ML20206K362

Revised McGuire Nuclear Station Selected Licensee Commitments (SLC) Manual

ML20206K362
Person / Time
Site:	McGuire, Mcguire
Issue date:	04/30/1999
From:	DUKE POWER CO.
To:
Shared Package
ML20206K360	List: ML20206K356 ML20206K362
References
PROC-990430, NUDOCS 9905130143
Download: ML20206K362 (94)

Text

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McGuire Nuclear Station Selected Licensee Commitments j List of Effective Pages Page Revision Date IOEP Tab List of Effective Pages 28 Tab 16.Q 16.0-1 4/99 Tab 16.1 l l 16.1-1 Tab 16.2 ,

, l 16.2-1 3/91 16.2-2 5/91 16.2-3 16.2-4 )

Tab 16.3 16.3-1 Tab 16.4 Tab 16.5 16.5-1 6/96 16.5-2 6/96 16.5-2a 6/96 16.5-3 3/93 16.5-4 5/97 16.5-5 3/93 16.5-6 3/93 16.5-7 3/93 16.5-8 3/93 16.5-9 3/93 j 9905130143 990413 l PDR ADOCK 05000369 P PDR m

,r'% Page Revision Date 16.5-10 3/93 Tab 16.6 Tab 16.7 16.7-1 8/95 16.7-2 8/95 16.7-3 8/95 16.7-4 8/95 16.7-5 8/95 16.7-6 8/95 16.7-7 8/95 16.7-8 8/95 16.7-9 9/98 o

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Table 16.9-3 12/90 Table 16.9-3 (cont'd) 12/90 I

Table 16.9-3 (cont'd) 12/90 '

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V TABLE OF CONTENTS SECTION TITLE PAGE 16.0 SELECTED LICENSEE COMMITMENTS 16.1

16.1 INTRODUCTION

16.1-1 16.2 APPLICABILITY 16.2-1 16.3 DEFINITIONS 16.3-1 16.4 COMMITMENTS RELATED TO REACTOR COMPONENTS 16.4-0 16.5 COMMITMENTS RELATED TO REACTOR COOLANT SYSTEM 16.5-0 16.6 COMMITMENTS RELATED TO ENGINEERED SAFETY FEATURES 16.6-0 (NON-ESF SYSTEMS) l 16.7 COMMITMENTS RELATED TO INSTRUMENTATION ,

16.7-1 16.7-1 ATWS MITIGATION SYSTEM 16.7-1 16.8 COMMITMENTS RELATED TO ELECTRICAL POWER SYSTEMS 16.8-0 16.9 COMMITMENTS RELATED TO AUXILIARY SYSTEMS 16.9-1 16.10 COMMITMENTS RELATED TO STEAM AND POWER CONVERSION 16.10-0 SYSTEMS (h

( _) 16.11 COMMITMENTS RELATED TO RADIOAC1'IVE WASTE MANAGEMENT 16.11-0 16.12-1 IN-PLANT IODINE MONITORING 16.12-1 l 16.14-1 STARTUP REPORTS 16.14-1 l 16.15 RELOCATED ITEMS FROM ITS IMPLEMENTATION 16.15-1 16.0-1 4/99 l Lj'

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ELECTRICAL EQUIPMENT PROTECTION DEVICES 16.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES COMMITMENT COMMITMENTS:

1 The containment penetration conductor overcurrent protective devices shown in Table 16.8-1 shall be maintained in compliance with the requirements of SLC 16.15-3.8.4.1, Containment Penetration Conductor l Overcurrent Protective Devices. l 1

BASES:

The tables listed in this commitment were relocated from the McGuire Technical Specifications with the approval of the U.S. Nuclear Regulatory Commission. Any additions, deletions, or revisions to the table are considered a change in a commitment and shall be performed pursuant to the Compliance Control or Selected Licensee Commitments Section Manual Procedure. Being a commitment, these tables can only be changed using the 10 CFR 50.59 evaluation process as described in the previously mentioned procedure.

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I 16.9 AUXILIARY SYSTEMS

[ h FIRE PROTECTION SYSTEMS i

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16.9-6 FIRE DETECTION INSTRUMENTATION COMMITMENTS:

i As a minimum, the fire detection instrumentation for each fire detection zone shown in Table 16.9-3 shall be OPERABLE.

APPLICABILITY:

l Wherever equipment protected by the fire detection instrument is required to be OPERABLE.

REMEDIAL ACTION:

a. With any, but not more than one-half the total in any fire zone, functions fire detection instruments shown in Table 16.9-3 inoperable, restore the inoperable instrument (s) to OP'ERABLE status within 14 days or within the next I hour established a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the instrument (s) is located inside the containment, then inspect that containment zone at least once per 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> or monitor the containment air temperature at least once per hour at the locations given in ITS SR 3.6.5.1 or SR 3.6.5.2. l

/N b. With more than one-half of the Function A fire detection instruments

() in any fire zone shown in Table ly6.9-3 inoperable, or with any Function B fire detection instruments shown in Table 16.9-3 inoperable, or with any two or more adjacent fire detection instruments shown in Table 16.9-3 inoperable, within 1 heur

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establish a fire watch patrol to inspect the zone (s) with the inoperable instrument (s) at least once per hour, unless the ins trument (s) is located inside the containment, then inspect that containment zone at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or monitor the containment air temperature at least once per hour at the locations given in ITS SR 3.6.5.1 or SR 3.6.5.2. With any Annulus Fire detection instrumentation listed in Table 16.9-3 inoperable do the l following:

i) Within one hour, perform a fire watch patrol of the annulus and, ii) Perform the following:

1) Perform an hourly fire watch by verifying that a least one adjacent zone is operable.

23

2) If no adjacent zones are operable then perform a fire watch patrol of the annulus at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, thereafter.

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16.11 RADIOLOGICAL EFFLUENT

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\w ') RADIOACTIVE EFFLUENTS 16.11-11 SOLID RADIOACTIVE hw i 9

j COMMITMENTS:

Radioactive wastes shall be solidified or dewatered in accordance with the PROCESS CONTROL PROGRAM to meet shipping and transportation requirements during transit, and disposal site requirements when received at the disposal site.

APPLICABILITY: At all times REMEDIAL ACTION:

a) With SOLIDIFICAIOTN or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of inadequately processed wastes and correct the PROCESS CONTROL PROGRAM, the procedures and/or the Solid Waste System as j necessary tc prevent recurrence.

b) With SOLIDIFICATION or dewatering not performed in accordance l with the PROCESS CONTROL PROGRAM, test the improperly processed waste in each container to ensure that it meets burial ground and j shipping requirements and take appropriate administrative action i

to prevent recurrence.

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TESTING REQUIREMENTS:

SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive wastes (e.g.,

fi]ter sludges, spent resins, evaporator bottoms, boric acid solutions and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM:

a) If any te.t specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, alternative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed using the alternative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM; 16.11-44

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( b) If the initial test specimen from a batch of waste fails to i

verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall I

provide for the collection and testing of representative test l specimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required, as specified in the BASES, to assure SOLIDIFICAITON of subsequent batches of waste; and, l l

c) With the installed equipment incapable of meeting SLC 16.11-11 or declared out of service, restore the equipment to operable status or provide for contract capability to process wastes as necessary to satisfy all applicable transportation and disposal requirements.

BASES:

This commitment implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part ,5,0. The process parameters included in establishing the PROCESS CONTROL PROGRAM may include, but are not limited to waste type, waste pH, waste / liquid / solidification agent / catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing time.

Licensee-initiated changes to the PROCESS CONTROL PROGRAM:

1. Shall be documented and records of reviews performed shall be j retained as required by the Quality Assurance Plan. This

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documentation shall contain:

a. Sufficient information to support the change together with the appropriate analyses or evaluations justifying the change (s) and

b. A determination that the change will maintain the overall j conformance'of the solidified waste product to existing requirements of Federal, State, or other applicable regulations.

2. Shall become effective upon review and acceptance by the Station Manager and a qualified individual / organization.

Written procedures shall be establisned, implemented, and maintained for the PROCESS CONTROL PROGRAM implementation.

REFERENCES:

1. Process Control Program

2. 10 CFR Part 50

3. 10 CFR Part 50, Appendix I l 1

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rs 16.12 RADIATION PROTECTION r \

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16.12-1 IN-PLANT IODINE MONITORING COMMITMENTS:

A program shall'be established, implemented, and maintained which will

! ensure the capability to accurately determine the airborne iodine l concentration in vital areas under accident conditions. This program i shall include the following:

1

1. Training of personnel,

2. Procedures for monitoring, and

3. Provisions for maintenance of sampling and analysis equipment.

l APPLICABILITY: At all times.

REMEDIAL ACTION: Not Applicable.

TESTING REQUIREMENTS: Not Applicable.

BASES:

This commitment is provided to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

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REFERENCES:

1. UFSAR 11.4, Process and Effluent Radiological Monitoring Systems, January 1, 1998.

2. Technical Specification 6.8.4.b, as amended through amendments 166/148.

3. Technical Specification 3.3.3.6 as amended through amendments 166/148.

4. NUREG-0737, III.D.3.3.

16.12-1 A

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N 16.14 REFUELING OPERATIONS '

(O 16.14-1 STARTUP REPORTS COMMITMENTS:

The following report shall be submitted in accordance with 10CFR50.4:

1. A summary report of plant STARTUP and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the License involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the plant.

2. The Startup Report shall address each of the tests identified in the UFSAR and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in License conditions based on other commitments shall be included in this report.

3. Startup Reports shall be submitted within: (1) 90 days following completion of the STARTUP test program, or (2) 90 days following

, resumption or commencement of commercial POWER OPERATION, or (3) 9 months following initial criticality, whic';3ver is earliest. If the Startup Report does not cover all three eve:.:s (i.e., initial criticality, completion of STARTUP test program, and resumption or commencement of commercial operation), supplementary reports shall be submitted at least every 3 months until all three events have been completed.

APPLICABILITY: At all times.

REMEDIAL ACTION: Not Applicable.

TESTING REQUIREMENTS: Not Applicable.

BASES:

This commitment is to satisfy the requirements of 10CFR50.4.

REFERENCES:

1. Technical Specifications 6.9.1.1, 6.9.1.2, and 6.9.1.3 as amended through amendment 166/148.

2. UFSAR 14.0, Initjal Tests and Operation, as revised through January 1, 1998.

16.14-1 4/99

16.15 CTS RELOCATED ITEMS

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s i 16.15-1.0 DEFINITIONS l

l 16.15-1.17 - MEMBER (S) OF THE PUBLIC shall include all persons who are 1 l

not occupationally associated with the plant. This category does not include employees of the licensee, its contractors of vendors. Also excluded from this category are persons who enter the site to service equipment or make deliveries. This category does not include personnel who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

16.15-1.23 - PROCESS CONTROL PROGRAM (PCP) shall contain the correct formulas, sampling, analyses, test, and determinations to be made to ensure that 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 10CFR Parts 20, 61, and 71, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

16.15-1.24 - PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain, cemperature, pressure, humidity, concentration or other operating condition in such a manner that replacement air or gas is required to purify the confinement.

16.15 (New) SECURED related to value position indicates that:

1. For MANUAL valves, the subject valve is locked in t..e desired

[ position Y-}l OR

2. for AUTOMATIC valves, the subject valve is de-energized and properly tagged 16.15-1.31 - SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by the licensre.

16.15-1.34 - SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity.

16.15-1.39 - UNRESTRICTED AREA shall be any area or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters, or for industrial, commercial, institutional, and/or recreational purposes.

