that total containment leakage exceeds L , within one hour action shall be initiated to bring the Plant to hot shutdown within the next six (6) hours and cold shutdown within the following thirty (30) hours.

i (3) If air lock door seal leakage is greater than 0.023 L.,

i repairs shall be initiated immediately to restore the door to

, less than specification 4.5.2.b(2). In the event repairs j cannot be completed within 7 days, the Plant shall be brought to a hot shutdown condition within the next six (6) hours and

cold shutdown within the following thirty (30) hours.

I If air lock door seal leakage results in one (1) door causing

total containment leakage to exceed 0.60 L , the door shall be declared inoperable and the remaining operable door shall be i

i' immediately locked closed and tested within four (4) hours.

As long as the remaining door is found to be operable, the l

. provisions of 4.5.2.c(2) do not apply. Repairs shall be

! initiated immediately to establish conformance with specification 4.5.2.b(1). In the event conformance to this l I

specification cannot be established within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months the Plant shall be brought to a hot shutdown within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and

cold shutdown within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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4-20 l Ch=;;; 7, Amendment No. 4M ,

~.5 4 CONTAINMENT TESTS 4.5.2 Local Leak Detection Tests (continued)

d. lest Freauggy (1) Individual penetrations and containment isolation valves shall be leak rate tested at a frequency of at least every six months prior to the first postoperational integrated leak rate test and at a frequency of at least every refueling thereafter, not exceeding a two-year interval, except as specified in (a) and (b) below:

(a) The containment equipment hatch and the fuel transfer tube shall be tested at each refueling shutdown or after each time used, if that be sooner.

(b) A full air lock penetration test shall be performed at six-month intervals. During the period between the six-month tests when containment integrity is required, a reduced pressure test for the door seals or a full air lock penetration test shall be performed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after either each air lock door opening or the first of a series of openings.

(2) Each three months the isolation valves must be stroked to the position required to fulfill their safety function unless it is established that such operation is not practical during plant operation. The latter valves shall be full-stroked during each cold shutdown.

l 4.5.3 Containment Isolation Valyg1

a. The isolation valves shall be demonstrated operable by performance of a cycling test and verification of isolation time for auto isolation valves prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit.

b. Each isolation valve shall be demonstrated operable by verifying that on each containment isolation right channel or left channel test signal, applicable isolation valves actuate to their required position during cold shutdown or at least once per refueling cycle,

c. The isolation time of each power operated or automatic valve shall be determined to be within its limit as specified in Table 3.6.1 when tested in accordance with Section XI of the ASME Boiler and Pressure Vessel Code.

4.5.4 Surveillance for Prestressina System Verify containment structural integrity in accordance with the Containment Structural Integrity Surveillance Program.

l 4-21 Amendment No. 4M, 4M, i

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,' 4.5 CONTAINMENT TESTS (continued)

EA111

The containment is designed for an accident pressure of 55 psig."'

i While the reactor is operating, the internal environment of the containment

will be air at approximately atmospheric pressure and a temperature of about l 104*F. With these initial conditions, following a LOCA, the temperature of j the steam-air mixture at the peak accident pressure of 55 psig is 283*F.

l i Prior to initial operation, the containment was strength-tested at 63 psig 4

and then leak rate tested. The design objective of this preoperational leak i rate test was established as 0.1% by weight per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 55 psig. This 1 leakage rate is consistent with the construction of the containment, which 1 i is equipped with independent leak-testable penetrations and contains channels l over all unaccessible containment liner welds, which were independently leak- )

j tested during construction. :

$ Accident analyses have been performed on the basis of a leakage rate of I 0.1% by weight per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With this leakage rate and with a reactor power l l level of 2530 MWt, the potential public exposure would be below 10 CFR 100 3

! guideline values in the event of the Maximum Hypothetical Accident. '

The performance of a periodic integrated leak rate test during plant life i provides a current assessment of potential leakage from the containment in l case of an accident that would pressurize the interior of the containment.

!' In order to provide a realistic appraisal of the integrity of the containment under accident conditions, this periodic leak rate test is to be performed

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without preliminary leak detection surveys or leak repairs and containment I isolation valves are to be closed in the normal manner.

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This normal manner is a coincident two-of-four high radiation or two-of-four

high containment pressure signals which will close all containment isolation

! valves not required for engineered safety features except the component l cooling lines' valves which are closed by CHP only. The control system is

designed on a two-channel (right and left) concept with redundancy and

! physicalsegaration.

Each channel is capable of-initiating containment l isolation.

The Type A test requirements including pretest test methods, test pressure, i acceptance criteria, and reporting requirements are in accordance with the

! Containment Leak Rate Testing Program."

l The frequency of the periodic integrated leak rate test is keyed to the ;

i refueling schedule for the reactor because these tests can best be performed

] during refueling shutdowns. The specified frequency is based on three major l

considerations. First is the low probability of leaks in the liner because l of (a) the test of the leak tightness of the welds during erection; (b)

{ conformance of the complete containment to a low leak rate at 55 psig during i preoperational testing which in consistent with 0.1% leakage at design basis l accident (DBA) conditions: and (c) absence of any significant stresses in the j liner during reactor operation.

1 j 4-22 l 4

! Amendment No. M9, MS i

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, 4.5 CONTAINMENT TESTS Basis (continued) )

Second is the more frequent testing, at the full accident pressure, of those portions of the containment envelope that are most likely to develop leaks during reactor operation (penetrations and isolation valves) and the low l value (0.60L.) of the total leakage that is specified as acceptable from penetrations and isolation valves. Third is the Containment Structural Integrity Surveillance Program which provides assurance that an important part, of the structural integrity of the containment is maintained.

The basis for specification of a total leakage rate of 0.60 L, from penetrations and isolation valves is specified to provide assurance that the integrated leak rate would remain within the specified limits during the intervals between integrated leak rate tests. This value allows for possible deterioration in the intervals between tests.

The basis for specification of an airlock door seal leakage rate of 0.023 L, ,

is to provide assurance that the failure of a single airlock door will not i result in the total containment leakage exceeding 0.6 L,. The seven (7) day LC0 specified for exceeding the airlock door leakage limit is acceptable i since it requires that the total containment leakage limit is not exceeded. 1 References I

(1) Updated FSAR Section 5.8.2. l l

(2) Updated FSAR Section 5.8.8 l l

(3) Updated FSAR 14.22 '

(4) Updated FSAR Section 8.5.1.2 (5) 10 CFR Part 50, Appendix J.

(6) Regulatory Guide 1.163, " Performance-Based Containment Leak-Test Program", September 1995.

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l 4-23 Amendment No. -14, M9, M6, M5,

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. 4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS l

I Surveillance Reauirements 4.6.1 Safety In.iection System l a. System tests shall be performed at each reactor refueling interval. l I A test safety injection signal will be applied to initiate l l

operation of the system. The safety injection and shutdown cooling

, system pump motors may be de-energized for this test. The system l will be considered satisfactory if control board indication and visual observations indicate that all components have received the safety injection signal in the proper sequence and timing (ie, the appropriate pump breakers shall have opened and closed, and all valves shall have completed their travel).

4.6.2 Containment Soray System

a. System test shall be performed at each reactor refueling interval. l l The test shall be performed with the isolation valves in the spray '

supply lines at the containment blocked closed. Operation of the system is initiated by tripping the normal actuation instrumentation.

b. The test will be considered satisfactory if visual observations l indicate all components have operated satisfactorily.

4.6.3 Pumps

a. The safety injection pumps, shutdown cooling pumps, and containment l spray pumps shall be started at intervals not to exceed three months. Alternate manual starting between control room console and the local breaker shall be practiced in the test program,

b. Acceptable levels of performance shall be that the pumps start, l reach their rated heads on recirculation flow, and operate for at least fifteen minutes.

