Text

Revised Oconee Nuclear Station Selected Licensee Commitments, List of Effective Pages
	ML20204B414
Person / Time
Site:	Oconee
Issue date:	03/27/1999
From:	; DUKE POWER CO.
To:	;
Shared Package
ML15112A425	List: ML20204B414;
References
	PROC-990327, NUDOCS 9903220115
	Download: ML20204B414 (600)
	v • d • e

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O U TABLE OF CONTENTS L SECTION M pgg

-g 16.0 SELECTED LICENSEE COMITMENTS 16.1-1 16.1' INTRODUCTION 16.1-1 1

16.2 APPLICABILITY 16.2-1 16.3 DEFINITIONS 16;3-1 16.4 COMITMENTS RELATED TO REACTOR COMPONENTS Pending 16.5 REACTOR COOLANT SYSTEM 16.5.1-1 16.5.1 Reactor Coolant System Vents 16.5.1-1 16.5.2 Low Temperature Overpressure Protection (LTOP) System 16.5.2.1'

'16.5.3 Loss of Decay Heat Removal 16.5.3-1 16.5.4 [ Deleted] 16.5.4-1 l 16.5.5 [ Deleted] 16.5.5-1 16.5.6 [ DELETED) 16.5.6-1 16.5-7. Chemistry Requirements 16.5.7-1 t

16.5.8 Pressurizer 16.5.8-1 l 16.5.9 Testing Following Opening of System (Core Barrel Bolt 16.5.9.1 l Inspections) 16.5.10 RCS Leakage 16.5.10-1 16.5.11 Subcriticality 16.5.11-1 16.Ts.12 RCS Leakage Testing Following Opening of System 16.5.12-1 16.5.13 High Pressure Injection and the Chemical Addition Systems 16.5.13-1 16.6 CGMITMENTS RELATED TO ENGINEERED SAFETY FEATURES (NON-ESF 16.6.1-1 SYSTEMS)

( 16.6-1 Containment Leakage Tests- 16.6.1-1 4

l_ - . . 16.6-2' Reactor Building Post-Tensioning System 16.6.2-1 16.6.3 Containment Heat Removal Verification Frequency 16.6.3-1 16.0-1 03/27/99 L

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TABLE OF CONTENTS (continued)

SECTION TITLE PAGE No 16.6.4 Low Pressure Injection System Leakage 16.6.4-1 !

16.6.5 Core Flood Tank Discharge Valve Breakers 16.6.5-1 l 16.6.6 Core Flooding System Test 16.6.6-1 16.6.7 BWST Outlet Valve Control 16.6.7-1 1

1 16.6.8 LPI System Valve Test Restrictions 16.6.8-1 l 16.6.9 Containment Purge Valve Testing 16.6.9-1 16.6.10 Containment Hydrogen Control Systems 16.6.10-1

'16.6.11 Reserved l 16.6.12 Additional High Pressure Injection (HPI) Requirements 16.6.12-1 16.7 INSTRUMENTATION 16.7.1-1 l 16.7.1 Accident Monitoring Instrumentation 16.7.2-1 16.7.2 Anticipated Transient Without Scram 16.7.2-1 16.7.3 Emergency Feedwater System 16.7.3-1 16.7.4 Deleted 16.7.4-1 l 16.7.5 Steam Generator Overfill Protection 16.7.5-1 16.7.6 Deleted 16.7.6-1 16.7.7 Position Indicator Channels 16.7.7-1 16.7.8 Incore Instrumentation 16.7.8-1 16.7.9 RCP Monitor 16.7.9-1 16.7.10 Core Flood Tank Instrumentation 16.7.10-1 16.7.11 Display Instrumentation 16.7.11-1 16.7.12 SSF Diesel Generator (DG) Air Start System Pressure 16.7.12-1 Instrumentation 16.7.13 SSF Instrumentation 16.7.13-1 i 16.0-2 03/27/99 l l

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L TABLE OF CONTENTS (continued) l SECTION TITLE PAGE

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'16.8 ELECTRIC POWER SYSTEMS 16.8.1-1 l

16.8.1' Control 'of Room Temperatures. for Station Blackout 16.8.1-1 16.8.2 Deleted 16.8.2-1 16.8.3 Power Battery Parameters 16.8.3-1 16.8.4 Keowee Operational Restrictions 16.8.4-1 16.8.5 125VDC Vital I&C' System Ground Locating Policy 16.8.5-1 I l

16.8.6 Lee / Central Alternate Power System- 16.8.6-1 l 16.8.7 Auctioneering Diodes 16.8.7-1 16.8.8 - External Grid Trouble Protection 16.8.8-1 l 16.9 AUXILIARY SYSTEMS 16.9.1-1 p 16.9.1 Fire Suppression Water System 16.9.1-1 1

16.9.2 Sprinkler and Spray Systems 16.9.2-1 16.9.3 Keowee CO2 Systems 16.9.3-1 l 16.9.4 Fire Hose Stations 16.9.4-1 i i 16.9.5 Fire Barriers' 16.9.5-1 16.9.6 Fire Detection Instrumentation 16.9.6-1 16.9.7 Keowee Lake Level 16.9.7-1 L

l- 16.9.8 HPSW Pump Requirement to Support LPSW 16.9.8-1 l 16.9.8a HPSW System Requirements to Support Loss of LPSW lb.9.8a-1 16.9.9 Auxiliary Service Water System and Main Steam Dump Valve 16.9.9-1 Operability requirements t ,

16.9.10 Component Cooling and HPI Seal Injection to Reactor 16.9.10-1 l Coolant' Pumps !

16.9.11 Turbine Building Flood Protection Measures 16.9.11-1 lO 16.0-3 03/27/99

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V TABLE OF CONTENTS (continued)

SECTION TITM PAGE

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16.9.12 Additional Low P~ ssure Service Water (LPSW) 16.9.12-1 System Operabil1 ; Requirements 16.9.13 Spent Fuel Cooling System 16.9.13-1 ,

16.9.14 SSF Diesel Generator (DG) Inspection Requirements 16.9.14-1 16.9.15 Radioactive Material Sources 16.9.15-1 l

16.9.16 Reactor Building Polar Crane and Auxiliary Hoist 16.9.16-1 (RCS System Open)-

16.9.17 Reactor Building Polar Crane (RCS at elevated 16.9.17-1 temperature and pressure) 16.9.18 Sn!Abers 16.9.18-1 16.10 COMMITMENTS RELATED TO STEAM & POWER CONVERSION SYSTEMS 16.10.1-1 16.10.1 Condensate Inventory Regtrements for Emergency Feedwater 16.10.1-1 O

(j 16.10.2 Steam Generator Secondary Side Pressure and 16.10.2-1 Temperature (P/T) Limits 16.10.3 Emergency Feedwater (EFW) Pump and Valve Testing 16.10.3-1 16.10.4 Low Presssure Service Water System Testing 16.10.4-1 16.10.5 Main Steam Line Break (MSLB) Feedwater Isolation Features 15.10.5-1

. 16.10.6 Emergency Feedwater Controls 16.10.6-1 16.11 RADIOLOGICAL EFFLUENTS CONTROL 16.11.1-1 16.11.1 Radioactive Liquid effluents 16.11.1-1 16.11.2 Radioactive Gaseous Effluents 16.11.2-1 16.11.3 Radioactive Effluent Monitoring Instrumentation 16.11.3-1 16.11.4 Operational Safety Review 16.11.4-1 16.11.5 Solid Radioactive. Waste 16.11.5-1 16.11.6 Radiological Environmental Monitoring 16.11.6-1 ]

L 16.11.7 Dase calculations 16.11.7-1 16.0-4 03/27/99 i

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d TABLE OF CONTENTS (continued)

SECTION TITLE EAG1 M

16.11.8 Reports 16.11.8-1 l 16.11.9 Radioactive effluent release report 16.11.9-1 16.11.10 Radiological Environmental Operating Reports 16.11.10-1 l 16.11.11 Iodine Radiation Monitoring Filters 16.11.11-1 l

l 16 11.12 Radioactive Material in Outside Temporary 16.11.12-1 l Tanks Exceeding Limit l 16.11.13 Radioactive Material in Waste Gas Holdup 16.11.13-1 Tank Exceeding Limit i

16.11.14 Explosive Gas Mixture 16.11.14-1 16.12 REFUELING OPERATIONS 16.12.1-1 16.12.1 Decay Time for Movement of Irradiated Fuel 16.12.1-1 16.12.2 Area Radiation Monitoring for Fuel Loading and Refueling 16.12.2-1 O 16.12.3 Communication Between Control Room and Refueling Personnel 16.12.3-1 16.12.4 Handling of Irradieted Fuel Assemblies 16.12.4-1 16.12.5 Loads Suspended over Spent Fuel in Spent fuel Pool 16.12.5-1 16.13 CONDUCT OF OPERATION 16.13.1-1 16.13.1 Fire Brigade 16.13.1-1 16.13.2 Technical Review and Control 16.13.2-1 16.13.3 Plant Operations Review Committee 16.13.3-1 l

16.13.4 Reactivity Anomaly 16.13.4-1 16.13.5 Additional Operating Shift Requirements 16.13.5-1 16.13.6 Retraining and Replacement of Station Personnnel 16.13.6-1 16.13.7 Procedures for Control of Ph in Recirculated 16.13.7-1 Coolant after Loss-of-coolant Accident & Long-term Ercrgency Core Cooling Systems 16.13.8 . Respiratory Protective Program 16.13.8-1

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TABLE-0F CONTENTS (continued)

SECTION TlT_ll . 11G1 ,

N0 l 1

16.13.9 Startup Report 16.13.9-1 16.13.10 Core Operating Limits Reports 16.13.10-1 16.13.11 Procedure for Station Survey Following an Earthquake 16.13.11-1 16.14 CONTROL RODS AND POWER DISTRIBUTION 16.14.1-1 ,

1 16.14.1 APSR Movement 16.14.1-1 16.14.2 Control Red Program Verification 16.14.2-1 16.14.3 Power Mapping 16.14.3-1 16.14.4' Control Rod Drive Patch Panels 16.14.4-1 16.15 VENTILATION FILTER TESTING PROGRAM 16.15.1-1 16.15.1 Penetration Room Ventilation Room System Testing 16.15.1,1 16.15.2 Control Rooir. Pressurization and Filtering System 16.15.2-1 0 16.15.3 Spent Fuel Pool Ventilation System 16.15.3-1 I

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INTRODUCTION Q 16.1 16.0 SELECTED LICENSEE COMMITMENTS

16.1 INTRODUCTION

This chapter providesLa single location in the UFSAR where certain selected licensee commitments are presented. The content of this chapter is based on the results of application of a' set of criteria to determine the content of technical specifications. For purposes of administrative ease, this c' iter is maintained in_ a separate manual, The Oconee Nuclear Station Selected

-Licensee Commitments Manual. Those previous technical specification requirements which did not meet the criteria are relocated in this chapter.

The control of the Oconee Nuclear Station selected licensee commitment program and manual shall be in accordance with approved directive NSD 221 Facility Operating License and Technical Specifications Amendments / Selected Licensee-Commitments / Technical Specifications Bases Changes. The manual is officially designated as Chapter 16 of the Oconee UFSAR. The original issue and subsequent revisions of the manual are approved by the station manager or his designee. Administrative requirements of the manual are the responsibility of the Site Regulatory Compliance Section. l Changes to_these Selected Licensee Commitments may be made, pursuant to

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10CFR50.59, only after the bases for the requirement have been clearly established and after a multidisciplinary review by Qualified Reviewers, including onsite Operations personnel.

Additional NRC commitments, as selected by the Station manager or designee may be located in this chapter. It is the intent of this chapter to provide information regarding systems that are a part of the licensing basis, as described in the UFSAR, but are not of such a level of importance that they need to be under the rigorous control provided by technical specifications.

This chapter includes Surveillance Requirements for.certain systems, and remedial actions to be taken in the event the system is inoperable. A bases for the commitment is also provided. Reference is also provided to specific sections of the UFSAR where the information relative to the commitment is further described.

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l N APPLICABILITY l'(V 16.2 1

16.2 APPLICABILITY This section provides the general requirements applicable to each of the Commitments within Section 16.0, Selected Licensee Comitments.

16.2.1 Commitments shall be met during the MODES or other specified conditions in the Applicability.

16.2.2 Upon discovery of a failure to meet a Comitment, the associated Actions shall be met. If the comitment is met or no longer required prior to expiration of the specified time intervals, completion of the Actions is not required.

16.2.3 When a Comitment is not met, except as provided in the associated Actions, the Station Manager and/or responsible Group Superintendent will determine any further actions.

Where corrective measures are completed that permit operation under the Actions, the action may be taken in accordance with the specified time limits as measured from the time of failure to meet the Comitment. Exceptions to these requirements are stated in the individual Comitments.

16.2.4 Entry into a MODE other specified condition in the Applicability

/O must be made with: (1) the full complement of required systems, equipment, or V components OPERABLE, and (2) all other parameters as specified in the Commitment being met without regard for allowable deviations and out-of-service provisions contained in the Actions.

The intent of this provision is to ensure that facility operation is not initiated with either required equipment or systems inoperable or other specified limits being exceeded without the approval of the Station Manager and/or the responsible Group Superintendent.

Exceptions to this provision have been provided for a limited number of Comitments when startup with inoperable equipment would not affect plant safety. These exceptions are stated i- the Actions of the appropriate Comitments.

When a comitment is not met, entry into a MODE or other specified condition in the Applicability shall not be made except when the associated Action to be entered permits continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time unless approved by the Station manager and/or responsible Group Superintendent. This comitment shall not prevent changes in MODE or other conditions specified in the Applicability that are required to comply with Actions. Exception to this comitment are stated in the individual Comitment.

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O APPLICABILITY 16.2 16.2.5 Commitments, including the associated Actions, shall apply to each individually unless otherwise indicated as~ follows:

Whenever the Commitment refers to systems or components which are ,

shared by units, the Actions will apply to all affected units -l simultaneously.

l Whenever the Commitment applies to only one unit', this will' be identified in the comitment section of the Comitment; and Whenever certain portions of a Comitment contain operating parameters, setpoints, etc. which are different for each unit, this will be identified in parentheses or footnotes.

-16.2.6 Surveillance Requirements shall be performed during the. MODE or other. specified condition in the Applicability for individual Comitments unless otherwise stated in an individual Surveillance Requirement or reference.

Failure to meet a. Surveillance Requirement, whether such failure is experienced during the performance of the Surveillance Requirement or between C performances of the Surveillance Requirement, shall be failure to meet the

commitme:st. Failure to perform a Surveillance Requirement within the Specified Frequency shall be failure to meet the Comitment except as provided in Commitment 16.2.8. Surveillance Requirements do not have-to be performed .

..on inoperable equipment or variables outside specified' limits.

Surveillance Requirements are necessary to ensure the Comitments are met and will be performed during the MODE or other specified condition in the Applicability. Provisions for additional Surveillance Requirements to be performed without regard to the applicable operational condition or MODE or other specified conditions in the Applicability are provided in the individual Surveillance Requirements. Surveillance Requirements for Special Conditions ]

need only be performed when the Special Condition is being utilized as an )

exception to an individual Comitment. ~

-16.2.7 . Each Surveillance Requirement shall be performed on its specified

. frequency with'a maximum allowable extension not to exceed 50% of the test frequency, unless specified dif.ferently in.the individual Comitment. ,

Allo'wable tolerances are provided for performing Surveillance Requirements beyond those 'specified in the nominal testing frequency. The tolerance is j necessary to provide operational flexibility because of scheduling and '

performance considerations. The phrase "at least" associated with a testing ;

frequency.does not negate this allowable tolerance value, and permits the i performance of more frequent surveillance activities. For frequencies specified as "once" the above extension does not apply to the first performance.

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. APPLICABILITY 16.2 16.2.8 Exceptions to these requirements are stated in the individual Commitments or may be approved by the Station Manager and/or responsible Group Superintendent.

' 16.2.9 Surveillance Requirements associated with a Commitment shall be performed within the specified time interval prior to entry into a MODE or other specified condition in the Applicability. The intent of this provision is to ensure that Surveillance Requirement have been satisfactorily demonstrated on a current basis as required to meet the OPERABILITY requirements of the Commitment unless otherwise approved by the Station manager and/or the responsible Group Superintendent.

Entry into a MODE or other specified condition in the Applicability shall not i

.be made unless the Surveillance Requirement associated with the Commitment have been performed within the specified frequency or as approved by the !

Station Manager and/or responsible Group Superintendent. These provisions shall not prevent entry into MODES or other specified conditions in the !

Applicability that are required to comply with Actions.

16.2.10 Surveillance Requirement shall apply to each unit individually unless otherwise indicated or whenever certain portions of a specification contain testing parameters different for each unit, which will he identified O 'in notes.

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DEFINITIONS

16.3 .

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l 16.3 DEFINITIONS The definitions in the Oconee Technical Specifications apply to defined terms used herein. The following additional defined terms are applicable throughout this Selected Licensee Connitment document:

16.3.1 DELETED j 16.3.2 SOLIDIFICATION - SOLIDIFICATION shall be the immobilization of wet radioactive wastes such as evaporator bottoms, spent resins, sludges, and reverse osmosis concentrations as a result of a process of thoroughly mixing the waste type with a solidification agent (s) to form a free standing monolith with chemical and physical characteristics specified in the Process Control Program (PCP).

16.3.3' A GASEOUS RADWASTE TREATMENT SYSTEM is any system designed and.

installed to reduce radioactive gaseous effluents by collecting primary 4 coolant system offgases from the primary system and providing for delay or ,

holdup for the purpose of reducing the total radioactivity prior to release to '

the environment. ,

4 Q 16.3.4 A VENTILATION EXHAUST TREATMENT SYSTEM is any system designed and V installed to reduce ~ gaseous radioiodine 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. Engineered Safety Feature (ESF) atmospheric cleanup systems are not considered to be Ventilation Exhaust Treatment System components.

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k Reactor Coolant System Vents 16.5.1

t 16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.1 Reactor Coolant System Vents COMMITMENT a. The following reactor coolant system vent paths shall be

' OPERABLE:

1) Reactor Vessel Head Vent

2) Pressurizer Steam Space Vent (through PORV)

-3) RCS Loop A High Point. Vent

4) RCS Loop B High Point Vent

b. For each vent path, two electrically-operated valves shall be capable of being opened, and all manual-valyc:

shall be open. 1 APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS I

' CONDITION REQUIRED ACTION COMPLETION TIME O A. One RCS vent path inoperable.

A.1 Restore to OPERABLE status.

30 days l

'B. Tim or more RCS vent B.1 Restore to OPERABLE 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months paths inoperable. status.

C. The Required Actions C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and associated Completion Times of AND Condition A or B not met. C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months l

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Reactor Coolant System Vents 16.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.5.1.1 Verify an open flow path for each RCS vent 18 months path by testing the head vents and loop high point vents.

SR 16.5.1.2 Perform high point vent valve testing. In accordance with ASME Section XI DbAS.fS Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core cooling. The RCS vents have two valves in series which are capable of '

O being powered from emergency buses. The valves are normally closed with power removed to prevent inadvertent opening of the valves. In order for a vent

' path to perform its intended safety function of venting, the two-electrically-operated valves in the flow path must be capable of being opened, and all manual valves must be open. ,

REFERENCES

1. NUREG 0737, Item II.B.1

2. Generic Letter 83-37 l

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r Low Temperature Overpressure Protection System

. 16.5.2 16.5 REACTOR COOLANT SYSTEM (RCS)

~

- 16.5.2 Low Temperature Overpressure Protection System COMITMENT The following controls are required to meet the provisions '

of ITS 3.4.12:

a. RCS pressure:

< 345 psig when RCS temp s 220*F, and

< 450 psig when RCS temp > 220*F and s 325'F.

b. Pressurizer level maintained within the following limits when RCS pressure is > 100 psig:

s 220 inches when RCS temp s 220*F, and

< 260 inches when RCS temp > 220*F and s 325'F.

c. Pressurizer level maintained within the following limits when RCS pressure is s 100 psig:

s 310 inches when one or more HPI pumps are running, and ;

when RCS temp a 150*F and s 220'F, and

s 380 inches when RCS temp s 160*F and no HPI pumps are running.

d. Makeup flow restricted with HP-120 travel stops to:

Units 1 & 2 s 102.3 gpm Unit 3 s 84.5 gpm. ,

e. Three audible Pressurizer level alarms a 225 inches, ,

a 260 inches, and a 315 inches from the temperature I compensated Pressurizer level indication.

f. Two audible RCS pressure alarms a 345 psig and a 450 psig.

g. High pressure nitrogen system administrative 1y controlled 3 to prevent inadvertent pressurization of the RCS.

APPLICABILITY: MODE 3 when any RCS cold leg temperature is s 325'F, ,

MODES.4, 5, and 6 when an RCS' vent path capable of mitigating the most limiting LTOP event is not open.

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l Low Temperature Overpressure Protection System 16.5.2 O,

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more controls A.1 Enter applicable Immediat21y not met. Condition of ITS 3.4.12.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.5.2.1 Verify travel stops limit flow through 18 months HP-120 to the specified limit.

SR 16.5.2.2 Perform Channel Calibration on pressurizer O' level and RCS pressure alarms.

18 months SR 16.5.2.3 Perform an inspection of the PORV. every 2 refueling cycles I

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Low Temperature Overpressure Protection Systea 16.5.2 l

{ (~~

SURVEILLANCE FREQUENCY l SR 16.5.2.4 -------------------NOTE--------------------

Only required to be met when vent (s) are being used for overpressure protection.

................__....__............s_.....

Verify valves in the flowpath for the RCS 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for vent (s) are open. valves not locked, sealed, or otherwise l j secured open I AND l l

31 days for valves locked, l sealed, or otherwise secured open l

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16.5.2-3 03/27/99

Low Temperature Overpressure Protection System 16.5.2 E8hli BACKGROUND

- The requirement (s) of Surveillance Requirement SR 16.5.2.3 and SR 16.5.2.4 were relocated from the CTS 4.2.6.c and 4.2.7 during the conversion to ITS.

The Low Temperature Overpressure Protection (LTOP) System protects the reactor vessel against damage due to excessive pressures at low temperatures by.

limiting the pressure of transients to below the limits of 10 CFR 50 Appendix G using a conservative safety. factor of 1.5.. The first train of the LTOP-system is the PORV. The second LTOP train is comprised of the controls which assure that 10 minutes are available for operator action to mitigate an LTOP event. This SLC.is provided to establish the requirements for operability of the. second train of the system in accordance with ITS 3.4.12.

APPLICABLE SAFETY ANALYSIS Analysis of LTOP events are described in Section 5.2 of the UFSAR. Scenarios having the potential to result in a LTOP event are described in the bases for ITS 3.4.12. The inadvertent addition of nitrogen has also been identified as a potential LTOP event.

.5LC1 The requirements of the SLC have been provided to assure 10 minutes are available for operator action to mitigate an LTOP event.

APPLICABILITY

- The SLC is applicable when the provisions of ITS 3.4.12 are applicable. The vahe for RCS temperature is' based on analyses used to develop the 10CFR50 Appendix G pressure temperature limits using a safety factor of 1.5. This limit is provided in ITS Figures 3.4.12-3, 3.4.12-6, and 3.4.12-9, the Inservice Leak and Hydrostatic Test heatup and cooldown limitations applicable for the first 21 EFPY for Units 1* and 3 and for the first 19 EFPY for Unit 2.

This SLC is not applicable for operating conditions above 325'F since the possibility of non-ductile failure is significantly diminished. Vent paths capable of. mitigating the most limiting LTOP event are specified in Operations procedures. If an LTOP event were to occur, violation of this SLC could l result in exceeding the brittle fracture pressure limits, overstressing the reactor vessel and closure head, or require reanalysis to demonstrate the i resulting stresses would not impair further operation. )

ACTIONS ad If one or more of the_ restrictions in the SLC is not met, 10 minutes will not '

p be available for operators to mitigate potential LTOP events. In addition, if

_ V the requirements for deactivation of the HP1 System and CFTs are not met, the

. PORV may not have sufficient relief capacity to mitigate the associated LTOP :

events. The Completion Time of "Immediately" simply requires that there be no i

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Low Temperature Overpressure Protection System 16.5.2 delay in implementing the requirements of ITS 3.4.12. ITS 3.4.12 reouires that compensatory measures be established within four hours to monitor for initiation of an LTOP event; or within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the RCS will be depressurized and a vent path capable of mitigating the most limiting LTOP event shall be opened. The compensatory measures must be sufficient to identify and mitigate potential LTOP events promptly. For example, establishing a dedicated LTOP operator will in many cases be an adequate compensatory measure.

SURVEILLANCE RE0UIREMENTS The identified surveillance requirements are provided to assure that the second train of LTOP is functioning properly and gives the operator 10 minutes to mitigate an LTOP event.

REFERENCES:

1. ITS 3.4.12

2. 10 CFR 50 Appendix G " Fracture Toughness Requirenients."

3. Calc. File OSC-5355 " Revision of Unit 1. 2. and 3 LTOP Operations Res+rictions for 21.0. 19.0. and 21.0 EFPY Brittle Fracture Limits, Rupertively: Issued June 9, 1993.

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Loss of Decay Heat Removal (Q 16.5.3 l V

16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.3 Loss of Decay Heat Removal COP 9ilTMENT The following conditions shall be met:

a. Conduct a containment closure survey to identify containment penetrations that would need to be closed in the event of a loss of decay heat removal capability and to ensure that containment closure can be achieved within 2 1/2 hours,

b. Two operable core exit thermocouple indications and alarm shall be available. The core exit temperature shall'be monitored and recorded at least once every two hours,

c. The LT-5 Reactor vessel level indication system shall be available and operable.

d. An ultrasonic Reactor vessel level detection system, or p other backup level indi ting system, shall be available y .

and operable in additi' to LT-5.

e. Both Main Feeder Buses (MFB) shall be energized.

f. Two sources of power shall be available to supply the Main Feeder buses.

g. Two of the following means of adding inventory to the RCS are avallable and operable:

1. A gravity flow path from the BWST

2. One Bleed Transfer Pump (BTP) and connecting piping

3. A High Pressure Injection (HPI) pump

h. Both steam generators upper primary side handhole covers, or equivalent RCS vent path, shall be removed.

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O 16.5.3-1 03/27/99

Loss of Decay Heat Removal 3 16.5.3 (G

i. Testing and maintenance activities shall be reviewed to ensure no adverse effects on systems and components l required for decay heat removal. Those activities which pose a substantial threat to decay heat removal capability will be prohibited.

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

1. Commitment b is required to be met only when the reactor ,

vessel head is in place. l

2. Commitment d is a Duke Power internal comitment, not a NRC commitment.

1 1

APPLICABILITY: MODE 5 and 6 with RCS level < 50 inches above centerline of I the reactor vessel hot leg and with irradiated fuel in the reactor vessel.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i

- 1 A. One or more commitment A.1 As determined by NA I not met. Station Manager or '

Responsible Group Superintendent.

I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.5.3.1 NA NA BASES Generic Letter 88-17, Loss of Dscay Heat Removal, was issued October 17, 1988 by the Nuclear Regulatory Commission (NRC) due to numerous events in the industry involving a loss of decay heat removal capability while the reactor vessel was in a drained down condition. The major concer.' is that substantial O

16.5.3-2 03/27/99

l Loss of Decay Heat Removal 16.5.3 core decay heat may pose a significant likelihood of a release due to a severe core damage. accident.

Babcock ar.d Wilcox (B&W) designed plants are not as sensitive to the loss of decay heat removal problems, but the 4dditional actions of this commitment provide added assurance of avoiding possible problems while the reactor vessel is in a reduced inventory condition..

REFERENCES:

1. Generic Letter 88-17

2. ONS responses to Generic Letter 88-17, dated Jar.uary 3,1989 and February 2, 1989.

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16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.4 ----

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DELETED 16.5.5 16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.5 -----DELETED-----

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16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.6 -----DELETED-----

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16.5.6-1 03/27/99

Chemistry Requirements 4

16.5.7 16.5 REACTOR COOLANT SYSTEM (RCS)

-16.5.7 Chemistry Requirements COMITMENT The concentration of oxygen, chloride and fluoride in the

RCS and pressurizer shall be maintained within limits.