16.15-0 O

V 4'99

16.15-1.40.- VENTIIATION EXHAUST TREATMENT SYSTEM shall be any system p designed and installed to reduce gaseous radiciodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble' gas effluents.

Engineered Safety-Feature (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

16.15-1.41 - VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING.

Vent, used in system names, does not imply a VENTING process.

16.15-1.42 - WASTE GAS HOLDUP SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting Reactor Coolant System off gases from the Reactor Coolant System and providing for delay or holdup for the purpose of reducing the total radioactivity prior to. release to the environment. ,

O V

16.15-00 4/99

16.15 CTS RELOCATED ITEMS 16.15-3.1.2.1 BORATION SYSTEMS FLOW PATH - SHUTDOWN COMMITMENT:

As a minimum, one of the following boron injection flow paths shall be OPERABLE and capable of being powered from an OPERABLE emergency power source:

a. A flow path from a boric acid tank via a boric acid transfer pump and a charging pump to the Reactor Coolant System if the boric acid storage tank in SLC 16.15-3.1.2.5a is OPERABLE, or

b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling woter storage tank in SLC 16.15-3.1.2.5b is OPERABLE.

APPLICABILITY: MODES 5 and 6. 1 REMEDIAL ACTION:

With none of_the above flow paths OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

TESTING REQUIREMENTS:

O 16.15-4.1-2.1 At least one of the above required flow paths shall be demonstrated OPERABLE:

l

a. At least once per 7 days by verifying that the temperature of the heat j traced portion of the flow path is greater than or equal to 65 'F when a j flow path from the boric acid tanks is used, and  !

b. At least once per 31 days by verifying that each valve (manual, power operated, or automatic) in the flo.w path that is not locked, sealed, or otherwise secured in position, is in its correct position. .

BASES:

The. Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps,

.(3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systems, and (6) an emergency power supply from OPERABLE diesel generators.

With' the RCS temperature below 200'F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS.and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

16.15-1 4/99

g- The OPERABILITY of one Boron Injection System during REFUELING ensures that I i this system is available for reactivity control while in MODE 6.

SLC 16.15-3.1.2.1 requires at least one boron injection flow path to be OPERABLE in MODES 5 and 6. This SLC requires that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or i otherwise secured in position be verified to be in its correct position every 31 days.

i For automatic valves and power operated valves which are OPERABLE and have an OPERABLE emergency power source, these valves may be repositioned as required to support other plant operations if the valves will move to their proper position on demand to establish the Boration Flow Path.

1 r

i 5

\s 16.15-2 4/99

l 16.15 CTS RELOCATED ITEMS

+

16.15-3.1.2.2 BORATION SYSTEMS FLOW PATH - OPERATING l

l l COMMITMENT:

At least two' of the following three boron injection flow paths shall be OPERABLE:

a. The flow path from a boric acid tank via a boric acid transfer pump and a charging pump to the Reactor Coolant System, and

b. Two flow paths from the refueling water storage tank via charging pumps to the Reactor Coolant System.

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

REMEDIAL ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at lL Ast two flow paths to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

O TESTING RFOUIREMENTS:

16.15 4.1.2.2 At least two of the above required flow paths shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the boric acid tanks is greater than or equal to 65'F when it is a required water source; l

b. At least once per 31 days by verifying that each valve (manual, power operated, or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position;

c. At least once per 18 months during shutdown by verifying that each automatic valve in the flow path actuates to its correct position on a safety Injection test signal; and

d. At least once per 18 months by verifying that the flow path required by SLC 16.15-3.1.2.2a delivers at least 30 gpm to the Reactor Coolant System,

' only one boron injection flow path is required to be OPERABLE whenever the temperature of one or more of the RCS cold legs is less than or equal to 300*F.

[ 16.15-3 C )/

4/99

j. BASES-The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to l perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systems, and (6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 200'F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% delta k/k after xenon decay and cooldown to 200*F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions.

I 16.15-4 4/99

l l 16.15 - CTS RELOCATED ITEMS Iq 16.15-3.1.2.3 BORATION SYSTEMS CHARGING PUMP - SHUTDOWN N))

l COMMITMENT:

l One' charging pump in the boron injection flow path required by SLC 16.15-3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source. j l

APPLICABILITY: MODES 5 and 6. '

REMEDIAL ACTION:

1 With no charging pump OPERABLE or capable of being powered from an  !

OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

TESTING REQUIREMENTS:

16.15-4,1.2.3.1 The above required charging pump shall be demonstrated OPERABLE by verifying a differential pressure across the pump of greater than 5.5.8, or equal to 2380 psid is developed when tested pursuant to ITS BASES:

The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources,

[T (2) charging pumps, (3) separate flow paths, (4) boric acid transfer

( ') pumps, (5) associated Heat Tracing Systems, and (6) an emergency power supply from OPERABLE diesel generators.

With the RCS temperature below 200 *F, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE below 300 *F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. Allowing two Centrifugal Charging pumps to operate simultaneously for 5 15 minutes increases the nargin of safety with respect to the Reactor Coolant pump seal failure resulting in a LOCA in that the Reactor Coolant pump seal injection flow is not interrupted during pump swap.

For the 15 minute period during which simultaneous Centrifugal Charging pump operation is allowed, the safety margins as related to the mass addition analysis are not appreciably reduced. ITS 3.4.12 requires two PORVs (or equivalent) to be operable during this period of operation, j thus a mass addition transient can be relieved as required assuming the two PORVs (or equivalent) function properly. [

" Two charging pumps may be operable and operating for less than or equal to 15 minutes to allow swapping charging pumps.

16.15-5 O

V 4/99

I

~s The OPERABILITY of one Boron Injection System during REFUELING ensures 1 1

that this system is available for reactivity control while in MODE 6.

[O s

SLC 16.15-3.1.2.3 required one charging pump in the boration flow path of SLC 16.15-3.1.2.1 to be OPERABLE and capable of being powered from an OPERABLE emergency power source. SLC 16.15-3.1.2.3 is applicable in MODES 5 and 6.

The ACTION statement requires ' suspend all operations involving CORE ALTERATIONS or positive reactivity changes.' The intent is that j I

specific evolutions or operations that involve positive reactivity changes (fuel movement, dilutions, control rod movements or sustained NC temperature changes adding positive reactivity) not continue if the conditions described above do not exist. There are operations (e.g.

swapping ND trains, swapping KC trains, some testing) that can result in temperature oscillations that have insignificant effects on shutdown margin and can continue.

Operational or testing activities that result in NC temperature swings of 20'r about an initial value have been judged not to constitute positive reactivity changes as intended in SLC 16.15-3.1.2.3 when in MODE 5. There must be at least 500 ppm of boron beyond,the required Shutdown Boron Concentration for this interpretation to remain valid.

This interpretation should not be used to establish sustained NC s'; stem heatups or cooldowns that result in sustained positive reactivity additions.

REFERENCES:

Nuclear / Reactor Engineering Memo to File R.F. 4.0.1, August 23, 1994 "NC l

's' Temperature Swings affect on Shutdown Margin as per Technical Specification 3.1.2.3. l 16.15-6 4'99

16.15 - CTS RELOCATED ITEMS 16.15-3.1.2.4 BORATION SYSTEMS CHARGING PUMPS - OPERATING COMMITMENT:

At least two' charging pumps shall be OPERABLE.

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

REMEDIAL ACTION:

With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

TESTING REQUIREMENTS:

16.15-4.1.2.4.1 At least two charging pumps shall be demonstrated OPERABLE by verifying a differential pressure across each pump of greater than or equal to 2380 psid is developed when tested pursuant to ITS 5.5.8.

BASES:

The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systems, and (6) an emergency power supply from OPERABLE diesel generators.

The limitation for a maximum of one centrifugal charging pump to be OPERABLE below 300 "F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. Allowing two Centrifugal Charging pumps to operable simultaneously for < 15 minutes increases the margin of safety with respect to the Reactor Coolant pump seal failure resulting in a LOCA in that the Reactor Coolant pump seal injection flow is not interrupted during pump swap. For the 15 minute period during which simultaneous Centrifugal Charging pump operation is allowed, the safety margins as related to the mass addition analysis are not appreciably reduced. Improved Technical Specification 3.4.12 requires two PORVs (or equivalent) to be operable during this period of l operation, thus a mass addition transient can be relieved as required assuming the two PORVs (or equivalent) function properly. l

• A maximum of one centrifugal charging pump shall be OPERABLE whenever the temperature of one or more of the RCS. cold legs is less than or equal to 300 F.

~

Two charging pumps may be operable and operating for 5 15 minutes to allow swapping charging pumps.

16.15-7 i

4/99

16.15 e CTS RELOCATED ITEMS 16.15-3.1.?.5 BORATED WATER SOURCE - SHUTDOWN COMMITMENT:

As a minimum, one of the following borated water sources shall be [

OPERABLE:

'a.

A Boric Acid Storage System and at least one associated Heat Tracing System with:

1. A minimum contained borated water volume as presented in the Core l Operating Limits Report,

2. A minimum boron concentration as presented in the Core Operating Limits Report, and

3. A minimum solution temperature of 65 F. .-

b. The refueling water storage tank with:

1. A minimum contained borated water volume as presented in the Core l Operating Limits Report,

2. A minimum boron concentration as presented in the Core Operating Limits Report, and

3. A minimum solution temperature of 70'F' .

APPLICABILITY: MODES 5 and 6.

REMEDIAL ACTION:

With no borated water source OPERABLE, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

TESTING REQUIREMENTS:

16.15-4.1.2.5 The above required borated water source shall be demonstrated l OPERABLE:

a. At least once per 7 days by:

1. Vorifying the boron concentration of the water,

2. Verifying the contained borated water volume, and l

3. Verifying '-he boric acid storage tank solu' tion temperature when it is l the source of borated water.

16.15-8 O

4/99

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when it is the source of borated water and the outside air temperature is less than 7 0'F. l BASES:

The Boron Injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate l flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systems,  !

and (6) an emergency power supply from OPERABLE diesel generators.  !

1 The boron capability required below 200 F is sufficient to provide a SHUTDOWN MARGIN of 1% delta k/k after xenon decay and cooldown from 200 'F to 140 "F. The minimum borated water volumes and concentrations required to maintain shutdown margin for the Boric Acid Storage system and the Refueling Water Storage Tank i

l are presented in the Core Operating Limits Report.

The SLC commitment value for the Boric Acid Storage Tank and the Refueling Water Storage Tank minimum contained water volume during Modes 5 and 6 is based on the l required volume to maintain shutdown margin, an allowance for unusable volume and additional margin as follows:

1 Boric Acid Storage Tank Requirements for Maintaining SDM - Modes 5 & 6 Required volume for maintaining SDM Presented in the COLR j Unusable volume (to maintain full suction pipe) 4,199 gallons Additional margin 4,100 gallons Refueling Water Storage Tank Requirements for Maintaining SDM - Modes 5 & 6 Required volume for maintaining SDM Presented in the COLR Unusable volume (to maintain full suction pipe) 16,000 gallons Additional margin 23,500 gallons The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST also '

ensure a pH value of between 7.5 and 9.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

l 16.15-9 4/99 l

~.