4.6.4 Valves

a. Each Safety Injection Tank flow path shall be verified OPERABLE l within 7 days prior to each reactor startup by verifying each motor operated isolation valve is open by observing valve position indication and valve itself, and locking open the associated circuit breakers.

b. The Low Pressure Safety Injection flow path shall be verified l OPERABLE within 7 days prior to each reactor startup by verifying l

flow control valve CV-3006 is open, and its air supply is isolated.

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4-24 l Amendment No. M, M, %, M, W, MB, l

,' 4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS Surveillance Reauirements (continued) 1 l

l 4.6.4 Valves (continued)

c. The safety injection recirculation path shall be verified OPERABLE within 7 days prior to each reactor startup by verifying valves CV-3027 and 3056 are open and their switches HS-3027A, HS-30278, HS-3056A, and HS-3056B are open,

d. Each Containment Spray Valve manual control shall be verified to be OPERABLE at least once each refueling by cycling each valve from the control room while observing valve operation at least each 18 months.

4.6.5 Containment Air Coolina System

a. Emergency mode automatic valve and fan operation will be checked for OPERABILITY during each refueling shutdown.

b. Each fan and valve required to function during accident conditions will be exercised at intervals not to exceed three months. ;

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l l 4-25 Amendment No. 69, M, M, H7, MB,

4.6 SAFETY INJECTION AND CONTAINMENT SPRAY SYSTEMS TESTS IA111 The safety injection system and the containment spray system are principal plant safety features that are normally inoperative during reactor operation.

Complete systems tests cannot be performed when the reactor is operating because a safety injection signal causes containment isolation and a containment spray system test requires the system to be temporarily disabled.

The method of assuring OPERABILITY of these systems is therefore, to combine systems tests to be performed during annual plant shutdowns, with more frequent component tests, which can be performed during reactor operation.

The refueling interval systems tests demonstrate proper automatic operation l of the safety injection and containment spray systems. A test signal is applied to initiate automatic action and verification made that the components receive the Safety Injection Signal in the proper sequence. The test demonstrates the operation of the valves, pump circuit breakers, and automatic circuitry.o.2i During reactor operation, the instrumentation which is depended on to initiate safety injection and containment spray is generally checked each shift and the initiating circuits are tested monthly. In addition, the active components (pumps and valves) are to be tested every three months to check the operation of the starting circuits and to verify that the pumps are in satisfactory running order. The test interval of three months is based on the judgment that more frequent testing would not significantly increase the reliability (ie, the probability that the component would operate when required), yet more frequent test would result in increased wear over a long period of time.

Other systems that are also important to the emergency cooling function are the SI tanks, the component cooling system, the service water system and the containment air coolers. The SI tanks are a passive safety feature. In accordance with the specifications, the water volume and pressure in the SI tanks are checked periodically. The other systems mentioned operate when the reactor is in operation and by these means are continuously monitored for satisfactory performance.

Referencg1 (1) FSAR, Section 6.1.3.

(2) FSAR, Section 6.2.3.

(Next Page is 4-42) l 4-26 l Amendment No. W , M 1, M B,

4.14 STEAM GENERATORS SURVEILLANCI l l 4.14.1 Verify Steam Generator tube integrity is acceptable in accordance with the Inservice Inspection and Testing Program, Specification 6.5.7, and ;

the Steam Generator Tube Surveillance Program, Specification 6.5.8.

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(Next page is 4-70) l 4-66 Amendment No. H, M, M, M, 39, M, M, M, M6, MB, MB, M1,

.' 4.16 INSERVICE INSPECTION PROGRAM FOR SH0CK SUPPRESSORS (Snubbers)

. i Acolicability Applies to periodic surveillance of safety-related snubbers as described per Specification 3.20.

4.16.1 Soecifications Each snubber shall be demonstrated OPERABLE by performance of the 4 following augmented inservice inspection program in addition to the !

requirements of Specification 6.5.7. As used in this specification, l

" type of' snubber" shall mean snubbers of the same design and manufacturer, irrespective of capacity.

a. Visual Insnection l Snubbers are categorized as inaccessible or accessible during reactor operation. Each of these categories (inaccessible and accessible) may be inspected independently according to the following paragraph:

l If one or more unacceptable snubbers are found, the next inspection '

interval shall be 2/3 (-25%) of the previous interval. If no unacceptable snubbers are found, the next interval may be doubled

(-25%), but not to exceed 48 months. The interval extension I

provisions of Technical Specification 4.0.2 are applicable for all inspection intervals up to and including 48 months.

Inspections performed before the interval has elapsed may be used as a new reference point to determine the next inspection.

However, the results of such early inspections, performed before l the original required time interval has elapsed (nominal time less 25%), may not be used to lengthen the required inspection interval.

Any inspection whose results require a shorter inspection interval will override the previous schedule.

b. Visual Insoection Acceptance Criteria l

Visual inspection shall verify that (1) the snubber has no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or supporting structure are functional, and (3) fasteners for the attachment of the snubber to the component and to the snubber anchorage are functional. Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers, irrespective of type, that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Technical Specification 4.16.1d or 4.16.le, as applicable. All snubbers found connected to an inoperable common hydraulic fluid reservoir shall be counted as unacceptable for determining the next inspection interval.

4-71 Amendment No. M, 69, M7, M8, M4,

, 4.16 INSERVICE INSPECTION PROGRAM FOR SH0CK SUPPRESSORS (Snubbers)

~ l 4.16.1 f. Snubber Service Life Monitorina A record of the service life of each snubber, the date at which the 1 designated service life commences and the installation and '

maintenance records on which the designated service life is based i l shall be maintained.

Concurrent with the first inservice visual inspection and at least once per 18 months thereafter, the installation and maintenance records for each safety related snubber in use in the plant shall be reviewed to verify that the indicated service life has not been exceeded or will not be exceeded prior to the next scheduled service life review. If the indicated service life will be exceeded prior to the next scheduled snubber service life review, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned so as to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement or reconditioning shall be indicated in the records.

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4-74 Amendment No. M, 69, M, M7, M4

6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The plant superintendent shall be responsible for overall plant l operation and shall delegate in writing the succession for this responsibility during his absence.

The plant superintendent or his designee shall approve, prior to implementation, each proposed test, experiment or modification to systems or equipment that affect nuclear safety. ;

6.1.2 The Shift Supervisor (SS) shall be responsible #or the control room j command function. During any absence of the SS i.'om the control room while the plant is above COLD SHUTDOWN, an individel with an active ,

Senior Reactor Operator (SRO) license shall be desigated to assume the l control room command function. buring any absence of the SS from the control room while the plant is in COLD SHUTD9WN, an individual with an i active SR0 license or Reactor Operator (RO) license shall be designated to assume the control room command function.

l 6.2 ORGANIZATION '

6.2.1 Onsite and Offsite Oraanizations l l Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting the safety of the Palisades plant.

a. Lines of authority, responsibility and communication shall be established and defined for the highest management levels through intermediate levels to and including all operating organization positions. These relationships shall be documented, and updated, as appropriate, in the form of organization charts, functional !

descriptions of departmental responsibilities and relationships, and job descriptions for key positions, or in equivalent forms of documentation. These requirements and the plant specific equivalent of those titles referred to in these Technical Specifications shall be documented in the FSAR.

b. The plant superintendent shall be responsible for overall plant l safe operation and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

c. A specified corporate executive shall have corporate rasponsibility l for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety.

d. The individuals who train the operating staff and those who carry out radiation safety and quality assurance functions may report to l the appropriate onsite manager; however, they shall have sufficient organizational freedom to ensure their independence from operating pressures.