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

Prior to exceeding 525'F, the limits of SR 16.5.7.1 shall be met.

APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more chemistry A.1 Initiate action to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months limits not met in MODE restore to within

3. limits.

AND A.2 Restore to within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months limits.

B. Required Action and B.1 Be in MODE 5. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months associated Completion l Time for Condition A not met.

C. One or more chemistry C.1 Initiate action to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months limits not met in MODE restore to within 1 or MODE 2. limits.

AND C.2 Restore to within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months limits. i O

16.5.7 1 03/27/99 i

Chemistry Requirements 16.5.7 l

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Actions and 0.1 Be in MODE 3. 2a hours associated Completion Times for Condition C AND not met.

D.2 Be in MODE 5. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months E. RCS oxygen level E.1 Be in MODE 3. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months

, exceeds 1.00 ppm.

AND O_8 i E.2 Be in MODE 5. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months RCS chloride or i

fluoride level exceeds 1.50 ppm. l O

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Chemistry Requirements 16.5.7 i

I' SURVEILLANCE REQUIREMENTS SURVEILLANCE !

FREQUENCY SR 16.5.7.1 --------------------NOTES------------------

1. Not applicable to oxygen concentration '

in the pressurizer in MODE 3 with RCS temperature s 525'F.

2. Not required to be performed for the pressurizer.

Verify fluoride, chloride and oxygen three times per concentrations in the RCS and the week pressurizer are s 0.15 ppm, s 0.15 ppm, and s 0.10 ppm, respectively. i i

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SR 16.5.7.2 --------------------NOTE-------------------

Not required to be performed for the pressurizer.

]

l Verify oxygen concentration in the N/A I pressurizer is s 0.20 ppm. 1 SR 16.5.7.3 Analyze the RCS tritium level. 92 days.

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Chemistry Requirements 16.5.7 I

BASES BACKGROUND l This commitment' incorporates the changes to the Technical. Specifications that ONS proposed.in a letter dated February 20, 1996. The Technical Specifications revision, approved on February 19, 1997, removed several chemistry requirements from the Oconee Technical Specifications. The limits and surveillance for the oxygen, chloride, and fluoride concentration levels were .

relocated to the SLC as stated in the proposed amendment. In addition, the relocation of the tritium analysis from the Technical Specifications to the SLCs was covered in the proposed amendment.

APPLICABLE SAFETY ANALYSIS By maintaining the chloride, fluoride, and oxygen concentration in the reactor coolant within limits, the integrity of the reactor coolant system is protected against potential stress corrosion attack.

Oxygen is normally expected to be below detectable limits. Oxygen is normally below limits since dissolved hydrogen is used when the reactor u critical and s a residual of hydrazine is used in the pressurizer when the reactor is subcritical to control the oxygen. The requirement that the oxygen s concentration not exceed 0.1 ppm in the RCS when in MODES I and 2 and when in i MODE 3 when exceeding 525'F provides added assurance that stress corrosion i cracks will not occur.

During the startup of an Oconee unit the Oconee unit will be held at MODE 4 for initial cleanup. The control of the chloride and fluoride is accomplished through the use of purification demineralizers and it is essential that this control be established during startup since it cannot be reasonably expected that normal power operation will change their performance in any way. This ;

approach, which is recommended by the EPRI PWR Primary Water Chemistry l Guidelines, during startup reduces the possible negative effects on material integrity at high temperatures by establishing these specifications at reasonably achievable low levels before entering MODE 3. i l

APPLICABILITY SLC 16.5.7 is applicable in MODES 1, 2, and 3.

During heatup from MODE 5, oxygen control is accomplished through the addition of hydrazine.to the pressuirizer or hydrogen to the RCS. The reduction of pressurizer oxygen during heatup is influenced by the following considerations.

. a. Venting of the pressurizer during startups below 600 psig is not

O permitted due to the low range pressure transmitter for the RCS being on the same line.

b. Pressurizer volume is small conipared to the total RCS volume.

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I Chemistry Requirements 16.5.7 j

c. Influx of water with oxygen. scavenging hydrazine is limited by the operating constraints placed on the pressurizer to control system pressure.

Significantly higher temperatures exist in the pressurizer than in the reactor coolant system in the preliminary heatup. Because of the problems associated with oxygen scavenging in the pressurizer, a maximum of 0.20 ppm oxygen (twice the specified normal power operation RCS concentration) is allowed prior to entering MODE 3 and time constraints are placed on operating at this concentration in MODES 1, 2, and 3 to ensure all possible effort is made to bring the value below 0.10 ppm as rapidly as reasonably achievable.

ACTIONS l A.1 and A.2 i i

If one or more chemistry limits are not met in MODE 3, actions will be initiated within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to restore the chemistry levels to within the l commitment requirements. The RCS chemistry levels shall be restored to within I the commitment requirements within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

1 B.1 and B.2 O If the Required Action and associated Completion Time for Condition A are not met, the unit shall be placed in MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

L.1 l If the oxygen, chloride, or fluoride limits are exceeded in MODES 1 and 2, ,

measures can be taken to correct the condition (e.g., switch to the spare '

demineralizer, replace the ion exchange resin, increase the hydrogen concentration in the makeup tank, etc.). The oxygen and halogen limits specified are at least an order'of magnitude below concentrations which could I

result in damage to materials found in the reactor coolant system even if l maintained for an extended period of time. Thus, the period of eight hours to

initiate corrective action and the period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to perform corrective action to restore the concentration within the limits have been established.

! The eight hour period to initiate corrective action allows time to ascertain that the chemical analyses are correct and to locate the source of contamination.

t D.1 and D.2 If the Required Action and associated Completion Time for Condition C are not met, the unit shall be placed in MODE 3 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and MODE 5 within 72 l hours.

E.1 and E.2 If the RCS chemistry levels exceed 1.00 ppm for oxygen and 1.50 ppm for chloride and fluoride, the unit shall be placed in MODE 3 within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and L

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Chemistry Requirements O 16.5.7-1 g l MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This requirement is consistent with the action level l 3 requirements in +he EPRI PWR Primary Water Chemistry Guidelines. '

SURVEILLANCE REQUIREMENTS SR 16.5.7.1 This'SR ensures that the RCS chemistry levels are within the SLC requirements.

The 3 time per week Frequency is considered adequate to provide such 1 assurance. '

SR 16.5.7.2 This SR establishes the limit for oxygen concentration in the pressurizer. No ,

specific frequency is required and the Note states the SR is not required to l be performed. Chemistry procedures provide appropriate controls for samplin9 the pressurizer. This presentation establishes the limit for this parameter l without requiring actual performance of the SR. '

SR 16.5.7.3 This SR analyzes the tritium level every 92 days. !

REFERENCES

1. UFSAR, Section 5.2.1.7

2. UFSAR, Section 9.3.1-2

3. Stress Corrosion of Metals, Logan

4. Corrosion and Wear Handbook, O. J. DePaul, Editor

5. EPRI PWR Primary Water Chemistry Guidelines

6. Duke letters to the NRC dated February 20, 1996, and October 16. 1996.

7. NRC Safety Evaluation dated February 19, 1997.

1 16.5.7-6 03/27/99

P- surizer i 16.5.8 l 16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.8 Pressurizer COMMITMENT a. The pressurizer shall have a steam bubble and a water level > 80 inches in MODES 1 and 2.

b. The pressurizer heatup and cooldown rates shall be s 100*F/hr. I 1

c. The pressurizer spray shall not be used if the l temperature difference between the pressurizer and the !

spray fluid is > 410*F.

APPLICABILITY: At all times. i ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME 1 4O A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.5.8.1 N/A N/A BASES The requirement (s) of this SLC section were relocated from CTS 3.1.2.6 and 3.1.3.4 during the conversion to ITS.

The requirement for pressurizer bubble formation and specified water level when the reactor is less than 1% subcritical will assure that the reactor

[' coolant system cannot become solid in the event of a rod withdrawal accident or a startup accident.

16.5.8-1 03/27/99

Pressuri;. r 16.5.8 O. 1 The major components of the reactor coolant pressure boundary have been analyzed in accordance with Appendix G to 10 CFR 50. Results of this analy- .

sis, including the actual pressure-temperature limitations of the reactor '

coolant pressure boundary, are given in BAW-1699 and BAW-1697.

The spray temperature difference is imposed to maintain the thermal stresses at the pressurized spray line nozzle below the design limit.

REFERENCES UFSAR Section 15.3 1

O

~

o 16.5.8-2 03/27/99

-_ .. . . - - . - -- - .- .-. - - _ , _ . _ - _ - =. . -

l Testing Following Opening of System (Core Barrel Bolt Inspections) g- 16.5.9 l V l 16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.9 Testing Following Opening of System (Core Barrel Bolt Inspections)

) l l ;

1 COMITMENT Two sets of main internal bolts (connecting the core barrel '

to the core support shield and to the lower grid cylinder) shall remain in place and under tension.

APPLICABILITY: MODES 1, 2, 3, 4, and 5. j ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 16.5.9.1 Visually inspect the core barrel to lower Whenever the grid cylinder welded bolt locking caps. internals are l removed from i the vessel 16.5.9.2 Visually inspect the core barrel to core ------NOTE-----

support shield welded bolt locking caps. The provisions of SLC 16.2.7 do not apply.

L 18 months +25%

AND i

Whenever the i

internals are

~i p removed from i

d the vessel 16.5.9-1 03/27/99

Testing Following Opening of System (Core Barrel Bolt Inspections) 16.5.9 O

BASES The requirement (s) of this SLC section were relocated from CTS 4.2.2 during the conversion to ITS.

To assure the structural integrity of the reactor internals throughout the life of the unit, the two sets of main internals bolts (connecting the core

, barrel to the core support shield and to the lower grid cylinder) must remain in place and under tension. This is verified by visual inspection to determine that the welded bolt locking caps remain in place.

REFERENCES N/A 4

3 O

3 4

e io 16.5.9-2 03/27/99

1 RCS Leakage j q 16.5.10 NJ 16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.10 RCS Leakage COMilTMENT RCS leakage, including loss of reactor coolant through reactor coolant pump seals and system valves to connecting systems which vent to the gas vent header and from which coolant can be returned to the reactor coolant system, shall !

. be :s; 30 gpm.

j APPLICABILITY: MODES 1, 2, 3, and 4 i ACTIONS 1, CONDITION REQUIRED ACTION COMPLETION TIME 1 I

. A. RCS Leakage evaluated A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i as unsafe.

pg 1 i A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months l 1 RCS leakage > limit.

\

! SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i- SR 16.5.10.1 Initiate evaluation of safety implications Once within 4 of RCS leakage. hours of detection u

l 1

! E8Efd i

The requirement (s) of-this SLC section were relocated from the CTS.I.6.2, 3.1.6.6, and 3.1.6.9 during the conversion to ITS

-' Water inventory balances, radiation monitoring equipment, boric acid l crystalline deposits, and physical inspections can disclose reactor coolant 16.5.10-1 03/27/99 1

j RCS Leakage 16.5.10 leaks. Any leak of radioactive fluid, whether from the reactor coolant system 4

primary boundary or not can be a serious problem with respect to in-plant

radioactivity contamination and cleanup or it could develop into a still more i serious problem; and therefore, first indications of such leakage will be i followed up as soon as practicable.

Although some leak rates on the order of GPM may be tolerable from a dose 3 point of view, especially if they are to closed systems, it must be recognized l

that leaks in the order of drops per minute through any of the walls of the 4

primary system could be indicative of materials failure such as by stress j corrosion cracking. If depressurization, isolation and/or other safety mea-

! sures are not taken promptly, these small breaks could develop into much

,. latser leaks, possibly into a gross pipe rupture. Therefore, the nature of j the leak, as well as the magnitude of the leakage must be considered in the

safety evaluation. The safety evaluation shall assure that the exposure of

offsite personnel to radiation is within the guidelines of 10 CFR 20.

The upper limit of 30 gpm is based on the contingency of a complete loss of

station power. A 30 gpm loss of water in conjunction with a complete loss of I

station power and subsequent cooldown of the reactor coolant system by the i turbine bypass system (set at 1,040 psia) and steam driven emergency feedwater

! pump would require more than 60 minutes to empty the pressurizer from the com-bined effect of system leakage and contraction. This will be ample time to O restore electrical power to the station and makeup flow to the reactor coolant system. 1 1

REFERENCES N/A l

O 16.5.10-2 03/27/99

l Subcriticallity t 16.5.11 J

16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.11 Subcriticallity I C0PetITNENT The reactor shall be maintained subcritical by an amount l greater than or equal to the calculated reactivity insertion due to depressurization.

APPLICABILITY: MODE 2 and 3 with Tavg < 525'F MODE 4, 5, and 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND l l

A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months i U-s SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.5.11.1 N/A N/A HA1Q The requirement (s) of this SLC section were relocated from CTS 3.1.3.3 during the conversion to ITS.

The potential reactivity insertion due to the moderator pressure coefficient (2i that could result from depressurizing the coolant from 2100 psia to saturation pressure of 900 psia is approximately 0.1% Ak/k.

If the specified shutdown margin is maintained, there is no possibility of an accidental criticality as a result of a decrease of coolant pressure.

) REFERENCES N/A 16.5.11-1 03/27/99

RCS LEAK TESTING FOLLOWING OPENING OF SYSTEM 16.5.12 16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.12 RCS LEAK TESTING FOLLOWING OPENING OF SYSTEM COMMITMENT Perform specified SR.

t APPLICABILITY: MODE 1, MODE 2 with Keff a: 1.0 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS

()

f SURVEILLANCE FREQUENCY SR 16.5.12.1 Perform RCS leakage test at not less than Once following 2200 psig. any opening of RCS 4

BASES The requirement (s) of this SLC section were relocated from CTS 4.3.2 during the conversion to ITS.

Repairs or modifications made to the Reactor Coolant System are inspectable and testable under applicable codes. The specific code and edition thereof shall be consistent with 10 CFR.55a.

-REFERENCES N/A O

16.5.12-1 03/27/99

HPI.and the Chemical Addition Systems 16.5.13 16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.13 High Pressure Injection (HPI) and the Chemical Addition Systems Comi!TMENT a. Two high pressure injection pumps per unit shall be OPERABLE except as specified in ITS 3.5.2.

b. The concentrated boric acid storage tank (CBAST) shall be OPERABLE.

1 l

APPLICABILITY: MODE 1, I MODE 2 with Keff a: 1.0 l

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. CBAST inoperable. A.1 Restore CBAST to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

AND BWST.0PERABLE.

B. Required Action and B.1 Be in MODE 2 with 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Keff < 1.0.

Time of Condition A not met.

l 1

1

.O i

i 16.5.13-1 03/27/99 >

l

1

] HPI and the Chemical Addition Systems 16.5.13 i

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.5.13.1 Verify boron concentration in CBAST is 7 days within the limit specified in the COLR.

BASES 4

The requirement (s) of this SLC section were relocated from CTS 3.2 and Table

, 4.1-3, Item 6 during the conversion to ITS.

! One source per unit of concentrated soluble boric acid in addition to the

borated water storage tank is available and OPERABLE.

4 This source shall be the concentrated boric acid storage tank (CBAST). The

CBAST is OPERABLE when volume and boron concentration are within the limits of

l. the Core Operating Limits Report (COLR)-with a temperature at least 10*F above the crystallization temperature. System piping and valves necessary to O establish a flow path from the tank to the high pressure injection system shall be OPERABLE and shall- have the same temperature requirement as the CBAST. At least one channel of heat tracing capable of meeting the above j temperature requirement shall be in operation. One associated boric acid pump ~

_ shall be OPERABLE.

, The high pressure injection system and chemical addition system provide-control of the reactor coolant system boron concentration. This is normally accomplished by using any of the three high pressure injection pumps in series with a boric acid pump associated with either the boric acid mix tank or the 4

, CBAST or a bleed transfer pump aligned to take suction from the CBAST. -The l boric' acid pump associated with the CBAST is normally used for small additions '

during operation and the bleed transfer pumps are utilized when larger volumes

are to be added. An alternate method-of boration will be the use of the high ;

l . pressure injection pumps taking suction directly from the borated water l storage tank (BWST). Part a to the commitment is relocated from CTS 3.2.2 for i completeness (i.e., verbatim relocation). The requirement for two HPI pumps

, except as specified in ITS 3.5.2 establishes no additional require'nents other than those specified in ITS 3.5.2. Conipliance with ITS 3.5.2 (LC0 and

, ACTIONS) for HPI pumps establishes compliance with this SLC for commitment

part a.

The quantity of boric acid in storage in the CBAST or the BWST is sufficient

to borate the reactor coolant system to a 1% Ak/k suberitical margin at 70'F with the maximum worth stuck rod and no credit for xenon at the worst time in i;

( core. life. The current cycles for each unit are analyzed with the limits presented in the COLR. The cycle specific analyses determine the volume and i ' boron concentration requirements for the BWST and CBAST necessary to borate to a cold shutdown condition (MODE 5). The volume requirements include a 10%

16.5.13-2 03/27/99

HPI and the Chemical Addition Systems 16.5.13

<O l

BAl[1 (continued) margin and, in addition, allow for a deviation of 10 EFPD in the cycle length.

The specification assures that two supplies are available whenever the reactor l 1s critical so that a single failure will not prevent boration to MODE 5. One l

of the supplies requires the operability of the CBAST with an associated pump and flow path to ensure the capability to borate the RCS to MODE 5. This requirement is not one which must be immediately available since the shortest required timeframe to reach MODE 5 is 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Thus the required boric acid from the CBAST can be added by either of the bleed transfer pumps manually aligned to take suction from the CBAST and discharging to the inlet of the makeup filters at nominal flow rates of 100 gpm. Since there is sufficient time to make the alignment, manual alignment of the bleed transfer pumps is acceptable. This flow path and the associated pumps are equivalent from safety-related and seismic criteria to that of the CBAST pump and are capable of adding the required volume from the CBAST well within the minimum 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Equivalent volumes with lower concentrations will take longer than those volumes with higher concentrations, however, both can be added within the required timeframe.

During operation, the CBAST pump is normally aligned to the CBAST and discharges to the inlet of the makeup filters. Each CBAST pump is capable of delivering the required boric acid to the RCS within the required timeframe at O a minimum flow of 7 gpm. Small volume additions from the CBAST will normally be added with the CBAST pump with the bleed transfer pumps being utilized for larger volume additions. An alternate method of addition is to inject borici acid from the BWST using the high pressure injection pumps.

The concentration of boron in the CBAST may-be higher than the concentrai. ion which would crystallize at ambient conditions. For this reason, and to assure a flow of boric acid is available when needed, these tanks and the associated piping for the flowpaths will be kept at least 10*F above the crystallization temperature for the concentration present. Once in the high pressure injection system, the concentrate is sufficiently well mixed and diluted so that normal system temperatures assure boric acid solubility.

REFERENCES

1. 'UFSAR, Sections 9.3.1, and 9.3.2.

2. UFSAR, Figure 6-1.

1 0 16.5.13-3 03/27/99

Containment Leakage Tests 16.6.1 O

Q 16.6 ENGINEERED SAFETY FEATURES 16.6.1 Containment Leakage Tests COM4ITNENT The local leak rate shall be measured for the containment penetrations listed in Table 16.6-1 in accordance with ITS SR 3.6.1.2.

1 APPLICABILIfY MODES 1, 2, 3, and 4.

l ACTIONS l

CONDITION REQUIRED ACTION COMPLETION TIME !

A. NA A.1 NA NA SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.6.1.1 NA NA BASES This commitment establishes the list of penetrations that require local leak rate testing in accordanc.e with ITS SR 3.6.1.2. This list was removed from the Technical Specifications in accordance with the guidance in NRC Generic Letter 91-08.

The requirement to leak test the blind isolatinn flanges on the containment Hydrogen Recombiner System permanent piping after each installation was relocated from CTS 4.4.3.1.b during conversion to the ITS.

REFERENCES

1. 10 CFR 50, Appendix J.

2. NRC Generic Letter 91-08.

3. UFSAR section 3.8.1.7.4, 6.2.3,and 6.2.4.

16.6.1-1 03/27/99 A

Containment Leakage Tests O 16.6.1 V

Table 16.6-1 List.of Penetrations With 10 CFR 50 Appendix J Requirements Penetration System Type A Test Local Leak Remarks Number System Test Condition 1 Pressurizer liquid sample line Note 1 Type C Notes 2, 7b (Unit 1 only) 2 OTSG A Sample line Note 1 Type C Note 7b 3 Component Cooling intet Note 1 Type C Note 3, 7d Line 4 OTSG B drain line Note 1 None required Note 7b 5a RB normal stap drain line Note 10 Type C Note 7s, 7b, 9 portion 5a hydrogen Recombiner dralss Note to Type C Note 7a, 7d, 7e portion

$b Post Accident Liquid Note 1 None Required Note 2, 7c Sample Line 6 Letdown line Note 1 Type C Note 2, 7b 7 RC Ptap seal return line Note 1 Type C (Units 2 & 3)

( Note 7b, 9 (Unit 1)

( Note 3, 7b, 9 Sa Pressurizer Aux. Spray Line Not Vented None Required Note 5, 7d 8b Loop A nozzle warming line Not Vented None Required Note 5, 7d 9 RCS normal makeup line Not Vented None Required Note 5 and HP injection "A" loop 10e RC Ptsup B1 seat injection Not Vented Type C Note 5, 7d, 9 l l

l 10b RC Ptap 82 seat injection Not vented Type C Note 5, 7d, 9 i l

11a Fuel transfer tube cover Not Verted Type B Note 64, 11 11b RC Makeup Ptsup suction Note 1 Type C Note 2 11e Fuel transfer tube drain Not Vented Type C Note 5 12e Fuel transfer ttbe cover Not Vented Type B Note 6a, 11 12b RC Makeup Puup discharg, note 1 Type C Note 2 13 RB Spray inlet line Not Vented None Required Note 5, 7d 14 R$ Spray intet line Not Vented None Required. Note 5, 7d i 15 LPI and DNR inlet line Not vented None Required Note 4, 5 16 LPI and DNR intet line Not Vented None Required Note 4, 5 17 OTSG B Emergency FDW Line yat Vented None RegJired Note 5, 7d 1B ouench tank vent line Note 1 Type C Note 3, 7b, 9

_\ 19 RB purge intet line Vented Type C Note Ta, 7b, 9 20 RB purge outlet line Vented Type C Note 7a, 7b, 9 21 LPSW to RC Ptap motors and Not Vented None Required Note 7b, 9 (the of t cooters intet l

16.6.1-2 03/27/99 i

Containment Leakage Tests p 16.6.1 i

l

(') Table 16.6-1 I

I List of Penetrations With 10 CFR 50 Appendix J Requirements Penetration System Type A Test Local Leak Remarks I

thaber System Test Condition 22 LPsW from RC Pump motors and Not Vented Type C Note 7b tthe oil cooters outlet l 23a RC Puup A1 seal injection Not Vented Type C Note 5, 7d, 9 23b RC Pump A2 seat injection Not Vented Type C Note 5, 7d, 9 l 24a RB N, Analyzer Train A Vented Type C Note 7c l 24b RB N, Analyzer Train A Vented Type C Note 7c 25 0TsG B Feedwater Line Not vented None Required Note 7d, 14 26 OTSG A Main steam line Not Vented None Required Note 5 27 OTSG A Feedwater line Not Vented None required Note 7d, 14 28 OTSG B Main steam line Not vented None required Note 5 29 Quench tank drain line Note 1 Type C Note 3, 7b, 9 30, 31, 32 LPSW for RB Cooling units Not Vented None required Note 5 inlet line

( 33, 34, 35 LPSW for RB cooling units Not Vented None required Note 5 outlet line 36, 37 RB emergency sump Not Vented None required Nota 5 recirculation line 38 Quench tank cooler inlet line Note 1 Type C Note 2, 7d 39a CFT Vent Line Note i None required Note 3 (Unit 2, 3 only) 39b NP Nitrogen supply Note 1 Type C Note 2, 3 40 RB emergency supp drain line Note 1 None required 41 Instrument air supply & ILRT Vented None required Note 3 verification line 42a RB H, Analyzer train B Vented Type C Note 7c 42b RB H, Analyzer Train B Vented Type C Note 7c 43 OTSG A drain line Note 1 None required Note 7b 44 Component cooling to control Note i Type C Note 3, 7d rod drive inlet line 45a ILRT instrument line Vented Type C Note 3, 7a 45b ILRT instrunent Line vented Type C Note 3, 7a 45c ILRT instrument Line Vented Type C Note 3, 7a (Units 2 & 3) .

46 Reactor head wash filtered Note 1 Type C Note 3, 9 water intet t 47 Domineralized water supply to Note 1 Type C Note 3, 7d l

(Unit 1 only) RC pump seat vents 48 Breathing air inlet Vented None required Note 3 16.6.1-3 03/27/99

s I

Containment Leakage Tests p

16.6.1 Table 16.6-1 List of Penetrations With 10 CFR 50 Appendix J Requirements Penetration System Type A Test Local Leak Remarks Ntaber System Test

'ondition l

49 LP Nitrogen supply Vented None required Note 3 (Unit 1 only) ,

50 OTSG A Emergency FDW tine Not Vented None required Note 5 51 ILRT Pressurization line vented None recpJired Note 6a, 7a 52 NP injection to 'B' loop Not Vented . None required Note 5 53a (All) NP Nitrogen supply to 'A' core Note 1 Type C Note 2, 3, 7d flood tank

$3b LP Nitrogen supply Vented None required Note 2, 3, 7d (units 2,3) 54 Component cooling outlet line Note 1 Type C Note 3, 7b, 9(8) 55 Domineralized water supply Note 1 Type C (Unit 1)

Note 3 (Unit 2, 3)

Note 3, 9 56 Spent fuel canst fill and Note 1 None required Note 3 drain

\ 57 DNR return line Not Vented None required Note 4 (Unit 1 only) 58e Pressurizer sample line Note 1 Type C Note 2, 7b I (Unit 2, 3) 5ab (Att) OTSG B sample line Note 1 Type C Note 7b 59 CF tank sample line Note 1 None required Note 2 60 Rs sample line (outlet) Note 1 Type C Note 2, 7b, 9, 15 61 RB sample line (inlet) Note 1 Type C Note 2, 7b, 9. 15 62 (Units DNR return line Not Vented None ra<pJired Note 4 2,3, only) 90 Personnel hatch Vented Type B Note 6b 91 Equipment hatch Vented Type B Note 6e

' 92 Emergency hatch Vented Type 8 Note 6b 101 through Electrical Penetrations vented Type 8 Note 6a 105 NOTE 1 Att vented systems shall be drained of water or other fluids to the extent necessary to assure exposure of the system contairunent isolation valves to contairunent atmos @ere and to assure they will be st&Jocted to the test differential pressure.

NOTE 2 Fluid system that is part of the reactor coolant pressure boundary or open directly to the contaltunent atmosphere under post accident conditions (vented to contsirunent atmosphere during Type A test).

NOTE 3 Closed system inside containment that penetrates containment and postulated to rupture as a result of a loss of coolant accident (vented to contairvnent atmosphere during Type A test).

N NOTE 4 System required to maintain the plant in a safe condition during the test (need not be vented).

NOTE 5 System normally fitted with water or under pressure and operating under post accident condition (need not be vented).

16.6.1-4 03/27/99

f Containment Leakage Tests

/~N 16.6.1 V)

(

Table 16.6-1 List of Penetrations With 10 CFR 50 Appendix J Requirements NOTE 6 a. Contairment penetration whose design incorporates resilient seals, gaskets, or sealant compomds, piping penetration filled with expansion bellows, and electrical penetrations fitted with flexible metal seat assenblies.

b. Air lock door seals including door opening mechanisms nSh are part of the contairveent pressure bomdary.

c. Doors with resilient seats or gaskets except for seat welded doors.

d. Components other than those above which must meet the acceptance criteria of Type B tests.

NOTE 7 e. Isolation ve;tes provide a direct connection between the inside and outside atmosphere of the primary reactor containment uMer normat operation, such as purge and ventitation, vacuta relief, and instru m t valves.

b. Isolation valves are required to close automatically upon receipt of a containment isolation signal in response to controts intended to affect containment isolation.

c. Isolation valves are required to operate intermittently under post accident conditions,

d. Check valve (s) used for containment isolation,

e. Valves are normatty closed but must be opened for hydrogen control.