~~

16.15 - CTS RELOCATED ITEMS 16.15-3.1.2.6 BORATED WATER SOURCES - OPERATING COMMITMENT:

As a minimum, the following borated water source (s) shall be OPERABLE as required by SLC 16.15-3.1.2.2:

a. Awith:

Boric Acid Storage System and at least one associated Heat Tracing System

1. A minimum contained borated water volume as presented in the Core Operating Limits' Report, l

2. AReport, minimum and boron concentration as presented in the Core Operating Limits

3. A minimum solution temperature of 65*F.

b. The refueling water' storage tank with: '

1. A minimum contained borated water volume as presented in the Core Operating Limits Report, l

2. AReport, minimum boron concentration as presented in the Core Operating Limits i

3. A minimum solution temperature of 70*F, and

4. A maximum solution temperature of 100 F.

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

REMEDIAL ACTION:

a. With the Boric Acid Storage System inoperable and being used as of the above required borated water sources, restore the storage system to (

OPERABLE next status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and borated to a SHUTDOWN MARGIN equivalent to at.least 1%

l delta k/k at 200*F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With the refueling water storage tank inoperable, restore the tank to OPERABLE. status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. .

i 1

16.15-10 v

4/99 1

i l

b( ,j TESTING REQUIREMENTS:

16.15-4.1.2.6 Each borated water source shall be demonstrated OPERABLE: l

a. At least once per 7 days by:

1. Verifying the boron concentration in the water,

2. Verifying the contained borated water volume of the water source, and

3. Verifying the Boric Acid Storage System solution temperature when it is the source of borated water.

b. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying the RWST temperature when the outside air temperature is either less than 70*F or greater than 100 F.

l BASES:

The Boron Injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water sources, (2) charging pumps, (3) separate flow paths, (4) boric acid transfer pumps, (5) associated Heat Tracing Systems, and (6) an emergency power supply from OPERABLE diesel generators.

The minimum borated water volumes and concentrations required to maintain shutdown margin for the Boric Acid Storage System and the Refueling water

,~ ,s 3

Storage Tank are presented in the Core Operating Limits Report. >

f

\~- The SLC commitment value for the Boric Acid Storage Tank and the Refueling Water l Storage Tank minimum contained water volume during Modes 1-4 is based on the required volume to maintain shutdown margin, an allowance for unusable volume and additional maroin as follows. ,

I Boric Acid Storage Tank Requirements for Maintaining SDM - Modes 1-4  !

Required volume for maintaining SDM Presented in the COLR Unusable volume (to maintain full suction pipe) 4,199 gallons Additional margin .

6,470 gallons Refueling Water Storage Tank Requirements for Maintaining SDM - Modes 1-4 Required volume for maintaining SDM Presented in the COLR l Unusable volume (to maintain full suction pipe) 16,000 gallons ,

Additional margin 23,500 gallons l The contained water volume limits include allowance for water not available I because of discharge line location and other physical characteristics.

l The limits on contained water volume and boron concentration of the RWST also '

ensure a pH value of between 7.5 and 9.5 for the solution recirculated within containment after a LOCA. -This pH band minimizes the evolution of iodine and i minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

16.15-11 p)

\~, i 4/99

I l

(N,

? 1 16.15 - CTS RELOCATED ITEMS i l

1 L ,/

l 16.15-3.1.3.3 ROD POSITION INDICATION SYSTEM-SHUTDOWN l COMMITMENT:

One rod position indicator (excluding demand position indication) shall be l OPERABLE and capable of determining the control rod position within + 12 steps l for each shutdown or control rod not fully inserted. [

APPLICABILITY: MODES 3*, 4* and 5*.

REMEDIAL ACTION:

With less than the above required position indicator (s) OPERABLE, immediately open the Reactor Trip breakers.

l BASES: .

l OPERABILITY of the control rod position indicators is required to determine I

control rod positions and thereby ensure compliance with the control rod alignment and insertion limits.

gx

• With the Reactor Trip breakers in the closed position.

l

() .

, 16.15-12

( )

%/

4/99 l

i l

% 16.15 - CTS RELOCATED ITEMS 16.15-3.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS l

COMMITMENT:

1 The radiation monitoring instrumentation channels for plant operations shown in l Table 3.3-6 shall be OPERABLE with their Alarm / Trip Setpoints within the I specified limits.

APPLICABILITY: As shown in Table 3.3-6.

REMEDIAL ACTION:

a. With a radiation monitoring channel Alarm / Trip Setpoint exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or declare the channel inoperable,

b. With one or more radiation monitoring channels inoperable, take the action shown in Table 3.3-6.

l TESTING REQUIREMENTS:

16.15-4.3.3.1 Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and ANALOG CHANNEL OPERATIONAL TEST operations for the modes and at the frequencies

( shown in Table 4.3-3.

The OPERABILITY of the radiation monitoring instrumention for plant operations ensures that: (1) the associated action will be initiated with the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincident logic is maintained, and (3) sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance.

The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded, If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions. Once the required logic combination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems.

16.15-13 O

4/99

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\' l l

16.15-3.3.3.2 MOVABLE INCORE DETECTORS l

COMMITMENT: I The Movable Incore Detection System shall be OPERABLE with:

l a. At least 75% of the detector thimbles, 1

b. A minimum of two detector thimbles per core quadrant, and i

c. Sufficient movable detectors, drive, and readout equipment to map these thimbles.

APPLICABILITY: When the Movable Incore Detection System is used for:

I a. Recalibration of the Excore Neutron Flux Detection System, or l

b. Monitoring the QUADRANT POWER TILT RATIO, or

c. Measurement of Fo(Z) and F AH REMEDIAL ACTION

With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions.

O l l \ s,,fI TESTING REQUIREMENTS:

1 16.15-4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by normalizing each detector output when required for:

a. Recalibration of the Excore Neutron Flux Detection System, or l b. Monitoring the QUADRANT POWER TILT RATIO, or j c. Measurement of Fo(Z) and F"AH l BASES:

The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this i system accurately represent the spatiaA neutron flux distribution of the core.  ?

The OPERABILITY of this system is demcnstrated by irradiating each detector used and determining the acceptability of i s voltage curve.

1 l

16.15-16

\

(N N_J l 4

4/99  !

I l

i

O d. For the purpose of measuring Fo(Z) or EAH a full incore flux map is used.

( Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be sued for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable.

The Testing Requirements for SLC 16.15-3.3.3.2 require that each detector be demonstrated Operable at last once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by normalizing detector output when the system is required for the specified activities.

The interval of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> begins when the normalization procedure for the detectors has been initiated, such that each detector is normalized at least once in a given 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.

O 16.15-17

/

(

, 4/99

l i

~s 16.15. - CTS RELOCATED ITEMS 16.15-3.3.3.9 I

EXPLOSIVE GAS MONITORING INSTRUMENTATION COMMITMENT:

l I

The gas monitoring instrumentation channels shown in Table 3.3-13 shall be  !

OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of ITS I 5.5.12.a are not exceeded.

APPLICABILITY: As shown in Table 3.3-13. i l

REMEDIAL ACTION:

i 1

a. With an explosive gas monitoring instrumentation channel Alarm / Trip I Setpoint less conservative than required by the above specification, j declare the channel inoperable and take the ACTION shown in Table 3.3-13.

l l

b. With less than the minimum number of explosive gas monitoring l instrumentation channels OPERABLE, take the ACTION shown in Table 3.3-13. l l Restore the inoperable instrumentation to OPERABLE status within the time I

specified in the ACTION, if unsuccessful, prepare and submit a Special Report to the Commission to explain why this inoperability was not corrected within the time specified.

I

' TESTING REQUIREMENTS:

16.15-4.3.3.9 Each explosive gas monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 4.3-9 BASES:

The gas instrumentation is provided for monitoring (and controlling) the concentrations of potentially. explosive gas mixtures in the WASTE GAS HOLDUP

~ SYSTEM.

SLC 16.15-3.3.3.9, Table 3.3-13, Item 2 requires that one hydrogen and two oxygen analyzers per station be OPERABLE in the Waste Gas (WG) System to ensure that explosive gas mixtures are not allowed in the WG System.

Only one recombiner train is in service at a time. In this case, the requirement for one hydrogen and two oxygen analyzers shall apply only to the train in service.

16.15-18 O

4/99

()

rN The requirement for oxygen analyzers may be satisfied for Train of the following three analyzers:

"A" by using two OWGMT5790 OWGMT6210 OWGMT6211 The requirement for oxygen analyzers may be satisfied for Train "B" by using two of the following three analyzers:

OWGMT5780 OWGMT6210 OWGMT6211 NOTE: The recombiner inlet oxygen analyzers (0WGMT5660/5670) shall not be used to satisfy the requirements of either train because these instruments measure the oxygen concentration after the addition of bulk oxygen. This is not representative of the Waste Gas System as defined in SLC 16.15-3.11.2.5. These analyzers will henceforth be used for the operation of the hydrogen recombiner but will not be used to satisfy the requirements of SLC 16.15-3.3.3.9.

REFERENCES:

Review of UFSAR, Section 11.3, System Description, and the Catalytic Hydrogen Recombiner Operational Manual, MNM-1201.04-0174.

x_J ,

f 16.15-19 p

4/99

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g-~3 16.15 - CTS RELOCATED ITEMS kN ')

16.15-3.4.2.1 REACTOR COOLANT SYSTEM SAFETY VALVES -SHUTDOWN COMMITMENT:

A minimum of one pressurizer Code safety valve shall be OPERABLE with a lift setting setting of 2485 psig + 3%, -2%*.

APPLICABILITY: MODES 4 and 5.

REMEDIAL ACTION:

With no pressurizer Code safety valve OPERABLE, immediately suspend all operations involving positive reactivity changes and place an OPERABLE RHR loop into operation in the shutdown cooling mode.

TESTING REOUIREMENTS:

16.15-4.4.2.1 No additional requirements other than those required by Specification 5.5.8. Following testing, lift settings shall be within + 1%.

BASES:

The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed

/

,_s to relieve 420,000 lbs per hour of saturated steam at the valve Setpcint. The

) relief capacity of a single safety valve is adequate to relieve any overpressure

\,,/ condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. SLC 16.15-3.4.2.1 & ITS 3.4.10 allow a

+

3% and - 2% setpoint tolerance for OPERABILITY; however, the valves are reset to + 1% during surveillance testing to allow for drift. [

• The lift setting pre .re shall correspond to ambient conditions of the valve at nominal operating temperature and pressure.

l 16.15-22

<- s .

I t ,!

v i

4/99 l

I l

, 16.15 - CTS RELOCATED ITEMS t

16.15-3.4.7 REACTOR COOLANT SYSTEM CHEMISTRY COMMITMENT:

The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 3.4-2.

-APPLICABILITY: At all times.

REMEDIAL ACTION:

MODES 1, 2, 3, and 4:

a. With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Limit, restore the parameter to within its Steady-State Limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or Oe in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN w1?hin the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />; and

b. With any one or more chemistry parameter in excess of its Transient Limit, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

At All Other Times:

With the concentration of either chloride or fluoride in the Reactor l

I O Coolant System in excess of its Steady-State Limit for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or in excess of its Transient Limit, reduce the pressurizer pressure to less than or equal to 500 psig, if applicable, and perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation prior to increasing the pressurizer pressure above 500 psig or prior to proceeding to MODE 4.

TESTING REQUIREME!1TS:

16.15-4.4.7 The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 4.4-3.

BASES:

The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor :oolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are' time and temperature dependent. Corrosion studies show that l

l 16.15-23 j l

4/99 I

('~'g operation may be continued with contaminant concentration levels in excess of {

l ( ,/ the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity 1 of the Reactor Coolant System. The time interval permitting continued operation

! within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.