6-1 Amendment No. M, M, MB, B9,

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6,0 ADMINISTRATIVE CONTROLS 6.2.2 Plant Staff

a. A non-licensed operator shall be assigned to each reactor I containing fuel and an additional non-licensed operator shall be !

assigned for each control room from which a reactor is operating l above COLD SHUTDOWN.

b. At least one licensed P.eactor Operator (RO) shall be present in the control room when fuel is in the reactor. In addition, while the unit is above COLD SHUTDOWN, at least one licensed Senior Reactor Operator (SRO) shall be present in the control room. )

c. Shift crew composition may be less than the minimum requirement of 10 CFR 50.54(m)(2)(i), and 6.2.2.a and 6.2.2.g for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the requirements.

d. A radiation safety technician shall be on site when fuel is in the reacter. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position,

e. Administrative procedures shall be developed and implemented to l limit the working hours of plant staff who perform safety-related functions (e.g., licensed SR0s, licensed P.0s, radiation safety personnel, auxiliary operators, and key maintenance personnel).

. In the event that overtima is used, the following guidelines shall be followed:

3

, 1. An individual should not be permitted to work more than 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months straight, excluding shift turnover time;

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2. An individual should not be permitted to work more than l 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months in any 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, nor more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months period, nor more than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in any 7 day period, all excluding shift turnover time;

3. A break of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months should be allowed between work periods, including shift turnover time;

4. Except during extended shutdown periods, the use of overtime l should be considered on an individual basis and not for the entire staff on a shift.

4 6-2 4

Amendment No. M, M, 60, 9, M, MB, 4N, M9, MB, 1

. \

',' 6.0 ADMINISTRATIVE CONTROLS 6.2.2.e Plant Staff (Continued) l Any deviations from the overtime guidelines shall be authorized in advance by the plant superintendent or his designee, in accordance with approved administrative procedures, or by higher levels of l management, in accordance with established procedures and with I documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the plant superintendent or his designee to ensure that excessive hours have not been assigned.

Routine deviation from the above guidelines is not authorized.

f. The operations manager or an assistant operations manager shall hold an SR0 license. The individual holding the SRO license shall be responsible for directing the activities of the licensed operators. )

g. The Shift Technical Advisor (STA) shall provide advisory technical I support to the Shift Supervisor (SS) in the areas of thermal i hydraulics, reactor engineering, and plant analysis with regard to the safe operation of the unit. If either SR0 on shift satisfies i the Shift Engineer qualification requirements, then the STA does not need to be stationed.

l 6.3 PLANT STAFF OVALIFICATIONS 6.3.1 Each member of the plant staff shall meet or exceed the minimum 1 qualifications of ANSI N18.1-1971 for comparable positions, i

6.3.2 The radiatan safety manager shall meet the qualifications of a '

Radiation Protection Manager as defined in Regulatory Guide 1.8, September 1975. For the purpose of this section, " Equivalent," as utilized in Regulatory Guide 1.8 for the bachelor's degree requirement, may be met with four years of any one or combination of the following:

(a) Formal schooling in science or engineering, or (b) operational or technical experience and training in nuclear power.

6.3.3 The Shift Technical Advisor shall have a bachelor's degree or equivalent and the Shift Engineer shall have a bachelor's degree in a scientific or engineering discipline. Specific training for both the Shift Technical Advisor and the Shift Engineer shall include plant design, operations, and response and analysis of the plant for transients and accidents.

The Shift Engineer shall hold a Senior Reactor Operator license.

6.3.4 The plant staff who perform reviews which ensure compliance with 10 CFR 50.59 shall meet or exceed the minimum qualifications of ANS 3.1-1987, Section 4.7.1 and 4.7.2. A Senior Reactor Operator license or certification shall be considered equivalent to a bachelors degree for the purpose of this specification. l 6-3 l Amendment No. M, M, W, 60, G, 68, M, M8, IN, H9,

. l

,' 6.0 ADMINISTRATIVE CONTR01.S j

\

l 6.4 PROCEDURES l Written procedures shall be established, implemented, and maintained I covering the activities referenced below:

a. The applicable procedures recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, Appendix A, February 1978.

b. Refueling operations. l

c. Surveillance and test activities of safety-related equipment. l

d. Site Fire Protection Program implementation. l

e. All programs specified in Specification 6.5. l l

l 6-4 l Amendment No. M, M, 69, M, M3, M7, M4,

,' 6.0 ADMINISTRATIVE CONTROLS 1

6.5 PROGRAMS AND MANUALS l

l The following programs shall be established, implemented, and maintained:

6.5.1 Offsite Dose Calculation Manual (ODCM) l l

a. The 00CM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program; and

b. The ODCM shall also contain (1) the radioactive effluent controls and radiological environmental monitoring activities and (2) descriptions of the information that should be included in the Radiological Environmental Operating Report, and Radioactive Effluent Release Report required by Specification 6.6.2. and Specification 6.6.3.

c. Changes to 00CM: l

1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain: l

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the changes, and

b. A determination that the change will maintain the level of radioactive effluent control required by 10 CFR 20.1302, 40 CFR 190, 10 CFR 50.36a, and 10 CFR 50, Appendix I, and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations. '

2. Shall become effective after approval by the plant l superintendent.

3. Shall be submitted to the NRC in the form of a complete. l 1egible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Release Report for the period of the report in which any change to the ODCM was made. Each change shall be identified by markings in the margin of the affected pages, clearly indicating the area of the page that was changed, and shall indicate the date (e.g., month / year) the change was implemented.

6-5 l Amendment No. 85, M4,

6.0 ADMINISTRATIVE CONTROLS 6.5.2 Primary Coolant Sources Outside Containment l l This program provides controls to minimize leakage to the engineered

! safeguards rooms, from those portions of systems outside containment i that could contain highly radioactive fluids during a serious transient i or accident, to as low as practical. The systems include the

! Containment Spray System, the Safety Injection System, the Shutdown j Cooling System, and the containment sump suction piping. This program shall include the following:

i a. Provisions establishing preventive maintenance and periodic visual l

. inspection requirements, and

b. Integrated leak test requirements for each system at a frequency l

j. net to exceed refueling cycle intervals.

c. The portion of the shutdown cooling system that is outside the l containment shall be tested either by use in normal operation or hydrostatically tested at 255 psig. l

. d. Piping from valves CV-3029 and CV-3030 to the discharge of the l

. safety injection pumps and containment spray pumps shall be j hydrostatically tested at no less than 100 psig. I i

! e. The maximum allowable leakage from the recirculation heat removal l s.ystems' components (which include valve stems, flanges and pump seals) shall not exceed 0.2 gallon per minute under the normal

] hydrostatic head from the SIRW tank (approximately 44 psig).

6.5.3 Post Accident S=nlina Proaram l 1 This program provides controls which will ensure the capability to accurately determine the airborne iodine concentration in vital areas and which will ensure the capability to obtain and analyze reactor i coolant, radioactive iodines and particulates in plant gaseous effluents, and containment atmosphere samples under accident conditions.

This program shall include the following

a. Training of personnel, I

b. Procedures for sampling and analysis, and l 1

c. Provisions for maintenance of sampling and analytic equipment. I d

i 6-6 l Amendment No. 9 , 400,

l l 6.0 ADMINISTRATIVE C0i4TROLS 6.5.4 Radioactive Effluent Controls Proaram l A program shall be provided conforming with 10 CFR 50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable. The program (1) shall be contained in the Offsite Dose Calculation Manual (0DCM), (2) shall be implemented by operating procedures, and (3) shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

a. Limitations on the operability of radioactive liquid and gaseous l monitoring instrumentation including surveillance tests and setpoint determination in accordance with the methodology in the ODCM,

b. Limitations on the concentrations of radioactive material released l in liquid effluents to unrestricted areas conforming to 10 CFR 20, Appendix B, Table 2, Column 2. l

c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM,

d. Limitation on the annual and quarterly doses or dose commitment to l a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas conforming to 10 CFR 50, Appendix I, l

e. Limitations on the dose rate resulting from radioactive material l released in gaseous effluents to areas beyond the site boundary conforming to the doses associated with 10 CFR 20, Appendix B, Table 2, Column 1. l

f. Limitations on the annual and quarterly air doses resulting from l noblo gases released in gaseous effluents from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I, l

g. Limitations on the annual and quarterly doses to a member of the l public from Icdine-131, Iodine-133, tritium and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released from each unit to areas beyond the site boundary conforming to 10 CFR 50, Appendix I, l

h. Limitations on the annual doses or dose commitment to any member of I the public due to releases of radioactivity and to radiation from uranium fuel cycle sources conforming to 40 CFR 190.