NOTE 8 OELETED.

NOTE 9 Reverse c.irection test of inside containment isolation valve authorfred. Leakage results are conservative.

Os b3TE 10 System is sLhmerged during post accident conditions and performance of Type A test. System will be drained to the extent possible.

NOTE 11 Type B test performed on the blind flanges inside the Reactor Building. Yatves outside the contalrunent are not tested.

NOTE 12 DELETED NOTE 13 DELETED NOTE 14 Closed system inside containment separated from ths Reactor Coolant System and not postulated to rteture as a result of a toss of coolant eccident.

NOTE 15 The blind isolation flanges on the Containment Hydrogen Recombiner System permanent piping shall

be test tested after each instattation to en=ure adequate isolation.

r 16.6.1-5 03/27/99

O Containment Tendon Surveillance Program (O 16.6.2 16.6 ENGINEERED SAFETY FEATURES 16.6.2 Containment Tendon Surveillance Program COPMITMENT The structural integrity of the containment shall be maintained.

The Reactor Building Post-Tensioning System shall meet the minimum required values (MRVs).and Prescribed Lower Limits (PLLs) as specified in this Selected Licensee Commitment.

The required MRVs and PLLs which shall be used as limits during the conduct of the SRs are specified in Figures 16.6.2-1, 16.6.2-2 and 16.6.2-3.

APPLICABILITY MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Abnormal degradation A.1 Restore containment 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months O

V of containment structural integrity to required level of structural integrity indicated by average of all measured OR prestressing forces for any group A.2.1 Verify that 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (corrected for average containment condition) is found to structural integrity be less than the is maintained, by minimum required performing an prestress level at engineering anchorage location for evaluation of the that group containment structural integrity.

AND A.2.2 Submit Tendon 30 days Surveillance Report in accordance with ITS 5.6.7.

L) 16.6.2-1 03/27/99

Containment Tendon Surveillance Progran 1 16.6.2 l COGITION REQUIRED ACTION COMPLETION TIME 1

)

B. Abnormal degradation B.1 Restore containment 15 days i of the containment to required level of l structural integrity structural integrity other than J Condition A. QR B.2.1 Performing an 15 days engineering evaluation to verify that containment structural integrity is maintained.

AND B.2.2 Submit Tendon 30 days Surveillance Report 4 in accordance with ITS 5.6.7.

I C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion ;

Time of Required AND Action A.1, A.2.1, B.1 or B.2.1 not met. C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months :

U) 16.6.2-2 03/27/99

. . - - . ~ - - - - - . - . - - . - - - - . - . . - . - - - . - . . - - . - . - . . . . -

4 4

4

- Containment Tendon Surveillance Program 16.6.2 SURVEILLANCE REQUIREMENTS f .

j SURVEILLANCE FREQUENCY

, SR 16.6.2.1 --------------------NOTE------------------- '

! 1. This SR may be conducted during MODE 1 l provided design conditions regarding loss of adjacent tendons are satisfied at all times.

2. Frequency may be modified in accordance with ASME Section XI, Subsection IWL.

Determir.e that a random, but 5 years representative, sample of at least eleven tendons (five hoop, three vertical, three dome) each have an observed lift-off force within the predicted limits established for each tendon group. '

O SR 16.6.2.2 Verify on a tendon from each group that 5 years I tendon wires are free of corrosion, cracks, I

and damage, and minimum tensile strength of 240,000 psi (guaranteed ultimate tensile strength of the wire material) exists for at least three wire samples (one from each end and one at mid-length) cut from the removed wire.

j SR 16.6.2.3 Verify for tendons detensioned for 5 years.

inspection that SR retensions to a force at least equal the force recorded prior to detensioning or the predicted value at the time of inspection, whichever is greater, but do not exceed 70% of the guaranteed ,

ultimate tensile strength of the tendon l wire material.

O 16.6.2-3 03/27/99

\

Containment Tendon Surveillance Program G 16.6.2 O

SURVEILLANCE FREQUENCY SR 16.6.2.4 Verify acceptability of the sheathing 5 years filler grease by assuring that:

1. No free water is present and no ,

changes in the presence or physical appearance of the sheathing filler grease occur.

2. Amount of grease replaced does not exceed 5% of the net duct volume when injected at +/-10% of the specified installation pressure.

3. Minimum grease coverage exists for the i different parts of the anchorage system.

4. Reactor building exterior surface does c not exhibit grease leakage that could affect reactor building integrity.

5. Chemical properties of the sheathing filler grease are within the following tolerance limits:

Water Content 0 - 10% (dry wt.) l Chlorides 0 - 10 ppm Nitrates 0 - 10 ppm Sulfides 0 - 10 ppm Reserve > 50% of installed Alkalinity value; (Base Numbers) > 0 (for older grease)

SR 16.6.2.5 Verify no abnormal degradation exists by 5 years visual examination of tendon anchorage assembly hardware (such as bearing plates, stressing washers, wedges, and buttonheads) of all tendons selected for inspection.

.O 16.6.2-4 03/27/99

Containment Tendon Surveillance Program 16.6.2 j SURVEILLANCE FREQUENCY j i

SR 16.6.2.6 --------------------NOTE-------------------

This inspection may be performed prior to the Type A-containment leakage rate test.

........................................... l t

, The exterier surface of the reactor 5 years bLilding(s) should be visually examined to i

detect areas of large spall, severe scaling, D-cracking in an area of 25 sq.

ft. or more, other surface deterioration or disintegration, or grease leakage.

W 3

4 l O

i i

h

=

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16.6.2-5 03/27/99

,, ,-, - -- , e--~ -,.,,ww ,-v~, -r--

Containment Tendon Surveillance Progran 16.6.2 Figure 16.6.2-1 Dome Tendon PLLs and MRVs O

DOME TENDONS PRESCRIBED LOWER UMITS 660 640 PLL=660-17.0781n(yr) 620

(*/ MRV = 577 kips

'*{:

.'i 560 ,

N 520 _-

1 10 100 YEiMt3 O i 16.6.2-6 03/27/99

Containment Tendon Surveillance Program 16.6.2 Figuie 16.6.2-2 Hoop Tendon PLLs and MRVs HOOP TENDONS PRESCRIBED LOWER LIMITS 800 640 '

i s20 .PLL = 642-16.8071n(yr) 1 g_ .uRv .

.xig.

.! N I

540

- 1 1 10 100 YEARS 16.6.2-7 03/27/99

Containment Tendon Surveillance Prcgran 16.6.2 Figure 16.6.2-3 p Vertical Tenden PLLs and MRVs V

VMUCAL TMDONS PRESCRIBED LOWER LIMITS 600 se0

_ PLL = 678-14.096tn(yr) se .

MrW'= 6iS Nips .

i .

600 1

' '

I 580

. 5 4 $

540 520

!i 500 '8 -

1 to 100 Y1!iMt3 O

16.6.2-8 03/27/99 1

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Containment Tendon Surveillance Program p

V 16.6.2 BASES Some requirements in this SLC were relocated from CTS 3.6.7 and 4.4.2 during the conversion to ITS. The Selected Licensee Commitment (SLC) provides the

. details of the Containment Tendon Surveillance Program required by ITS 5.5.7.

This SLC prescribes Minimum Required Values-(MRV's) and Prescribed Lower Limits (PLL's). In a letter dated July 2,1997, Duke committed to provide a SLC which presc.ribes MRVs and PLLs in support of the Reactor Building Post-Tensioning (RBPT) System surveillances which are performed in accordance ,

with ITS SR 3.6.1.3 and ITS 5.5.7, Pre-Stressed Concrete Containment Tendon Surveillance Program. In letters dated October 30, 1996, and April 22, 1997, Duke requested a Technical Specification amendment to convert from a Reactor Building Post Tensioning (RBPT) System surveillance methodology of testing pre-designated tendons to a more industry-wide methodology as prescribed in Regulatory Guide 1.35 Revision 3. Regulatory Guide 1.35 Revision 3 requires testing of tendons which are randomly selected from the population of in-service tendons.

Acceptance criteria are given in terms of PLL's and hRV's. The required MRVs and PLLs which shall be used as limits during the conduct of the surveillances are provided in Figures 16.6.2-1, 16.6.2-2, and 16.6.2-3. These figures contain the dome, hoop, and vertical tendon MRVs and PLLs, respectively, for-all three units.

Provisions have been made for an inservice inspection program intenaed to provide sufficient evidence that the integrity of the Reactor Building is being preserved. This program will be conducted-in accordance with the l guidance of Regulatory Position C of Regulatory Guide 1.35, Inservice i Inspection of Ungrouted Tendons in Prestressed Concrete Containments. ..tvision 3 dated July 1990. Regulatory Guide 1.35 describes a basis acceptable to the

~ NRC staff for developing an appropriate inservice inspection and surveillance program for ungrouted tendons in prestressed concrete reactor buildings of l light-water-cooled reactors. The inservice inspection program will be subject to review and revision as warranted based on studies and on results obtained for this and other prestressed concrete reactor buildings throughout the life of the plant.

Prior to implementation of Regulatory Guide 1.35 methodology in accordance I with this specification, RSPT System surveillances were performed by examining

. specific, pre-designated test tendons. Therefore, this specification l conservatively identifies'the date of the last surveillance performed for each !

-anit under the superseded CTS 4.4.2, and measures the periodicity of future ,

inspections from these dates. )

Seating forces for all tendons were documented at the time of installation, thus providing one data point. A second point will be obtained from data ;

obtained during the initial tendon surveillance for each unit. The data from the initial surveillance is considered reliable since any error due to i tensioning and retensioning had not been introduced. This data will be '

O averaged on a per unit basis and used in the trend analysis along with new data obtained from the new proposed surveillance program in accordance with Regulatory Guide 1.35. l 16.6.2-9 03/27/99

Containment Tendon Surveillance Program 16.6.2 AMD.

Some requirements in this SLC were relocated from CTS 3.6.7 and 4.4.2 during

the conversion to'ITS. The Selected Licensee Commitment (SLC) provides the i- details of the Containment Tendon Surveillance Program required by ITS 5.5.7.

This SLC prescribes Minimum Required Values (MRV's) and Prescribed Lower Limits (PLL's). In a letter dated July 2,1997, Duke committed to provide a L SLC which prescribes'MRVs and PLLs in support of the Reactor Building l' Post-Tensioning (RBPT) System surveillances which are performed in accordance

' with ITS SR 3.6.1.3 and ITS 5.5.7, Pre-Stressed Concrete Containment Tendon l Surveillance Program. In. letters dated October 30, 1996, and April 22, 1997, i

Duke requested a Technical Specification amendment to convert from a Reactor i Building Post Tensioning (RBPT) System surveillance methodology of testing i pre-designated tendons to a more industry-wide methodology as prescribed in 4 Regulatory Guide 1.35 Revision 3. Regulatory Guide 1.35 Revision 3 requires testing of tendons which are randomly selected from the population of in-service tendons.

Acceptance criteria are given in terms of PLL's and MRV's. The required MRVs and PLLs which shall be used as limits during the conduct of the surveillances

! are provided in Figures 16.6.2-1, 16.6.2-2, and 16.6.2-3. These figures l- contain the dome, hoop, and vertical tendon MRVs and PLLs, respectively, 'or

all three units.

! Provisions have been ruade for an insarvice ir'oection progran ' .,tended to l

.; 4 provide .ufficient evidence that the integri./ of the Reactor N.fing is i

being preserved. This program will.be conducted in accordance u ;h m guidance of Regulatory Position'C of Regulatory Guide 1.35, Inserv e Inspection of Ungrouted Tendons in Prestressed Concrete Containments, Revision 3 dated July 1990. Regulatory Guide 1.35 describes a basis acceptable to the NRC staff for developing an appropriate inservice inspection and surveillance program for ungrouted tendons in predressed concrete reactor buildings of light-water-cooled reactors. The-inservice inspection program will be subject to review and revision as warranted based on studies and on results obtained for this and other prestressed concrete reactor buildings throughout the life of the plant.

Prior to implementation of Regulatory Guide 1.35 methodology in accordance with this specification, RBPT System surveillances were performed by examining specific, pre-designated test tendons. Therefore, this specification conservatively identifies the date of the last surveillance performed for each unit under the superseded CTS 4.4.2, and measures the periodicity of future inspections from these dates. l Seating forces for all tendons were documented at the time of installation, thus providing one data point. A second point will be obtained from data obtained during the initial tendon surveillance for each unit. The data from the initial surveillance is considered reliable since any error due to tensioning and retensioning had not been introduced. This data will be averaged on a per unit basis and used in the trend analysis along with new data obtained from the new proposed surveillance program in accordance with O Regulatory Guide 1.35.

16.6.2-9 03/27/99 l

Containment Tendon Surveillance Program 16.6.2 O(N SR 16.6.2.1 This SR determines that a random, but representative, sample of at least eleven tendons (five hoop, three vertical, three dome) each have an observed lift-off force within the predicted limits established for each tendon group.

For each subsequent inspection, one tendon from each group shall be kept unchanged to develop a history and to correlate the observed data. The procedure of inspection and the tendon acceptance criteria shall be as follows:

1. If the measured prestressing force of the selected tendon in a group lies above the prescribed lower limit, the lift-off test is considered to be a positive indication of the sample tendon's acceptability.

2. If the measured prestressing force of the selected tendon in a group ,

lies between 95% of the prescribed lower limit and 90% of the l prescribed lower limit, two tendons, one on each side of this tendon, '

shall be checked for their prestressing forces. If the prestressing !

forces of these two tendons are above 95% of the prescribed lower j limits for the tendons, all three tendons shall be restored to the i required level of integrity, and the tendon group shall be considered acceptable. If the measured prestressing forces of any two adjoining tendons fall below 95% of the prescribed lower limits of the tendons, O

V additional lift-off testing shall be done to detect the cause and J extent of such occurrence. The conditions shall be considered as an indication of abnormal degradation of the reactor building (s).

3. If the measured prestressing force of any tendon lies below 90% of the prescribed lower limit, the defective tendon shall be fully investigated and additional lift-off testing shall be done so as to cetermine the cause and extent of such occurrence. The condition shall be considered as an indication of abnormal degradation of the reactor building.

4. If the average of all measured prestressing forces for any group (corrected for average condition _) is found to be less than the minimum required prestress level at anchorage location for that group, the condition shall be considered as abnormal degradation of the reactor building.

5. If the measured prestressing forces from consecutive surveillances for the same tendon, or tendons in a group, indicate a trend of prestress loss laeger than expected and the resulting prestressing forces are likely to be less than the minimum required for the group before the next scheduled surveillance, additional lift-off testing shall be done so as to determine the cause and extent of such occurrence. The condition shall be considered as an indication of abnormal degradation of the reactor building.

O V

16.6.2-10 S/27/99

Containment Tendon Surveillance Program r's 16.6.2 lj 1

SR 16.6.2.2 l l

This SR performs tendon detensioning, inspections, and material. tests on a tendon from each group. A randomly selected tendon from each group shall be l completely detensioned in order to identify any broken or damaged wires and to i determine the following conditions over the entire length of a removed tendon !

wire sample (this wire sample should be the broken wire if so identified)- 1 I

1. Tendon wires are free of corrosion, cracks, and damage, and

2. Minimum tensile strength of 240,000 psi (guaranteed ultimate tensile strength of the wire material) exists for at least three wire samples (one from each end and one at mid-length) cut from the removed wire.

1 Failure to meet these requirements shall be considered as an indication of ~

abnormal degr:dation of the reactor building.

i SR 16.6.2.3 This SR retensions tendons detensioned for inspection to a force at least i equal the force recorded prior to detensioning or the predicted value at the time of inspection, whichever h greater, but do not exceed 70% of the guaranteed ultimate tensile strangth of the tendon wire material. Tendon

]v seating force toleranco shall b : -0 / +6%. During retensioning of these tendons, change in load versus elongation should be measured at varying levels of fcrce. The following table provides levels of force, pressure, and elongation at which measurements should be taken: 1 Force (Kips) Pressure (psi) Elongation (In)

PTF Step 1 Step 2 LOF '

OSF l Where:

Total Elongation (actual) = (LOF-PTF) Elongation PTF -Pretensioning Force necessary to bring the tendon into a slightly stressed condition to remove slack and seat the buttonheads.

Step 1 An intermediate force approximately equally spaced between PTF and LOF, LOF - Lock Off Force at which the tendon is seated on the shims.

OSF - Overstress Force at which the maximum elongation is measured. l If the elongation corresponding to a specific load differs by more than 10%

from that recorded during the original installation, an investigation should ,

be made to ensure that the difference is not related to wire failures or slip l of wires at anchorages. This condition shall be considered as an indication I

/s of abnormal degradation of the reactor building. I U

16.6.2-11 03/27/99 l

Containment Tendon Surveillance Program A 16.6.2 SR 16.6.2.4 This SR verifies acceptability of the sheathing filler grease by assuring that:

1. No free water is present and no changes in the presence or physical appearance of the sheathing filler grease occur.

2. Amount of grease replaced does not exceed 5% of the net duct volume when injected at +/-10% of the specified installation pressure.

3. Minimum grease coverage exists for the different parts of the anchorage system.

4. Reactor building exterior surface does not exhibit grease leakage that could affect reactor building 1.tegrity.

5. Chemical properties of the sb,athing filler grease are within the specified tolerance limits.

Failure to meet these requirements shall be considered an indication of potential abnormal degradation of the reactor building.

SR 16.6.2.5 As an assurante or i,ne structural integrity of the reactor building (s), tendon anchorage assembly hardware (such as bearing plates, stressing washers,

,Q b

wedges, and buttonheads) of all tendons selected for inspection shall be visually examined. Tendon anchorages selected for inspection shall be visually examined to the extent practical without dismantling the load bearing components of the anchorages. Top and bottom grease caps of all vertical tendons shall be visually inspected to detect grease leakage or grease cap deformations. The surrounding concrete should also be checked visually for indication of any abnormal condition.

Siijnificant grease leakage, grease cap deformation or abnormal concrete condition shall be considered as an indication of abnormal degradation of the reactor building.

SR 16.6.2.6 The exterior surface of the reactor building (s) snould be visually examined to detect areas of 1.m spall, severe scaling, D-cracking in an area of 25 sq.

ft. or more, otnf a rface deterioration or disintegration, or grease leakage.

Each of these t' iit ons can be considered as evidence of abnormal degradation of structural in city of the reactor building (s). This inspection may be performed prior to the Type A containment leakage rate test.

4

>V 16.6.2-12 03/27/99

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Containment Tendon Surveillance Program 16.

6.2 REFERENCES

1. Duke letter to NRC dated 10/30/97

2. Duke letter to NRC dated 4/22/97

3. Duke letter to NRC dated 7/2/97

4. Duke letter to NRC dated 9/3/97

5. Duke letter to NRC dated 9/4/97 l

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16.6.2-13 03/27/99

I Containment Heat Removal Verification Frequ:ncy 16.6.3 16.6 ENGINEERED AFETY FEATURES 16.6.3 Containment' Heat Removal Verification Frequency COMITMENT Performed required SRs.

l

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

l ACTIONS CONDITION REQUIRED A ....i COMPLETION TIME i

1 i A. N/A. A.1 N/A. N/A i

i j SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

- O SR 16.6.3.1 Verify containment heat removal capability 1s sufficient to maintain post accident As determined by LPI and RBCU l

j conditions within design limits. fouling rate 1

i BAiff

. The requirement (s) of this SLC section were relocated from CTS 4.5.3.1.b

during the conversion to ITS.

] The safety functions of the LPI system, RB Spray system, and RBCUs include maintaining containment pressure and temperature below design limits following an accident. This surveillance assures that containment heat removal capability is adequate assuming a worst case single failure. ITS SR 3.6.5.4 i requires that the containment heat removal capability be verified on an 18

month frequency. Since service induced fouling can reduce containment heat removal capability, a fouling rate must be determined in oider to establish a more frequent. test interval if required.

j REFERENCE' i

a e

16.6.3-1 03/27/99 a.

. _,_m.e

LPI Syst:m Leakaga 16.6 4 I p 16.6 ENGINEERED SAFETY FEATURES l

16.6.4 Low Pressure Injection (LPI) System Leakage C009tITMENT The maximum allowable leakage from the LPI System components (which includes valve stems, flanges and pump seals) shall not exceed two gallons per hour.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.6.4.1 Verify leakage from the portion of the LPI ------NOTE-----

System, except piping from the containment The provisions l

emergency sump to the low pressure of SLC 16.2.7 l injection pump suction isolation valve, do not apply.

that is outside the containment is within ---------------

the limit either by use in normal operation 18 months +25%

i or by hydrostatically testing at a 350 psig.

SR 16.6.4.2 Verify leakage from piping from the ------NOTE-----

containment emergency sump to the LPI pump The provisions suction isolation valve is within limit of SLC 16.2.7 when tested at 2 59 psig. do not apply.

18 months +25%

, O) 16.6.4-1 03/27/99

t u LPI System Leakage 16.6.4 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY-SR 16.6.4.3 -Verify leakage is within limit by visual ------NOTE-----

inspection for. excessive leakage from The provisions components of the system. of SLC 16.2.7 <

do not apply.

18 months +25%-

E8H1 The requirement (s) of this SLC section were relocated from CTS 4.5.5 and Technical Specification Interpretetion (TSI) 4.5.2/3.3.2 during the conversion to ITS.

Excessive leakage shall be measured by collection and weighing or by another s equivalent method. The leakage rate limit for the Low Pressure Injection System is a judgement value based on assuring that the components can be expected'to operate with-out mechanical failure for a period on the order of ,

200 days after, a loss of coolant accident. The test pressure (350 psig) {

achieved either by normal system operation or by hydrostatically testing, '

gives an adequate margin over the highest pressure within the system after a

. design basis accident. Similarly, the pressure test for the return lines from the containment to the Low Pressure Injection System (59 psig) is equivalent to the design pressure of the containment. The dose to the thyroid calculated as a result of this leckage is 0.76 rem for a two-hour exposure at the site boundary.

SURVEILLANCE TEST PRESSURE Leakage is measured during refueling outage surveillance testing. The surveillance testing may be parformed as a hydrostatic test, or measured l- during system operation. Hydrostatic testing is used for the RBES suction lines, from the sump to LP-19/20. Required testing for.the remaining portions of the system is accomplished through leakage measurement during alignment and/or operation of the LPI System, test pressure may be below 350 psig as

, . necessary to meet operating limits. The test pressure of the rest of the l LPI System.is'at or above the'specified 59 psig or 350 psig.

, Leakage rates corrections can be made, as described in the Bases, for the i actual ' pressure versus the specified test pressure. These corrections are i- small - less than 10%.

i 1

1 16.6.4-2 03/27/99 L ,

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! LPI System Leakage 16.6.4 The following practices are used to ensure that requirements are met:

a) The test pressure for the discharge side of the LPI System is ,

established as near to 350 psig as practical, within the operating i limits in plant procedures.

b) If the total measured leakage-(uncorrected) is above 1.8 gph, the measured. leakage will be evaluated for pressure differences prbr to l comparison to the 2 gph limit.

LEAKAGE OBSERVED DURING UNIT OPERATION LPI System leakage found.during operation in MODES 1, 2, 3 or 4, which would be in excess of 2 gph at the specified pressures of this SLC, requires entry into the appropriate Actions of ITS 3.5.3 or ITS LC0 3.0.3.

The affected train, based on the location of the leakage, must be declared out l of. service in accordance with the appropriate Actions of ITS 3.5.3. In addition, the point- of leakas. 'nust be isolated. As an alternative to isolating the leakage, provisions may be made to take steps, in the event of a LOCA, which would isolate the leakage or prevent it from exceeding 2 gph during emergency sump recirculation. If the LPI train remains capable of l performing its safety functions after isolating the leakage, or if the planned steps in the event of a LOCA would still ' allow all subsequent safety functions :

O- to be performed by each train, the Action of ITS 3.5.3 can be exited.

If leakaga exists in each train, and would be greater than 2 gph from each train, ITS LCO 3.0.3 must be entered. Also, if there is a point of leakage which would exceed 2 gph, and this leakage could not be isolated or reduced to less than 2 gph during emergency sump recirculation of at least one operable LPI train, ITS LCO 3.0.3 must be entered.

The 2 gph limit is based upon assuring that there is no existing leakage which could indicate that mechanical components might not continue to operate during long term core cooling. It is also based upon limiting the amount of off site dose due to leakage of highly contaminated primary coolant that is recirculated from the RB emergency sump. The 59 psig / 350 psig pressures are conservative values for which leakage is to be evaluated in the LPI suction /

discharge piping.

SURVEILLANCE TEST PRESSURE Performance of the_ leakage measurement during LPI System operation is a valid test method. This method is used for required portions of the LPI System other than the-lines between the RBES and LP-19/20. However, operational

limits, fluid dynamics characteristics, and other considerations prevent L achieving the exact 59 psig or 350 psig pressures. In the pump suction piping, the measurements are performed at > 59 psig. When an LPI pump is on, l the pressure between the pump and throttle valve (s) LP-12/14 may be higher l than the required 350 psig. The pressure downstream of the throttle valve (s) is- approximately equal to RCS pressure.

16.6.4-3 03/27/99

l LPI System Leakage 16.6.4 L

i O The fact that portions of.the discharge side of the LPI System cannot be pressurized to 350 psig does not invalidate the testing. The specified 350 ,

, psig test pressure is an arbitrary value for evaluation of leakage, rather '

! than a true hydrostatic test pressure. (The lowest design pressure in this

-portion of the system'is 470 psig. Leakage rate, not structural integrity, l

1s being evaluated.) However, the discharge side measurements are performed

! with the LPI System as near as practical to its operating limit (300 psig for !

l Unit I and 2, 290 psig for unit 3). Given any particular flow path, flow rate ,

. varies primarily with the square root of the pressure drop across the flow

)

path. -Therefore, the measured leakage rate can be adjusted for a difference l between the actual test pressure and the specified test pressure as follows:

l

! MEASURED LEAKAGE X SQUARE ROOT (SPECIFIED PRESS / ACTUAL PRESS)

Neglecting the adjustment is conservative where the actual pressure is higher l l than specified. Neglecting the adjustment is slightly non-conservative where the actual pressure is lower than the specified pressure. For example, if 1 actual test pressure were 310 psig at a point of leakage measured as 1.0 gph, ,

the normalized leakage for 350 psig would be about 1.06 gph. Because the 1 correction is small, adjustment of the measured leakages is not necessary when I uncorrected leakage is well within the 2 gph-limit (i.e., when uncorrected leakage is < l.8 gph). ;

Water temperatures are different during testing of different portions of the '

lO system, so that density corrections could be considered. Because higher-temperatures result in conservative offsite dose results due to evaporation rate, the UFSAR, assumes temperatures of 252 degrees and 115 degrees on the LPI System suction and discharge sides, respectively. However, the UFSAR !

evaluation expresses leakage in drops / minute, and converts drops per minute to cc/hr without considering temperatures. Actual leakage measurements are made j by a combination of drop counting and/or collection of ' leakage-in a container 1 (which approaches room temperature). Therefore, correction of measured :

-leakage for temperature / density effects goes beyond the precision assumed in !

the analysis, and would conflict with accepted measurement techniques. It is ;

concluded that density corrections are unnecessary. l l LEAKAGE OBSERVED-DURING UNIT OPERATION During operation in MODES 1, 2, 3 or 4, if it is determined that leakage exists such that the limit of this SLC would not be met, two actions must be taken. The affected LPI train (s) must be declared out of service, and provisions must be made to limit any leakage of recirculating RBES water, outside of containment, from exceeding 2 gph during an accident. It is not I sufficient-to simply declare an affected LPI train out of service. If both !

l- LPI trains are affected, such that it is not possible to take measures which

! would limit leakage to.< 2 gph during recirculation, while also maintaining an operable LPI train; then ITS LCO'3.0.3 must be entered.

l

. These requirements apply to the outside-of-containment leakage boundaries of 1 4

tnose portions of the LPI System which would be used during emergency sump ;

recirculation. Declaration of affected LPI train (s) as being out-of-service must be made upon determination that the limits of this SLC would be exceeded.