I The Surveillance Requirements provide adequate assurance that concentrations in l excess of the limits will be detected in sufficient time to take remedial l 1 ACTION.

l

'ss _ / l i

16.15-24

/~T 4/99

A i

TABLE 3.4-2 1

REACTOR COOLANT SYSTEM CHEMISTRY LIMITS PARAMETER STEADY - STATE LIMIT l TRANSIENT LIMIT Dissolved Oxygen' 5 0.10 ppm 51.00 ppm Chloride 5 0.15 ppm 51.50 ppm Fluoride 5 0.15 ppm 51.50 ppm TABLE 4.4-3 REACTOR COOLANT SYSTEM CHEMISTRY LIMITS SURVEILLANCE REOUIREMENTS PARAMETER l SAMPLE AND ANALYSIS FREQUENCY Dissolved Oxygen " At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Chloride At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

(}

'O' Fluoride At least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 4

l

• Limit not applicable with T.v,less than or equal to 250 F.

• *Not required with T.,,less than or equal to 250 F.

gw 16.15-25

\g 4/99

_ _.m_

} 16.15 - CTS RELOCATED ITEMS 16.15-3.4.9.2 PRESSURIZER COMMITMENT:

The pressurizer temperature shall be limited to:

a. A maximum heatup of 100'F in any 1-hour period,

b. A maximum cooldown of 200'F in any 1-hour period, and

c. A maximum spray water temperature differential of 320'F.

{

APPLICABILITY: At all times. '

REMEDIAL ACTION:

With the pressurizer temperature limits in excess of any of tHe above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer remains acceptable for continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the pressurizer pressure to less than 500 psig.

within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

TESTING REQUIREMENTS:

16.15-4.4.9.2 The pressurizer temperatures shall be determined.to be within the limits at least once per 30 minutes during system heatup or cooldown. The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.

BASES:

The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G: *

?

The pressurizer heatup and cooldown rates shall not exceed 100"F/hr and 200*F/hr, respectivels The spray shall not be used if the temperature l difference between the pressurizer and the spray fluid is greater than 320 "F, and System preservice hydrotests and inservice leak and hydretests shall be performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance within the ASME Code requirements.

16.15-26 s

4S9

1 1

l 1

16.15. - CTS RELOCATED ITEMS 16.15-3.4.10 REACTOR COOLANT SYSTEM STRUCTURAL INTEGRITY l

I COMMITMENT:

)

The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained in accordance with Testing Requirement 16.15-4.4.10. {

l APPLICABILITY: All MODES.

REMEDIAL ACTION:

a. With the structural integrity of any ASME Code Class 1 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s ) prior to increasing the Reactor Coolant System temperature more than 50*F above the minimum temperature required by NDT considerations.

b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) prior to increasing the Reactor Coolant System temperature above 200 F.

c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service.

1 I

TESTING REQUIREMENTS:

16.15-4.4.10 In addition to the requirements of ITS 5.5.8, each reactor coolant I pump flywheel shall be inspected in accordance with ITS 5.5.7.

BASES:

The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and cperational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g) (6) (i) .

Components of the Reactor Coolant System were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1971 Edition and Addenda through Winter 1972.

16.15-27 4/99

16.15 - CTS RELOCATED ITEMS 16.15-3.4.11 REACTOR COOLANT SYSTEM REACTOR VESSEL HEAD VENT SYSTEM

' COMMITMENT:

1 Two rear. tor .*essel head vent paths, each consisting of two valves in series l powered from emergency buses shall be OPERABLE and closed.

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

REMEDIAL ACTION:

i

' a.-With one of the above reactor vessel head paths inoperable, STARTUP and/or POWER OPERATION may continue provided the inoperable vent path is maintained closed with power removed from the valve actuator of all the valves in the inoperable vent-path; restore the inoperable vent path to OPERABLE status within 30 days or be in HOT STANDBY witMin 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With both of the above reactor vessel head vent paths inoperable; maintain the inoperable vent path closed with power removed from the valve actuators of all the valves in the inoperable vent paths, and restore at least two of the vent paths to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

TESTING REQUIREMENTS:

16.15-4.4.11 Each reactor vessel head vent path shall be demonstrated OPERABLE at least once per 18 months by:

1. Cycling each val.ve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING.

l

2. Verifying flow through the reactor vessel head vent paths during venting during COLD SHUTDOWN or RETJELING.

BASES:

Reactor Vessel Head Vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core ,

cooling. The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists'to perform this function. (Operability of the pressurizer steam space vent path-is provided by ITS 3.4.11 and 3.4.12).

l The valve redundancy of the reactor coolant system vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a

- single failure of a vent valve, power supply or control system does not prevent isolation of the vent path.

16.15-28

,)

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The surveillance to verify Reactor Vessel Head Vent flowpath is qualitative as

( no specific size or flow rate is required to exhaust noncondensible gases. The function, capabilities, and testing requirements of the reactor coolant system vent systems are consistent with the requirements of Item II.B.1 of NUREG-0737,

" Clarification of TMI Action Plan Requirements", November 1980.

t O

i 16.15-29

.U 4/99 i

1

/""s 16.15 CTS RELOCATED ITEMS 16.15-3.5.2 ECCS SUBSYSTEMS - TAVG > 350*F COMMITMENTS: See ITS LCO 3.5.2 APPLICABILITY: Modes 1, 2, and 3.

l TESTING REQUIREMENTS:

16.15-4.5.2.c Each ECCS subsystem shall be demonstrated OPERABLE by visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during I.OCA conditions. This visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and .*

2. Of the areas affected within containment at the completion of each containment entry when CONTAINMENT INTEGRITY is established.

BASES:

None f\

x-_/

16.15-30

.D\

U 4'99

T 16.15 - CTS RELOCATED ITEMS

( 16.15-3.6.5.2 - ICE BED TEMPERATURE MONITORING SYSTEM COMMITMENT:

The Ice Bed Temperature Monitoring System shall be OPERABLE with at least two OPERABLE RTD channels in the ice bed at each of three basic elevations (10'6",

30'9" and 55' above the floor of the ice condenser) for each one-third of the ice condenser.

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

REMEDIAL ACTION:

a. With the Ice Bed Temperature Monitoring System inoperable, POWER OPERATION may continue for up to 30 days provided:

1. The ice compartment lower inlet doors, intermediate deck doors, and top deck doors are closed;

2. The last recorded mean ice bed temperature was less than or equal to 20*F and steady; and

3. The ice condenser cooling system is OPERABLE with at least:

a) 21 OPERABLE air handling units, b) 2 OPERABLE glycol circulating pumps, and c) 3 OPERABLE ref rigerant units; Otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN

'within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

b. With the Ice Bed Temperature Monitoring System inoperable and with the Ice Condenser Cooling System not satisfying the minimum components OPERABILITY requirements of ACTION a. 3. . above, POWER OPERATION may continue for up to 6 days provided the ice compartment lower inlet doors, intermediate deck doors, and top deck doors are closed and the last recorded mean ice bed temperature was less than or equal to 15*F and steady; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

TESTING REQUIREMENTS:

~16.15-4.6.5.2 The Ice Bed Temperature Monitoring System shall be determined OPERABLE by performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

16.15-31 4/99

BASES:

p i

The OPERABILITY of the Ice Bed Temperature Monitoring System ensures that the capability is available for monitoring the ice temperature. In the event the system is inoperable, the ACTION requirements provide assurance that the ice bed heat removal capacity will be retained within the specified time limits.

The Ice Bed Temperature Monitoring System shall be operable with at least two OPERABLE RTD channels in the ice bed at each of three basic elevations ( 10' 6",

30' 9" and 55' above the floor of the ice condenser) for each one-third of the ice condenser.

Operations needs a list of applicable RTDs with their respective locations to make appropriate operability evaluations.

Location of RTDs in Ice Condenser l

DukeID# Westing- Radial Degree Elevation Recorder Region houseID# Location l Location Pen # .

NPRD5000 TE801-1 54'E 310 823 1 A NPRD5010 TE801-2 54'E 310 799 2 A NPRD5020 TE801-3 54'E 310 768 Bay 1 3 A Fir Slb NPRD5030 TE801-4 50' 347 823 4 A NPRD5040 TE801-6 50' 347 799 5 A NPRD5050 TE801-5 50' 347 779 6 A Q NPRD5060 TE801-7 46' 25 823 7 A l

(/ NPRD5070 NPRD5080 TE801-8, TE801-5 46' 46' 25 799 8 A 25 779 9 A NPRD5090 TE601-10 50' 25 823 10 A NPRD5100 TE801-11 50' 25 799 11 A NPRD5110 TE801-12 50' 25 779 12 A NPRD5120 TE801-13 54'E 25 823 13 A NPRD5130 TE801-14 54'E 25 799 14 A NPRD5140 TE801-15 54'& 25 779 15 A NPRD5150 TE801-16 50' 62 823 16 A NPRD5160 TE801-17 50' 62 799 17 8 NPRD5170 TE801-18 50' 62 779 18 B NPRD5180 TE801-19 46' 86 823 19 8 NPRD5190 TE801-20 46' 86 799 20 B NPRD5200 TE801-21 46' 86 779 21 B NPRD5210 TE801-22 50' 86 823 22 B NPRD5220 T E801-23 50' 86 799 23 B NPRD5230 TE801-24 50' 86 779 24 B NPRD5240 TE801-25 54'E 86 823 25 B NPRD5250 TE801-26 54'E 86 799 26 B NPRD5260 TE801-27 54'& 86 768 Bay 27 B 12 Fir Slb NPRD5270 TE801-28 50' 122 823 28 B NPRD5280 TE801-29 50' 122 799 29 B NPRD5290 TE801-30 50' 122 779 30 B NPRD5300 TE801-31 46' 160 823 31 C 16.I S-32 4/99

l I

DukeID# Westing- Radial Degree Elevation Recorder Region g houseID# Location Location Pen #

NPRD5310 TE801-32 46' 160 799 32 C

NPRD5320 TE801-33 46' 160 779 33 C NPRD5330 TE80134 50' 160 823 34 C NPRD5340 TE801-35 50' 160 799 35 C' NPRD5350 TE801-36 50' 160 779 36 C NPRD5360 TE801-37 54'E 160 823 37 C NPRD5370 TE801-38 54'E 160 799 38 C NPRD5380 TE801-39 54'E 160 779 39 C NPRD5390 TE801-40 50' 195 823 40 C NPRD5400 TE801-41 50' 195 799 41 C NPRD5410 TE801-42 50' 195 779 .42 C NPRD5420 TE801-43 54'E 232 823 43 C NPRD5430 TE801-44 54'6" 232 799 44 C NPRD5440 TE801-45 54'E 232 768 45 C NPRD5450 TE801-46 83 828 46 B NPRD5460 TE801-47 83 828 47 B NPRD5470 TE801-48 SPARE 48 Miscellaneous information

Each third of the Ice Condenser covers approximately 100 degrees of the containment structure:

REGION A: 305 degrees to 45 degrees

/~N REGION B: 45 degrees to 145 degrees REGION C: 145 degrees to 245 degrees The 3 basic elevations above the floor of the ice condenser and their relative elevations are:

55 ft. - 823 ft.

30 ft. 9 in. - 799 ft.

i 10 ft. 6 in. - 779 ft.

1 48 total points available for monitoring; #48 not normally used (spare); should be bypassed.

5 points are not addressed by the SLC: #3,27 & 45 are at or near the ice condenser floor slab; #46 & 47 are in the ice condenser upper plenum at or near the intermediate deck doors.

CHART RECORDER POINTS Elevation Ice Condenser Ice Condenser Ice Condenser l Section A Section B Section C 823 1,4,7,10,13 16,19,22,25,28 31,34,37,40,43 799 2,5,8,11,14 17,20,23,26,29 32,35,38,44,41 779 6,9,12,15 18,21,24,30 33,36,39,42 l 1

l l

A 16.15-33

V 4/99 I

l

g 16.15 - CTS RELOCATED ITEMS 16.15-3.6.5.4 ICE CONDENSER INLET DOOR POSITION MONITORING SYSTEM COMMITMENT:

The Inlet Door Position Monitoring System shall be OPERABLE.