6-7 l Amendment No. H4,

~

,- 6.0 ADMINISTRATIVE CONTROLS 6.5.5 Containment Structural Intearity Surveillance Proaram l This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.

The program shall include baseline measurements prior to initial operations. The Containment Structural Integrity Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3, July 1990.

The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the I Containment Structural Integrity Surveillance Program inspection frequencies.

6.5.6 Primary Coolant Puma Flywheel Surveillance Proaram l ,

Surveillance of the primary coolant pump flywheels shall consist of a l 100% volumetric inspection of the upper flywheels each refueling.

6.5.7 Inservice Inspection and Testina Proaram This program provides controls'for inservice inspection and testing of ASME Code Class 1, 2, and 3 components including applicable supports.

The program shall include the following:

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

B&PV Code terminology Required interval for inservice testing for performing inservice activities testina activities Weekly s 7 days Monthly s 31 days Quarterly or every 3 months s 92 days Semiannually or every 6 months s 184 days Every 9 months s 276 days Yearly or annually s 366 days Biennially or every 2 years s 731 days

b. The provisions of Surveillance Requirement 4.0.2 are applicable to the above required intervals for performing inservice testing activities;

c. The provisions of Surveillance Requirement 4.0.3 are applicable to inservice testing activities; and

d. Nothing in the B&PV Code shall be construed to supersede the requirements of any Technical Specification.

6-8 Amendment No.

L l .

l 6.0 ADMINISTRATIVE CONTROLS 6.5.8 Steam Generator Tube Surveillance Proaram l This program provides controls for surveillance testing of the Steam Generator (SG) tubes to ensure that the structural integrity of this portion of the Primary Coolant System (PCS) is maintained. The program shall contain controls to ensure:

a. Steam Generator Tube Samole Selection and Insoection l i The inservice inspection may be limited to one SG on a rotating schedule encompassing 6% of the tubes if the results of previous inspections indicate that both SGs are performing in a like manner.

If the operating conditions in one SG are found to be more severe than those in the other SG, the sample sequence shall be modified to inspect the most severe conditions.

The SG tube minimum sample size, inspection result classification,

! and the corresponding action required shall be as specified in 1 l Table 6.5.8-1. The tubes selected for each inservice inspection l l shall include at least 3% of the total number of tubes in all SGs; the tubes selected for these inspections shall be selected on a random basis except:

, 1. blhere experience in similar plants with similar water l chemistry indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from these critical areas.

2. The first sample of tubes selected for each inservice l inspection of each SG shall include

a) All nonplugged tubes that previously had detectable wall l .

penetrations greater than 20%.

b) Tubes in those areas where experience has indicated l potential problems.

c) A tube inspection shall be performed on each selected l I tube. If any selected tube does not permit the passage of the eddy current probe for a tube inspection, this shall be recorded and an adjacent tube shall be selected ,

and subjected to a tube inspection. 1 i

1 6-9 l

Amendment No. H, M, M, M, 49, M, M, M, M6, He, MG, M1

6.0 ADMINISTRATIVE CONTROLS l

l 6.5.8 Steam Generator Tube Surveillance Procram (continued) l l

3. The tubes selected as the second and third samples (if required by Table 6.5.8-1) during each inservice inspection may be subjected to a partial tui:e inspection provided:

a) The tubes selected for these samples include the tubes l from those areas of the tube sheet 7.rray where tubes with imperfections were previously found.

b) The inspections include those portions of the tubes where l imperfections were previously found.

4. The results of each sample inspection shall be classified into l-one of the following three categories:

Cateaory Insoection Results C-1 Less than 5% of the total tubes inspected are degraded tubes and none of the inspected tubes are defective.

C-2 One or more tubes, but not more than 1% of the total tubes inspected are defective, or between 5% and 10% of the total tubes inspected are degraded tubes.

1 More than 10% of the total tubes inspected are degraded C-3 tubes or more than 1% of the inspected tubes are defective.

Note: In all inspections, previously degraded tubes must exhibit significant (greater than 10%) further wall penetrations to be included in the above percentage calculations.

b. Insoection Freauencies l l The above required inservice inspection of SG tubes shall be i

4 performed at the following frequencies:

l 1. Inservice inspections shall be performed at intervals of not l L less than 12 nor more than 24 calendar months after the

! previous inspection. If two consecutive inspections following I' service under AVT conditions, not including the preservice

[

inspection, result in all inspections results falling into the C-1 category or if two consecutive inspections demonstrate i that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval j may be extended to a maximum of once per 40 months.

i i

6-10 Amendment No. 39, 46, M, M6, 4M, MG, W A

4

6.0 ADMINISTRATIVE CONTROLS 6.5.8 Steam Generator Tube Surveillance Proaram (continued) l I 2. If the results of the inservice inspection of a SG conducted in accordance with Table 6.5.8-1 at 40 month intervals fall into. Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy ;

the criteria of Specification 6.5.8.b.1; the interval may then l l be extended to a maximum of once per 40 months.

l

3. Additional, unscheduled inservice inspections shall be l performed on each SG in accordance with the first sample inspection specified in Table 6.5.8-1 during the shutdown 1 i subsequent to any of the following conditions: !

a) Primary-to-secondary tube leaks (not including leaks l l originating from tube-to-tube sheet welds) in excess of i the limits of Specification 3.1.5. '

b) A seismic occurrence greater than the Operating Basis l Earthquake.

c) A loss-of-coolant accident resulting in initiation of l flow of the engineered safeguards.

d) A main steam line or main feedwater line break. l

c. Acceptance Criteria l

1. As used in this Specification: l ,

a) Imoerfection means an exception to the dimensions, finish l :

or contour of a tube from that required fabrication drawings or specifications. Eddy-current testing i indications below 20% of the nominal tube wall thickness, l if detectable, may be considered as imperfections. ,

b) Hearadation means a service-induced cracking, wastage, l wear or general corrosion occurring on either inside or outside of a tube, c) Dearaded Tube means a tube containing imperfections l greater than or equal to 20% of the nominal wall thickness caused by degradation.

d) % Dearadation means the percentage of the tube wall l thickness affected or removed by degradation.

6-11 Amendment No. 39, M, M6, 4H, MB, 44

. 6.0 ADMINISTRATIVE CONTROLS

't l 6.5.8 Steam Generator Tube Surveillance Proaram (continued) l i e) Defect means an imperfection of such severity that it l

! exceeds the plugging limit. A tube containing a defect j is defective.

i f) Pluacina limit means the imperfection depth at or beyond l i which the tube shall be removed from service and is equal l to 40% of the nominal tube wall thickness.

j g) Unserviceable described the condition of a tube if it l leaks or contains a defect large enough to affect its 4

structural integrity in the event of an Operating Basis Earthquake, a loss-of-coolant accident, or a steam line 1[ or feedwater line break as specified in 6.5.8.b.3, above. .l

( h) Tube Insoection means an inspection of the SG tube from l l the point of entry (hot leg side) completely around the j U-bend to the top support of the cold leg.

1) Preservice Insoection means an inspection of the full l

. length of each tube in SG performed by eddy current '

i techniques prior to service to establish a baseline

! condition of the tubing. This inspection shall be i

i performed after the shop hydrostatic test and prior to initial POWER OPERATION using the equipment and techniques expected to be used during subsequent j

inservice inspections.