16.6.4-4 03/27/99 1

LPI System Leakage 16.6.4 O

Actions to isolate leakage or provide steps to limit leakage to < 2 gph, should be completed promptly as dictated by ITS 3.5.3 Actions or ITS LC0

'3.0.3.

If the LPI train (s) remain capable of performing all safety functions after isolating the leakage, or if the planned steps in' the event of a LOCA would still allow all. subsequent safety functions to be performed by each train, the affected train (s) can be declared operable after the isolation / provision of planned steps is completed.

Because recirculation would not be required for some accidents, and because an affected train might be needed prior to beginning recirculation, physically securing an affected train may not always be the best method for controlling leakage. Example-2 below illustrates this point. Situations in which more complex provisions are necessary should be addressed by a procedure.

The basis for these requirements is that-leakage above 2 gph could cause the offsite doses during an accident to exceed those which have been evaluated in the UFSAR. Because of the high concentration of fission products assumed to be present in the primary coolant after a large break LOCA significant off I site dose is associated with even small amounts of primary coolant leakage outside containment.

These practices, and the examples below, represent a conservative application O- of this SLC during operation in MODES 1, 2,'3 or 4, when ITS LC0 3.5.3 also applies. The 2 gph limit is relatively restrictive. Therefore, the

.Com>liance Section should be contacted when marginal rates of LF1 System leatage could cause a unit shutdown.

Example 1:

During power operation with the LPI pumps not running, leakage around the '

shaft of the 'A' LPI pump is observed. The leakage is determined to be about i 3 gph (from BWST head). To repair the leakage, the pump must be isolated. I The required action is to declare LPI Train ' A' out of service, and to enter l the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Required Action A.1 for ITS 3.5.3. In addition, the A' pump must !

be isolated. !

Example 2: !

During power operation with the LPI pumps not running, leakage is observed =

around the packing of LP-9. The leakage is determined to be about 1 gph. It )

is also determined that this leakage would exceed 2 gph at 350 psig, but be !

less than 2 gph at 59 psig. Repairs do not require isolating LP-9. j

1 The required action is to declare LPI Train -A' out of service, and to enter the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Required Action A.1 for ITS.3.5.3. Action must also be taken to control the leakage. Because isolating LP-9 is not necessary for repairs, it i is preferable to leave the 'A' LPI train aligned for ES actuation even though !

it is declared out of service. To' prevent leakage in excess of 2 gph in )

O. recirculation mode, verbal instructions and a turnover sheet item could be provided to turn off the 'A' LPI pump, prior to initiating the recirculation mode during an accident. j 16.6.4-5 03/27/99 j l

LPI System Leakage O 16.6.4 Example 3:

During power operation, the 3/8" line between LP-38 and LP-39 (PALS LINE ISOLATION VALVES) becomes disconnected. Leakage which would exceed 2 gph at 350 psig is observed coming from both directions (i.e., > 2 gph from 'A' and >

2 gph from 'B' The leakage cannot be controlled by closing LP-38 and LP-39.

The required action is to enter ITS LC0 3.0.3, and to attempt to reduce leakage.

Example 4: !

Leakage is observed around LP-28. '

This problem is not applicable to this SLC, because LP-28 is neither in the emergency sump recirculation flow path nor a boundary of this flow path.

References:

1)- UFSAR Sections 6.0.3, 6.0.3.1, 6.0.3.2, 6.0.3.4, and 6.0.3.5 .

1

2) UFSAR Section 15.15.4 and 6.3.3.2.2

2) PT/1/A/0203/04 LPI System Leakage PT/2/A/0203/04 LPI System Leakage PT/3/A/0203/04 LPI System Leakage

3) OP/1/A/1104/04 LPI System Operation OP/2/A/1104/04 LPI System Operation OP/3/A/1104/04 LPI System Operation

4) PIR 3-090-0019 Leakage on 3LP-9 Exceeded Tech Specs 16.6.4-6 03/27/99

I CFT Discharge Valve Breakers 16.6.5 A

\g 16.6 ENGINEERED SAFETY FEATURES 16.6.5 Core Flood Tank (CFT) Discharge Valve Breakers COMITMENT The breakers associated with the CFT discharge valves shall be locked open and tagged.

APPLICABILITY:. MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) pressure

> 800 psig. l ACTIONS !

i CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND A.2 Be in MODE 3 with RCS 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months pressure :s; 800 psig.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.6.5.1 N/A. N/A hag 1 The requirement (s) of this SLC section were relocated from CTS 3.3.3 during the conversion to ITS.

REFERENCE N/A G

V 16.6.5-1 03/27/99

Core Flooding System Test 16.6.6

( 16,6 ENGINEERED SAFETY FEATURES 16.6.6 Core Flooding System Test COP 91ITMENT Perform required SRs.

APPLICABILITY: MODES 1 and 2, MODE 3 with Reactor Coolant System (RCS) pressure

> 800 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS l 1

SURVEILLANCE FREQUENCY SR 16.6.6.1 Verify that the check and isolation valves ------NOTE-----

in the core flooding tank discharge lines The provisions operate properly during pressurization of of SLC 16.2.7 the Reactor Coolant System. do not apply, i 18 months +25%

BASES The requirement (s) of this SLC section were relocated from CTS 4.5.1.1.3 during the conversion to ITS.

l With the reactor shut down, the valves in each core flooding line are checked j for operability by reducing the Reactor Coolant System Pressure until the l indicated. level in the core flood tanks verify that the check and isolation valves have opened. The test will be considered satisfactory if control board indication of core flood tank level verifies that all valves have opened.

l p REFERENCE f

1 N/A 16.6.6-1 03/27/99

BWST Outlet Valve Control i 16.6.7 l- 16.6 ENGINEERED SAFETY FEATURES 16.6.7 Borated Water Storage Tank (BWST) Outlet Valve Control COMITMENT Manual valve LP-28 on the BWST discharge line shall be locked open.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I

A. Commitment not met. A.1 Enter applicable ITS N/A Condition for BWST inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.6.7.1 N/A N/A BASES The requirement (s) of this SLC section were relocated from CTS 3.3.4.b during the conversion to ITS.

REFERENCE N/A l

l l

[

16.6.7-1 03/27/99 u

LPI System Valve Test Restrictions 16.6.8 A

U 16.6 ENGINEERED SAFETY FEATURES 16.6.8 Low Pressure Injection (LPI) System Valve Test Restrictions COMMITMENT Perform specified SR.

' 'LICABILITY: MODE 5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY O

V SR 16.6.8.1 Test valves LP-17 and LP-18. After entry into MODE 5 from MODE 4 unless tested within the previous 92 days BASES

'The requirement (s) of this SLC section were relocated from CTS 4.5.1.2.1.a during the conversion to ITS. SR 16.6.8.1 is normally performed after entering MODE 5 but prior to exiting MODE 5 unless performed in the previous 92-days.

Power Operated Valves LP-17 and LP-18, are boundary valves between high pressure and low pressure design piping. As such, functional testing of these valves is performed during cold shutdown conditions when the Reactor Coolant SystempressureisgelowthedesignpressureoftheLowPressureInjection System' piping and the potential for over-pressurization of the low pressure

. system is eliminated.

p REFERENCE

.N/A 16.6.8-1 03/27/99

l Containment Purge Valve Testing 16.6.9

-(,

O) 16.6 ENGINEERED SAFETY FEATURES 16.6.9 Containment Purge Valve Testing COMMITMENT Perform specified SRs.

APPLICABILITY: MODES 1, 2, 3 and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY n

k,,) SR 16.6.9.1 ------------------NOTE---------------------

1. Not required to be performed until prior to going above MODE 3.

2. Not required to be performed for purge valves that have not been operated if conducted within the proceeding 184 days.

3. Perform after final closing when the purge valves have been operated.

Perform leakage integrity tests. After every entry into MODE 3 from MODE 2 i

E 16.6.9-1 03/27/99 l

l

1 Containment Purge Valve Testing 16.6.9 1

()

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 16.6.9.2 Visually-inspected and adjust or replace ------NOTE-----

the valve seals of the purge isolation The provisions valves. of SLC 16.2.7 do not apply.

18 months +25%

BASES :

The requirement (s) of this SLC section were relocated from (TS 4.4.4.1 and 4.4.4.3 during the conversion to ITS.

Leakage integrity tests of the purge supply. and isolation valves are conducted in order to identify excessive degradation of the resilient seals. Excessive leakage past resilient seals is typically caused by severe environmental conditions and/or wear due to frequent use. :

REFERENCE N/A 1

1 7

i O

16.6.9-2 03/27/99

i Containment Hydrogsn Recombiner System 16.6.10 16.6 ENGINEERED SAFETY FEATURES 16.6.10 Containment Hydrogen Recombiner System COMITMENT The Containment Hydrogen Recombiner System shall be OPERABLE.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Hydrogen Recombiner A.1 Restore to OPERABLE 7 days System inoperable. status.

B. Required ~ Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS

' SURVEILLANCE FREQUENCY SR 16.6.10.1 Verify the post-LOCA flow path by ------NOTE-----

connecting and operating the Hydrogen The provisions Recombiner through its flow path. The of SLC 16.2.7 Hydrogen Recombiner. flow path shall do not apply.

circulate Reactor Building atmosphere at a ---------------

flow > 50 SCFM.

18 months +25%

O 16.6.10-1 03/27/99

Containment Hydrcgen Recombiner System 16.6.10 f SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY l

4 SR 16.6.10.2 Visual inspection of the hydrogen ------NOTE-----

recombiner unit. The provisions of SLC 16.2.7 do not apply.

18 months +25% !

SR 16.6.10.3 CHANNEL CALIBRATION of recombiner ------NOTE-----

instrumentation channels. The provisions of SLC 16.2.7 do not apply.

18 months +25%

1 o

g SR 16.6'.10.4 Operate a recombiner unit at design flow rate t 10% and allow unit to reach recombination temperature.

NOTE-----

The provisions of SLC 16.2.7 i

i do not apply.

18 months +25%

BASES The requirement (s) of this 5'.C section were relocated from CTS 3.16.1 and 4.4.3 during the conversion to ITS.

The Containment Hydrogen Recoinbiner System shall be OPERABLE in MODES 1, 2, 3 and 4. The Recombiner System consists of one OPERABLE hydrogen recombiner unit available for connection to the associated flow path for each Oconee unit.

The Containment Hydrogen Recombiner System is required at approximately 7 days following c LOCA to limit hydrogen concentration to 4.0 percent by volume.

The Containment Hydrogen Recombiner System is utilized to maintain the post-accident containment atmosphere hydrogen concentration below its lower Q(,j flammability limit of 4.0 percent by volume. The Containment Hydrogen 16.6.10-2 03/27/99

Containment Hydrogen Recombiner System 16.6.10 O

Recombiner System includes a portable hydrogen recombiner which will be moved to the affected unit following a LOCA, anchored to its foundation, and connected to piping penetrations. Also included is a portable control panel, which will be locally mounted near the recombiner, anchored to its foundation and connected to its motor control center and the recombiner.

The control panel mounted near the recombiner enables the operator to control and monitor system parameters for all functions of the recombiner system except containment isolation valve operation. The control and monitor functions include: process temperature indications, temperature control, flow indication, start /stop switch, low temperature timer and various annunciators.

Therefore, the operational performance testing ensures operability.

The penetrations to and from the hydrogen recombiner are shared with the gaseous radiation monitoring pump. Since this pump is normally in operation i and since there is no system isolation valve on the supply branch to the recombiner, the blind flanges are the only means of system isolation. ;

Therefore, these flange joints shall be leak tested after each removal and installation to ensure adequate isolation.

The hydrogen recombiner unit operational performance test should be conducted with full flow and with the heaters energized. The capability of the recom-p d

biner to achieve the required recombination temperature and flow rate is considered an adequate test of recombination efficiency. Gas inlet and outlet sampling is not required.

REFERENCE UFSAR, Section 15.16 O

16.6.10-3 03/27/99 I

Additional HPI Requirements 16.6.12 h

a 16.6 ENGINEERED SAFETY FEATURES 16.6.12 Additional High Pressure Injection (HPI) Requirements COMMITMENT: Two HPI trains shall be OPERABLE.

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

1. Three HPI pumps and the HPI discharge crossover valves shall be OPERABLE and the suction headers shall be cross-connected.

2. The HPI discharge headers shall be hydraulically separated.

APPLICABILITY: MODES I and 2, MODE 3 with Reactor Coolant System (RCS) temperature

> 350*F.

ACTIONS CONDITION COMPLETION TIME REQUIRED ACTION A. One HPI pump A.1 Restore HPI pump to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

M AND One or more HPI A.2 Restore HPI discharge 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months discharge crossover crossover valve (s) to valve (s) inoperable. OPERABLE status.

@ AND HPI suction headers A.3 Cross-connect HPI 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months not cross-connected. suction headers.

B. HPI discharge B.1 Hydraulically separate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months headers cross- the HPI dischage connected. headers.

(continued)

O V

OCONEE UNITS 1, 2, & 3 16.6.12-1 03/27/99

Additional HPI Requirements 16.6.12 (O

ACTIONS (conti..nued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One HPI train C.1 Restore capability to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months incapable of being automatically actuate automatically train.

actuated but capable of being manually actuated.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of AND Condition A, B or C not met. D2 Reduce RCS temperature 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months to s 350*F.

E. One HPI train E.1 Enter LC0 3.0.3 Immediately incapable of being O automatically actuated and incapable of being manually actuated.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.6.12.1 Perform CHANNEL CHECK for each HPI 31 days discharge crossover valve flow instrument.

SR 16.6.12.2 Perform CHANNEL CALIBRATION for each HPI 18 months discharge crossover valve flow instrument.

SR 16.6.12.3 Cycle each HPI discharge crossover valve. 18 months OCONEE UNITS 1, 2, & 3 16.6.12-2 03/27/99

Additional HPI Requirements 16.6.12 BASES This SLC imposes operability requirements regarding the high pre:;sure injection (HPI) system in addition to those imposed by Limiting Condition for Operation (LCO) 3.5.2, "High Pressure Injection (HPI)." These additional l requirements are:

1. The third HPI pump and the HPI discharge crossover valves are required to be OPERABLE and the suction headers are required to be cross-connected when THERMAL POWER is s 60% RTP; 4

2. The HPI discharge headers are required to b hyde-"lically separated

, whenever the plant is operating in a MODE or condit,on which requires the HPI system to be OPERABLE; and

3. Surveillance Requirements have been added to confirm the OPERABILITY of the HPI discharge crossover valves and their associated flow instruments.

These additional requirements are necessary, because the requirements of the Improved Technical Specifications (ITS) were based on the previous Technical Specifications (referred to as the CTS) which were deficient. Due to the time limitations associated with the completion of the review of the ITS, the NRC and Duke agreed to resolve the deficiencies regarding these requirements in ,

another License Amendment Request. )

In the following letters, Duke informed the NRC of deficiencies with the CTS i requirements regarding the HPI System: 1) Licensee Event Report (LER) 50-269/90-15 dated December 20, 1990 (Reference 1), 2) License Amendment Request dated March 31,1997 (Reference 7), 3) Special Report dated October 22, 1990 (Reference 9), and 4) License Amendment Request dated December 16, 1998. The ,

License Amendment Request submitted on December 16, 1998 (Reference 8),

resolves the deficiencies identified in the above letters.

In addition, another deficiency was discovered regarding the lack of Surveillance Requirements to demonstrate the OPERABILITY of the flow instruments associated with the HPI discharge crossover valves prior to the implementation of the ITS.

Commitment Two HPI trains are required to be OPERABLE when in MODES 1 and 2, and MODE 3 with reactor coolant system (RCS) temperature > 350*F. This requirement is consistent with LCO 3.5.2.

f 4

OCONEE UNITS 1, 2, & 3 16.6.12-3 03/27/99

Additional HPI Requirements 16.6.12 '

O. Commitment Note 1 This Commitment Note establishes that the third HPI pump and the HPI discharge crossover valves (i.e., HP-409 and HP-410) shall be operable and the HPI suction headers shall be cross-connected whenever the plant is operating with i RCS temperature > 350*F. The requirement is consistent with the Note in ITS 3.5.2 for operation with THERMAL POWER > 60% RTP, and it replaces the interpretation of CTS 3.3.1 approved on November 26, 1990 (Reference 2),

regarding operation with THERMAL POWER < 60% RTP.

ITS 3.5.2 provides requirements for the HPI System when the RCS temperature is

> 350*F. The ITS requirements captured the CTS requirements which were based i on the analysis of a Small Break Loss of Coolant Accident (SBLOCA) which assumed a break on the discharge side of the reactor coolant pumps. The analysis concluded that one HPI train had sufficient capacity to mitigate SPLOCAs when reactor power was < 60% full power. As reported in LER 269/90-15 (Reference 1), Duke discovered that the analysis was non-conservative. It ascamed:

1) an even flow split between the injection line connected to the broken cold leg and the injection line connected to the intact cold leg. The even flow

~

split resulted from the assumption that the back pressure on each line was equal to RCS pressure; and

2) HPI flow from the injection line connected to the broken leg is injected into the reactor coolant pump discharge volume. A computer model then determined how much of the injection flow is lost out the break.

In the LER, Duke reported that one HPI train was inadequate to mitigate a break of an HPI injection line when reactor power was < 60% full power. In this case, the appropriate back pressure assumption would be containment pressure for the broken injection line, and RCS pressure for the intact injection lines. Additionally, none of the HPI flow through the broken injection line would reach the RCS. The resulting flow split from this asymmetric pressure boundary condition would cause less injection flow to reach the reactor. As a result, Duke imposed additional requirements upon the operation of Oconee Nuclear Station Units 1, 2,'and 3 with reactor power < 60%

full power (Reference 2). These additional requirements were equivalent to the requirements for operation with reactor power > 60% full power (i.e., a i third HPI pump and HPI discharge crossover valves were required to be operable, and the HPI suction headers were required to be cross-connected).

Commitment Note 2 This Commitment Note replaces an interpretation of CTS 3.3.1.a(1) approved on January 21, 1999, which limited operation with HPI discharge headers cross-connected to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (Reference 3).

4 The operability of the HPI System must be maintained to ensure that no single active failure can disable both HPI trains. Additionally, while the HPI System was not designed to cope with passive failures, the HPI trains must be

, maintained independent to the extent possible during normal operation. The OCONEE UNITS 1, 2, & 3 16.6.12-4 03/27/99

l i

, Additional HPI Requirements p 16.6.12 only NRC approved exception to this principle is cross-connecting the HPI suction headers during normal operation (Reference 4). Thus, hydraulic separation of the HPI discharge headers is required during normal operation to maintain defense-in-depth (i.e., independence of the llPI discharge headers).

i

- Actions Reauired Actions A.I. A.2, and A.3 Witi one HPI pump inoperable, one or more HPI discharge crossover valve (s) inoperable, or the HPI suction headers not cross-connected, the HPI pump and discharge crossover valve (s) must be restored to OPERABLE status and the HPI suction headers must be cross-connected within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The Completion Time is reasonable, because the HPI System continues to be capable of mitigating an accident. It is consistent with the Completion Times established in Action A of ITS 3.5.2, and the CTS interpretation dated November 26, 1990 (Reference 2).

Rauired Acticn B.1 In the event the HPI discharge headers are cross-connected, the HPI trains are not maintained independent to the extent possible; thus, Condition B of this SLC shall be entered. Required Action B.1 requires the HPI discharge headers to be hydraulically separated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . It is consistent with the O Completion Time established in the CTS interpretation dated January 21, 1999.

Reauired Action C.1 As stated in the discussion of Commitment Note 1, ITS 3.5.2 is based on the CTS requirement which utilized a non-conservative SBLOCA analysis which determined that only one HPI train was required to mitigate accidents when THERMAL POWER was < 60% RTP. As identified in References 1 and 2, both HPI trains are required to mitigate the consequences of an accident regardless of the THERMAL POWER at which the plant is operating. The Completion Time is reasonable, because the HPI system continues to be capable of mitigating an accident. It is consistent with the Completion Time established in the CTS interpretation dated November 21, 1990 (Reference 5).

In an SER dated December 13, 1978, (Reference 4) the NRC approved the use of manual operator action to place a second HPI train in operation within 10 l minutes of accident initiation. Originally, this manual action was only !

required when THERMAL POWER was > 60% RTP. An interpretation to CTS 3.3.l(c)(2) dated November 21, 1990, limits operation with one HPI train incapable of automatic actuation but capable of being manually aligned to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with THERMAL POWER > 60% full power (Reference 5). This CTS interpretation was incorporated into Condition D of ITS 3.5.2.

Action C of this SLC is consistent with Required Action D.1 of ITS 3.5.2. !

This is appropriate, because: 1) the requirements with operation with THERMAL Og POWER > 60% RTP have not changed; and 2) the interpretation of CTS 3.3.1 approved on November 26, 1990 (Reference 2) expanded the applicability of CTS 3.3.1.c(1) to operation at < 60% full power. In the event an HPI train is OCONEE UNITS 1, 2, & 3 16.6.12-5 03/27/99 j

)

l' L

Additional HPI Requirements 16.6.12 O incapable of automatic actuation but capable of manual actuation, Required Action' C.1 of this SLC requires the restoration of the ability to automatically actuate the HPI train within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Reauired Actions D.1 and DJ In the event a Required Action and associated Completion Time of Condition A, l B, or C is not met, the plant must be placed in a condition for which the SLC l does not apply. Required Actions D.1 and D.2 require the plant to be placed l in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and RCS temperature reduced to s 350*F within 60 l l hours. These Completion Times are consistent with those provided in ITS '

3.5.2, Condition E.

Reauired Action E.1 In the event one HPI train is incapable of being automatically actuated and incapable of being manually actuated, the plant is required to enter LC0 3.0.3 i i

immediately. In this condition, the HPI System is unable to perform its !

safety function. The immediate Completion Time is appropriat' given the safety significance of the condition.

l !

Surveillance Recuirements SR 16.6.12.1 and SR 16.6.12.2 i

l This SLC specifies Surveillance Requirements for the flow instruments

l. associated with the HPI discharge crossover valves. SLC SR 16.6.12.1 and SR l 16.6.12.2 require a CHANNEL CHECK and CHANNEL CALIBRATION be performed for

these flow' instruments at a Frequency of 31 days and 18 months, respectively.

If one or both of these SRs is not met, the associt.ted HPI discharge crossover valve (i.e., HP-409 or HP-410) must be declared inoperable, because the Bases for Technical Specification 3.5.2 requires that the associated flow instrument be OPERABLE to support the valve's OPERABILITY.

i SR 16.6.12.3 Periodic stroke testing of the HPI discharge crossover valves (i.e., HP-409 and HP-410) is required to ensure tha;. the valves can be manually cycled from j the Control Room. This test is performed on an 18-month Frequency.

I References

1. Licensee Event Report 269/90-15, dated December 20, 1990.

2. Interpretation of CTS 3.3.1 approved on November 26, 1990.

! 3. Interpretation of CTS 3.3.1.a(1) approved on January 21, 1999.

! 4. Letter from R. W. Reid (NRC) to W. O. Parker (Duke), NRC Safety Evaluation Report on the Oconee modification adding HP-409 & HP-410, l dated December 13, 1978.

l OCONEE UNITS 1, 2, & 3 16.6.12-6 03/27/99 l

Additional HPI Requirements l i 16.6.12

5. Interpretation of CTS 3.3.1 approved on November 21, 1990. l

6. UFSAR Sections 5.4.7.2, 6.3.1, 6.3.2.2.1, and 9.3.2, and Chapter 15.

7. Letter from J. W. Hampton (Duke) to U. S. NRC, License Amendment Request regarding CTS requirements for the HPI System, dated March 31, 1997.

8. Letter from W. R. McCollum (Duke) to U. S. NRC, License Amendment Request regarding CTS requirements for the HPI System, dated December 16, 1998.

9. Letter from H. B. Barron (Duke) to U. S. NRC, "Special Report Concerning High Pressure Injection Train Rendered Inoperable Due To Inappropriate Operator Actions," dated October 22, 1990.

10. ITS 3.3.8 and 3.5.2.

l O

i O

OCONEE UNITS 1, 2, & 3 16.6.12-7 03/27/99

_ _ _ _ _ _ . _ _. _._ . _ _. _ _. _ .. _ _ ._ _ _ _ . _ _ _ _ _ ._. _ _ _ _ _ . . _ .m _ _

Accident Monitoring Instrumentation - Noble Gas Effluent Monitor (RIA-56) 16.7.1

' 16.7 INSTRUMENTATION 16.7.1 Accident Monitoring Instrumentation - Noble Gas Effluent Monitor (RIA-56)

~COMITMENT The noble gas effluent monitor shall be OPERABLE.

APPLICABILITY:' MODES 1 AND 2 ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME A. Noble Gas Effluent A.1 Institute alternative 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

, Monitor (RIA-56) noble gas monitorina inoperable. program.

M.

SURVEILLANCE REQUIREMENTS i SURVEILLANCE- FREQUENCY SR 16.7.1.1 Perform CHANNEL FUNCTIONAL TEST. 31 days 5

SR 16.7.1.2 Perform CHANNEL CALIBRATION. 18 months.

EMfd The Noble Gas Effluent Monitor (RIA-56) is utilized for detection of significant releases and release assessment.

The Alternative methods for monitoring noble gas effluent during inoperability of RIA-56 shall include one or more of the following methods:

- RIA-45 normal range noble gas monitor on the unit vent.

- RIA-46 high range noble gas monitor on the unit vent.

- Actual vent sample.

- Direct radiation readings on RIA-45 and RIA-46 sample line.

(

16.7.1-1 03/27/99

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

.a

, Accident Monitoring Instrumentation - Noble Gas Effluent Monitor (RIA-56) 16.7.1 jlMfl (continued) 1 -

REFERENCES:

l

l. Generic Letter 83-37

2. Regulatory Guide 1.97, Rev. 2 i

1 1

i 4

)

?

.[

e 16.7.1-2 03/27/99

Anticipated Transients Without Scram p 16.7.2

'O 1 16.7 INSTRUMENTATION I 16.7.2 Anticipated Transients Without Scram 1

.COM41TMENT The ATWS Mitigation Systems Actuation Circuitry (AMSAC) and Diverse Scram System (DSS) shall be OPERABLE.

APPLICABILITY: MODE 1, MODE 2 when Keff 2: 1.0 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME !

A. One or both channels A.1 Restore AMSAC to 7 days I of AMSAC inoperable. OPERABLE status. !

B. One or both channels B.1 Restore DSS to 7 days of DSS inoperable. OPERABLE status.

C. Required Action and -------------NOTE------------

associated Completion When initiated the Required ?

Time not met. Action must be completed.

C.1 Submit a written 30 days report to the NRC outlining the cause of the channel (s) or system (s) malfunction and the plans for restoring the channel (s) or system (s) to OPERABLE l status.

t

.U i

16.7.2-1 03/27/99 l

Anticipated Transients Without Scram

,q 16.7.2 V

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.7.2.1 Perform a Channel Logic Test of AMSAC. 184 days SR 16.7.2.2 Perform a Channel Logic Test of DSS. 184 days SR 16.7.2.3 Perform an Actuation Test of AMSAC. 18 months SR 16.7.2.4 Perform an Actuation Test of DSS. 18 months.

/7 BASES U

The AMSAC and DSS are provided to mitigate the consequences of anticipated transient without scram. These anticipated transients are beyond the design basis for the plant. These events are associated with a failure 6f the reactor to normally trip when required as defined in the references below.

The AMSAC/ DSS consists of two channels and uses a two-out-of-two coincidence logic to actuate. Each channel has an AMSAC portion and a DSS portion.

The AMSAC circuitry of each channel receives input signals on low Feedwater pump Turbine (FDWPT) control oil pressure or low Feedwater pump (FDWP) discharge pressure. Upon a valid input signal to the AMSAC portions of the two AMSAC/ DSS channels, an output is generated to trip the Main Turbine and start all operable Emergency Feedwater Pumps.

The DSS circuitry of each channel receives an input signal from the Inadequate Core Cooling Monitoring System RCS pressure signals. Upon a valid signal (RCS Pressure Very High / a: 2450 psig) to both DSS portions of each channel an output is' generated to interrupt power to the Control Rod Drive System gate drives for regulating rod groups 5 through 7 and the auxiliary gate drives.