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

REMEDIAL ACTION:

With.the Inlet Door Position Monitoring System inoperable, POWER OPERATION may continue for up to 14 days, provided the Ice Bed Temperature Monitoring System l 1s OPERABLE and the maximum ice bed temperature is less than or equal to 27 F when monitored at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; otherwise, restore the Inlet Door Position Monitoring System to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least l HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

TESTING REQUIREMENTS:

16.15-4.6.5.4 The Inlet Door Position Monitoring System shall be determined

' OPERABLE by:

a. Performing a CHANNEL CHECK at least once per 7 days and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> af ter receiving an " Ice Condenser Inlet Door Open" alarm on the control room annunciator portion of the system,

b. Performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least once per 18 months, and c, Verifying that the Monitoring System correctly indicates the status of each inlet door ,as the door is opened and reclosed during its testing per ITS 3.6.13.

l BASES:

The OPERABILITY of the Inlet Door Position Monitoring System ensures that the capability is available for monitoring the individual inlet door position. In the event the system is inoperable, the ACTION requirements provide assurance that the ice bed heat removal capacity will be retained within the specified time limits.

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4/99 l

16.15 - CTS RELOCATED ITEMS 1

16.15-3.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION COMMITMENT:

The temperatures of both the reactor and secondary coolants in the steam generators shall be greater than 70*F when the pressure of either coolant in the l steam generators greater than 200 psig. l APPLICABILITY: At all times.

REMEDIAL ACTION:

With the requirements of the above specification not satisfied:

a. Reduce the steam generator pressure of the applicable side to less than or equal to 200 psig within 30 minutes, and

b. Perform an engineering evaluation to determine the effect of the overpressurization on the structural integrity of the steam-generator.

Determine that the steam generator remains acceptable for continued operation prior to increasing its temperatures above 200*F.

TESTING REQUIREMENTS: l l

16.15-4.7.2 The pressure in each side of the steam generator shall be determined O

g to be less than 200 psig at least once per hour when the temperature of either the reactor or secondary coolant is less than 70 F.

BASES:  ;

The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70'F and 200 psig are. based on a steam generator RTun of 10*F and are sufficient to prevent brittle fracture.

16.15-35 O

4/99

l i

l 16.15 - CTS RELOCATED ITEMS (f~)

I 16.15-3.7.6 CONTROL AREA VENTILATION SYSTEM COMMITMENTS:

Control Room Area AHU, Switchgear Room AHU, and Battery Room Exhaust shall be operable.

APPLICABILITY: ALL MODES REMEDIAL ACTION:

1. If either A or B train Control Room Area AHU fails, becomes unavailable or is unable to maintain the temperature as stated on Table 1, restore within 7 days or declare the AHU INOPERABLE.

2. A SWGR-AHU shall be declared INOPERABLE if:

a) The AHU unit of any sub-equipment is inoperable or unavailable, or, b) The inability to maintain the temperature below the maximum in the area served as listed on Table 1.

With any ONE SWGR-AHU (or support system) of any one of four SWGR f-~g rooms INOPERABLE, restore the inoperable equipment to operable l I within 30 days or declare the affected SWGR INOPERABLE. With

\~ / both trains of SWGR AHU per SWGR room INOPERABLE, immediately restore one train of SWGR-AHU and comply with the above.

The redundant SWGR-AHU shall be immediately started following a single SWGR unit being declared INOPERABLE and the respective i SWGR room temperature verified trending towards or below Table 1, or else declare the affected SWGR INOPERABLE.

If any -+ these components becomes INOPERABLE, a priority E work request shall be written, requiring prompt and continuing attention.

The components listed in this Commitment share the Control Room Area Chilled Water System with the Control Room Area Ventilation System but are not governed by ITS 3.7.9.

16.15 36

,-~

f

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4/99

f I

3. With one train of BR-XF INOPERABLE, restore the equipment to operable within 30 days or declare the batteries INOPERABLE.

With both battery room exhaust fans INOPERABLE, within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> lock open all battery room BR-XF check dampers (1VC23, IVC 24, IVC 25, IVC 26) and at least every 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> verify the ambient conditions of temperature s 104' F and hydrogen concentration 12%. Restore at least one BR-XF to OPERABLE "with all check dampers l unlocked / restored" within 7 days.or declare all the batteries l INOPERABLE.

If either BR-XF becomes INOPERABLE, a priority E work request shall be written requiring prompt and continuing attention.

I I BASES:

1. In order to maintain Control Room Area equipment operability, the following ventilation equipment shall be OPERABLE:

CRA-AHU-l CRA-AHU-2

2. In order to maintain SWGR equipment operability, the following ventilation equipment shall be OPERABLE:

SWGR-AHU-1A SWGR-AHU-2A SWGR-AHU-1B SWGR-AHU-2B SWGR-AHU-1C SWGR-AHU-2C

[\~ -) SWGR-AHU-lD SWGR-AHU-2D

3. For clarity, each SWGR room has its own 30 day clock, i.e. lETA's 30 day clock is independent of IETB's 30 day clock.

NOTE: A hand held pyrometer, etc. is acceptable if the computer points for temperature are not available.

4. In order to maintain battery operability, the following Equipment must be OPERABLE. For conservatism, assume that all battery rooms are affected by a Battery Room Exhaust fan incperability:

BR-XF-1 BR-XF-2 I

REFERENCES:

Discussions with Mechanical Engineering and M.A. Tartaglia November 2, 1992 MEMO TO FILE. This interpretation is part of the corrective actions of PIR 0-M91-0114.

16.15-37 v

4/99

l

{

A Table 1 EQUIPMENT AREA (S) SERVED MAXIMUM ALLOWABLE TEMPERATURE

• l SGR-AHU-1A ROOM 803, EL. 750 90 SGR-AHU-1B l SGR-AHU-2A ROOM 805, EL. 750 90 SGR-AHU-2B SGR-AHU-1C ROOM 705, EL. 733 90 SG3-AHU-1D SC'R- AHU-2 C ROOM 716, EL. 733 90 SGR-AHU-2D CRA-AHU-1 ELECTRICAL PEN RM. EL.767 104 CRA-AHU-2 ELECTRICAL PEN RM. EL.767 104 l

104 BATTERY ROOMS 707, 708, 104 706, 710, 711 EL. 733 104 MCC-2EMXA (821) 104 MCC-1EMXA (808) 104 MCC-2EMXA 104 RESTRICTED INST. SHOP f"'s (807) 104 INSTRUMENT STORAGE 104 CABLE RM. EL. 750 104 MECH. EQ. RM. EL. 767 104 1EMXB 104 2EMXB 104 CONTROL ROOM 90 i The maximum allowable temperature indicates the temperature for  !

continuous duty rating for equipment and instrumentation found in these zones. These temperatures are provided to assure that the equipment will have an acceptable service life.

16.15-38 p

O l

4/99

i

/"'N 16.15 - CTS RELOCATED ITEMS b 16.15-3.7.8 SNUBBERS COMMITMENTS:

All snubbers shall be OPERABLE. The only snubbers excluded from the requirements are those installed on non-safety-related systems and then only if the f ailure or the failure of the system on which they are installed would not have an adverse effect on any safety-related system.

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

MODES S and 6 for snubbers located on systems required OPERABLE in those MODES. l REMEDIAL ACTION: l With one or more snubbers inoperable, within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> replace . cestore the inoperable snubber (s) to OPERABLE status and perform an engineerAng evaluation per SLC 16.15-4.7.8.g. on the attached component or declare the attached system inoperable and follow the appropriate ACTION statement for that system.

1 TESTING REQUIREMENTS:

16.15-4.7.8 Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program.

I a. Inspection Types

]

As used in this specification, type of snubber shall mean snubbers of the

'same design and manufacturer, irrespective of capacity.

b. Visual Inspections l

Snubbers are categorized as inaccessible or accessible during reactor operation and may be treated independently. Each of these categories (inaccessible and accessible) may be inspected independently according to the. schedule determined by Table 4.7-2. The visual inspection interval for i each category of snubber shall be determined based upon the criteria l provided in Table 4.7-2 and the first inspection interval determined using  !

this criteria shall be based upon the previous inspection interval as established by the requirements in effect before Amendment No. 126.

l

.16.15-39 l 4/99 1

O h TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL NUMBER OF UNACCEPTABLE i SNUBBERS Population or Column A Column B Column C Category Extended Interval Repeat Interval Reduce Interval (Notes 1 and 2) (Notes 3 and 6) (Notes 4 and 6) (Notes 5 and 6) 1 0 0 1 80 0 0 2 100 0 1 4

~

150 0 3 8 200. 2 5 13 300 5 12 25 400 8 18 36 500 12 24 48 750 20 40 78 1000 or greater 29 56 109 Note 1: The next visual inspection interval for a snubber population or category size shall be determined based upon the previous inspection interval and the number of unacceptable snubbers found during that interval. Snubbers may be categorized, based upon their accessibility during power operation, as accessible or inaccessible. The categories may be examined separately or jointly. However, the licensee must make and document that decision before any inspection and shall use that decision as the basis upon which to determine the next inspection interval for that category.

Note 2: Interpolation between population or category size and the number of unacceptable snubbers is permissible. Use next lower integer for the value of the limit for Columns A, B, or C if that integer includes a fractional value of unacceptable snubbers as described by interpolation.

16.15-40 4/99

Note 3: If the number of unacceptable snubbers is equal to or less than the y number in Column A, the next inspection interval may be twice the previous interval but not greater than 48 months.

Note 4: If the number of unacceptable snubbers is equal to or less than the number in Column B but greater than the number in Column A, the next inspection shall be the same as the previous interval.

Note 5: If the number of unacceptable snubbers is equal to or greater than the number in Column C, the next inspection interval shall be two-thirds of the previous interval. However, if the number of unacceptable snubbers is less than the number in Column C but greater than the number in Column B, the next interval shall be reduced proportionally by interpolation, that is, the previous interval shall be reduced by a factor that is one third of the ratio of the difference between the number of unacceptable snubbers found during the previous interval and the number in Column B to the difference in the numbers in Columns B l and C. 1 Note 6: The provisions of SLC 16.2.7 are applicable for all inspection intervals up to and including 48 months.

c. Refueling Outace Inspections At e ih refueling, the systems which have the potential for a severe dyr c event, specifically, the main steam system (upstream 'of the mai , team isolation valves) the main steam safety and power-operated re f valves and piping, auxiliary feedwater system, main steam supply

[,, ) to . e auxiliary feedwater pump turbine, and the letdown and charging

\s / portion of the CVCS system shall be inspected to determine if there has been a severe dynamic event. In case of a severe dynamic event, mechanical snubbers in that system which experienced the event shall be inspected during the refueling outage to assure that the mechanical snubbers have freedom of movement and are not frozen up. The inspection shall consist of verifying freedom of motion using one of the following: (1) manually induced snubber movement; (2) evaluation of in-place snubber piston setting; (3) stroking the mechanical snubber through its full range of travel. If one or more mechanical snubbers are found to be frozen up during this inspection, those snubbers shall be replaced or repaired before returning to power. The requirements of SLC 16.15-4.7.8b are independent of the requirements of this SLC.

d. Visual Inspection Acceptance Criteria Visual inspections shall verify: (1) that there are no visible indica-tions of damage or impaired OPERABILITY, and (2) attachments to the foundation or supporting structure are secure. 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, 16.15-41 O

V 4'99

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 SLC 16.15-4.7.8f. A hydraulic snubber found )

with the fluid port uncovered and all hydraulic snubbers found connected to an inoperable common reservoir shall be classified as unacceptable and may be reclassified acceptable by functionally testing each snubber starting with the piston in the as-found setting, extending the piston rod in the tension direction.

e. Functional Tests During the first refueling shutdown and at least once per refueling thereafter, a representative sample of snubbers shall be tested using one of the following sample plans. The large bore steam generator hydraulic snubbers shall be treated as a separate population for func-tional test purposes. A 10% random sample from previously untested snubbers shall be tested at least once per refueling outage until the entire population has been tested. This testing cycle.shall then begin anew. For cach large bore steam generator hydraulic snubber that does not meet the functional test acceptance criteria, at least 10% of the remaining population of untested snubbers for that testing cycle shall be tested. The sample plan shall be selected prior to the test period and cannot be changed during the test period. The NRC shall be notified of the sample plan selected prior to the test period.