I 2. The SG shall be determined OPERABLE after completing the l I

, corresponding actions (plug all tubes exceeding the plugging i limit and all tubes containing through-wall cracks) required j by Table 6.5.8-1. l i

I i l I

i i

! 6-12 l

Amendment No. 39, 64, M6, W, IBB, M1

l TABLE 6.So8-1 STEAM GENERATOR TUBE INSPECTION ,

IST SAMPLE INSPECTION 2ND SAMPLE INSPECTION 3RD SAMPLE INSPECTION ;

Sample Size Result Action Required Result Action Required Result Action Required i l

A minimum of C-1 None N/A N/A N/A N/A S Tubes per t S.G. s C-2 Plug defective tubes C-1 None N/A N/A !

and inspect additional <

2S tubes in this S.G. C-2 Plug defective tubes C-1 None and inspect additional 4S tubes in this S.G. C-2 Plug defective tubes l C-3 Perform action for C-3 result of first Sample "

C-3 Perform action for C-3 result of first N/A N/A Sample C-3 Inspect all tubes in All other None N/A N/A this S.G., plug de- S.G.s are i fective tubes and C-1 inspect 2S tubes in each other S.G. Some S.G.s Perform action for N/A N/A C-2 but no C-2 result of second additional sample 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal S.G. are notification to NRC C-3 with written follow up within next Additional Inspect all tubes 30 days S.G. is each S.G. and plug C-3 defective tubes. N/A N/A l S - 6/n % Where n is the number of steam generators inspected during an inspection 6-13 Amendment No. 444-

l 6.0 ADMINISTRATIVE CONTROLS 6.5.9 Secondary Water Chemistry Proaram l

A program shall be established, implemented and maintained for monitoring of secondary water chemistry to inhibit steam generator tube degradation and shall include:

a. Identification of a sampling schedule for the critical variables l and control points for these variables,

b. Identification of the procedures used to measure the values of the l critical variables,

c. Identification of process sampling points, which shall include l monitoring the discharge of the condensate pumps for evidence of condenser in-leakage,

d. Procedures for the recording and management of data, l

e. Procedures defining corrective actions for all off-control point l chemistry conditions, and

f. A procedure identifying (a) the authority responsible for the l interpretation of the data, and (b) the sequence and timing of administrative events required to initiate corrective actions.

6-14 Amendment No.

6.0 ADMINISTRATIVE CONTROLS

~

6.5.10 Ventilation Filter Testina Proaram A program shall be established to implement the following required and Fuel Pool Ventilation testing of Control (FPV systems Room at the Ventilation frequencies (CRV)d in Regulatory Guide 1.52, specifie Revi)s ion 2 (RG 1.52), and in accordance with RG 1.52 and ASME N510-1989 at the system flowrates and tolerances specified below*:

a. Demonstrate for each of the ventilation systems that an inplace test of the high efficiency particulate air penetration and system bypass < 0.05% for the(HEPA) CRV andfilters shows

< 1.00% for a the FPV when tested in accordance with RG 1.52 and ASME N510-1989:

Ventilation Systgm Flowrate (CFM)

V-8A or V-8B 7300 1 20%

V-8A and V-8B 10,000 1 20%

V-95 or V-96 12,500 1 10%

b. Demonstrate for each of the ventilation systems that an inplace test of the charcoal adsorber shows a penetration and system bypass

< 0.05% for the CRV and < 1.00% for the FPV when tested in accordance with RG 1.52 and ASME N510-1989.

Ventilation System Flowrate (CFM)

V-8A and V-8B 10,000 1 20%

V-26A and V-26B 3200 +10% -5%

c. Demonstrate for each of the ventilation systems that a laboratory test of a sam le of the charcoal adsorber when obtained as describedinkG1.52showsthemethyliodidepenetrationlessthan I the value specified below when tested in accordance with !

ASTM D3803-1989 at a temperature of s 30*C and equal to the relative humidity specified as follows:

Ventilation System Penetration Relative Humidity VF-66 6.00% 95%

VFC-26A and VFC-268 0.157% 70%

d. For each of the ventilation 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 RG 1.52 and ASME N510-1989:

Ventilation System Delta P (In H 291 Flowrate (CFM)

V-8A and V-8B 6.0 10,000 1 20%

VF-26A and VF-268 8.0 3200 +10% -5%

e. Demonstrate that the heaters for each of the ventilation systems dissipate the following specified value i 20% when tested in accordance with ASME N510-1989:

Ventilation System Wattaae VHX-26A and VHX-268 15 kW The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Ventilation Filter Testing Program frequencies.

Should the 720-hour limitation on charcoal adsorber operation occur during a plant operation requiring the use of the charcoal adsorber . such as refueling testing may be delayed until the completion of the plant operation or up to 1,500 hours0.00579 days 0.139 hours 8.267196e-4 weeks 1.9025e-4 months of filter operation; whichever occurs first.

6-15 Amendment No.

l

,- 6.0 ADMINISTRATIVE CONTROLS 6.5.11 Reserved l l

6.5.12 Technical Soecifications (TS) Bases Control Proaram 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 involve 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 involves an unreviewed safety question as defined in 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the FSAR.

d. Proposed changes that meet the criteria of Specification 6.5.12 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 I with 10 CFR 50.71(e). ,

6-16 Amendment No. 440,

l 6.0 ADMINISTRATIVE CONTROLS l 6.5.13 Reserved 1

I l

l 6.5.14 Containment Leak Rate Testina Procram Programs shall be established to implement the leak 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. The Type A test program shall meet the requirements of 10 CFR 50, Appendix J, Option B and shall be in accordance with the guidelines of Regulatory Guide 1.163,

" Performance-Based Containment Leakage-Test Program, dated September 1995." The Type B and Type C test program shall meet the requirements of 10 CFR 50, Appendix J, Option A, as modified by the exemption from !

certain requirements of 10 CFR 50 Appendix J which was granted in an NRC letter to Consumers Power Company dated December 6, 1989.

The peak calculated containment internal pressure for the design basis loss of coolant accident, P., is 52.64 psig (FSAR Table 14.18.1-4).

The maximum allowable containment leak rate, L., at P., shall be 0.1% of containment air weight per day.

Leak rate acceptance criteria are:

a. Containment leak rate acceptance criteria is s 1.0 L,. During the first plant startup following testing in accordance with this program, the leak rate acceptance criteria are s 0.60 L, for the Type B and Type C tests and s 0.75 L, for Type A tests;

b. Air lock leak rate acceptance criteria is s 0.023 L, for each door, when pressurized to ;t 10 psig.

The Surveillance interval extensions of LC0 4.0.2 are not applicable to the Containment Leak Rate Testing Program requirements.

The provisions of LC0 4.0.3 ACg applicable to the Containment Leak Rate Testing Program requirements.

6-17 Amendment No.

6.0 ADMINISTRATIVE CONTROLS 6.5.15 Process Control Proaram l

a. The Process Control Program shall contain the current formula, l sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20,10 CFR 71, Federal and State regulations, and other requirements governing the disposal of the radioactive waste.

b. Changes to the Process Control Program: l

1. Shall be documented and records of reviews performed shall be retained as required by the Quality Program, CPC-2A. This I documentation shall contain:

a) Sufficient information to support the change together with the appropriate analyses or evaluation justifying the change (s) and b) A determination that the change will maintain the overall conformance of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

2. Shall become effective after approval by the plant superintendent.

i i 4 l

1 i

i i

i 6-18 l Amendment No. 85, 164, 162, l

~

l 6.0 ADMINISTRATIVE CONTROLS 6.6 REPORTING REQUIREMENTS d

l l I

The following reports shall be submitted in accordance with 10 CFR 50.4. l 1

l l 6.6.1 Occupational Radiation Exoosure Report l l

j This report shall include a tabulation on an annual basis of the number l l of station, utility and other personnel (including contractors)

! receiving exposures greater than 100 mrenVyear and their associated i man rem exposure according to work and job functions (e.g., reactor

operations and surveillance, inservice inspection, routine maintenance, '

! special maintenance [ describe maintenance], waste processing and i

refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignment to various duty functions may be

, estimates based on pocket dosimeter, electronic dosimeter, TLD, or film 2 badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources shall be assigned to specific major work functions. The report shall be submitted by April 30 of each year.