An AMSAC/ DSS channel is considered operable if it has niet the surveillance criteria of this commitment and the AMSAC/ DSS enabled light (located in the control room) is on and the AMSAC/ DSS Ch. I and Ch. 2 bypassed lights (also

, q located in the control room) are not on and "Sy Max" Programmable Controllers 4

y; RUN Lights (ON) and HALT Lights (0FF) for AMSAC/ DSS Ch. 1 AND AMSAC/ DSS Ch. 2.

16.7.2-2 03/27/99

Anticipated Transients Without Scram 16.7.2 An Actuation Test consists of a complete test from input sensors through output actuation relays. ;

REFERENCES:

1. Code of Federal Regulations, Section 10 CFR 50.62 "The ATWS Rule".

2. B&WOG Generic ATWS Design Basis Document 47-1159091-00, October 9, 1985.

3. NRC Safety Evaluation Report on 47-1159091, June 30, 1988.

4. AMSAC and DSS Final Design Description, August 30, 1988.

5. NRC Safety Evaluation Report for Final Design of Oconee ATWS Modification (TACS 59119/59120/59121), November 29, 1989. '

l J

O O l 16.7.2-3 03/27/99

1 Emergency Feedwater - Low Level Initiation 16.7.3 16.7 INSTRUMENTATION 3

16.7.3 Emergency Feedwater - Low Level Initiation 1

COMMITMENT Automatic low level initiation of both MDEFW pumps shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3 ACTIONS

) -

i CONDITION REQUIRED ACTIO.1 COMPLETION TIME

)

A. One or both channels A.1 Restore to OPERABLE 7 days j inoperable, status.

i Il B. Required Action and -------------NOTE------------

i associated Completion When initiated the Required I

i Time not met. Action must be completed.

3 B.1 Submit a written 30 days report to the NRC -

outlining the cause of the channel (s) or system (s) malfunction !

3 and the plans for j restoring the channel (s) or

system (s) to OPERABLE status.

O 16.7.3-1 03/27/99

l Emergency Feedwater - Low Level Initiation

( 16.7.3

-SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY i

l SR 16.7.3.1 Perform a CHANNEL CALIBRATION. 18 months um The Steam Generator Level Control System (SGLCS) receives four OTSG level signals. Each train receives one signal from each OTSG (OTSG A & B Level to Train A and the same to Train B). These level signals are used to start the l MDEFWPs upon 2 out of 2 low level in either OTSG. A level signal indicating l below the initiation setpoint or failed low is considered to be operable.

The most limiting transient for the EFW system is the Loss of Main Feedwater .

(LMFW), (Ref. UFSAR Section 10.4.7). The primary success path to mitigate the LMFW includes initiation of the EFW system. The UFSAR evaluation credits automatic initiation of EFW on loss of both main feedwater pumps as sensed by low hydraulic oil pressure. In addition, for plant conditions in which automatic initiation circuitry must be disabled (i.e., turbine header pressure A < 850 psig) adequate time is available for manual initiation of EFW. Thus, 4

. initiation of EFW on low OTSG level is not credited for any DBA or transient.

EFW initiation on low OTSG 1evel has been included as a SLC in response to GL 89-19 and USI A-47 and provides additional prei.ection from OTSG dryout.

EFW initiation on low OTSG 1evel is applicable =bove 250*F, although it is not required for operability of the EFW System.

In order to provide additional protection from OTSG dryout, RCS temperature may not be increased above 250*F with low level initiation of MDEFW '

inoperable. However, if the Unit is above 250*F, shutdown is not required since low level initiation is not credited for any DBA or transient.

REFERENCES:

1. Generic Letter 89-19, Safety Implication of Control Systems O .

16.7.3-2 03/27/99

DELETED G 16.7.4 V

16.7 INSTRUMENTATION 16.7.4 -----DELETED-----

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16.7.4-1 03/27/99

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Steam Gentrator Overfill Prstection 16.7.5 (D

(,) 16.7 INSTRUMENTATION l

1 16.7.5 Steam Generator Overfill Protection C0fetITMENT The steam generator overfill protection system shall be OPERABLE.

APPLICABILITY: MODES I and 2, MODE 3 when RCS T,,, > 325'F i

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME !

l A. Steam generator A.1 Restore to OPERABLE 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months overfill protection . status.

system inoperable.

/N B. Required Action and -------------NOTE------------

U associated Completion Time not met.

When initiated the Required Action must be completed.

B.1 Submit a written 30 days report to the NRC outlining the cause of the channel (s) or system (s) malfunction and the plans for restoring the channel (s) or system (s) to OPERABLE status.

16.7.5-1 03/27/99

. . -. ... ~ . . ..= - - . - - _ _ . - _ . - - - - - . - ~ . .

j Steam Generator Overfill Protection 16.7.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ,

.SR 16.7.5.1 Perform Trip Test (SV6). 92 days

}

SR 16.7.5.2 Perform Trip Test (SV12). 18 months

I j SR 16.7.5.3 Perform CHANNEL FUNCTIONAL TEST. 18 months 1

SR 16.7.5.4 Perform CHANNEL CALIBRATION. 18 months.

BACKGROUND i

! This commitment supports closure of Generic Letter 89-19 by providing limiting i conditions for operation, actions, and surveillances for steam generator overfill protection.

l APPLICABLE SAFETY ANALYSES 1

j Steam generator overfill protection (e.g.. the high steam generator level f

feedwater pump trip) plays an important role in the mitigation of Main

Feedwater (MFW) overfill events that could lead to Pressurized Thermal Shock

(PTS). Studies have been performed by Duke, the B&W Owners Group, and Oak Ridge National Laboratory (ORNL) to assess the probability of vessel failure

i due to PTS events. The high level trip is credited in many of these studies

, to mitigate overfill transients; thus, the PTS results are highly dependent on the functioning of the high level trip. However, the PTS sequences which lead

- to core melt contribute less than 1% to the overall calculated core melt frequency. ;

APPLICABILITY The overfill protection system is required to be operable for RCS temperatures above 325'F to assure that an overcooling event due to steam generator 16.7.5-2 03/27/99 m w - . - -

- - - - + -

l Steam Generator Overfill Protection l; -

16.7.5 l

l overfill will not lead to pressurized thermal shock of the reactor vessel.

l .For RCS temperatures s; 325' F, The Low Temperature Overpressure Protection (LTOP) system provides protection against overpressure concerns.

, COMITMENT '

l l Steam generator overfill protection is provided through the ICS to terminate main feedwater when the high level setpoint is reached. Two transmitters per l steam generator monitor steam generator water level. Protection is provided I

l by 2 out of 2 logic on either steam generator which actuater two trip devices.

The high level monitoring circuits deenergize to trip: thus a deenergized module is operable. Two trip devices (SV6 and SV12) are provided on each MFWPT. For example, 2 out of 2 logic on the "A" steam generator will actuate both trip devices-on both MFWPTs. Since either steam generator can cause an overcooling event, then the overfill protection logic for both steam generators are required to be operable for the overfill protection system to be considered operable.

ACTIONS The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months completion time in Required Action A.1 provides an adequate level of availability of the overfill protection system for performing its function

, while allowing reasonable time to permit necessary maintenance on the system.

SURVEILLANCE REQUIREMENTS SR 16.7.5.1 This surveillance verifies that the SV6 trip device will trip the associated liFWPT. SV6 is exercised by the " oil trip" test. When the oil trip is exercised, SV6 is pnergized thus tripping the overspeed governor which trips the mechanical trip ~ mechanism of the MFWPT. This surveillance can be performed on line and is part of the secondary system protection test. The 92 day frequency for this Surveillance was determined to be adequate based on operating experience.

.SR 16.7.5.2 1 This surveillance verifies that the SV12 trip device will trip the associated MFWPT. This Surveillance can only be performed when the MFWPT is out of service. The 18 month frequency for.this Surveillance was determined to be adequate based on-operating experience. '

SR 16.7.5.3 This surveillance requires a CHANNEL FUNCTIONAL TEST which verifies a trip signal is provided in response to high steam generator level. The 18 month

l. frequency for this Surveillance was determined to be ;deqtate based on operatkg experience.

v

(

l 16.7.5-3 03/27/99 l - . . _ -

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

Steam Generator Overfill Protection 16.7.5 SR 16.7.5.4 l

l This surveillance requires a CHANNEL CALIBRATION which verifies the channel l responds to steam generator level with the necessary range and accuracy. This surveillance is also required by ITS SR 3.3.8-1 for Item 12. The 18 month frequency for this surveillance was determined to be adequate based on operating experience, l

REFERENCES:

i 1. Generic Letter 89-19, Request for Action Related to Resolution of Unresolved Safety Issue A-47 " Safety Implication of Control Systems in LWR Nuclear Power Plants." i

2. H.'B._ Tucker (Duke) to NRC Document Control Desk, Response to GL 89-19, l March 19, 1990 l

O 16.7.5-4 03/27/99

_so J < es A-.ma_ a ....aan - i ._.a4 __-Jm a.a.g .gs-- 4 _.h_a :m.s4_t-.. .A.A aua_-a--.w4 44 _aAAma s.J A%.h.--w. Ae,.A---A -se4.4_asa me_. m_d DELETED 16.7.6 J 16.7 INSTRUMENTATION l 16.7.6 -----

DELETED-----

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f 16.7.6-1 03/27/99

i Position Indicator Channel Testing 16.7.7 16.7 ' INSTRUMENTATION f' 16.7.7 Position Indicator Channel Testing

! 1 l

COMITMENT Perform specified SR.

l APPLICABILITY: . MODES I and 2.

-l l

l ~ ACTIONS 1

l CONDITION REQUIRED ACTION COMPLETION' TIME Il A. N/A. A.1 N/A. N/A l

SURVEILLANCE REQUIREMENTS .

O SURVEILLANCE FREQUENCY l

SR 16.7.7.1 Perform CHANNEL CALIBRATION of the Absolute ------NOTE-----

and Relative Position Indication Channels The provisions for each CONTROL R00. of SLC 16.2.7 ;

do not apply.

18 months +25%

E83.El

-The requirement (s) of this SLC section were relocated from CTS Table 4.1-1, Item 23 and 24 during the conversion to ITS.

Calibration of the CONTROL R00 Absolute and Relative position indication channels supports OPERABILITY of the CONTROL R00 position indication channels required by ITS LC0 3.1.7 i

Lo l

l 16.7.7 1 03/27/99 l

Position Indicator Channel Testing 16.

7.7 REFERENCES

1. ITS B 3.1.7

2. UFSAR, Section 7.6 l

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O 16.7.7-2 03/27/99

___.___.___.~.____ _ _ _ ___ _ _ __.__ _ _ _..___ _ _ _._ _._

l Incore Instrumentation 16.7.8 16.7 INSTRUMENTATION 16.7.8 Incore Instrumentation l COMITMENT Incore detectors shall be OPERABLE as follows:

a. At least three detectors in each of at least three strings shall. lie in the same axial plane, with one plane -in each axial core half. The axial planes in each core half shall be symmetrical abcut the core mid-plane.

The detector strings shall not have radial symmetry.

b. At'least two sets of at least four detectors shall lie in each axial core half. Each set of detectors shall lie in the same axial plane. The two sets in the same core half may lie in the same axial plane. Detectors in the same plane shall have quarter core radial symmetry.

APPLICABILITY: MODE 1 with ALLOWABLE THERMAL POWER a 80% RTP ACTIONS CONDITION- REQUIRED ACTION COMPLETION TIME A. Incore detectors A.1 Discontinue using Immediately inoperable. incore detectors to determine axial imbalance or quadrant power tilt.

AND A.2 Reduce ALLOWABLE 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months THERMAL POWER < 80% 1 RTP. i I

"O 16.7.8-1 03/27/99 l

1

Incore Instrumentation 16.7.8 V(h SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.7.8.1 Check functioning of incore detectors, 31 days including a check of the computer readout or recorder readout.

BASES The requirement (s) of this SLC section were relocated from CTS 3.5.4 and Table 4.1-1, Item 34 during.the conversion to ITS.

The OPERABILITY of the incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor Core.

The safety of reactor operation at or below 80 percent of the power allowable for the reactor coolant pump combination (Ref.1) without the axial imbalance

('~s s trip system has been determined by extensive 3-D calculations, and was verified during the physics startup testing program.

REFERENCES ,

1. UFSAR, Section 5.1.2.3 f

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1 16.7.8-2 03/27/99

4 1

RCP Nonitor 16.7.9 16.7 INSiRUMENTATION 16.7.9 RCP Nonitor 1

COMMITMENT Four RCP monitor channels shall be OPERABLE.

[ APPLICABILITY: H0 DES I and 2 ACTIONS i

CONDITION REQUIRED ACTION COMPLETION TIME i

1 A. One channel A.1 Place channel in Immediately inoperable. trip.

() SURVEILLANCE REQUIREMENTS .

SURVEILLANCE FREQUENCY "T

SR 16.7.9.1- N/A. N/A AAg}

The requirement (s) of this SLC section were relocated from CTS Table 3.5.1-1, 1

Note h during the conversion to ITS.

The RCP monitors provide inputs to the RCP pump to power trip channel logic.

l For operability to be met either all RCP monitor channels must be operable or 3 operable with the remaining channel in the tripped state.

. REFERENCES 1

UFSAR, Section 7.2 O

16.7.9-1 03/27/99

i CFT Instrumentation 16.7.10

( 16.7 INSTRLMENTATION 16.7.10 Co. flood Tank (CFT) Instrumentat#on COMMITMENT One level instrument channel and one pressure instrument channel shall be OPERABLE for each CFT.

APPLICABILITY: MODES I and 2, MODE 3 with Reactor Coolant System (RCS) pressure

> 800 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND A.2 BE in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months O

V SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.7.10.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 16.7.10.2 Perform CHANNEL CALIBRATION. ------NOTE-----

The provisions of SLC 16.2.7 do not apply.

18 months +25%

t

\

16.7.10-1 03/27/99

l CFT Instrumentation 16.7.10 i

MS.E.S The requirement (s) of this SLC section were relocated from CTS 3.3.3 and Table 1 4.1-1, item 25 during the conversion to ITS.

l REFERENCES l N/A I

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16.7.10-2 03/27/99

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Display Instrumentation 16.7.11-16.7 INSTRUMENTATION 16.7.11 Display Instrumentation COMITMENT , Perform specified Surveillance Requirements for each Function in Table 16.7.11-1.

APPLICABILITY: According to Table 16.7.11-1. I ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A l

l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.7.11.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

SR 16.7.11.2 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 16.7.11.3 Perform CHANNEL CHECK. 31 days

, SR 16.7.11.4 Perform battery check. 31 days 4

SR 16.7.11.5 Perform functional test. 31 days

O 4

(continued) 16.7.11-1 03/27/99

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

d;)

$- Display Instrumentation r

16.7.11 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

. . ~;

12 months SR 16.7.11.6 Perform CHANNEL CALIBRATION.

NOTE-----

SR 16.7.11.7 Perform CHANNEL CALIBRATION. The provisions of SLC 16.2.7 do not apply.

18 months +25%

SA.Sf}.

The requirement (s) of this SLC section were relocated from TCTS Table 4.1-1, Items 22, 27, 31, 32, 33, 35, 36, 38, 40, and 50 during the conversion to ITS.

REFERENCES N/A a

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l 16.7.11-2 03/27/99

J Display Instrumentation 16.7.11 Table 16.7.11-1 (page 1 of 1) ,

01 splay Instrumentation Function APPLICABLE MODES SURVEILLANCE OR OTHER SPECIFIED REQUIREMENTS CONDITIONS

1. Pressurizer temperature 1, 2. 3 SR 16.7.11.1 I SR 16.7.11.7

2. Letdown storage tank level 1. 2. 3 SR 16.7.11.2 I SR 16.7.11.7

3. Boric Acid Mix Tank Level 1. 2. 3 SR 16.7.11.6

4. Boric Acid Mix Tank Temperature 1. 2. 3 SR 16.7.11.3 SR 16.7.11.6

5. CBAST Level 1. 2, 3 SR 16.7.11.6 l 1

6. CBAST Temperature 1. 2. 3 SR 16.7.11.3 SR 16.7.11.6.

7. Containment Temperature 1. 2. 3 SR 16.7.11.7

8. Emergency Plant Radiation Instruments At all times SR 16.7.11.4 SR 16.7.11.7

9. Environmental Monitors At all times SR 16.7.11.5 SR 16.7.11.7

10. Reector Building Emergency Sump Level 1. 2. 3 SR 16.7.11.7

11. Turbine Overspeed Trip 1. 2. 3 SR 16.7.11.7

12. PORV Position 1. 2, 3 SR 16.7.11.3 SR 16.7.11.7

13. Primary System Safety Relief Valve Position 1. 2. 3 SR 16.7.11.3 SR 16.7.11.7 5

l 16.7.11-3 03/27/99 I l

SSF DG Air Start System Pressure Instrumentation q 16.7.12 N.)

16.7 INSTRUMENTATION 16.7.12 SSF Diesel Generator (DG) Air Start System Pressure Instrumentation COMilTMENT One SSF DG Air Start System Pressure instrument channel shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

ACTIONS

r-CONDITION REQUIRED ACTION COMPLETION TIME A. One instrument channel A.1 Restore instrument 7 days inoperable. channel to OPERABLE status.

O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.7.1. 1 Perform CHANNEL CHECK. 7 days SR 16.7.12.2 Perform CHANNEL CALIBRATION. 12 months BASES The requirement (s) of this SLC section were relocated from CTS Table 3.18-1, Item 6 and Table 4.20-1, Item 9 during the conversion to ITS.

m The surveillance requirements for the SSF Instrumentation are based on I,") experience in operation of both conventional and nuclear systems. The minimum checking frequency stated is deemed adequate for SSF Instrumentation.

Calibration is performed to assure the presentation and acquisition of 16.7.12-1 03/27/99

SSF DG Air Start System Pressure Instrumentation 16.7.12

O 1 accurate information. Process system instrumentation errors induced by drift j can be expected to remain within acceptable tolerances if recalibration is 1

perfonned at the intervals specified.

I REFERENCES UFSAR Section 9.6.1.

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16.7.12-2 03/27/99 I

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SSF Instrumentation !

4 16.7.13

() 16.7 INSTRUMENTATION 16.7.13 SSF Instrumentation COMMITMENT Perform specified Surveillance Requirements for each ;

Function in Table 16.7.13-1. '

APPLICABILITY: MODES 1, 2 and 3.

1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A 4

a 1

g SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.7.13.1 -------------------NOTE--------------------

This Surveillance shall be performed when i the associated pump is operated during IST.

Perform CHANNEL CHECK. 92 days SR 16.7.13.2 Perform CHANNEL CALIBRATION. 12 months SR 16.7.13.3 Perform CHANNEL CALIBRATION. ------NOTE-----

The provisions of SLC 16.2.7 do not apply.

18 months +25%

O V

16.7.13-1 03/27/99

SSF Instrumentation 16.7.13 am The requirement (s) of this SLC section were relocated from Technical Specification Table 4.20-1, Items 2, 5, 7 and 8 during the conversion to Improved Technical Specifications.

The surveillance requirements for the SSF Instrumentation are based on experience in operation of both conventional and nuclear systems. The minimum checking frequency stated is deemed adequate for SSF Instrumentation.

Calibration is performed to assure the presentation and acquisition of accurate information. Process system instrumentation errors induced by drift can be expected to remain within acceptable tolerances if recalibration is performed at the intervals specified.

REFERENCES N/A O

16.7.13-2 03/27/99

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SSF Inst ~rNntation m 16.7.13 Table 16.7.13-1 (page 1 of 1)

SSF Instrumentation Function SURVEILLANCE ,

RE0VIREMENTS

1. SSF RC Makeup Pap Suction Pressure SR 16.7.13.1 SR 16.7.13.3

2. SSF RC Makeup Fe p Discharge Pressure SR 16.7.13.1 SR 16.7.13.3

3. SSF RC Makeup Pump Suction Temperature SR 16.7.13.1 SR 16.7.13.3

4. SSF RC Makeup Pump Discharge Flow SR 16.7.13.!

SR 16.7.13.3 S. SSF Auxilimry Service Water Pump Suction Pressure SR 16.7.13.1 SR 16.7.13.2

6. SSF Auxiliary Servier Water Pump Discharge Pressure SR 16.7.13.1 SR 16.7.13.2

7. SSF Auxiliary Service Water Pump Unit 1 Discharge Pressure SR 16.7.13.2

8. SSF Auxiliary Service Waler Pump Unit 2 Discharge Pressure SR 16.7.13.2 T

9. SSF Auxiliary Service Water Pump Unit 3 Discharge Pressure SR 16.7.13.2

10. SSF Auxiliary See vice Water Pump Discharge Test Flow SR 16.7.13.1 SR 16.7.13.2

11. $$F Auxiliary Service Water Pump Suction Temperature SR 16.7.13.1 SR 16.7.13.2

12. Underground Fuel Oil Storage Tank Inventory SR 16.7.13.2

13. D/G Service Water Pump Discharge Flow SR 16.7.13.1 SR 16.7.13.2

14. D/6 Service Water Pump Discharge Pressure SR 16.7.13.1 SR 16.7.13.2 4

a 16.7.13-3 03/27/99

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Control 'of Room'Terperature for Station Blackout

' 16.8.1 r 16.8 ELECTRIC POWER SYSTEM 16.8.1 Control of Room' Temperature for Station Blackout i

COMITMENT Control Room, Cable Room and Electrical Equipment Room i temperatures shall be within limits.

APPLICA8ILITY: At all times -

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. .One or more room A.1 Initiate action to Immediately temperatures exceeding restore temperature limit. to within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.8.1.1 Verify Control Room temperature is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months nominally s 85'F.

SR 16.8.1.2 Verify Cable Room temperature is nominally 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s 85'F.

SR 16.8.1.3 Verify Electrical Equipment Room 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months temperature is nominally s 90*F.

O 16.8.1 1 03/27/99

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Control of Room Temperature for Station Blackout 16.8.1 mu BACKGROUND This commitment _ provides guidance on the need to ensure that the initial room temperatures assumed in the room heat up calculations for the Oconee's response to the Station Blackout Rule,10 CFR 50.63, are maintained. The requirements of the Station Blackout Rule are that the control rooms not exceed 120*F during the entire coping duration of a Station Blackout. The requirements for the other rooms are that they not exceed temperatures for which the equipment contained in them would not reliably operate. It has been shown by room heat up calculations that Oconee can meet the temperature requirements assuming initial starting temperatures are not greater than specified above which are greater than normal operating temperature for these rooms.

If these room temperatures are not maintained at the normal operating value below those specified, then the ability to cope with a Station Blackout for the entire 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months duration would be affected. The room temperatures could exceed the assumed limits prior to the end of the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . When normal room ,

cooling cannot maintain normal room temperatures, then immediate action should be taken to supplement room cooling or provide a backup cooling method. This will reduce the vulnerability to exceeding the upper temperature limits assumed in the event of a Station Blackout.

APPLICABLE SAFETY ANALYSES Maintaining normal room temperatures assures that in the event of a Station Blackout, Operators can remain in the control rooms for the duration of the required 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months coping time period. Additionally, equipment located in the cable rooms and electrical equipment rooms will continue to function throughout the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. This commitment has been identified to the NRC ,

in a letter dated July 1, 1992, " Revised Response to 10 CFR 50.63, Requirements for Station Blackout."

REFERENCES:

1. 6/12/92 letter from the NRC to Duke Power Company, Oconee Nuclear Station Units 1, 2, and 3, ' Summary of June 4, 1992, Meeting on Station Blackout Response for the Oconee Nuclear Station."

2. 7/01/92 letter from J. W. Hampton to the NRC, " Revised Response to 10 CFR 50.63 Requirements for Station Blackout."

3. 10 CFR 50.63, Loss of All Alternating Current Power.

4. OSC 4747, Attachment 7, Rev. 3.

5. PIP 0-094-0668.

O 16.8.1-2 03/27/99

ea- ,. .., - - s z' n .w. a > .'m

. s.a+a m. 6 --a., aa.---..a a s,,u. .m m.'. .ams-A.-_ s Deleted 16.8.2

( 16.8 ELECTRIC POWER SYSTEM 16.8.2 -----DELETED-----

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O 16.8.2-1 03/27/99 1

Power Battery Parameters I 16.8.3 l 16.8 ELECTRIC POWER SYSTEM 16.8.3 Power Battery Parameters

C000ilTMENT Power Battery parameters shall be within specified limits. I APPLICABILITY: MODES 1, 2 and 3.

l 3

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Electrolyte level A.1 Declare associated Immediately below top of cell battery inoperable, plates.

M I

Battery cell float i voltage < 2.06 volts. ,

O =

Electrolyte Temperature < 60*F.

E l No battery chargers are available to a battery.

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O 16.8.3-1 03/27/99

Power Battery Parameters I r 16.8.3 CONDITION REQUIRED ACTION COMPLETION TIME B. A single battery B.1 Declare associated Immediately inoperable. distribution center (IDP, 2DP, 3DP) inoperable.

AND l

l B.2 --------NOTE---------

! Not required when l associated buses (PA I or PB) are cross-tied l for all ONS units.

l l

Declare Turbine Immediately Driven Emergency Feedwater (TDEFW) l System and l- Anticipated

! , Transients Without Scram (ATWS) System l

inoperable.

l AND l

B.3 Initiate action to Immediately cross-tie the associated buses (PA

! or PB) for all ONS Units.

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l 16.8.3-2 03/27/99 l .__ _

-Power Battery Parameters 16.8.3 O CONDITION REQUIRED ACTION COMPLETION TIME C. Two or more batteries C.1- Declare associated Immediately inoperable. distribution center (lDP,2DP,3DP) inoperable.

AND C.2 Declare-Turbine- Immediately Driven Emergency Feedwater(TDEFW)

System and Anticipated Transients Without Scram (ATWS) System inoperable.

I l

D. One battery charger D.1 Initiate action to Immediately l inoperable. . connect the standby I O charger to the associated bus, i

i E. Electrolyte level < E.1 Restore electrolyte 90 days j minimum or > maximum level to within !

level indication limits.

marks. ;

F. Battery cell float F.1 Restore cell float 90 days voltage <.2.13 Volts voltage to within

.and ;a: 2.06 Volts. limits.

G.. Required Action and G.1 Declare associated Immediately associated Completion battery inoperable.

Time not met.

!O 1

16.8.3-3 03/27/99

Power Battery Parameters 16.8.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.8.3.1 Verify pilot cell float voltage a 2.13 VDC. 7 days SR 16.8.3.2 Verify pilot cell electrolyte level > 7 days minimum and < maximum level indication marks.

SR 16.8.3.3 Verify each cell float voltage a 2.13 VDC. 92 days 1

SR 16.8.3.4 Verify each cell electrolyte level > 92 days minimum and < maxiinum level indication fs)

%J marks.

l SR 16.8.3.5 Verify temperature of every sixth connected 92 days cell > 60*F.

16.8.3-4 03/27/99

, Power Battery Parameters 16.8.3 O =

BACKGROUND

, This SLC on the 250VDC Power Battery Cell Parameters utilizes the l'imits on electrolyte level, float voltage, and temperature for the 250VDC Power Batteries to determine operability of the batteries. Float voltage is the voltage that is required-to be continuously applied to the battery which is sufficient to maintain a constant state of charge. The limits for the designated pilot cell's float voltage, electrolyte level, and temperature is ,

characteristic of a charged cell with adequate capacity. The limits for each connected cell's float voltage, electrolyte level and temperature ensures the OPERABILITY'and capability of the battery.

In addition, the SLC provides the required actions for restoring the system to an OPERABLE status should a battery or charger become inoperable. ,

' APPLICABLE SAFETY ANALYSIS The 250VDC Power Batteries provide DC power for the Turbine Driven Emergency Feedwater (TDEFW) System. The OPERABILITY of the 250VDC Power Batteries is required to ensure the operability and the capability of the TDEFW system.

The TDEFW system is required to be operable in accordance with ITS 3.7.5 In addition, the Anticipated Trip Without Scram (ATWS) system is supported by the power batteries. Selected Licensee Commitment 16.7.2 provides the operability ,

requirements of the ATWS system. In order to maintain the required 250VDC O Power Batteries OPERABLE, Battery Cell Parameters must be maintained within specific limits.