1) At least 10% of the snubbers required by SLC 16.15-3.7.8 shall be

\ functionally tested either in place or in a bench test. For each snubber that does not meet the functional test acceptance criteria of SLC 16.15-4.7.8f, an additional 10% of the snubbers shall be functionally tested until no more failures are found or until all snubbers have been functionally tested; or

2) A representative sample of the snubbers required by SLC 16.15-3.7.8 shall be functionally tested in accordance with Figure 4.7-

1. "C" is the total number of snubbers found not meeting the acceptance requirements of SLC 16.15-4.7.8f (failures). The cumulative number of snubbers tested is denoted by "N." Test results shall be plotted sequentially in the order of sample assignment (i.e., each snubber shall be plotted by its order in the random sample assignments, not by the order of testing). If at any time the point plotted falls in the " Accept" region, testing of snubbers may be terminated. When the point plotted lies in the " Continue Testing" region, additional snubbers shall be tested until the point falls in the " Accept" region, or all the snubbers required by SLC 16.15-3.7.8 have been tested. Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later time, providing all snubbers tested with the failed equipment during the day of equipment failure are retested; or 16.15-42 f

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b G -3) An initial representative sample of 55 snubbers shall be functionally tested.

For each snubber which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until the total number tested is equal to the initial sample size multiplied '

by the factor, 1 + C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. This 1 can be plotted using an " Accept" line which follows the equation  !

N = 55(1 + C/2). Each snubber should be plotted as soon as it is tested. If the point plotted falls on or below the " Accept" line, testing may be discontinued. If the point plotted falls above the

" Accept" line, testing must continue unless all snubbers have been tested, 10 9

8 7

C 6 5

CONTINUE 4

TESTING 3 C = 0.055N- 2.007 f l 2

N /

7

-/ ACCEPT

/

0 10 20 30 40 50 60 70 80 90 100 N

FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST 16.15-43 4/99

l The representative samples for the functional test sample plans shall l I

i

' be randomly selected from the snubbers required by SLC 16.15-3.7.8 and j reviewed before beginning the testing. The review shall ensure as far as practical that they are representative of the various ,

configurations, operating environments, range of sizes, and capacities. Snubbers placed in the same locations as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan.

l If during the functional testing, additional sampling is required due to failure of only one type of snubber, the functional testing results shall be reviewed at that time to determine if additional samples should be limited to the type of snubber which has failed the functional testing.

f. Functional Test Acceptance Criteria l

The snubber functional test shall verify that:

1) Activation (restraining action) is achieved within the specified I range in both tension and compression, except.that inertia depen-dent, acceleration limiting mechanical snubbers may be tested to verify only that activation takes place in both directions of I travel;

2) Snubber bleed, or release rate wher,e required, is present in both tension and compression, within the specified range;

3) Where required, the force required to initiate or maintain metion

[ \ of the snubber is within the specified range in both directions

\s.s# of travel; and

4) For snubbers specifically required not to displace under con-tinuous load, the ability of the snubber to withstand load with-out displacement.

Testing meth'ods may be used to measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods.

g. Functional Test Failure Analysis An engineering evaluation shall be made of each failure to meet the functional test acceptance criteria to determine the cause of the failure. The results of this evaluation shall be used, if applicable, in selecting snubbers to be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode.

For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are s

16.15-44 4/99

/N attached were adversely affected by the inoperability of the snubbers

( ,) in order to ensure that the component remains capable of meeting the designed service.

If any snubber selected for functional testing either fails to activate or fails to move, i.e., frozen-in-place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be evaluated in a manner to ensure their OPERABILITY. This testing requirement shall be independent of the requirements stated in SLC 16.15-4.7.8e for snubbers not meeting the functional test acceptance criteria.

h. Functional Testing of Repaired and Replaced Snubbers l

Snubbers which fail the visual inspection or the functional test acceptance criteria shall be repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test result shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and freedom-of-motion test must have been performed within i 12 months before being installed in the unit.

i. Snubber Seal Replacement Program The seal service life of hydraulic snubbers shall be monitored to ensure that the service life is not exceeded between surveillance I h s_ ) inspections. The expected service life for the various seals, seal materials, and applications shall be determined and established based on engineering information and the seals shall be replaced so that the expected service life will not be exceeded during a period when the snubber is required to be OPERABLE. The seal replacements shall be documented and the documentation shall be retained in accordance with QA Topical Report BASES:

I All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their f ailure or failure of the system on which they are installed, would have no adverse effect on any safety-related system.

Snubbers are classified and grouped by design and manufacturer but not by size.

For example, mechanical snubbers utilizing the same design features of the 2 kip, 10 kip, and 100 kip capacity manufactured by Company "A" are of the same t ype . The same design mechanical snubbers manufactured Company "B" for the pur-poses of this specification would be of a different type, as would hydraulic snubbers from either manufacturer.

16.15-45 O

V 4/99

i The visual inspection frequency is based upon maintaining a constant level of

( snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to deter-l mine the next inspection. However, the results of such early inspections per-formed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspec-l tion whose results require a shorter inspection interval will override the l previous schedule.

To provide assurance of snubber functional reliability one of the three sampling and acceptance criteria methods are used:

1 1. Functionally test 10% of a type of snubber with an additional 10%

tested for each functional testing failure, or

2. Functionally test a sample size and determine sample acceptance or continue testing

• using Figure 4.7-1, or I l

l 3. Functionally test a representative sample size and determine sample acceptance or rejection using the stated equation.

l Figure 4.7-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in " Quality Control and Industrial Statistics" by Acheson J. Duncan.

Permanent or other exemptions from the surveillance program for individual

\ snubbers may be granted by the Commission if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubber for the applicable design conditions at either the completion of their fabrication or at a subsequent date.

The service life of a snubber is established vic manufacturer input and information through consideration of the snubber service conditions and asso-ciated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to l ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical i

bases for future consideration of snubber service life. The requirements for the maintenance of records and the snubber service life review not intended to affect plant operation.

l l If testing continues to between 100-200 snubbers (or 1-2 weeks) and still the accept region has not been reached, then the actual % of population quality (C/N) should be used to prepare for extended or 100% testing.

16.15-46 b

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1 l

fs 16.15 - CTS RELOCATED ITEMS i i

- ') 16.15-3.7.9 SEALED SOURCE CONTAMINATION l

COMMITMENT:

Each sealed source containing radioactive material either in excess of 100 microCuries of beta and/or gamma emitting material or 5 microCuries of alpha emitting material shall be free of greater than or equal to 0.005 microcurie of removable contamination.

APPLICABILITY: At all times.

REMEDIAL ACTION:

With a sealed source having removable contamination in excess of the above s

{

limits, immediately withdraw the sealed source from use and either:

1. Decontaminate and repair the sealed source, or

2. Dispose of the sealed source in accordance with Commission Regulations.

TESTING REQUIREMENTS:

16.15-4.7.9.1 Test Requirements - Each sealed source shall be tested for leakage and/or contamination by:

[%/ a. The licensee, or

b. Other persons specifically authorized by the Commission'or an Agreement .

State.

The test method shall have a detection sensitivity of at least 0.005 microcurie per test sample.

16,15-4.7.9.2 Test Frequencies - Each category of sealed sources (excluding startup sources and fission detectors previously subjected to core flux) shall be tested at the frequency described below,

a. Sources in use - At least once per 6 months for all sealed sources containing radioactive materials:

1) With a half-life greater than 30 days (excluding Hydrogen 3), and, j

2) In any form other than gas.

16.15-47 l

l

/N l

\)~- i 4!99

I 7- s

(, b. Stored sources not in use - Each sealed source and fission detector shall be tested prior to use or transfer to another licensee unless tested within the previous 6 months. Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed into use; and,

c. Startup sources and fission detectors - Each sealed startup source and fission detector shall be tested within 31 days prior to being subjected to core flux or installed in the core and following repair or maintenance to the source.

16.15-4.7.9.3 Reports - A report shall be prepared and submitted co the Commission on an annual basis if sealed source or fission detector leakage tests t reveal the presence of greater than or equal to 0.005 microcurie of removal I contamination. l BASES:

The limitations on removable contamination for sources requir.1ng leak testing, l including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium, g This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with '

Surveillance Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required  ;

to be tested more often than those which are not. Sealed sources which are

[

continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

l 16.15-48 A

4'99

16.15 - CTS RELOCATED ITEMS i k

16.15-3.7.12 AREA TEMPERATURE MONITORING COMMITMENT: I l

The temperature of each area shown in Tcble 3.7-6 shall be maintained within '

the limits indicated in Table 3.7-6.

APPLICABILITY: Whenever the equipment in an affected area is required to be OPERABLE. l l

REMEDIAL ACTION:

With one or more areas exceeding the temperature limit (s) shown in Table 3.7-6:

a. For more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, prepare and submit a Special Report to the Commission within the next 30 days providing a record of the amount by which and the cumulative time the temperature in the affected area j exceeded its limit and an analysis to demonstrate the continued OPERABILITY of the affected equipment.

b. By more than 30*F, in addition to the Special Report required above, within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either restore the area to within its temperature limit or declare the equipment in the affected area inoperable.

TESTING REQUIREMENTS:

(

16.15-4.7.12 The temperature in each of the areas shown in Table 3.7-6 shall be determined to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

TABLE 3.7-6 AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT (*F)

1. Containment Spray Pump Rooms 145

2. Miscellaneous Terminal Cabinets

a. TB208-209 (Turbine Bldg.) 150

b. TB496 (Fuel Bldg.) 150

3. Residual Heat Removal Pump Rooms 145

4. Diesel Generator Rocms 125

5. Spent Fuel Pool Cooling Pump Room 145 16.15-49 O

4/99

O BASES:

The area temperature limitations ensure that safety related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The temperature limits include an allowance for instrument error of 3.9'F.

h V

i

)

16.15-50 o

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4/99

/N 16.15 - CTS RELOCATED ITEMS u,

16.15-3.7.13 GROUNDWATER LEVEL l COMMITMENT:

l The groundwater level shall be maintained at elevations less than the values in '

Table 3.7-7 for the five (5) Auxiliary Building monitors listed in Table 3.7-7.

APPLICABILITY: At all times.

REMEDIAL ACTION: For Units 1 and 2.

If groundwater level for any three (3) of the five (5) monitors is above the values shown in Table 3.7-7, take the following actions:

1. Within one hour, reduce the groundwater level to below the values shown in Table 3.7-7; or,

2. Be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

TESTING REQUIREMENTS:

16.15-4.7.13.1 During each shift, the groundwater level shall be demonstrated to be within the values of Table 3.7-7 by the absence of alarms or by visual observation of the monitor lestl gauge.

\s / 16.15-4.7.13.2 Each groundwater level monito- instrument / loop for locations listed in Table 3.7-7 shall be demonstrated 2ERABLE at least once per year by the performance of a loop calibration and operational test.