6.6.2 Radioloaical Environmental Doeratina Report l The Radiological Environmental Operating Report covering the operation I of the unit during the previous calendar year shall be submitted before l May 15 of each year. The report shall include summaries, l interpretations, and analysis of trends of the results of the radiological environmental monitoring program for the reporting period. l The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual and in 10 CFR 50, Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

6.6.3 Radioactive Effluent Release Reoort l The Radioactive Effluent Release Report shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual and Process Control Program, and shall be in conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I, Section IV.B.I.

6.6.4 Monthly Ooeratina Report l Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the NRC to arrive no later than the fifteenth of each month following the calendar month covered by the report.

6-19 Amendment No. M, M, M, 85, M8, M4,

6.0 ADMINISTRATIVE f.0NTROLS 6.6.5 Core Operatina Limits Report (COLR) l

a. Core operating limits shall be established prior to each reload l cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

3.1.1 ASI Limits.

3.10.5 Regulating Group Insertion Limits 3.23.1 Linear Heat Rate (LHR) Limits 3.23.2 Radial Peaking Factor Limits

b. The analytical methods used to determine the core operating limits l shall be those approved by the NRC, specifically those described in the latest approved revision of the following documents:

1. XN-75-27(A), " Exxon Nuclear Neutronics Design Methods for l l Pressurized Water Reactors," and Supplements 1(A), 2(A),

3(P)(A), 4(P)(A), and 5(P)(A); Exxon Nuclear Company. 1 (LCOs 3.1.1, 3.10.1, 3.10.5, 3.23.1, & 3.23.2) l

2. ANF-84-73(P)(A), " Advanced Nuclear Fuels Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events,"

and Appendix B(P)(A) and Supplements 1(P)(A), 2(P)(A); !

Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.10.5, 3.23.1, & 3.23.2)

3. XN-NF-82-21(P)(A), " Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," ;

Exxon Nuclear Company. (LCOs 3.1.1, 3.23.1, & 3.23.2)

4. ANF-84-093(P)(A), "Steamline Break Methodology for PWRs," and l Supplement 1(P)(A); Advanced Nuclear Fuels Corporation.

(LCOs3.10.1,3.10.5,3.23.1,&3.23.2)

5. XN-75-32(P)(A), " Computational Procedure for Evaluating Fuel Rod Bowing," and Supplements 1(P)(A), 2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company. (LCOs 3.1.1, 3.10.5, 3.23.1, )

& 3.23,2) ;

6. EXEM PWR Large Break LOCA Model as defined by:

(LCOs 3.10.5, 3.23.1, & 3.23.2) a) XN-NF-82-20(A), " Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," and Supplements 1(P)(A),

2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company.

b) XN-NF-82-07(P)(A), " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company.

c) XN-NF-81-58(A), "RODEX2 Fuel Rod Thermal-Mechanical l Response Evaluation Model," and Supplements 1(P)(A),

2(P)(A), 3(P)(A), and 4(P)(A); Exxon Nuclear Company.

6-20 '

Amendment No. M9,

,- 6.0 ADMINISTRATIVE CONTROLS ,

6.6.5 1018 (continued) l d) XN-NF-85-16(A), "PWR 17x17 Fuel Cooling Tests Program,"

Volume 1 and Supplements 1(P)(A), 2(P)(A), and 3(P)(A),

and Volume 2 and Supplement 1(P)(A); Exxon Nuclear Company.

e) XN-NF-85-105(A), " Scaling of FCTF Based Reflood Heat ;

Transfer Correlation for other Bundle Designs," and Supplement 1(P)(A); Exxon Nuclear Company.

7. XN-NF-78-44(A), "A Generic Analysis of the Control Rod i

' Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company. (LCOs 3.10.5, 3.23.1, & 3.23.2)

8. ANF-1224(P)(A), " Departure from Nucleate Boiling Correlation l for High Thermal Performance Fuel," and Supplement 1(P)(A);

Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.23.1, &

3.23.2)

9. ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.10.5, 3.23.1, &

3.23.2)

10. EMF-92-153(P)(A),"HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power

. Corporation. (LCOs3.1.1,3.23.1,&3.23.2)

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

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

6.6.6 Reserved l t 6.6.7 Accident Monitorina Instrument Report When a report is required by Condition 3.17.4.7c, " Accident Monitoring Instrumentation," a report shall be submitted within the following 30 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 to OPERABLE status.

6-21 Amendment No.

,- 6.0 ADMINISTRATIVE CONTROLS l

l l

6.6.8 Containment Structural Intearity Surveillance Report l

j Reports shall be submitted to the NRC covering Prestressing, Anchorage, and Liner and Penetration tests within 90 days after completion of the tests. l 6.6.9 Steam Generator Tube Surveillance Report l

The following reports shall be submitted to the Commission following 4

each inservice inspection of steam generator tubes:

a. The number of tubes plugged in each steam generator shall be l

reported to the Commission within 15 days following the completion '

of each inspection, and

b. The complete results of the steam generator tube inservice I inspection shall be reported to the Commission within 12 months following completion of the inspection. This report shall include:

1. Number and extent of tubes inspected.

2. Location and percent of wall-thickness penetration for each indication of an imperfection.

3. Identification of tubes plugged.

c. Results of steam generator tube inspections that fall into Category C-3 shall require 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months verbal notification to the NRC prior to resumption of plant operation. A written followup within the next 30 days shall provide a description of investigations and corrective measures taken to prevent recurrence.

i 1

1 1

6-22 Amendment No. M B, M9,

6.0 ADMINISTRATIVE CONTROLS

~

6.7 HIGH RADIATION AREA 6.7.1 Pursuant to 10 CFR 20, paragraph 20.1601(c), in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as defined in 10 CFR 20, in which the intensity of radiation is > 100 mrem /hr but < 1000 mrem /hr, shall be barricaded and conspicuously posted as a high radiation area i and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g., health physics technicians) or personnel continuously escorted by such individuals may be exempt from the RWP l

issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates < 1000 arem/hr, provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

l Any individual or group of individuals permitted to enter such areas l shall be provided with or accompanied by one or more of the following:

l

a. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

b. A radiation monitoring device that continuously integrates the l radiation dose rate in the area and alarms when a preset i integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

c. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified by the Radiation Work Request.

6.7.2 In addition to the requirements of Specification 6.7.1, except as allowed by 6.7.3, areas with radiation levels 21000 mrem /hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of the stay time specification of the RWP, direct or remote (such as closed circuit TV cameras) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

6.7.3 For individual high radiation areas with radiation levels of 2 1000 )

ares /hr, accessible to personnel, that are located within large areas i such as reactor containment, where no enclosure exists for purposes of j

locking, or that cannot be continuously guarded, and where no enclosure l l can be reasonably constructed around the individual area, that i individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

6-23 Amendment No. 48, M 8, M4, l

e 4

1 j ATTACHMENT 3 i

j CONSUMERS POWER COMPANY PALISADES PLANT DOCKET 50-255 1

i 1

l

TECHNICAL SPECIFICATION CHANGE REQUEST ADMINISTRATIVE CONTROLS

+

3 ADDITIONAL CHANGES Revised Pages Marked to Show Changes From Prior Submittal 2

i I

i 10 Pages l

/ 4.0 SURVEILLANCE RE0VIREMENTS 4

4.0.1 Surveillance requirements shall be applicable during the reactor operating conditions associated with individual Limiting Conditions for Operation unless otherwise stated in an individual surveillance

! requirement.

t

, 4.0.2 0:1::: Oth:rwi:: ::::if t:d, :::h : rv:ill:::: r:;;ir::: t :h:11 bc

;;rf:rn:d within t:: :;;;ifi;d ti : int:re:1 with:

} :.