APPLICABILITY ,

The Power Battery Cell Parameters are required to be within limits when the associated'DC sources are required to be OPERABLE.

ACTIONS The pilot cells are monitored closely as a measure of battery performance.

Because pilot cells lose more electrolyte than the other cells, the designation of the pilot cell should be rotated among all cells in the battery. The Completion Times are based on engineering judgment considering operating experience, and the time required to complete the Required Actions.

M.

If the electrolyte. level is below the. top of the cell plates, the entire battery is conservatively assumed to be inoperable, because the cell's discharge capacity would be reduced, and the plates may suffer permanent

, damage. The battery may be restored to OPERABLE status by restoring the -

electrolyte level in accordance with the Required Actions of the SLC.

. If the float voltage of a battery cell is < 2.06 volts, the battery is assumed

to be inoperable, because. battery voltage may not be adequate to carry 16.8.3-5 03/27/99

Pcwer Battery Parameters 16.8.3 I required loads. The battery may be restored to OPERABLE status by restoring the float voltage to a 2.06 volts in accordance with the Required Actions of the SLC.

1 If the electrolyte temperature of a connected cell is < 60*F, the associated ,

battery must be declared inoperable and the Required Actions taken as i appropriate. With temperature < 60'F, the battery's capability may not be I sufficient to meet the design basis load demand.

If no battery charger is available to a battery, then the associated battery !

shall be declared inoperable. The associated DC buses on all ONS units can be cross-tied to ensure operability of the system. l B.1. B.2. and B.3 i If a single battery is inoperable, then the associated DC buses (PA or PB) on ;

all ONS units can be cross-tied to ensure operability of the system. The )

TDEFW system and ATWS are considered operable in this configuration. If the DC '

buses are not cross-tied then the associated distribution center (1DP, 2DP, !

3DP) is inoperable. The TDEFW system and ATWS on the associated unit are NOT '

considered operable in.this configuration.

i C.1 and C.2 If two or more batteries are inoperable, then the associated distribution centers (IDP. 2DP, 3DP) are inoperable. The TDEFW system and ATWS on the O associated units are NOT considered operable in this configuration. Inadequate battery capacity is available to operate the PA or PB buses cross-tied with two PA or two PB batteries unavailable. In addition, excessive fault current (greater than protective device ratings) is available with a PA and PB battery unavailable and both PA and PB buses cross-tied.

M If a battery charger is inoperable, then the Standby Charger can be connected to the associated DC bus to ensure operability of the system. The TDEFW system and ATWS are considered operable in this configuration.

E.1 The limits on electrolyte level ensures no physical damage to the plates occurs and adequate electron transfer capability is maintained.

F.1 A float voltage limit of greater than or equal to 2.13 volts will ensure the j cell remains fully charged with adequate capacity.

l t

16.8.3-6 03/27/99

l Power Battery Parameters A 16.8.3 b

sa If the appropriate parameters cannot be restored in accordance with the Required Actions, the associated battery is assumed to be inoperable.

SURVEILLANCE RE0VIREMENTS SR 16.8.3.1 This Surveillance is consistent with the recommendations of Reference 1. The reference indicates that the battery be demonstrated to meet limits on a regularly scheduled interval.

i SR 16.8.3.2 This Surveillance is consistent with the recommendations of Reference 1. An adequate electrolyte level ensures that there will be a proper conductivity and capacity of the bat'.ery cell.

SR 16.8.3.3 l

This Surveillance is consistent with the recommendations of Reference 1. A minimum voltage is established to ensure adequate voltage to maintain cells in a constant state of charge.

i SR 16.8.3.4 i

This Surveillance is consistent with the recommendations of Reference 1 and l the battery manufacturers. An adequate electrolyte level ensures that there will be a proper conductivity path and capacity of the battery cell.

SR 16.8.3.5 t

This Surveillance is consistent with the recommendations of Reference 1. The electrolyte must be maintained above a minimum temperature for the battery to p deliver designed power.

REFERENCES:

1. A IEEE Standard 450-1975, Recommended Practice for Maintenance, Testing, i and Replacement of Large Lead Storage Bdteries for Generating Stations i and Substations.

2. ITS 3.7.5, Emergency Feedwater System.

3. Selected Licensee Commitment 16.7.2, Anticipated Transient Without Scram.

O l

16.8.3-7 03/27/99

i Keowee Operational Restrictions 16.8.4

.16.8 ELECTRIC POWER SYSTEM 16.8.4 Keowee Operational Restrictions COMITMENT Keowee station output, and the combination of.Keowee Lake Level and-Operating Tailrace level shall be within the limits specified in the applicable Figures 16.8.4-1, 16.8.4-2, 16.8.4-3 and 16.8.4-4.

l APPLICABILITY: Modes 1, 2, 3 and 4, during periods of commercial power generation by one or both Keowee Hydro Units (KHUs).

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Keowee station output A.1 Enter applicable ITS Imediately not within limit. Condition (s) and Required Actions for inoperable i

~

QB KHU(s),

O Combination of Keowee lake Level and operating tailrace AND A.2 Initiate action to Immediately level not within' restore within limits.

limits.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.8.4.1 Verify combination of Keowee station During

output, Keowee lake level and operating commercial tailrace level is within limits for the generation acceptable operating area specified in the applicable Figures 16.8.4-1, 16.8.4-2, 16.8.4-3 and 16.8.4-4.

l

!O I

16.8.4-1 03/27/99 l-

?

Keowee Operational Restrictions 16.8.4 l Figure 16.8.4-1 i

Keowee Operational Restrictions x -

Unit 1 ONLY 84 Mw Keowee Operating Chart Accepude Operanno Area era gy,, \ '

872 871 670 ese ._

tot g E es7 _ , ;;_ , ;u

) 888 SSS 664.25

.. s g \

- gg; . \

gSO l'96 35 .

659 058 957 066 _

799 85 065 788 700 787 704 789 790 791 792 793 794 795 796 797 790 799 000 Keowee Lake Level (ft)

!,/

y/

16.8.4-2 03/27/99

i I.

l Keow:e Operational Restrictions L 16.8.4 Figure 16.8.4-2 g Keowee Operational Restrictions 0 1 Unit 2 ONLY 84 Mw ;

Keowee Operating Chart hpatde Operating Area

\g 672.5 m \

s .

,,, \ I ass _ _ _ . _ . . _

, )

!- i 3

164.7 \ '

U 3"-

N ~ l

\ A ggg &$t20 "-

eso -

gge t.Metectsble Operdog ArjL_ ,_

l 791.14 008

, 796.sh i

ses 785 795 787 700 789 790 791 792 793 764 795 798 )ct 798 790 000 Keewee Labe level lRl G

16.8.4-3 03/27/99

Keowee Operational Restrictions 16.8.4 Figure 16.0.4 3 Keowee Operational Restrictions

-(O) 2 Units 74 Mw Keowee Operating Chart Acceplatdo Opera 41ng Aree 073 -. -.

\

672 5 672 - . . . . . . - . .. -. .

.7, __

X, k

670 _=

060 tog 668.1 E te? ..-

ope. , ~ . . s e

h A 805 l es4 s e64 7M.74 eos _ _ _

062 . -

881 _. . --.-- . ZD9.72 see - .. .

659 . -- - -- - - .

454 -. . . ..L 657 . _ _ . .

666 466 l 788 708 787 788 789 790 791 792 793 794 795 796 797 794 790 000 Keowee Late Level (ft) j l

j l

f%

G :

16.8.4 4 03/27/99

Keowee Operational Restrictions 16.8.4 Figure 16.8.4-4 Keowee Operational Restrictions

[v_,)

1

. l 2 Units 64 Mw l

. Keowee Operating Chart AC'*ptable operaung Area 673 \

672.5 x 672 % l 671 \

670 t 669 666

"' " N E 667 u, m.su % ,.:,,5 a,..

866 \

O\

T 665 . . ..

N 864 . w a. A 663 662 795. Mi 661 660 656 739 58 656 -

657 656 655 _

785 786 787 788 789 790 791 792 793 794 795 796 797 798 799 800 Keowee Lake Level (ft) l O

!d 16.8.4-5 03/27/99 M

1 3

Keowee Operational Restrictions 16.8.4 4

BACKGROUND

( Portions of this SLC are ralocated from CTS 3.7.1 TS Note 3.

During periods of commercial power generation, the operability of the Keowee Hydro units shall be based on -lake levels and the power level of the Keowee i Hydro units. The Keowee Hydro operating restrictions for commercial power j generation shall be contained in the ONS Selected Licensee Commitment manual.

4 This SLC is used to determine Keowee Hydro unit operability as an Oconee i j

Esargency Power source when Keowee is generating to the commercial grid. It I

specifies the range of acceptable Keowee lake and tailrace elevations for various Keowee power generation levels. The acceptable region of the

[ operating restrictions was determined by reference 1.

4 Figure 16.8.4-1 specifies the maximum operating limits of the Keowee Hydro units. This is applicable only for single unit operation of Keowee unit 1. i a

This refers to occasions when Keowee unit 1 is operating' and Keowee unit 2 is not operating. This figure allows for operation of Keowee Hydro unit I at a i maximum of 84MW. Also, any operation of Keowee Hydro unit 1 below 84MW is

allowed in accordance with this figure.

a Figure 16.8.4-2 is applicable only for single unit operation of Keowee unit 2.

i- This refers to occasions when Keowee unit 2 is operating and Keowee unit 1 is

, not operating. This figure allows for operation of Keowee Hydro unit 2 at a

-maximum of 84 MW. Also, any operation of Keowee Hydro unit 2 below 84 MW is l allowed in accordance with this figure.

Figures 16.8.4-3 and 16.8.4-4 apply to simultaneous commercial generation with

both Keowee units. In addition, the figures apply to single unit operation of l Keowee unit 1 or 2. In Figure 16.8.4-3, commercial generation is allowed up to

, a maximum of 74MW. Figure 16.8.4-4 contains the operating restrictions for

commercial generation up to a maximum of 64MW. The lake levels on the i operating charts are operating lake levels. Therefore, verification that the j operation of the Keowee Hydro units is within the acceptable region of the

charts will have to be performed during operation of the Keowee Hydro units.

APPLICABLE SAFETY ANALYSIS i' The Keowee Hydro units provide emergency power for Oconee Nuclear Station on j the aopropriate emergency power path. The operability of the Keowee Hydro

units is required to ensure the operab;11ty and the capability of Je Emergency Power System. Nuclear Station Modification (NSM) ON-52966 installed

. frequency protection and revised the runaway governor protection logic circuits which ensure the operability of the Keowee Hydro units during periods

) of commercial generation. This SLC will ensure that the Keowee Hydro units are l operated within the acceptable limits.

LO 4

16.8.4-6 03/27/99

- --a., . ,n - . .- y..., ..n., ,,,- - , n--.-, . . , , . - - ,

Keowee Operational Restrictions p 16.8.4 U

APPLICABILITY During periods of cosuercial power generation, the Keowee Hydro units are required to be within the acceptable regions of the operating restrictions

'when one or more Oconee Nuclear units are in MODES 1, 2, 3 or 4.

ACTIONS l

The operability of the Keowee Hydro units during periods of commercial !

generation is ensured when the Keowee Hydro units operate within the acceptable region of Figures 16.8.4-1, 16.8.4-2, 16.8.4-3 and 16.8.4-4.

If the Keowee Hydro units are determined to be outside the limits of the acceptable region, action will be taken to restore commercial generation of

, the Keowee Hydro units to within the limits of the acceptable region. In addition, the applicable ITS Condition shall be entered since the Keowee Hydro Unit may not be able to perform its design function. Once the commercial operation of the Keowee Hydro unit (s) is restored to within the limits of the acceptable region, the ITS Condition shall be exited. It is not necessary to perfom an operability test of Keowee Hydro units prior to exiting the Condition as long as no maintenance is performed on the units in order to return them to an acceptable operating region.

SURVEILLANCE REQUIREMENTS l

SR 16.8.4.1 This surveillance will ensure that the operating conditions are within the limits of the acceptable region of the operating restrictions in Figure 16.8.4-1, 16.8.4-2, 16.8.4-3 and -16.8.4-4 during commercial generation by the Keowee Hydro units. Since the lake levels in Figures 16.8.4-1, i 16.8.4-2, 16.8.4-3 and 16.8.4-4 are operating lake levels, verification that the operation of the Keowee Hydro units is within the acceptable regions will be performed during operation of the Keowee Hydro units.

REFERENCES:

1. Calculation KC-UNIT 1-2-0106

2. 04/19/95 letter from J. W. Hampton to the NRC, " Response to NRC Questions 4

on the Proposed Emergency Power Modification Action Plan."

3. 03/15/95 letter from J. W. Hampton to the NRC, " Proposed Emergency Power Modification Action Plan."

2 4 08/15/95 letter from the NRC to J. W. Hampton, " Issuance of Amendmencs"

^

5. 03/20/97 Safety Evaluation Report from the NRC to add OPERABILITY A requirements and surveillances to the Technical Specifications.

Q 4

. 16.8.4-7 03/27/99

125Vdc Vital I&C Syst:m Ground Locating Policy 16.8.5

() 16.8 ELECTRIC POWER SYSTEM 16.8.5 125 Vdc Vital I&C System Ground Locating Policy COMMITMENT Grounds on the 125 Vdc Vital Instrumentation and Control System will be pursued in accordance with this ground locating policy.

APPLICABILITY: At all times i

) ACTIONS i

CONDITION REQUIRED ACTION COMPLETION TIME f A. A continuous ground A.1 Determine the ground -----NOTE------

4 alarm is received. magnitude. If the buses cannot be separated due to extenuating circumstances, then the ground O magnitude will be determined within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months once the buses are separated.

8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months B. One or more ground B.1 Measure ground voltage ------NOTE-----

alarms are present. and bus voltage. A 50% interval extension QB applies to the Completion One ground alarm is Times.

inoperable. --------------

Once within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months AND 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months O thereafter 16.8.5-1 03/27/99

125Vdc Vital I&C System Ground Locating Policy 16.8.5 CONDITION REQUIRED ACTION COMPLETION TIME C. Ground resistance C.1 Initiate efforts to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from I

< 2.8V to ground locate the ground. receipt of (s 500 Ohms). continuous ground alarm E

C.2 Perform engineering 7 days from evaluation of safety receipt of l system vulnerability to continuous the ground using the ground alarm available data.

1 S \

C.3 Request PORC approval of 7 days from l

the evaluation. receipt of continuous ground alarm O

D. Ground resistance D.1 Initiate efforts to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from a: 2.8V and < 6V locate the ground. receipt of

(> 500 Ohms and continuous s 2,000 Ohms). ground alarm AND D.2 If ground. is not located, 14 days from perform engineering receipt of evaluation of safety continuous system vulnerability to ground alarm the ground using the available data.

AND D.3 Request PORC approval of 14 days from i.he evaluation. receipt of continuous ground alarm O

16.8.5-2 03/27/99

l. e 125Vdc Vital I&C System Ground Locating Policy '

! 16.8.5

! CONDITION REQUIRED ACTION COMPLETION TIME i

E. Ground resistance E.1 Initiate efforts to 128 hours0.00148 days 0.0356 hours 2.116402e-4 weeks 4.8704e-5 months from a: 6V and s 18V locate the ground. receipt of

(> 2,000 Ohms and continuous s 10,000 Ohms). ground alarm AND E.2 If ground is not located, 728 hours0.00843 days 0.202 hours 0.0012 weeks 2.77004e-4 months from perform engineering receipt of evaluation of safety continuous system vulnerability to ground alarm the ground using the available data. l AND E.3 Request PORC approval of 728 hours0.00843 days 0.202 hours 0.0012 weeks 2.77004e-4 months from the evaluation. receipt of continuous ground alarm SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.8.5.1 NA NA BASES BACKGROUND ine DC ground locating process was identified as a weakness in NRC inspection report 50-269,270,287/93-26. During December 1993, a pressure switch failed which resulted in a ground on the DC system and the inoperability of the 2A Motor Driven Emergency Feedwater Pump. This inoperability exceeded the allowed outage time in the CTS and resulted in NRC Violation 50-270/94-08-02.

The response to this violation indicated that guidelines would be developed for locating a DC ground, evaluating the significance of the ground, and O- removing the ground. The 9,'idelines have been devrioped and are contained in this SLC.

16.8.5-3 03/27/99

125Vdc Vital I&C System Ground Locating Policy 16.8.5 The primary concern with grounds is that the interaction of two or more grounds may cause the malfunction of equipment. The actions contained in this St.C are based on the recognition that lower resistance grounds have a greater probability to induce equipment malfunctions. In addition, the risk of ground-induced malfunctions is decreased as the ground resistance increases.

This ' decrease in risk is associated with the lower probability of affecting individual relays and the reduced number of relays which are vulnerable.

APPLICABILITY At all times, DC grounds on the 125Vdc Vital Instrumentation and Control System will be located in.accordance with this ground locating policy.

ACTIONS M

When a continuous valid ground alarm is recei'.>ed, Operations should evaluate the possible source of the ground. Information may be provided in the Alarm Response Procedure to aid in this evaluation. As part of the evaluation, a work request will be generated. SP0C/I&E Maintenance will take bus-to-ground Voltage measurements before the panelboards are separated. Following the completion of the voltage measurements, the panelboards will be separated.

O However, separation of the panelboards may not be allowed if Operations determines that extenuating circumstances exist. If the buses cannot be separated due to extenuating circumstances, then the ground magnitude will ba determined within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months once the buses are separated. The ground magnitude will be determined by SP0C/I&E Maintenance with the panelboards separated.

M Whenever valid ground alarms are present or a ground alarm is inoperable, ground voltage measurements will be made every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . Any changes in the positive-to-ground or negative-to-ground voltages will be evaluated to ensure that there is no additional system degradation. In addition to the ground voltage measurements, the bus voltage measurements will be made to assure proper charger operation, ful If the ground magnitude is determined to be s 500 Ohms, then locating efforts will begin within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from receipt of the ground alarm. If determination of ground magnitude is delayed due to extenuating circumstances as described above, groand locating efforts will begin within 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months after the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period for determination of the ground magnitude. The ground will be located within 7 days after the receipt of a continuous ground alarm, or an evaluation of the safety system vulnerability using the available ground data will be performed. This 7 day action statement is based on 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to initiate the

, locating efforts and 6 days to locate the ground or perform an evaluation. The Plant Operations Review Committee (PORC) will be contacted to approve the evaluation.

16.8.5-4 03/27/99

l 125Vdc Vital I&C System Ground Locating Policy l 16.8.5 D.d If the ground magnitude is determined te be > 500 Ohms end s 2,000 Ohms, then fewer relays are vulnerable. Locating efforts will begin within 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months .

This.is based on s' total time period of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from receipt of the ground alarm until locating efforts begin (40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months plus the initial 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ). If determination of ground magnitude is delayed due to extenuating circumstances as described above, ground locating efforts will begin within 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months after the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period for determination of the ground magnitude. The ground will be located within 14 days after the buses are separated, or an evaluation of the safety system vulnerability using the available ground data will be performed. This 14 day action statement is based on 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to initiate the locatir.g efforts and 12 days to locate the ground or perform an evaluation.

The Plant Operations Review Committee (PORC) will be contacted to approve the evaluation.

Ed If the ground magnitude is. determined to be > 2,000 Ohms and s 10,000 Ohms, then locating efforts will begin within 128 hours0.00148 days 0.0356 hours 2.116402e-4 weeks 4.8704e-5 months . If determination of ground magnitude is delayed due to extenuating circumstances as described above, i' ground locating efforts will begin within 5 days after the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period for determination of the ground magnitude. The ground will be located within 728 hours0.00843 days 0.202 hours 0.0012 weeks 2.77004e-4 months after receipt _of a continuous ground alarm, or an evaluation of the safety system vulnerability using the available ground data will be performed. 1 This 728 hour0.00843 days 0.202 hours 0.0012 weeks 2.77004e-4 months action statement is based on 5 days and 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to initiate the locating efforts and 25 days to locate the ground or perform an evaluation.

The Plant Operations Review Committee (PORC) will be contacted to approve the evaluation.

REFERENCES:

1. LER 270/94-01 dated March 10, 1994, " Technical Specification Limit Exceeded Due to Equipment Failure"

2. NRC Inspection report 50-269,270,287/93-26

3. NRC Inspection report 50-269,270,287/94-08

4. 5/11/94 letter from J. W. Hampton to NRC Documeat Control Desk, " Reply to Notice of Violation"

5. 6/23/94 letter from J. W. Hampton to NRC Document Control Desk, " Reply to Notice of Violation" 6, 2/9/95 memo from L. S. Underwood to C. A. Little, "DC Ground Locating Policy" O

16.8.5-5 03/27/99

l i

i Lee / Central Alternate Power System '

16.8.6

'16.8 ELECTRIC POWER SYSTEM 16.8.6 Lee / Central Alternate Power System COMITNENT Two Lee. Combustion Turbines (LCTs) shall be available for supplying power to the Oconee Standby. Buses through a i i separated 100 kV power path within one hour of a loss of i

both On-Site Emergency Power Paths. Requirements for energizing the Oconee Standby buses are found in the ITS.

l APPLICABILITY: MODES 1, 2, 3, and 4 ACTIONS C0NDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 --------NOTE--------

LCTs not available for Lee / Central Power System l

supplying power to the is considered Standby Buses within unavailable as input to one hour of demand. Maintenance Rule Risk Assessment and Unavailability j Monitoring. l Log unavailaisility ;a NA the Operations Log. ,

I l

O 16.8.6-1 03/27/99 l

i i

Lee / Central Alternate Power System l 16.8.6 4

l CONDITION REQUIRED ACTION COMPLETION TIME B. CT-5 is not available B.1 --------NOTE--------

within one hour of Lee / Central Power System demand. is considered unavailable as input to E Maintenance Rule Risk Assessment and 100 kV power path Unavailability from Lee is not Monitoring.

available within one --------------------

hour of demand.

l Log unavailability in NA M

the Operations Log. '

Both SL breakers are not available within one hour of demand. I C. Power from a LCT is C.1 --------NOTE----------

lost while supplying LCT is considered to power to the Standby have had a run failure Buses, as input into Maintenance Rule Failure E Monitoring.

Failure of a required LCT to start within Log unavailability in NA one hour of demand. Operations Log.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.8.6.1 Verify with Lee Steam Station the 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months availability status of required LCTs for supplying power to the standby buses.

O 16.8.6-2 03/27/99

_ _ _ . - _ _ _ . _ ___ . _. _ .. . . ~ . _ . - . __ --

Lee / Central Alternate Power System 16.8.6 mu BACKGROUND The three Lee Combustion Turbines 40, SC, and 6C; the separated 100 kv line through CT-5; and the SL Breakers serve as an alternate electrical power source for Oconee Nuclear Station when its on-site emergency power sources are unavailable. Only one Lee Combustion Turbine and one SL Breaker is required to supply hot shutdown loads for two Oconee units plus LOCA loads for one Oconee t. nit. The availability of two Lee Combustion Turbities allows for redundancy.

ACTIONS When less than two Lee Combustion Turbines or its separated pawer path are available within one hour of the loss of both on-site emergency power paths, the Oconee units are more susceptible to a station blackout event (SB0).

Adherence to Maintenance Rule Risk Assessment guidelines reduces the probability of a blackout event and increases the availability of SB0 mitigation equipment. Unavailability of this equipment is logged in the Operations Log. Requirements for energizing the Oconee Standby busses are found in the ITS.

SURVEILLANCE REQUIREMENTS The surveillance requires daily communication between Lee and Oconee, keeping Oconee personnel informed of the availability of the Lee Combustion Turbines for supplying power to the Oconee Standby Buses.

REFERENCES

1. Oconee Nuclear Station ITS 3.8

2. Work Process Manual Section 607, " Maintenance Rule Assessment of Equipment Removed From Service."

3. OSC-5771 "PRA Risk Significant SSC's for the Maintenance Rule."

4. 0S5-0254.00-00-2011 100KV Alternate Power System Design Basis Document.

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O 16.8.6-3 03/27/99 1

E Auctiensering Diodes 16.8.7 ,

16.8 ELECTRIC POWER SYSTEMS 16.8.7 Auctioneering Diodes l 1

COMITMENT Perform specified SR. l l \

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l

A. N/A A.I.1 N/A. N/A ,

l l

l- SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

i SR 16.8.7.1 Verify peak inverse voltage capability of 184 days each I&C auctioneering diode is within limits.

BASES The requirement (s) of this SLC section were relocated from CTS SR 3.7.8.2 during the conversion to ITS.

Each panelboard receives its DC power through an auctioneering network of two isolating diode assemblies. One assembly is supplied from the unit's 125 volt distribution system,-and the other assembly is supplied from another unit's (the backup unit) 125VDC Vital Distribution System. The diode assemblies permit the two distribution systems to supply current to the Vital I&C DC Panelboard connected to the output of the diode assemblies, and block the flow ~

of current from one DC distribution system to the other. Measuring peak inverse voltage capability of each abetioneering diode ensures the diodes are capable of isolating a fault on one source from the other source. The 184 day frequency is based on engineering judgement and operating experience.

REFERENCES i N/A 16.8.7-1 03/27/99

External Grid Trouble Protection System l 16.8.8 (Oj 16.8 ELECTRIC POWER SYSTEMS l 16.8.8 External Grid Trouble Protection System COMITMENT Perform specified SR.

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.I.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

{T v/

R 16.8.8.1 Verify the External Grid Trouble Protection System logic provides an isolated power 92 days i path between Keowee and Oconee.

l BASES The requirement (s) of this SLC + tion were relocated from CTS 4.6.5 during the rewrite of CTS 3.7 (Amendment Nos. 232,232,231).

I REFERENCES N/A 1

i V

16.8.8-1 03/27/99

Fire Suppression Water Supply Systems 16.9.1 Os 16.9 AUXILIARY SYSTEMS

.16.9.1 Fire Suppression Water Supply Systems i

COMITMENT The Fire Suppression Water Supply Systems shall be OPERABLE as follows:

l Oconee

a. High Pressure Service Water (HPSW) pumps A and B with automatic initiation logic, and associated piping and valves supplying water to the sprinkler system and fire hose stations.

b. The HPSW pumps shall be aligned to the high pressure fire header.

Keowee

c. The Fire Protection Pump, automatic initiation logic. the i associated piping and valves supplying water to the Main Transformer water spray system and hose stations listed in

\

l SLC 16.9.4 with the exception of the Mechanical Equipment Gallery stations

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

The Oconee High Pressure Service Water (HPSW) system is used both in support of the Oconee Low Pressure Service Water (LPSW) system and for Fire Suppression. The most restrictive requirements for the HPSW system are derived from the support function for.LPSW.

(See SLC 16.9.8 - HPSW requirement to support LPSW.)

APPLICABILITY: At all times O

16.9.1-1 03/27/99

i l

l Fire Suppression Water Supply Systems .

16.9.1 ACTIONS i CONDITION l REQUIRED ACTION COMPLETION TIME A. Equipment inoperable A.1 Restore inoperable 7 days in the Oconee Fire equipment to OPERABLE Suppression Water status.

! Supply System.

l 0_R l l

A.2 Develop guidance 7 days outlining plans and procedures to be used to -

compensate for the loss of redundancy in this system.

B. Equipment inoperable B.1 Restore inoperable 7 days in the Keowee Fire equipment to OPERABLE Suppression Water status.

l O

i Supply System.

08 B.2 Develop guidance 7 days outlining plans and procedures to be used to compensate for the loss of redundancy in this system.

C. No Oconee Fire C.1 Establish backup Oconee 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Suppression Water Fire Suppression Water Supply System Supply System.

OPERABLE.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met for AND Condition C.

D.2 Be in MODE 5. 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months O

16.9.1-2 03/27/99

~ .- ._.

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l

,. Fire Suppression Water Supply Systems l 16.9.1 CONDITION REQUIRED ACTION COMPLETION TIME l

l L

E. No Keowee Fire E.1 Establish backup Keowee 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Suppression Water Fire Suppression Water Supply System' Supply System.

OPERABLE.