TABLE 3.7-7 AUXILIARY BUILDING GROUNDWATER LEVEL MONITORS LOCATION INTERIOR / EXTERIOR ELEVATION UNIT PP-51 Interior 731' - 0" MSL 1 00-56 Interior 731' - 0" MSL 1&2 PP-61 Interior 731' - 0" MSL 2 West Wall Exterior 731' - 0" MSL 1 East Wall Exterior 731' - 0" MSL 2

,,s 16.15-51 4/99

~~ ~

l l

)

I i

s i

,,) BASES:

j This SLC is provided to ensure that groundwater levels will be monitored and prevented from rising to the potential failure limit for the McGuire Units 1 and i

2 Auxiliary Buildings. This potential failure limit is based on engineering calculations that have determined that the Auxiliary Buildings are susceptible to overturning due to buoyance at elevation 737 feet Mean Sea Level (MSL).

Under the requirements of this SLC, if groundwater level exceeds elevation 731 feet MSL, (3 out of 5 SLC groundwater monitor alarms), and cannot be reduced in one (1) hour, McGuire must begin reducing Units 1 and 2 to Mode 5, Cold Shutdown.

Analysis performed by Design Engineering determined that the Reactor and Diesel Generator Buildings are designed to withstand hydrostatic loadings due to groundwater levels up to elevation 760 feet MSL; therefore, no SLC requirements are specified for these structures.

Elevation 731 feet MSL is the action level of the five groundwater monitors listed in Table 3.7-7. The East Wall exterior monitor alarm At elevation 731 feet MSL is the Alert alarm. The other four (4) monitors are Hi-Hi alarms at elevation 731 feet MSL.

The East Wall exterior monitor was originally on the exterior of the Unit 2 i Auxiliary Building and subsequently was enclosed by the construction of the l Equipment Staging Building.  !

fs As required by Operations procedures, any alarms on non-SLC groundwater monitors

) will also be investigated. Additionally, if three (3) out of the five (5) l groundwater monitors alarm at levels below the action levels, Cperations will contact Duke Engineering (Civil) for investigation and resolution of the increased groundwater level.

If one or more of the 5 SLC groundwater monitors is determined to be inoperable, the monitor (s) will be considered to be indicating above the 731'-0" MSL until repaired and returned 'to an operable status.

16.15-52

,O V

4/99

7 16.15 - CTS RELOCATED ITEMS

'" 16.15-3.8.1.1 A.C. SOURCES - OPERATING V

COMMITMENTS: See ITS 3.8.1 APPLICABILITY: Modes 1,2,3 and 4 TESTING REQUIREMENTS:

16.15-4.8.1.1.2 Each diesel generator shall be demonstrated OPERABLE:

b. By removing accumulated water:

1) From the day tank after each occasion when the diesel is operated for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />

e. At least once per 18 months, by:

1) Subjecting the diesel to an inspection, during shutdown, in accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

12) Verifying that the fuel transfer pump transfers fuel from each fuel storage tank to the day tank of each diesel via the installed cross-connection lines.

7- s 14) Verifying, during shutdown (#), that the following diesel

( ) generator lockout features prevent diesel generator starting

'O only when required:

a) Turning gear engaged, or b) Emergency stop.

g. At least once per 10 years by:

2) Performing a pressure test of those portions of the diesel fuel oil system designed to Section III, subsection ND of the ASME code at a test pressure equal to 110% of the system design pressure, a This Surveillance Requirement may be performed in conjunction with periodic preplanned preventative maintenance activity that causes the diesel generator to be inoperable provided that performance of the surveillance requirement does not increase the time the diesel generator would be inoperable for the PM activity alone.

16.15-53

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4/99

i i

\

16.15-4.8.1.1.4 Each diesel generator 125-volt battery bank and charger shall be demonstrated OPERABLE:

a. At least once per 7 days by verifying that:

1) The electrolyte level of each battery is above the plates, and

2) The overall battery voltage is greater than or equal to 125 volts under a float charge g

b. At least once per 18 months by verifying that:

1) The batteries, cell plates, and battery racks show no visual indication of physical damage or abnormal deterioration;

2) The battery-to-battery and terminal connections are clear, tight, free of corrosion and coated with anti-corrosion material; and

3) The battery capacity is adequate to supply and maintain in OPERABLE status its emergency loads when subjected to a battery service test. .-

BASES:

The Surveillance Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides 1.9, " Selection of Diesel Generator Set Capacity for Standby Power Supplies, " March 10, 1971, 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear

, Power Plants, " Revision 1, August 1977, and 1.137, " Fuel-Oil Systems

\,

for Standby Diesel Generators," Revision 1, October 1979; Generic Letter 84-15, which modified the testing frequencies specified in Regulatory Guide 1.108; Generic Letter 93-05, which reduced the surveillance requirements for testing of Diesel Generators during power operation; also, Generic Letter 94-01, which removed the accelerated testing and special reporting requirements for Emergency Diesel Generators.

Some of the Surveillance Requirements for' demonstrating the operability of the diesel generators are modified by a footnote. The Specifications state the Surveillance Requirements are to be performed during shutdown, with the unit in mode 3 or higher. The footnote allows the particular surveillance to be performed during preplanned Preventative Maintenance (PM) activities that would result in the diesel generator being inoperable. The surveillance can be performed at that time as long as it does not increase the time the diesel generator is inoperable for the I;M activity that is being performed.

The footnote is only applicable at that time. The provision of the footnote shall not be utilized for operational convenience.

Since the McGuire emergency diesel generator manufacturer (Nordberg) is no longer in business, McGuire engineering is the designer of record.

Therefore, in the absence of manufacturer recommendations, McGuire engineering will determine the appropriate actions required for nuclear class diesel service taking into account McGuire diesel generator maintenance and operating history and industry experience where applicable.

V 16.15-54 4/99

/i 16.15 - CTS RELOCATED ITEMS N~Y 16.15-3.8.1.2 A.C. SOURCES-SHUTDOWN COMMITMENT: See ITS LCO 3.8.2 APPLICABILITY: Modes 5 and 6, During movement of irradiated fuel assemblies.

REMEDIAL ACTION:

With less than the above minimum required AC electrical sources OPERABLE, immediately suspend all operations involving crane opration with loads over the fuel storage pool.

's 16.15-55

,rm)

(

'%/

4/99

(

g 16.15 - CTS RELOCATED ITEMS (U 16.15-3.8.4.1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES I

COMMITMENT:

All containment penetration conductor overcurrent protective devices  ;

shown in SLC 16.8-1 shall be OPERABLE. l APPLICABILITY: MODES 1, 2, 3, and 4.

REMEDIAL ACTION:

1 With one or more of the containment penetration conductor overcurrent  !

l protective device (s) shown in SLC 16.8-1 inoperable: '

I l

a. Restore the protective device (s) to OPERABLE status ,or de-energize '

the circuit (s) by tripping the associated backup ci~rcuit breaker or racking out or removing the inoperable circuit breaker within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, declare the affected system or component inoperable, and verify the backup circuit breaker to be tripped or the inoperable circuit breaker racked out or removed at least once per 7 days thereafter; the provisions of SLC 16.2.4 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, their inoperable circuit breakers racked out, or

> removed, or

%/

b. Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

I l TESTING REOUIREMENTS:

1 16.15-4.8.4.1 All SLC 16.8-1 containment penetration conductor l overcurrent protective devices shall be demonstrated OPERABLE:

l

a. At least once per 18 months:

1) By verifying that the medium voltage (4-15 kV) circuit breakers are OPERABLE by selecting, on a rotating basis, at least 10% of the circuit breakers of each voltage level, and performing the following:

a. CHANNEL CALIBRATION of the associated protective relays,

b. An integrated system functional test which includes simulated A automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits 1

1 function as designed and as specified in SLC 16.8-1, and l

16.15-56 t

O 4/99

['

t c. For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least 10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested.

2) By selecting and functionally testing a representative sample of at least 10% of each type of lower voltage circuit breakers.

l Circuit breakers selected for functional testing shall be selected on a rotating basis. For the lower voltage circuit breakers the nominal Trip Setpoint and overcurrent response times are listed in SLC 16.8-1. Circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation. For each circuit breaker found inoperable during these functional tests, an ,

additional representative sample of at least 10% of all the '

l circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all circuit breakers of that type have been functionally tested; 3 I

and

)

3) A fuse inspection and maintenance program will be maintained to l ensure that:

l

1) The proper size and type of fuse is installed, '

2) The fuse shows no sign of deterioration, and

3) The fuse connections are tight and clean.

b. At least once per 60 months by subjecting each circuit breaker to an inspection and preventive maintenance in accordance with procedures prepared in conjunction with its manufacturer's recommendations.

1 l pASES:

Containment electrical penetrations and penetration conductors are protected by either de-energizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic surveillance.

The Surveillance Requirements applicable to lower voltage circuit breakers provide assurance of breaker reliability by testing at least one representative sample of each manufacturer's brand of circuit breaker. Testing of these circuit breakers consists of injecting a current in excess of the breaker's nominal setpoint and measuring the response time. The measured response time is compared to the 16.15-57 O

4/99

l t

l l

l

,e- manuf acturer's data to ensure that it is less than equal to a value 1 specified by the manufacturer. Each manufacturer's molded case and k_

s metal case circuit breakers are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers j are tested. If a wide variety exists within any manuf acturer's brand of circuit breakers, it is necessary to divide that manuf acturer's I

' breakers into groups and treat each group as a separate type of breaker )

j for surveillance purposes.

Fuse testing is in accordance with IEEE Standard 242-1975. This program will detect any significant degradation of the fuses or improperly sized fuses. Safety is further assured by the " fail safe" nature of fuses, that is, if the fuse fails, the circuit will de-energize.

l A

l l

l 1

l l

16.15-58 l

O 4/99

7 l

l

^

f s 16.15 - CTS RELOCATED ITEMS 16.15-3.9.3 DECAY TIME COMMITMENT:

The reactor shall be suberitical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />.

APPLICABILITY: During movement of irradiated fuel in the reactor vessel.

REMEDIAL ACTION:

With the reactor suberitical for less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, suspend all operations involving movement of irradiated fuel in the reactor vessel.

l TESTING REQUIREMENTS:

16.15-4.9.3 The reactor shall be determined to have been subcritical for at least 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> by verification of the date and time of suberiticality prior to movement of irradiated fuel in the reactor vessel.

BASES:

The minimum requirement for reactor suberiticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products.

f~x This decay time is consistent with the assumptions used in the accident analyses.

(V) 16.15-59 A

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4/99 l

(N 1 16.15 - CTS RELOCATED ITEMS l \- /

, 16.15-3.9.5 COMMUNICATIONS I

COMMITMENTS:

Direct communications shall be maintained between the control room and personnel I at the refueling station.

i APPLICABILITY: During CORE ALTERATIONS. )

l REMEDIAL ACTION:

When direct communications between the control room and personnel at the (

l refueling station cannot be maintained, suspend all CORE ALTERATIONS.

TESTING REQUIREMENTS:

l 16.15-4.9.5 Direct communications between the control room add personnel at the  ;

refueling station shall be demonstrated within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> prior to the start of and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during CORE ALTERATIONS.

BASES:

The requirement for communications capability ensures that refueling station g personnel can be promptly informed of significant changes in the facility status

, j or core reactivity conditions during CORE ALTERATIONS.