A :::t r, :110:251:

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t:::ica :t t: :::::d 25% of th: :erveill::::

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b. A t:t:1 :::i tr ::: tin:d int:rv:1 ti:: f:r : y thr:: :::::: tiv:

: rveill:::: int:rv:1: ::t t: :::::d 2.25 ti :: th :p::ified l :er :111:::: inter;:1.

14[0Xf2tiif$sHU1111H&E'RsistNiiiehCsh~all! tie"p~~e7fidiid iiritiili'ths'spidtfisd j

4 survatilance'_

15 patcent_of the suryet))ance interval with~a[ changed' jnterval.: maximum allowable by^ Amendment ^ 171)extension i

j 4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined by Specification 4.0.2, shall constitute 1

noncompliance with the operability requirements for a Limiting Condition for Operation. The time limits of the action requirements are applicable at the time it is identified that a Surveillance Requirement has not been performed. The action requirements may be delayed for up j to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to permit the completion of the surveillance when the

allowable outage time limits of the action requirements are less than

, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Surveillance Requirements do not have to be performed on inoperable equipment.

4.0.4 Entry into a reactor operating condition or other specified condition j shall not be made unless the Surveillance Requirements associated with a Limiting Condition of Opt ation has been performed within the stated l surveillance interval or as otherwise specified. This provision shall

! not prevent passage through or to plant conditions as required to comply

with action requirements.

1 i

i 1

i l

l 4

i 1

)

i l Amendment No. 30, M , B 0, M 2, p ,

1 i 4-1 I

i f 4.0 BASIS Specifications 4.0.1 through 4.0.4 establish the general requirements applicable to Surveillance Requirements. These reauirements are based on the Surveillance requirements stated in the code of Fedoral Regulations, 10 CFR 50.36(c)(3).

J

" Surveillance reauirements are requirements relating to test, cal bration, or inspection to ensure that the necessary qua'ity of systems and components is maintained that Tacility operation will be within safety limits, and that the limiting conditions of operation will be met."

l Specification 4.0.1 establishes the requirement that surveillances must be l performed during reactor operatino conditions or other conditions for which the requirements of the limiting Conditions for Operation a unless otherwise stated in an individual Surveillance Requirement.pply,The purpose of I this specification is to ensure that surveillances are performed to verify l the operational status of systems and components and that parameters are l within specified limits to ensure safe operation of the facility when the ;

plant is in a reactor operatina condition or other specified condit on for '

which the associated Limiting Conditions for Operation are applicab'e.

Surveillance Requirements do not lave to be performed when the facility is in an operational condition for whic1 the requirements of the associated Limiting Condition for Operation do not apply unless otherwise specified.

The Surveillance Requirements associated with,a Special Test Exception are only applicable when the Special Test Exception is used as an allowable exception the requirements of a specification.

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Specification 4.0.3 establishes the failure to perform a Surveillance defined by the Requirement within the allowed surveillance as a condition interval,t tha constitutes a failure provisions to meet the of Specification operability reauiremen4.0.2,ts for a Limiting Condition for Operation.

Under the provisions of this specification, systems and components are assumed to be operable when Surveillance Requ'irements have Amendment No. M G, M B, H f, 4-3

4.1 OVERPRESSURE PROTECTION SYSTEM TESTS l

)

) '

Surveillance Reauirements l

l i In addition to the requirements of The Inservice Inspection and Testing !

1 Program, Specification 6.5.7, each PORV flow path shall be demonstrated OPERABLE by:

1. Testing the PORVs in accordance with the inservice inspection i

requirements for ASME Boiler and Pressure Vessel Code,Section XI, j

4 Section IWV, Category B valves.

l 2. Performance of a CHANNEL CALIBRATION on the PORV actuation channel at least once per 18 months.* [ Changed by Amendment 171]

i

3. When the PORV flow path is required to be OPERABLE by Specification 3.1.8.1:

(a. Performing a complete cycle of the PORV with the plant above COLD SHUTDOWN at least once per 18 months.

(b. Performing a complete cycle of the block valve prior to heatup from COLD SHUTDOWN, if not cycled within 92 days.

4. When the PORV flow path is required to be OPERABLE by Specification !

3.1.8.2:

(a. Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, at least once per i 31 days.

(b. Verifying the associated block valve is open at least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

5. Both High Pressure Safety Injection pumps shall be verified incapable of injection into the PCS at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , unless the reactor head is removed, when either PCS cold leg temperature is < 300*F, or when both shutdown cooling suction valves, M0-3015 and M0-3016, are open.

Basis With the reactor vessel head installed when the PCS cold leg temperature is less than 300*F, or if the shutdown cooling system isolation valves H0-3015 and M0-3016 are open, the start of one HPSI pump could cause the Appendix G or the shutdown cooling system pressure limits to be exceeded; therefore, both pumps are rendered inoperable.

For Cycl: 11 :nly, thi: : rv:ill: :: n::d :t be perf :d until prior t: :t:rtup f:r Cy:1: 12. [ Changed by Amendment 171)

Amendment No. 440, 449, MG, Ma, M3, M4, $,

4-6

l l

.' 4.2 EQUIPMENT SAMPLING AND TESTS Table 4.2.3 VENTILATION SYSTEM TESTS The Control Room Ventilation and Isolation System and the Fuel Storage Area HEPA/ Charcoal Exhaust System shall be demonstrated to be OPERABLE by the following tests: I l

1. Performing required Control Room Ventilation and Fuel Storage Area filter testing in accordance with the Ventilation Filter Testing Program.

2. At least once per refueling cycle by:

a. Verifying that on a containment high-pressure and high-radiation test signal, the Control Room Ventilation system automatically switches into the emergency mode of operation with flow through the HEPA filter and charcoal adsorber bank.

b. Verifying that the Control Room Ventilation system maintains the Control Room at a positive pressure 21/8 inch WG relative to the outside atmosphere during system emergency mode operation. ,

I C" hit

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1

3. Verifying that the Control Room temperature is s 90*F; once per i 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4. Y:rifying th:t the Nel P;;l Y: til:ti= Sy:t= i: OPEPJaLE by initi ting fl= thr=gh the P.EM filter =d th: rec 1 2d:crber: fr=

th: :=trcl rc=.

I5N Amendment No. 81, M2,

~

1 I

l

.' 4.5 CONTAINMENT TESTS (continued)

'4

, Basis 4

i The containment is designed for an accident pressure of 55 psig."'

! While the reactor is operating, the internal environment of the containment i~ will be air.at approximately atmospheric pressure and a temperature of about 104*F. With these initial conditions, following a LOCA, the temperature of

the steam-air mixture at the peak accident pressure of 55 psig is 283*F.

! Prior to initial operation, the containment was strength-tested at 63 psig

! and then leak rate tested. The design objective of this preoperational leak i rate' test was established as 0.1% by weight per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at 55 psig. This l 1eakage rate is consistent with the construction of the containment ~, which i is equipped with independent leak-testable penetrations and contains channels 4

over all unaccessible containment liner welds, which were independently leak-

) tested during construction.

Accident analyses have been performed on the basis of a leakage rate of l 0.1% by weight per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With this leakage rate and with a reactor power i level of 2530 MWt, the potential public exposure would be below 10 CFR 100 l guideline values in the event of the Maximum Hypothetical Accident.

The performance of a periodic integrated leak rate test during plant life

provides a current assessment of potential leakage from the containment in i case of an accident that would pressurize the interior of the containment.

{ In order to provide a realistic appraisal of the integrity of the containment i' under accident conditions, this periodic leak rate test is to be performed without preliminary leak detection surveys or leak repairs and containment l

{

isolation valves are to be closed in the normal manner.