10 E.2 Develop plan to restore 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Keowee primary Fire Suppression Water Supply l System. -

F. Required Action and F.1 Energize both Standby 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated Completion Buses from Lee Time not met for Combustion Turbine on .

Condition E. the dedicated line.

AND F.2 Restore Fire Suppression 14 days Water Supply System or alternate capability.

G. Required Action and G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time not met for AND

. Condition F.

G.2 Be in MODE 5. 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months O

16.9.1-3 03/27/99

Fire Suppression Water Supply Systems 16.9.1 O

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.9.1.1 Functionally test the Keowee Fire 7 days Protection Pump.

SR 16.9.1.2 Functionally test the Oconee HPSW pumps, 31 days power supplies and associated automatic valve.

SR 16.9.1.3 Verify proper alignment of valves for 31 days Keowee and Oconee.

SR 16.9.1.4 Verify flow for the Oconee HPSW pumps. 12 months SR 16.9.1.5 Perform a performance test of the Keowee 12 months Fire Protection Pump.

SR 16.9.1.6 Flow test the Keowee Fire Water Suppression 12 months System by actuation of the Main Transformer water spray system.

SR 16.9.1.7 PerMrs Oconee Fire Suppression Water 36 months Supply System flow test in accordance with Chapter 5, Section 11 of the Fire Protection Handbook, 14 Edition, NFPA.

O 16.9.1-4 03/27/99

Fire Suppression Water Supply Systems 16.9.1 4 mm Portions of Surveillance SR 16.9.1.2 involving the HPSW pumps and power supplies were relocated from CTS Table 4.1-2, Item 8 during the conversion to 4

the ITS.

The OPERABILITY of the Fire Suppression System ensures that adequate fire suppression capability is available to confine and extinguish fires occurring

in any portion of the facility where safety-related equipment is located. The Fire Suppression System Consists of the Water Supply System, spray and/or sprinklers, Keowee CO, and fire hose stations. The collective capability of j the Fire Suppression Systems is adequate to minimize potential damage to 4 safety-related equipment and is a major element in the facility fire protection program. In the event that portions of the Fire Suppression l Systems are inoperable, alternate backup fire-fighting equipment is required 1 to be made available in the affected areas until the inoperable equipment is

{

restored to service.

The Testing Requirements provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met. In the event the Fire Suppression Water Supply System becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. This Selected Licensee Commitment is part of the

, Oconee Fire Protection Program and therefore subject to the Provisions of Oconee Facility Operating License conditions.

l

REFERENCES:

1) Oconee UFSAR, Chapter 9.5.1

2) Oconee Fire Protection SER dated August 11, 1978

3) Oconee Fire Protection Review, as revised.

4) Oconee Plant Design Basis Specification for Fire Protection as revised.

I,

O 4

16.9.1-5 03/27/99

l L

Sprinkler And Spray Systems 16.9.2 16.9 AUXILIARY SYSTEMS 16.9.2 Sprinkler And Spray Systems l

COMITMENT Sprinkler and Spray Systems in safety related areas listed in

, Table 16.9.2-1 shall be OPERABLE.

1 APPLICABILITY: At all times, j ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l

l A. Cae or more required A.1 Establish continuous I hour l Sprinkler or Spray fire watch with backup Systems inoperable. fire suppression i

equipment in the area.

AND

(' Affected Area (s) has no OPERABLE fire

! detection.

l l B. One or more required B.1 Establish hourly fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months l Sprinkler or Spray watch with backup fire j Systems inoperable. suppression equipment in the area.

AND Affected Area (s) has l

OPERABLE fire detection. ,

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16.9.2-1 03/27/99 I

Sprinkler And Spray Systems 16.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l I

SR 16.9.2.1 -----------------NOTE---------------------

Not required to be performed for systems in the cable spreading room, equipment rooms and cable shafts.

Functionally test each required Sprinkler 12 months or Spray System.

SR 16.9.2.2 Inspect each required Sprinkler Systems 12 months

. spray headers and nozzles.

SR 16.9.2.3 Verify by visual inspection each nozzle's 18 months spray area to ensure spray pattern is not obstructed.

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16.9.2-2 03/27/99

- - - - ~ - . - . - - - - - - . - - -. - . - - - .-. . - - - -

l Sprinkler And Spray Systems 16.9.2 Table 16.9.2-1 Sprinkler and Spray Systems

a. Oconee Ngglear Station

1. Turbine Driven Emergency FDW Units 1, 2, and 3 Pump

11. Transformers I CT-1, CT-2, CT-3, CT-4, CT-5 iii. Cable Room Units 1, 2, and 3 iv. Equipment Room Units 1, 2, and 3

v. Cable Shaft (3rd Level) Units 1, 2, and 3 1

vi. Cable Shaft (4th & Sth Level) Units 1, 2, and 3 ;

1

b. Keowee Hydro Station

1. Main Lube Oil Storage Room ii. Main Transformer

1. The transformers do not have fire detection devices. They have Activation devices that actuate the deluge valve of the fire suppression systems only.

O l 16.9.2-3 03/27/99

Sprinkler And Spray Systems 16.9.2

\

BhSEft

!. The OPERABILITY of the NRC committed Fire Suppression System ensures that j l

adequate fire suppression capability is available to confine and extinguish 1

, fires-occurring at the Oconee or Keowee facilities. The regulatory l l

requirement is to have NRC committed Sprinkler and Spray Systems OPERABLE only l when the equipment it is protecting is required OPERABLE for_ plant safety.

However, to protect the equipment for property conservation and minimize equipment loss due to fire; the Oconee and KeowerNRC committed Sprinkler and l Spray Systems will be rquired to be OPERABLE at all times.

The-Oconee CT-1, 2, 3, and 5 transformers do not have fire detection devices.

They have fire actuation devices that actuate the deluge valve of the fire suppression systems. These actuation devices do not directly annuciate to the l 1 Control Rooms. When the deluge valve trips, the flow pressure switch is the l sensor that activiates the Control Room alarms. With HPSW deactivated for l l maintenance or testing, there is no form of annucation of a fire in the )

1 Control Room. '

During periods of time when the Sprinkler or Spray system is not operable and I detection instrumentation is operable, a hourly fire watch patrol will be .

required to inspect the affected area frequently as a precaution. If the _

sprinkler or spray system in the area is not operable and no detection O instrumentation is operable, a continuous fire watch is required to be maintained in the vicinity of the affected sprinkler or spray system until the system is restored to operable status.

In the event that portions of the Fire Suppression Systems are inoper:bic, alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service.

The test requirements provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met.

This. Selected Licensee Commitment is part of the Oconee Fire Protection Program and therefore subject to the provisions of Oconee Facility Operating License conditions.

REFERENCES

1. Oconee UFSAR, Chapter 9.5-1.

2. Oconee Fire Protection SER dated August 11, 1978.

' 3.. Oconee Fire Protection Review, as revised.

4. Oconee Plant Design Basis Specification for Fire Protection, as revised.

l l

0 16.9.2-4 O/27/99

{?

1 Keowee CO Systems 2

16.9.3 16.9 AUXILIARY SYSTEMS 16.9.3 Keowee CO2 Systems l COMITMENT. The automatic CO, system provided for the generators at Keowee l Hydro Station shill be OPERABLE.

APPLICABILITY: 'At all times. l ACTIONS CONDITION- REQUIRED ACTION COMPLETION TIME

% )

A. Keowee C0 A.1 Establish continuous I hour :

inoperabl$. System fire watch with backup j fire suppression j equipment in the area. '

O SURVEILLANCE REQUIREMENTS-SURVEILLANCE FREQUENCY SR 16.9.3.1 Verify each valve in the flow path is in 31 days its correct position.

SR 16.9.3.2 Verify C07storage tank weight is 2: 90% 184 days full charge weight.

SR 16.9.3.3 Verify system actuates manually and 18 month automatically upon receipt of a simulated D action signal.

SR 16.9.3.4 Perform flow test through headers and 18 months O nozzles to assure no blockage.

I 16.9.3-1 03/27/99 ;

1 Keowee CO Systems 2

A 16.9.3 l

O ,

I l BASES The OPERABILITY of the NRC committed Fire 2,uppression system ensures that adequate fire suppression capability is available to protect safety-related equipment by confining and extinguishing fires occurring in the Keowee eletric generators. The Fire Suppression System consists of the water system, spray and/or sprinklers, Keowee 007 system and fire hose stations. The collective capability of the Fire Suppression Systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program.

In the event that portions of the Fire Suppression Systems are inoperable, alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. The Testing Requirements provide assurance that the minimum OPERABILITY 1 requirements of the Fire Suppression Systems are met.

This Selected Licensee Commitment is part of the Oconee Fire Protection Program and therefore subject to the provisions of Oconee Facility Operating License Conditions.

REFERENCES:

1. Oconee UFSAR, Chapter 9.5-1.

\]v 2. Oconee Fire Protection SER dated August 11, 1978.

3. Oconee Fire Protection Review, as revised.

4. Oconee Plant Design Basis Specification for Fire Protection, as revised.

16.9.3-2 03/27/99

a 4

i. Fire Hose Stations 16.9.4 1

f 16.9 AUXILIARY SYSTEMS 16.9.4 Fire Hose Stations

! COMITMENT The Fire Hose Stations listed in Table 16.9.4-1 shall be OPERABLE. l

APPLICABILITY: At all times. !

l ACTIONS i

. CONDITION REQUIRED ACTION COMPLETION TIME i A. Required Fire Hose A.1 Provide additional ~ I hour

, Station outside equivalent capacity fire

reactor building hose of length to reach j inoperable. unprotected area at

OPERABLE hose station.

i l- B. Required Fire Hose B.1 Ensure availability of.4 NA 4

Station inside portable fire i

reactor building extinguishers outside

inoperable (water not containment in the 1

available to personnel hatch area of

! isolation valves the auxiliary building

] LPSW-563 and for fire brigade use upon i

LPSW-564). entering reactor i building.

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16.9.4-1 03/27/99 i

1 Fire Hose Stations :

16.9.4 l

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

i SR 16.9.4.1 Perform visual inspection, including 31 days i inspection of coupling gaskets, of the fire !

hose stations located outside the reactor building and inside reactor building that l l are accessible during power operation.

l

, SR 16.9.4.2 Perform visual inspection, including 18 months inspection of coupling gaskets, of reactor building fire hose stations that are .

l inaccessible during power operation.

SR 16.9.4.3 Partially stroke test Fire Hose Station 36 months Valves. l i

i SR 16.9.4.4 Subject each fire hose to hydrostatic test 36 months at pressure a 50 psig greater than the maximum pressure at the station.

SR 16.9.4.5 Perform maintenance inspection including 36 months I removal and reracking the hoses and inspection of coupling gaskets.

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lO 16.9.4-2 03/27/99

Fire Hose Stations 16.9.4 Table 16.9.4-1 Fire Hose Stations

a. Oconee Nuclear Station Location No. Valve No. Area or Component Protectgd 3-D-28 2HPSW-194 1&2 Blockhouse.1 & 2 3rd Floor Switchgear M-35 1HPSW-436 #1 Cable Spread Room M-32 2HPSW-436 f2 Cable Spread Room M-33 2HPSW-437 1 & 2 Cable Spread Room M-30 3HPSW-436 f3 Cable Spread Room M-31 3HPSW-437 #3 Cable Spread Room 5-N-31 2HPSW-304 1 & 2 Control Room, 1 & 2 Emergency shutdown Panels TOH-3 3HPSW-338 f3 Control Room, f3 Emergency Shutdown Panels 1-J-28 2HPSW-242 fl First Floor MCCs HPSW Pumps.1 & 2 LPSW Pumps i 1-J-43 3HPSW-344 f31st Floor Motor Control Canters 1 1-8-19 1HPSW-283 #1 EFWP 1-D-39 2HPSW-236 #2 EFWP 1-D-53 3HPSW-336 #3 EFWP M-13 1HPSW-448 1 & 2 HP1 Pumps. 1 & 2 LPI Pumps M-14 3HPSW-449 3 HP1 Pumps, 3 LPI Pumps 1-J-47 3HPSW-348 3 LPSW Pumps M-36 1HPSW-445 #1 West Penetration Room M-45 1HPSW-444 fl East Penetration Room M-42 2HPSW-444 #2 East Penetration Room !

M-43 2HPSV-445 f2 West Penetration Room l M-29 3HPSW-444 f3 East Penetration Room M-44 3HPSW-445 f3 West Penetratton Room M-21 HPSW-457 1 & 2 Equipment Room g M-19 3HPSW-458 3 Equipment Room 3-N-24 HPSW-176 1 Equipment Room 3-N-29 2HPSW-245 2 Equipment Room 3-N-43 3HPSW-339 3 Equipment Room 3-J-28 2HPSW-241 1 & 2 3rd Floor Switchgear 3-N-43 3HPSW-339 3 3rd Floor Switchgear 600V Load Center M-22 1HPSW-440 1 Battery Room M-20 2HPSW-440 2 Battery Room M-18 3HPSW-440 3 Battery Room 1RBH1 ILPSV-471 Ground Floor Level - East Side 2RBH1 2LPSW-471 Basement Floor Level - East Side 3R8H1 3LPSW-471 Basement - East side IR8H2 ILPSW-473 Intermediate Floor Level - East Side

-2RBH2 2LPSW-473 Intermediate Floor Level - East Side 3RSH2 3LPSW-473 Intermediate Floor Level - East Side IR8H3 ILPSW-475 Top of shielding Floor Level - East Side 2RBH3 2LPSW-475 Top of Shielding Floor Level - East Side 3RBH3 3LPSW-475 Top of Shielding Floor Level - East Side IRBH4 1LPSW-465 Top of Shielding Floor Level - West Side 2RBH4 2LPSW-465 Top of Shielding Floor Level - West Side 3RBH4 3LPSW-465 Top of Shieldtra Floor Level - West $1de IRBHS ILPSW-467 Intermediate F.uor Level - West Side O

16.9.4-3 03/27/99

l l l Fire Hose Stations 16.9.4 Table 16.9.4-1 Fire Hose Stations l Location No. Valve No. Area or Conconent Protected 228HS 2LPSW-467 Intermediate Floor Level - West $1de 3ROHS 3LPSW-467 Intermediate Floor Level - West Side IR8H6 ILPSW-469 Ground Floor Level - West Side 2RSH6 2LPSW-469 Basement Flcor Level - West Side 3R8H6 3LPSW-469 Basement - West Side V8H-1 HPSW-916 Essential Siphon Vacuum Building

, VBH-2 HPSW-917 Essential Siphon Vacuum Building l Basement -

EL. 777' 6" Ground -

EL. 197* 6" Intermediate -

EL. 825' 0" Top of Shielding -

EL. 861' 0"

b. Keowee Hydro Station 1

Location No. Valve No. Area or Component Protected '

l Operating Deck (W) KH-1 Operating Floor Operating Deck (NE) KH-2 Operating Floor Operating Dock (SW) KH 4 Operating Floor !

Operating Deck (SE) KH-3 Operating Floor j Control Room KH-6 Control Room

\

Hech. Equip. Gallery KH-5 Hech. Equip. Gallery l

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l 16.9.4-4 03/27/99

i 1

Fire Hose Stations l 16.9.4 mu The OPERABILITY of the NRC committed Fire Suppression System ensures that adequate fire suppression capability is available to confine and extinguish fires occurring at the Oconee or Keowee facilities. The regulatory l requirement is to have NRC committed Fire Hose Stations OPERABLE only when the

equipment it is protecting is required OPERABLE for plant safety. However, to protect the equipment for property conservation and minimize equipment loss due to fire; the Oconee and Keowee NRC committed Fire Hose Stations will be rquired to be OPERABLE at all times.

1 In the event that portions of the Fire Suppression Systems are inoperable, j alternate backup fire-fighting equipment is required to be made available for the affected areas until the inoperable equipment is restored to service.

The testing requirements provide assurance that the minimum OPERABILITY requirements of the Fire Suppression System are mot.

This Selected Licensee Commitment is part of the Oconee Fire Protection Program and therefore subject to the provisions of Oconee Facility Operating License Conditions.

REFERENCES:

O 1. Oconee UFSAR, Chapter 9.5-1.

2. Oconee Fire Protection SER dated August 11, 1978.

3. Oconee Fire Protection Review, as revised.

i

4. Oconee Plant Design Basis Specification for Fire Protection, as required.

lO 16.9.4-5 03/27/99

1 Fire Barriers 16.9.5 0

16.9 AUXILIARY SYSTEMS 3

16.9.5 Fire Barriers j

4 COMITMEhT All Fire Barriers (including mechanical and electrical penetrations, fire doors, fire dampers, walls, ceilings and floors) boundaries, as shown on the 0-310-K and 0-310-L series

drawings shall be operable. I j APPLICABILITY: At all times

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME f A. Required Fire Barrier A.1 Determine OPERABILITY 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. status of fire detection 1 instrumentation for the affected area (s). '

O B. Required Fire Barrier B.1 Establish continuous I hour inoperable. fire watch on one side of affected penetration fire barrier.

AND !

Affected Area (s) has inoperable fire ;

detection '

instrumentation.

C. Required Fire Barrier C.1 Establish hourly fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable, watch patrols to inspect area.

AND Affected Area (s) has OPERABLE fire detection instrumentation.

i 16.9.5-1 03/27/99

Fire Barriers p 16.9.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.9.5.1 Visually inspect exposed surfaces of each 18 months fire rated barrier.

SR 16.9.5.2 Visually inspect at least 10% of all fire 18 months dampers. If apparent changes in appearance or abnormal degradation is found, a visul inspection of an additional 10% of the dampers shall be made. This inspection process shall continue until a 10% sample with no apparent changes in appearance or i

abnormal degradation is found. Samples shall be selected such that each fire damper will be inspected every 15 years. i SR 16.9.5.3 Visually inspect at least 10% of each type 18 months of sealed penetration. If apparent cianges in appearance or abnormal degradationa are found, a visual inspection of an additional 10% of each type of sealed penetration shall be made. This inspection process shall continue until a 10% sample with no apparent changes in appearance or abnormal 1 degradation is found. Samples shall be selected such that each penetration seal will be inspected every 15 years.

l 16.9.5-2 03/27/99

I Fire Barriers 16.9.5 RhifS The functional integrity of the penetration fire bcrriers ensures that fires will be confined or adequately retarded from spreading to adjacent portions of-the facility. This design feature minimizes the possibility of a single fire rapidly involving several areas of the facility prior to detection and extinguishment. .The penetration fire barriers are a passive element in the facility fire protection program and are subject to periodic inspections and sampling.

The~0PERABILITY of a-NRC committed fire barrier ensures that fires will be confined or adequately retarded from spreading to adjacent portions of the facility. The regulatory requirement is to have NRC committed Fire Barriers OPERABLE only when the equipment it is protecting is required OPERABLE for plant safety. However, to also protect the equipment for property conservation and minimize equipment loss due to fire; the Oconee and Keowee NRC committed Fire Barriers will be required to be OPERABLE at all times.

During periods of time when a barrier is not functional, a fire watch patrol will be required to inspect the affected area frequently as a precaution in addition to the fire detection instrumentation in the area. If fire detection ,

instrumentation in the area is nat operable, a continuous fire watch is !

required to be maintained in the vicinity of the affected barrier until the )

barrier is restored to functional status.

This Selected Licensee Commitment is part of the Oconee Fire Protection Program and therefore subject to the provisions of Oconee Facility Operating License Conditions. ;

REFERENCES:

1. Oconee UFSAR, Chapter 9.5-1.

2. Oconee Fire Protection SER dated August 11, 1978. j

3. Oconee Fire Protection Review, as revised.

4. Oconee Plant Design Basis Specification for Fire Protection, as revised.

0 G

16.9.5-3 03/27/99

Fire Detection Instrumentation 16.9.6 16.9' AUXILIARY SYSTDtS 16.9.6 Fire Detection Instrumentation ;

l COMMITMENT The provided Fire Detection Instrumentation for each !

cc,uipment/ location shall be OPERABLE as listed in Table 16.9.6-1. l

.........................--N0TE-----------------------------------

Fire Detection Instrumentation located within containment is not l required to be OPERABLE during the performance of Type A i Containment Leakage Rate Tests.

l APPLICABILITY: At all times.

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i 16.9.6-1 03/27/99 l 4

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Fire Detection Instrumentation !

16.9.6 )

ACTIONS

........................------NOTE---------------------------------------

OPERABILITY of fire detection instrumentation for adequate equipment / location coverage may also be determined by the Site Fire Protection Engineer or designee.

............................._................................................ l CONDITION REQUIRED ACTION COMPLETION TIME A. > 50%,of required A.1 ----------NOTE----------

detectors for one or An nourly firewatch is more Oconee not required for

equipment / location inaccessible inoperable, equipment / locations such as the Reactor Building QB at power operation.

Periodic inspections 2 required adjacent using a 1V camera (if detectors for one or available) are permitted more Oconee as described in Site g equipment / location Directives, or, the g inoperable. inaccessible equipment condition may be monitored by remote indications which would provide early warning of a fire.

Establish hourly fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months watch patrol (or as permitted by Site Directives) to inspect i the accessible area with the inoperable ,

instrumentation. I O

16.9.6-2 03/27/99

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

l-l Fire Detection Instrumentation 16.9.6 l

CONDITION REQUIRED ACTION COMPLETION TIME !

I B. > 50% of required B.1 Establish hourly fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months detectors for one or watch patrol to inspect i more Keowee the recessible area with equipment / location the inoperable inoperable. instrumentation.

9E l

2 required adjacent detectors for one or more Keowee equipment / location I l inoperable.

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16.9.6-3 03/27/99

Fire Detection Instrumentation 16.9.6 i

SURVEILLANCE REQUIREMENTS I

SURVEILLANCE FREQUENCY I 1

4 SR 16.9.6.1 Perform CHANNEL FUNCTIONAL TEST of Oconee 31 days Fire Detection Instruments using Fire Detection Instrumentation Control Board Panel Test Switch.

SR 16.9.6.2 Visually inspect Oconee Fire Detection 184 days l Instruments accessible during power !

operation. i

, SR 16.9.6.3 Visually inspect Keowee Fire Detection 184 days Instruments.

SR 16.9.6.4 Test each Oconee fire detector for 12 months sensitivity.

1 SR 16.9.6.5 Perform CHANNEL FUNCTIONAL TEST of Keowee 12 months Fire Detection Instruments.

SR 16.9.6.6 -------------------NOTE--------------------

Not required to be performed for Keowee Generator Detectors.

l Test each Keowee fire detector for 12 months I

sensitivity.

I SR 16.9.6.7 Visually inspect Oconee Fire Decection 18 months Instruments not accessible during power operation.

v 1

16.9.6-4 03/27/99 l

Fire Detection Instrumentation

- (~T 16.9.6

() Table 16.9.6-1 FIRE DETECTION INSTRUMENTATION OCONEE NUCLEAR STATION Units 1. 2. and 3 Reactor Buildinos Eautoment Detectors Provided Reactor Buildino Penetrations 8 (each unit)

Reactor Buildino Coolino Units 6 (each unit)

Reactor Coolant Pumos 8 (each unit)

Units 1. 2. and 3 Auxiliary Buildina EL 822' +0 3.95EL.g. Eauipment Detectors Provided 71-0 Unit 1 Cable Shaft 2 510 Unit I and 2 Control Room 10 75-0 Unit 2 Cable Shaft 2 552 Unit 3 Control Room 8 90-0 Unit 3 r ?le Shaft 2 EL. 809' + 3" Room No. Eauininent Detectors Provided 400 Unit 1 Control Battery Room 5 402 Unit 1 East Penetration Room 12 403 Unit 1 Cable Room and Cable Shaft 19 404 Unit 2 Cable Room and Cable Shaft 18 407 Unit 2 East Penetration Room 20 408 Unit 2 Control Battery Room 5 409 Unit 1 West Penetration Room 5 410 Unit 2 West Penetration Room 5 450 Unit 3 Cable Room 28 452 Unit 3 East Penetration Room 10 455 Unit 3 Ventilation Equipment 2 456 Unit 3 West Penetration Room 5 458 Unit 3 Control Battery Room 2 P

d 16.9.6-5 05/27/99

l l

l Fire Detection Instrumentation 16.9.6 s

pd Table 16.9.6-1 FIRE DETECTION INSTRUMENTAriON

'EL. 796' +6" h Eautoment Detectors Provided 300 Unit 1 Work Area 9 310 Unit 1 Equipment Room and Cable Shaft 13 311 Unit 2 Equipment Room and Cable Shaft 15 313 Janitor's closet (Unit 1) 1 314 Clean Protective Clothing Storage (Unit 1) 1 322 Protective Clothes Storage (Unit 2) 1 329 Hot Lab 1 330 Cold Lab 1 331 Counting Room (Unit 2) 1 333 Health Physics (Unit 2) 1 334 Office (Unit 2) 1 335 Environmental Lab (Unit 2) 1 337 Laundry Sorting (Unit 2) 1 338 Laundry Storage (U.iit 2) 1 339 Laundry (Unit 2) 2 347 Work Area (Unit 2) 8 354 Unit 3 Equipment Room and Cable Shaft 21 357 Janitor's Storage (Unit 3) 1 364 Towel Storage (Unit 3) 1 365 Janitor's Storage (Unit 3) 1 366 Protective Clothing (Unit 3) I b 369 HP Office (Unit 3) 1 V 369A 3698 Supv. Technicians Office Secondary Chemistry Lab 1

1 369C 1.C. Computer 1 376 Unit 3 Work Area 10 EL. 771* + 0 Room No. Eautoment Detectors Provided 119 Unit 1 and 2 LPI Hatch Area 3 159 Unit 3 LPI Hatch Area 2 EL. 838"+0 Room No. Eavioment Detectors Provided 611 Prote..tive Clothing Storage (Unit 2) 1 658 Protective Clothing Storage (Unit 3) 1 EL. 783* + 9" h Eauioment Detectors Provided 204 Storage (Unit 1) 1 207 Chemical Handling and Storage (Unit 1) 1 220 Hot Instrument Shop (Unit 2) 1 l 224 Storage (Unit 2) 1

~[ 264 Storage (Unit 3) 1

(

16.9.6-6 03/27/99

Fire Detection Instrumentation O

(,) Table 16.9.6-1 16.9.6 FIRE DETECTION INSTRUMENTATION EL. 758' +0 M. Eautoment Detectors Provided 54 Unit 1 High Pressure injection Pumps 1 56 Unit 1 and 2 High Pressure Injection Pumps 1 58 Unit 2 High Pressure Injection Pumps 1 61 Unit 1 Low Pressure Injection Pumps 2 62 Unit 1 and 2 Low Pressure Injection Pumps 2 63 Unit 2 Low Pressure Injection Pumps 2 1 76 Unit 3 High Pressure Injection Pumps 1 77 Unit 3 High Pressure Injection Pumps 1 81 Unit 3 Low Pressure injection Pumps 2 82 Unit 3 Low Pressure Injection Pumps 2 Units 1.2. and 3 Turbine Buildinos El. 775'+0 Eauioment Detectors Provided MCC IXC. 1XD. IXE. IXF: Unit 1 FDW 10 Turbines: Unit 1 Emergency Feedwater Turbine; Unit 1 H2 Panel: Unit 1 EHC Unit MCC 2XB. 2XC. 2XD 2XE. 2XF: Unit 2 FDW 11 Turbine: Unit 2 Emergency Feedwater Turbine; (s Unit 2 H2 Panel: Unit 2 EHC Unit MCC 3XC. 3XD. 3XE. 3XF; Unit 3 FDW 10 Turbines: Unit 3 Emergency Feedwater Turbine; Unit 3 H2 Panel: Unit 3 EHC Unit J S . 796' + 6*'

Eauipment Detectors Provided Switchgear ITA. 1TB 2TA 2TB; Load Centers B IX1, 1X2, 1X3. IX4. 1X5. 1X6. 2XI 2X2. ZX3.