%J 16.15-60 s -)

4/99

I i

16.15 - CTS RELOCATED ITEMS

, ~x 16.15-3.9.6 MANIPULATOR CRANE I

COMMITMENTS:

The reactor building manipulator crane and an auxiliary hoist shall be used for movement of fuel assemblies or control rods and shall be OPERABLE with:

a. The manipulator crane used for movement of fuel assemblies having:

1. A minimum capacity of 3250 pounds, and

2. An overload cutoff limit less than or equal to 2900 pounds.

b. Auxiliary hoists used for latching, unlatching and drag load testing of control rods having:

l

1. A minimum capacity of 1000 pounds, and  !

2. A load indicator which shall be used to prevent applying a lifting force in excess of 600 pounds on the core internals.

i APPLICABILITY: During movement of fuel assemblies and control rods within the j reactor vessel. '

REMEDIAL ACTION:

<x

( ) With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend  !

\m / use of any inoperable manipulator crane and/or auxiliary hoist from operations involving the movement of fuel assemblies and control rods within the reactor vessel.

TESTING REQUIREMENTS:

16.15-4.9.6.1 Each manipulator crane used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by performing a load test of at least 3250 pounds and demonstrating an automatic load cutoff when the crane load exceeds 2900 pounds.

16.15-4.9.6.2 Each auxiliary hoist and associated load indicator used for movement of control rods or control rod drag load testing within the reactor vessel shall be demonstrated OPERABLE within 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> prior to the start of such operations by performing a lead test of at least 1000 pounds.

BASES:

The OPERABILITY requirements for the manipulator cranes ensure that: (1) manipulator cranes will be used for movement of drive rods and fuel assemblies, (2) each crane has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

(~'$

5 /

16.15-61 4S9

I

/N 16.15 - CTS RELOCATED ITEMS 16.15-3.9.7 CRANE TRAVEL - SPENT FUEL STORAGE POOL BUILDING COMMITMENTS:

Loads in excess of 300 pounds

• shall be prohibited form travel over fuel assemblies in the storage pool. Truck casks shall be carried along the path outlined in Figure 3.9-1 in the fuel pit and fuel pool area.

l APPLICABILITY: With fuel assemblies in the storage pool.

REMEDIAL ACTION:

With the. requirements of the above specification not satisfied, place the crane load in a safe condition.

I TESTING REQUIREMENTS: .*

16.15-4.9.7, The weight of each load, other than a fuel assembly and control rod, shall be verified to be lets than 3000 pounds prior to moving it over fuel assemblies.

BASES:

The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analysis.

(See Figure 3.9-1 on next page)

Weir gates of the spent fuel pool may be moved by crane over the stored fuel provided the spent fuel has decayed for at least 17.5 days since last being part of a core at power, 16.15-62 b

t 4/99 l

F I-

/ REFUELING OPERATIONS Extreme of Main Hook I

i 1

1 1

A l m

r

/ T& r i

e

/s n M

l "l l H

.a 0

I

( .

s?

E 9

. v e n

.f 5' '

l h/

f 5 \/

ha

• Cask Pit b l &y$m s 2'-6 s

3'-0 i

l u .c K--> > Fuel Pool

! o g '- '

9,4 .x 6 .f= = ( --- l SU EM

=c

[ a o*

a f ne .

"E .

h j  ! 3 9 O h ,

I N Q n, u\ fX -

l

/-  ! \/

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i

. . _ _ . _ - _ _ _ _ _ . . . . . . . _ _ . . ... . .>. . .V.l ' - 1 1/2 l  !

! i i

l @ j l

FIGURE 3.9-1 REQUIRED PATH FOR MOVEMENT OF TRUCK CASKS l

l 16.15-63 i

1 l

4/b !

l l

l

/S 16.15 - CTS RELOCATED ITEMS 16.15-3.9.10 WATER LEVEL - STORAGE POOL i

COMMITMENT: i At least 23 feet of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

APPLICABILITY: Whenever irradiated fuel assemblies are being stored in the storage pool.

REMEDIAL ACTION:

With the requirements of the above SLC not satisfied, suspend all movement of fuel assemblies and crane operations with loads in the fuel storage areas and restore the water level to within its limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

TESTING REQUIREMENTS:

• 1 16.15-4.9.10 The water level in the storage pool shall be determined to be at or above its minimum required depth at least once per 7 days when irradiated fuel l are being stored in the fuel storage pool.

BASES:

7si Tne restrictions on minimum water level ensure that sufficient water depth is g available to remove 99% of the assumed 10% iodine gap activity released from the

\~ rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the accident analysis.

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g 16.15 - CTS RELOCATED ITEMS l

\s 16.15-3.10.5 ROD POSITION INDICATION SYSTEM - SHUTDOWN i

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d COMMITMENT:

The limitations of SLC 16.15-3.1.3.3 may be suspended during the s performance of individual full-length shutdown and control rod drop time '

measurements provided.

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a. Only one shutdown or control bank is withdrawn from the fully inserted position at a time, and

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b. The rod position indicator is OPERABLE during the withdrawal of the rods.*

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APPLICABILITY: MODES 3, 4, and 5 during performance of rod drop time  !

measurements. j i

REMEDIAL ACTION:

With the Position Indication System inoperable or with more than one I bank of rods withdrawn, immediately open the Reactor trip breakers. i TESTING REQUIREMENTS:  !

- 16.15-4.10.5 The above required Rod Position Indication Systems shall '

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be determined to be OPERABLE within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to the start of and  !

at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Rod Position Indication Systems agree:

a. Within 12 steps when the rods are stationary, and

b. Within 24 steps during rod motion.

BASES:

This special test exception permits the Position Indication Systems to be inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE.

• This requirement is not applicable during the initial calibration of the Rod Position Indication System provided: (1) Keff is maintained less than or equal to 0.95, and (2) only one shutdown or control rod bank is withdrawn from the fully inserted position at one time.

16.15-65 A 1 f

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16.15 - CTS RELOCATED ITEMS 16.15-3.11.1.4 LIQUID HOLDUP TANKS COMMITMENTS:

The quantity of radioactive material contained in each unprotected outdoor radwaste tank shall be limited to less than or equal to 10 Curies, excluding tritium and dissolved or entrained noble gases.

APPLICABILITY: At all times.

REMEDIAL ACTION:

With the quantity of radioactive material in any of the above tanks exceeding the above limit, immediately suspend all additions of radioactive material to the tank, within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit, and describe the events leading to this condition in the next Annual Radioactive Effluent Release Report. .

I TESTING REQUIREMENTS:

16.15-4.11.1.4 The quantity of radioactive material contained in each of the above tanks shall be determined to be within the above limit by analyzing a representative sample of the tank's contents at least once per 7 days when 1 radioactive materials are being added to the tank. '

i BASES:

The tanks applicable to this SLC include all those outdoor radwaste tanks that l

are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table II, Column 2, at the nearest potable water supply and the nearest surface water supply in an UNRESTRICTED AREA.

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/"' 16.15 - CTS RELOCATED ITEMS 16.15-3.11.2.5 EXPLOSIVE GAS MIXTURE COMMITMENTS:

The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to less than or equal to 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

APPLICABILITY: At all times.

REMEDIAL ACTION:

a. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 2% by volume but less than or equal 4% by volume, reduce the oxygen concentration to the above limits within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

b. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greater than 4% by volume and the hydrogen concentration greater than 4% by volume, immediately suspend all additions of waste gases to the system and reduce the concentration of oxygen to less than or equal to 4% by volume, and immediately take ACTION a. above.

TESTING REOUIREMENTS:

O 16.15-4.11.2.5 The concentrations of hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM shall be determined to be within the above limits by monitoring the waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and oxygen monitors required OPERABLE by Table 3.3-13 of SLC 16.15-3.3.3.9.

BASES:

This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. Automatic control features are included in the system to prevent the hydrogen and oxygen concentrations from reaching these flammability limits. These automatic control features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, or injection of dilutants to reduce the concentration below the flammability limits. Maintaining the concentration of hydrogen and oxygen below their flammability limits provides assurance that the releases of radioactive materials will be controlled in conformance with the requirements of General Design Criterion 60 of Appendix A to 10 CFR Part 50.

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16.15 - CTS RELOCATED ITEMS 16.15-3.11.2.6 GAS STORAGE TANKS COMMITMENTS:

The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal 'to 49,000 Curies noble gases (considered as Xe-133).

APPLICABILITY: At all times.

AEMEDIAL ACTION:

'With the quantity of radioactive material in any gas storage tank exceeding the above limit,.immediately suspend all additions of radioactive material to the tank and within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> reduce the tank contents to within the limit.

TESTING REQUIREMENTS: I 16.15-4.11.2.6 The quantity of radioactive material contained in each gas storage tank shall be determined to be within the above limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when radioactive materials are being added to the tank.

BASES:

O The specification considers postulated radioactive releases due to a Waste Gas System leak or failure, and limits the quality of radioactivity contained in l

each pressurized gas storage tank in the WASTE GAS HOLDUP SYSTEM to assure that a release would be substantially below the dose guideline values of 10 CFR Part 100 for the postulated event.

Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting total body exposure to a MEMBER OF THE PUBLIC at the nearest exclusion area boundary will not exceed 0.5 rem. This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure," in NUREG-0800, July 1981.

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<~s 16.15 - CTS RELOCATED ITEMS

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16.15-6.8.4.g RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM i i

l COMMITMENT: l l

The following program shall be established, implemented, and maintained:

A program shall be provided to monitor the radiation and radionuclides in the environs of the plant. The program shall provide (1) representative measurements of radioactivity in the highest potential exposure pathways, and (2) verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program shall (1) be contained in SLC, (2) conform to the guidance of Appendix I to 10 CFR Part l 50, and (3) include the following:

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1. Monitoring, sampling, analysis, and reporting of radiation anc radionuclides in the environment in accordance with the methodology and  !

parameters in the ODCM,

2. A Land Use Census to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the monitoring program are made if required by the results of this census, and n

3. Participation in an Interlaboratory Comparison Program to ensure that s independent checks on the precision and accuracy of the measurements of

[ radioactive materials in environmental sample matrices are performed as

\s_J part of the quality assurance program for environmental monitoring.

APPLICABILITY: At all times.

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16.15 CTS RELOCATED ITEMS A

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16.15-6.15.1 MAJOR CHANGES TO RADIOACTIVE LIQUID, GASEOUS, AND SOLID WASTE TP.EATMENT SYSTEMS

• COMMITMENT:

Licensee initiated major changes to the Radioactive waste Systems

-(liquid, gaseous and solid);

a. Shall be reported to the Commission in the Annual Radioactive ,

Effluent Release Report for the period in which the evaluation '

was reviewed by the Station Manager. The discussion of each l change chall contain:

1. A summary of the evaluation that led to the determination that the change could be made in accordance with 10 CFR Part 50.59;

2. Sufficient detailed information to totally support the j l reason for the change without benefit of additional or I

l supplemental information; i l -

l 3. A detailed description of the equipment, components, and processes involved and the interfaces with other plant systems;

4. An evaluation of the change, which shows the predicted l [ releases of radioactive materials in liquid and gaseous I \/ effluents and/or quantity of solid waste that differ from those previously predicted in the License application and amendments thereto;

5. An evaluation of the change, which shows expected maximum exposures to individual in the UNRESTRICTED AREA and to the general population that differ from t hose previously estimated in the License application and amendments thereto;

6. A comparison of the predicted releases of radioactive l materials, in liquid and gaseous effluents and in solid waste, to the actual releases for the period prior to when the changes are to be made;

7. An estimate of the exposure to plant operating personnel as a result of the change; and

8. Documentation of the fact that the change was reviewed and found acceptable by the Station Manager or the Chemistry Manager.

Licensees may choose to submit the information called for in this SLC as part of the annual UFSAR update

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b. Shall become effective upon review and acceptance by a qualified individual / organization.

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16.15-71 4/99 i

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