I b This normal manner is a coincident two-of-four high radiation or two-of-four i high containment pressure signals which will close all containment isolation valves not required for engineered safety features except the component j, cooling lines' valves which are closed by CHP only. The control system is j designed on a two-channel (right and left) concept with redundancy and '

4 physical segaration. Each channel is capable of initiating containment i isolation. '

i The Type A test requirements including pretest test methods, test pressure, acceptance criteria, and reporting reg ContainmentLeakRateTestingProgram.grementsareinaccordancewiththe

The frequency of the periodic integrated leak rate test is keyed to the j refueling schedule for the reactor because these tests can best be performed

. during refueling shutdowns. The specified frequency is based on three major i j considerations. First is the low probability of leaks in the liner because '

of (a) the test of the leak tightness of the welds during erection; (b) conformance of the complete containment to a low leak rate at 55 psig during preoperational testing which in consistent with 0.1% leakage at design basis i accident (DBA) conditions: and (c) absence of any significant stresses in the

! liner during reactor operation.

j 4-22 i Amendment No. M9, ME l

a

~

} , 6.0 ADMINISTRATIVE CONTROLS 2

6.5.2 Primary Coolant Sources Outside Containment

This program provides controls to minimize leakage to the engineered

! safeguards rooms, from those portions of systems outside containment 3 that could contain highly radioactive fluids during a serious transient

, or accident, to as low as practical. The systems include the

! Containment S end-the Safety Injection system i $isMid3yA ~ ystesl in:L. ding the containment sump suctio$)GI56iiWiiiiln l program shall e the following:

4

a. Provisions establishing preventive maintenance and periodic visual inspection requirements, and

b. Integrated leak test requirements for each system at a frequency not to exceed refueling cycle intervals.

i c. The portion of the shutdown cooling system that is outside the

containment shall be tested either by use in normal operation or j hydrostatically tested at 255 psig.

d. Piping from valves CV-3029 and CV-3030 to the discharge of the

safety injection pumps and containment spray pumps shall be i hydrostatically tested at no less than 100 psig.

}

e. The maximum allowable leakage from the recirculation heat removal i systems' components (which include valve stems, flanges and pump i seals) shall not exceed 0.2 gallon per minute under the normal

hydrostatic head from the SIRW tank (approximately 44 psig).

i 6.5.3 Post Accident Samnlina Proaram i

l This program provides controls which will ensure the capability to i

, accurately determine the airborne iodine concentration in vital areas

and which will ensure the capability to obtain and analyze reactor i coolant, radioactive iodines and particulates in plant gaseous i effluents, and containment atmosphere samples under accident conditions.

This program shall include the following:

l a. Training of personnel,

b. Procedures for sampling and analysis, and f c. Provisions for maintenance of sampling and analytic equipment.

4 i

1 a

l 6-6 i Amendment No. 9, MG, i

.? 6.0 A M TIVE CONTROLS e

6.5.5 Containment Structural Intearity Surveillance Proaram This program provides controls for monitoring any tendon degradation in pre-stressed concrete containments, including effectiveness of its corrosion protection medium, to ensure containment structural integrity.

The program shall include baseline measurements prior to initial operations. The Containment Structural Integrity Surveillance Program, inspection frequencies, and acceptance criteria shall be in accordance with Regulatory Guide 1.35, Revision 3, M 89 @ M 996]

The provisions of Specifications 4.0.2 and 4.0.3 are applicable to the Containment Structural Integrity Surveillance Program inspection frequencies.

6.5.6 Primary Coolant Pumn F1vwheel Surveillance Proaram Surveillance of the primary coolant pump flywheels shall consist of a 100% volumetric inspection of the upper flywheels each refueling.

6.5.7 Inservice Insoection and Testina Proaram This program provides controls for inservice inspection and testing of ASME Code Class 1, 2, and 3 components including applicable supports.

The program shall include the following:

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

B&PV Code terminology Required interval for inservice testing for performing inservice activities testina activities Weekly s 7 days )

Monthly s 31 days Quarterly or every 3 months s 92 days j Semiannually or every 6 months s 184 days Every 9 months s 276 days Yearly or annually s 366 days Biennially or every 2 years s 731 days

b. The provisions of Surveillance Requirement 4.0.2 are applicable to the above required intervals for performing inservice testing activities;

c. The provisions of Surveillance Requirement 4.0.3 are applicable to inservice testing activities; and

d. Nothing in the B&PV Code shall be construed to supersede the requirements of any Technical Specification.

6-8 Amendment No.

.' 6.0 ADMINISTRATIVE CONTROLS

. 6.5.13 Reserved 6.5.14 Containment Leak Rate Testina Proaram l Programs shall be established to implement the leak 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. The Type A test program shall meet the requirements of 10 CFR 50, Appendix J, Option B and shall be in accordance with the guidelines of Regulatory Guide 1.163,

" Performance-Based Containment Leakage-Test Program, dated September 1995." The Type B and Type C test program shall meet the requirements of 10 CFR 50, Appendix J, Option A, as modified by the exemption from certain requirements of 10 CFR 50 Appendix J which was granted in an NRC letter to consumers Power Company dated December 6, 1989.

The :: t:in=nt i; d::!gned f;r = :: id=t prc=ure, P., cf 55 p;i9t !

th: =xt r ;11:=ble :=t&=nt l=k r:te, L., :t P., 05:11 be 0.1% ef

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$9!! i:::..: ..jgegh,tppgj:q.j:.ayj Leak rate acceptance criteria are:

a. Containment leak rate acceptance criteria is s 1.0 L,. During the ,

first plant startup following testing in accordance with this program, the leak rate acceptance criteria are s 0.60 L, for the Type B and Type C tests and s 0.75 L, for Type A tests;

5. ^ir 10th tuting = cept =ce criteri: cre:

1) Over:ll ir leek luk r:t: i: ; 0.50 L, wh= t=ted :t i P,+

2) Br =ch dur, luk r;te b ; 0.023 L, wh= prc=urized t:

10 p:!g.

31N ESIIAEENIN k$$2M5d@tssM[@ijenM!islOlig$NNISII5iM55bN3($f5NNE30Ab 89pp i The Surveillance interval extensions of LC0 4.0.2 are not applicable to the Containment Leak Rate Testing Program requirements.

1 The provisions of LC0 4.0.3 Art applicable to the Containment Leak Rate Testing Program requirements.

6-17 Amendment No.

i I

i

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si 6.0 ADMINISTRATIVE CONTROLS 6.6 REPORTING RE0VIREMENTS The following reports shall be submitted in accordance with 10 CFR 50.4.

6.6.1 Occuoational Radiation Exoosure Reoort This report shall include a tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 arem/ year and their associated man rem exposure according to work and job functions (e.g., reactor operations and surveillance, inservice inspection, routine maintenance,

, special maintenance [ describe maintenance], waste processing and refueling). This tabulation supplements the requirements of 10 CFR 20.2206. The dose assignment to various duty functions may be

estimates based on pocket dosimeter, electronic dosimeter, TLD, or film j badge measurements. Small exposures totaling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources

shall be assigned to specific major work functions. The report shall be j submitted by April 30 of each year.

J 6.6.2 Radioloaical Environmental Ooeratina Reoort 4 The Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 15 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the i radiological environmental monitoring program for the reporting period.

The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (0DCM) and in 10 CFR 50,

Appendix I, Sections IV.B.2, IV.B.3, and IV.C.

6.6.3 Radioactive Effluent Release Report i

The Radioactive Effluent Release Report shall be submitted in accordance

, with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material rovided shall be consistent with the objectives outlined in the OffsifiDissiCilssiiMs MshGil i shdIPfoEsssiC6ntFo1ItPraiFsiii? an'd~ihi1r brTi^c6hrersin[ee 10~CFR"50 36Tihd*10"CFR"50, Appendix I,Section IV.B.1.

6.6.4 Monthly Doeratino Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the NRC to arrive no later than the fifteenth of each month following the calendar month covered by the report.

6-19 Amendment No. M, M, M, M, MB, M4,

ML20117K958

Text

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