2X4. 2X5. 2X6 Switchgear BIT. 82T: Transformer CT4 5 Switchgear 3 BIT 3B2T 3 MCC IXA 1 ITTC5 and ITTC6 1 Unit 1 Main Turbine Dil Tank 1 Unit 2 Main Turbine Dil Tank 1 Unit 3 Main Turbine Dil Tank 2 i DC Distribution Center IDA: Switchgear ITC. 7 j ITD. ITE MCC 1XGA 1 DC Distribution Center 2DA; Switchgear 2TC. 7 2TD 2TE MCC 3XGA 1 MCC 2XGB 1 Load Center 3XI. 3X2, 3X3. 3X4: MCC 3XGA: 5 Switchgear 3TC. 3TD. 3TE MCC 3XGB 1 S\

V 16.9.6-7 03/27/99

Fire Detection Instrumentation

("%

' 16.9.6 Table 16.9.6-1 FIRE DETECTION INSTRUMENTATION l

I EL. 822' + 0 Eauinment Detectors Provided Bearing 011 Lift Pumps for All Units 4 ea unit High Pressure Unit for All Units 2 es unit '

KE0 WEE HYDRO STATION Eautoment Detectors Provided Control Room 4 Battery Room 4 Mechanical Equipment Gallery 3 Main tube Oil Storage Room 1 Generators 1 and 2 6 ea Operating Floor 6 ESSENTIAL SIPHON VACUUM BUILDING 6 O

O 16.9.6-8 03/27/99

- . _ _ ~ - - -- - _ __

~ .I. ..~. . '

- ,u ...,,,.~..

O Fire Detection Instrumentation 16.9.6 Q

MSES OPERABILITY of the NRC committed Fire Det areas containing safety related and important to safety equipment at O Prompt detection of fires will reduce the potential and Keowee Facilities.

for damage to safety related equipment and is an integral element in overall facility fire protection program.have NRC committed However, toFire De equipment it is protecting is required OPERABLE for plant safety.

also protect the equipment for property conservation and minimize loss due to fire; the Oconee and Keowee NRC committed Fire Detection Instrumentation will be required to be OPERABLE at all times.

In the event that a portion of the Fire Detection Instrumentation is required to provide detection capability until is restored to operability.

This Selected Licensee Commitment is part of License Conditions.

REFERENCES:

1. Oconee UFSAR, Chapter 9.5-1.

2. Oconee Fire Protection SER dated August II, 1978.

3. Oconee Fire Protection Review, as revised.

4.

Oconee Plant Design Basis Specification for Fire Protection, a 5.

Oconee Plant Design Basis Specification for Fire Detection, as r n.

R h

03/27/99 16.9.6-9

l i

l Keowee Lake Level 16.9.7 16.'9 AUXILIARY SYSTEMS

- 16.9.7 Keowee Lake Level ,

'l COMITNENT a. '

Keowee lake level shall be > 794.15 ft to ensure that the l l requirements of ITS 3.7.7 (LPSW System) are met for all l three Units. '

b. One siphon sources shall be OPERABLE to ensure that the requirements of ITS 3.7.7 (LPSW System) are met for Unit 1. :

l

c. The HPSW system shall be OPERABLE to supply sealing water to l l the CCW pumps to ensure that the requirements of ITS 3.7.7 (LPSW System) are met for Unit 1. 1

d. Maintain lake ' level 2 784.15 ft to assure that the Keowee l Oil Storage Room Water Spray System shall be OPERABLE.

i e. Maintain lake level 2 781.15 ft to assure that adequate water supply shall be available for 7 days of Keowee 1 emergency operation. .

f. Maintain lake level 2 780.60 ft to assure that the Keowee Step-up Transformer Mulsifyre System shall be OPERABLE.

.......__.......____..__....-N0TES--------------------------------

1. The requirements of Commitment a, b and c do not apply in MODE 5.

1

2. The requirements of Commitments b and c do not apply to Unit 2 and 3.

3. Commitment f does not apply in MODE 5 when the Keowee step up transformer is not required to be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, 4, MODE 5 when a Keowee Hydro Unit (KHU) is required to be OPERABLE l

l .

l l

l s

16.9.7-1 03/27/99

Keowee Lake Level l

16.9.7 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME ;

1

.A. 'HPSW System not .

A.1 Verify lake level is Immediately:

available to supply greater than minimum sealing water to CCW level required for l

pumps. gravity (non-siphon) flow per Table 16.9.7-1.

E OB HPSW System not available to supply A . 2 - - - - - - - - - - - NCT E - - - - - - - - - -

sealing water to Unit 2 cannot supply both !

Unit 3 CCW pumps. Units 1 & 2 LPSW numps I and Unit 3 LPSW pumps simultaneously.

Verify LPSW pumps are Immediately capable of being supplied suction by Unit 2 ECCW per ITS 3.7.8.

! B. Required Action and B.1 Enter ITS LC0 3.0.3. Immediately associated Completion Time of Condition A

! not met.

L iO I-16.9.7-2 03/27/99

1 Keowen Lake Level 16.9.7 CONDITION REQUIRED ACTION COMPLETION TIME C. One required siphon C.1 Establish an additional Immediately source from Unit I siphon source by starting l inoperable. additional CCW pump (s) on l Unit 1. '

M C . 2 - - - - - - - - - - - NOT E - - - - - - - - - -

Unit 2 cannot supply both Units 1 & 2 LPSW pumps and Unit 3 LPSW pumps i simultaneously.

i Verify LPSW pumps are Immediately capable of being supplied suction by Unit 2 ECCW per ITS 3.7.8 M '

C.3 Verify lake level greater O than minimum level required for gravity Immediately (non-siphon) flow per Table 16.9.7-1. 1 D. Required Action and D.1.1 Verify one ECCW Immediately associated Completion siphon header Time of Condition C OPERABLE on Unit 2.

not met.

AND D.I.2 Enter applicable Immediately Condition for one required LPSW pump .

inoperable in accordance with ITS 3.7.7.

M D.2 Enter ITS LC0 3.0.3. Immediately 16.9.7-3 03/27/99

l Keowee Lake Level 16.9.7

(%

1 b CONDITION REQUIRED ACTION COMPLETION TIME E. Keowee Lake Level E.1 Enter applicable Immediately

< 794.15 ft. Condition for one required LPSW pump inoperable in accordance with ITS 3.7.7.

F. Keowee Lake Level F.1 Declare the Keowee Oil Immediately

< 784.15 ft. Storage Room Water Spray System inoperable.

G. Keowee Lake Level G.1 Cease commercial power Immediately

< 781.15 ft. generation using KHUs.

AND G.2 Notify the Plant Immediately Operations Review O' Committee (PORC) per NSD-308 and Request plant operation (and reportability) guidance. ,

H. Keowee Lake Level H.1 Declare Keowee Step-up Immediately

< 780.6 ft. transformer Mulsifyre inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.9.7.1 Verify Keowee lake level is within limits. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

\

16.9.7-4 03/27/99

. . _ . . _ ... _ _ _.m. _ ._ _ . . . _ _ _ _ _ . _ . . _ - . _ _ _ . . . _ . . . _ _ _ . _ _ _ _ _ . _ . _ _ . _ . . _ . _ _._ _ _.

Keowee Lake Level l 16.9.7 I

O TA8LE 16.9.7-1

. MINIMUM LAKE LEVEL FOR GRAVITY FLOW TO LPSW PUMPS SUCTION If Keowee lake level 2: minimum lake level in the following table, then gravity flow will provide adequate suction for the LPSW pumps without relying on the ECCW siphon:

l Number of CCW Pump Discharge Minimum Lake Level for Gravity Flow ,

l Valves currently Open (feet absolute)

1 805.06 2 801.79 l 3 800,85 L

4 800.37 5 800.08 6 799.87 7 799.71 8 799.59 I 9 799.48 10 799.40 11 799.33 i

12 799.26 Note: These lake levels are based on the assumption that all CCW crossover isolation valves (ICCW-40, 2CCW-41, 3CCW-42 and 3CCW-94) are open. If any of these valves are closed. contact Mechanical Systems i Engineering to determine the minimum lake level required for gravity

fl ow.-

!O 16.9.7-5 03/27/99 I

l l

1 Keowee Lake Level 16.9.7 185f1: ' !

An instrument error of 1.15 ft has been added to the absolute lake level to

.obtain the indicated lake levels identified in this SLC. The indicated lake levels in this SLC are based on the use of a conguter point to verify level.

Absolute lake level can be determined at the Keowee Hydro intake structure.

LPSW System Commitments:

UNITS 2 and 3: j The'CCW system provides the source of water to the CCW crossover piping which supplies suction to the LPSW system. Normally, this crossover header is ,

4 aligned to all three Oconee units, and CCW pumps provide adequate flow for the 1

, requirements of the'LPSW systems for all 3 units. To meet the requirements of l

, ITS 3.7.7, the Emergency CCW (ECCW) system must be capable of supplying '

1 suction to the_LPSW pumps in the event of a Loss of Off-site Power (LOOP).

The ECCW supply to LPSW must be capable of withstanding a single active l failure.

J 1

After a loss of power to the CCW pumps,. the ECW System is designed to supply l

suction to the LPSW pumps using a siphon assisted by the Essential Siphon l 4

[ -

Vacuum (ESV) pumps. ITS 3.7.8 establishes requirements for the EC 'W and ESV Systems.

l 1

9 lgilT 1:

The CCW system provides the source of water to the CCW crossover piping which

supplies suction to the LPSW system. Normally, this crossover header is

] aligned to all three Oconee units, and CCW pumps provide adequate flow for the i requirements of the LPSW systems for all 3 units. To meet the requirements of

ITS 3.7.7, tne Emergency CCW (ECCW) system must be capable of supplying i

suction to the LPSW pumps in the event of a Loss of Off-site Power (LOOP).

The ECCW supply to LPSW must be capable of withstanding a single active failure.

After a loss of power,to the CCW pumps, the ECCW System is designed to supply suction to the LPSW pumps using an unassisted siphon. To maintain siphon flow a capability, the ECCW piping must be relatively air-free and leak-tight. At i high lake levels, such as those depicted in Table 16.9.7-1, gravity flow may j -be adequate to supply suction to the LPSW pumps without relying on the siphon.

4 To help maintain ECCW siphon flow capability, HPSW must supply seal water to the CCW pump shafts to prevent air inleakage that may defeat the siphon. The Elevated Water Storage Tank (EWST) through valve HPSW-25 provides the seal

water necessary to the CCW pumps immediately following a LOOP. For longer-term CCW pump restart capability following a LOOP, refer to SLC 16.9.8 for HPSW pump requirements.

If the lake level is > 799.26 feet, it is possible to provide adequate suction pressure to the LPSW pumps due to gravity flow without dependance upon siphon 16.9.7-6 03/27/99

Keowee Lake Level 16.9.7 flow. The minimum lake level for gravity flov depends on the number of open CCW pump discharge valves before and during the LOOP event. Since the CCW

,* pump discharge valves remain as is after a LOOP event, the number of open CCW pump discharge valves during a LOOP is the same as the number of open CCW pump discharge valves before the LOOP event. Table 16.9.7-1 provides the minimum lake level for gravity flow as a function of the number of open CCW pump discharge valves.

To ensure siphon capability will be established in the event forced flow is stopped, the CCW inlet piping from the intake structure to the CCW crossover must be-maintained water-solid. Since the Continuous Vacuum Priming connections to the CCW inlet piping are normally isolated, the CCW piping is maintained water-solid by requir!ng a minimum number of CCW pumps operating on a given unit. " Water-solid" is defined as sufficient positive pressure to prevent gases from coming out of solution and sufficient flow to ensure accumulated gases will'be swept away. The CCW flowpath is maintained 3 water-solid by operating at least three CCW pumps on each Oconee unit being used as a siphon source.

One siphon sources shall be capable of providing siphon flow to the LPSW pumps. A " siphon source" for Unit 1 is defined as a water-solid flow path consisting of two 3 ft. CCW pump discharge valves open to a common 11 ft. CCW inlet header. One 11 ft. CCW inlet header being supplied by two CCW pumps alone does not constitute a siphon source. This is because a third CCW pump O must be running to supply sufficient back-pressure through the other 11 ft.

CCW inlet header. Therefore, whenever at least three CCW pumps are operating on a given unit, a water-solid flow path is assured in the 11 ft. CCW inlet header being fed by the two pumps. Running four CCW pumps does not result in two siphon sources on one unit.

The failure of a siphon source for Units 1 is not postulated since the siphon

. sources contain no active components. If the Unit 2 ECCW System is fully l

' OPERABLE per ITS 3.7.8, then Unit 2 may be credited for supplying either the 1 Units l'& 2 LPSW pumps or the Unit 3 LPSW pumps, and an ACTION need not be '

declared for those pumps being supplied by Unit 2 ECCW. If a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ACTION has been declared because lake level has fallen below 794.15 ft, then at least i one siphon source must continue to be maintained to avoid entering ITS 3.0.3.

Since ITS 3.7.8 has been implemented for Units 2 and 3, this SLC les not Opply to Unit 2 and 3. However, Unit 2 can help in meeting the requirement to shintain two siphon sources for Units 1. For example, if Unit I cannot operate three CCW pumps, then Unit 2 may be credited as one of the two required siphon sources provided that both ECCW siphon headers for Unit 2 are OPERABLE under ITS 3.7.8. Since the Unit 2 ECCW siphon headers rely on active components, they are vulnerable to active single failures. Therefore, both ECCW headers for Unit 2 must be OPERABLE for Unit 2 to qualify as a single siphon source for Unit 1.

, If only one ECCW siphon header is OPERABLE for Unit 2 and Unit 1 cannot be its

own siphon source, then the Units 1 and 2 LPSW pumps are not single failure i proof and an appropriate Action for one required LPSW pump inoperable entered in accordance with ITS 3.7.7 for Unit 1 16.9.7-7 03/27/99

. . = . . . .- .- . _- - . - . _ _ . - - - .- . .

' Keowee Lake Level q 16.9.7 Q

A Q THREE UNITS:

With lake level below 794.15 ft, calculations show that the LPSW pumps could experience inadequate NPSH with assisted siphon flow if a single failure 1 causes only the minimum number of LPSW pumps (two for the shared Unit I and 2 I LPSW System) to be available' during a design basis event. Therefore, the LPSW system must be considered unable to withstand a single failure for lake level below 794.15 ft and a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ACTION must be entered per ITS 3.7.7 by declaring one required LPSW pump inoperable.

Keowee Oil Storaae Room Commitment:

Should lake level fall below 784.15 ft, the Keowee Oil Storage Room water spray system may not provide the required flow rates because the system is dependent on lake level for driving head. For this reason, the spray system ,

l should be declared inoperable. '

Keowee Hydro Station Commitment:

With lake level below 781.15 ft, the water supply (for Keowee Hydro Station to provide emergency power to the overhead path at 46.5 MVA and the underground 1 path at 22.35 MVA) could be inadequate for 7 days of continuous operation at i these levels. Neither Keowee Hydro or Oconee Nuclear Station should be l considered inoperable at this lake level. Keowee Hydro ~should not generate to O the grid at lake levels below 781.15 ft in order to ensure ample water capacity for emergency power operation.

Keowee Main Start-up Transformer Commitment:

Should lake level fall below 780.60 ft, the Keowee main Step up Transformer Mulsifyre system may not provide the required flow rates because the system is dependent upon lake level for driving head. For this reason, the Mulsifyre should be declared inoperable.

REFERENCES:

1. PIR 0 092 0535, Potential Insufficient NPSH for LPSW pumps

2. LER 269/93 04, Rev. O and Rev. 1

3. 055-0254.00-00-1003, Rev. 8, Design Basis Specification for the CCW System

4. -OSS-0254.00-00-1039, Rev. 10, Design Basis Specification for the LPSW System

5. Calculation OSC 2895, Rev. 4, Hydraulic Calculations for Keowee Deluge Systems

6. Calculation OSC 5325, Rev. O, Keowee Lake Level Uncertainty Calculation

7. Calculation OSC 5304, Rev.1, Minimum Lake Level for Radwaste Equipment s Cooling System Isolation

8. Calculation OSC 5022, Rev.1, USQ Evaluation for Operability Evaluation of PIR 0-092-0535

9. Calculation OSC 2280, Rev. 10, LPSW NPSH and Minimum Required Lake Level 16.9.7-8 03/27/99

d Keowee Lake Level 16.9.7 O

10. Calculation OSC 5349, Rev.1, Minimum Lake level Required to Maintain Sufficient NPSH to the LPSW pumps via Gravity Flow

11. Calculation OSC 5670, Rev. 5, Required Number of CCW Intake Flow Paths

12. Calculation OSC 5461, Rev.1, Isolation of the Continuous Vacuum Priming System to the CCW Intake Piping

13. Calculation OSC-5409, Rev. 4, Single Failure Analysis of the ECCW System Supply to the LPSW Supply

14. Calculation OSC-3528, Rev. 3, Keowee Lake Level Minimum Administartive Limits

15. CTS 3.19, Emergency Condensor Circulating water, Ammendment Nos. !

229/230/226 !

1 l

l O

l f

A '

V l 16.9.7-9 03/27/99

_ _ -_ __ _ . _ _ _ . . . . _ _ - . . _ _ _ _ ~.._ ___ . . . _ . . _ _ . _ .

I HPSW Pump Requirements to Support LPSW

- 16.9.8

~

16.9 AUXILIARY SYSTEMS 16.9.8 HPSW Pump Requirements to Support LPSW COMMITMENT The HPSW System shall be OPERABLE as follows:

a. Two HPSW Pumps (combination of two of either HPSW A, HPSW B or Jockey Pump) shall be OPERABLE and capable of restarting prior to complete drain of the Elevated Water Storage Tank (EWST),

AND b.1 The EWST with HPSW-25 shall be OPERABLE to provide sealing / cooling water for CCW Pumps.

E b.2 Lake level shall be above the minimum lake level necessary to provide gravity flow to the suction of the LPSW Pumps without dependency on siphon flow.

/"'N -----------------------------NOTE--------------------------------

' _,)

\ This Selected Licensee Commitment does not apply to Units 2 and 3.

4 APPLICABILITY: Any time the LPSW system is required to be operable.

1 i

t I

!O .

16.9.8 1 03/27/99

HPSW Pump Requirements to Support LPSW 16.9.8 l

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I

A. Jockey pump unable to A 1 Declare Jockey Pump Immediately i

maintain EWST level. inoperable.

QR

Jockey pump unable to fill EWST. '

1 B. One required HPSW B.1 Declare required LPSW Immediately pump inoperable. pump not supplied by Unit 2 ECCW inoperable.

8HE l --------NOTE--------

l Unit 2 ECCW cannot i supply both Units 1 &

l 2 LPSW pumps and Unit l 3 LPSW pumps simultaneously.

I Required LPSW pump not supplied by Unit l 2 ECCW per ITS 3.7.8.

I j --

i i

I 16.9.8-2 03/27/99

i. . - . . . - -- __

HPSW Pump Requirements to Support LPSW 16.9.8 O

CONDITION REQUIRED ACTION COMPLETION TIME

,1

C. One Unit 1 Main C.1 Declare affected LPSW Immediately i Feeder Bus pump inoperable.

j inoperable.

4 E

4 --------NOTE--------

Unit 2 ECCW cannot j supply both Units 1 &

2 LPSW pumps and Unit

., 3 LPSW pumps j simultaneously.

1 Required LPSW pump not supplied from Unit 2 ECCW per ITS 3.7.8.

O D. Two required HPSW D.1 Enter ITS LCO 3.0.3. Immediately pumps inoperable.

E i --------NOTE--------

Unit 2 ECCW cannot supply both Units 1 &

2 LPSW pumps and Unit 3 LPSW pumps simultaneously.

LPSW pumps not supplied from Unit 2 i ECCW per ITS 3.7.8. !

)

1 4

O i l

16.9.8-3 03/27/99

_i

l l

HPSW Pump Requirements to Support LPSW 16.9.8 O

CONDITION REQUIRED ACTION COMPLETION TIME E. EWST inoperable E.1 Enter ITS LCO 3.0.3. Immediately '

8HE Lake Level not adequate to support gravity flow per SLC 16.9.7.

AND

NOTE--------

Unit 2 ECCW cannot supply both Units 1 &

2 LPSW pumps and Unit 3 LPSW pumps simultaneously.

O LPSW pumps not supplied from Unit 2 ECCW per ITS 3.7.8.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.9.8.1 NA. NA BASES Surveillance per SLC 16.9.1 and the Appendix B testing program is adequate to demonstrate the availability of the equipment and syst/ams discussed here.

The HPSW system provides support for the LPSW system suction, station fire suppression capability, back-up service water to the TDEFW Pump Bearing and l Oil coolers. HPSW system make-up demands are normally met by the HPSW Jockey Pump.

l 1

16.9.8-4 03/27/99

1 I

HPSW Pump Requirements to Support LPSW l q 16.9.8 D

The HPSW system must provide upper guide bearing sealing and motor cooling to ,

the Unit 1 CCW Pumps to ensure the CCW system can provide a suction supply to l the LPSW system. This is required to provide sealing water for CCW Pump shafts to prevent loss of siphon flow and to provide cooling for the CCW '

Pump (s) after restart of the CCW Pump (s) for Units 1 and 3. Sealing is required any time the CCW system is in the siphon flow mode of operation.

Cooling to the CCW Pump motors is required any time the CCW Pumps are required to operate. The Unit 2 and Unit 3 CCW pumps no longer rely upon the HPSW system for these functions, since the Siphon Seal Water (SSW) System, supplied from the LPSW system, fulfills these functions. Under some conditions, the Unit 2 ECCW System can supply adequate suction supply to the LPSW pumps for either Units 1 and 2 or Unit 3 per ITS 3.7.8. Therefore, the action statements for SLC 16.9.8 allow credit for the Unit 2 CCW supply to the LPSW system if ITS 3.7.8 requirements are met.

At certain lake levels unassisted gravity ' flow may be possible. If so, the EWST is not required to support siphon flow by providing sealing of the CCW Pump Upper Guide Bearing to prevent some of the air in-leakage that could defeat the ECCW siphon. However, HPSW is still required to support operation of the Unit I and Unit 3 CCW Pumps since procedures require that the Unit 1 CCW pumps must be restarted following a LOCA/ LOOP.

LPSW takes suction from the CCW crossover header. During certain analyzed accident conditions, a loss of power to the CCW Pumps for all three units must be assumed. This results in a loss of forced flow to the CCW crossover header.

Initially, the sealing requirements for Unit 3 1 are met via the EWST. The duration of the event may last beyond the capability of the inventory of the EWST. Therefore the HPSW Pumps must be capable of being started following a loss of power in order to meet the cooling and seal lubrication requirements of the Unit 1 CCW Pumps.

The HPSW Jockey Pump is supplied by " load shed" power and would not be available until after the load shed is reset. The CCW Design Basis Document (Section 20.1.1.3) requires a restart of a CCW Pump (for Unit 1) within one and one-half hours (for Unit 1). The load shed must be reset to restart the CCW pump (for Unit 1), thus the power would also be available to the Jockey Pump within that time frame. The Jockey Pump is of smaller capacity, would not meet fire protection capacity requirements, and would take longer to refill the EWST. Therefore, the Jockey Pump is considered as a substitute for an HPSW Pump only for purposes of supporting the siphon or the restart of a Unit 1 CCW Pump and not for Fire Protection.

The HPSW Jockey Pump is of smaller capacity than HPSW Pumps A and B.

Calculation OSC-5945, "HPSW Pump and Fire Protection Flow Test Acceptance Criteria" calculates the accident loads and concludes the HPSW Jockey Pump has sufficient capacity to supply those loads plus system leakage provided it is able to maintain the EWST level or fill the EWST in normal usage. Accident loads plus system leakage are calculated to be approximately the same as normal loads plus normal system leakage.

s' All three HPSW pumps are powered from the Unit 1 Main Feeder Busses. Backup power to the Unit 1 Main Feeder Busses is not available from another unit.

16.9.8-5 03/27/99

i l HPSW Pump Requirements to Support LPSW l

16.9.8 O Therefore, if one of the two available Unit 1 Main Feeder Busses is removed l from service, then the remaining HPSW pumps are vulnerable to a single failure of the other Unit 1 Main Feeder Bus. This would also result in LPSW not being single failure proof since HPSW is necessary for LPSW operation in the l conditions described above.

An EWST level of 70,000 gallons is chosen as the minimum level for EWST operability since this is the lowest level which would exist during normal daily operation. An EWST level of 70,000 gallons is the setpoint at which an

HPSW pump in " base" would start to make up to the EWST. This situation would i

! not be expected to occur during normal system operation since the HPSW Jockey pump is capable of maintaining EWST level at 100,000 gallons. The EWST is

.! considered out of service if HPSW-25 is out of service or if in any way water cannot be supplied from the EWST via the sealing water path to CCW Pumps or if EWST level cannot be maintained > 70,000 gallons.

REFERENCES:

l l

1. OSC-5409, Rev. 3, " Single Failure Analysis of the ECCW System Supply to the LPSW System".

2. OSC-5349, Rev.1, " Minimum Lake Level Required to Maintain Sufficient NPSH to the LPSW Pumps via Gravity Flow".

3. OSC-5945, Rev. O, "HPSW Pump and Fire Protection Test Acceptance ;

O Criteria". l

,) 4. PIP 0-094-0952

5. PIP 0-094-0995

6. PIP 0-095-0307

7. PIP 0-095-0174 l

8. Oconee UFSAR Sections 9.2.2, 9.5.1, 15.0, Table 9-4, Figure 9-9 through l 9-12; 12/31/96 update.

! 9. Selected Licensee Commitments 16.9.1, 16.9.7, as amended.

10. OSS-0254.00-00-1002, Rev. 7, " Design Basis Specification for HPSW".

11. OSS-0254.00-00-1039, Rev.10, " Design Basis Specification for LPSW".

12. OSS-0254.00-00-1003, Rev. 8, " Design Basis Specification for CCW". '

13. Letter dated 4/20/94 from J. W. Hampton (Duke) to NRC regarding '

supplemental information for revision to Tech. Spec. 3.4.

l

o l- 16.9.8-6 03/27/99 l

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

HPSW Requirements to Support Loss of LPSW 16.9.8a I

16.9 AUXILIARY SYSTEMS 16.9.8a HPSW Requirements to Support loss of LPSW COMMITMENT HPSW should be available to support loss of LPSW as follows:

a. HPSW should be available to provide backup cooling water to the Turbine Driven Emergency Feedwater Pump.

b. HPSW should be available to provide the backup cooling water to HPI Pump motor coolers.

APPLICABILITY: Any time the Turbine Driven Emergency Feedwater Pump or an HPI pump is required to be OPERABLE.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l t A. HPSW not available. A.1 Operations perform Risk NA l Assessment considering l l equipment out of service. I AND l A.2 Log unavailability NA l duration in the l Operations Log for Maintenance Rule performance monitoring.

l

!O l 16.9.8a-1 03/27/99 L

HPSW Requirements to Support Loss of LPSW 16.9.8a k O i SURVEILLANCE REQUIREMENTS ;

SURVEILLANCE FREQUENCY j

SR 16.9.8a.1 NA. NA i

j BASES Surveillance per SLC 16.9.1 and the Appendix B testing program is adequate to demonstrate the availability of the equipment and systems discussed here.

)

For loss of all AC power (Station Blackout, SB0), HPSW will provide cooling water to the Turbine Driven Emergency Feedwater Pump. However, the Standby 4 Shutdown Facility (SSF) remains the licensing and design basis commitment for decay heat removal during a Station Blackout. (Ref. 2, Section 20.1.2; Ref. 4, 7, 8 & 11).

For a loss of normal LPSW supply during a Turbine Building Flood, HPSW will provide cooling water to the HP1 Pump motor coolers. (Ref. 2, Section 20.1.2). 4 For both of these events, the source of HPSW inventory is the EWST since power will not be available for Station Blackout and the HPSW Pumps may be rendered i inoperable'in a Turbine Building flood. No operator action is required 4

because HPSW is supplied via self-contained pressure regulators and/or 2

air-operated fail open valves which open when needed to maintain cooling water supply whenever the normal cooling water supply (LPSW) fails. (Ref. 2, Section

20.1.2).

This SLC does not apply for the tornado scenario. For that scenario, the 4

Auxiliary Service Water (ASW) system supplies backup cooling water to the HPI Pumps since LPSW (primary cooling source) and HPSW (secondary cooling source) are vulnerable to damage by tornado. (Ref. 2, Section 20.1.4.1).

I In the Turbine Building Flood event, the SSF provides plant shutdown capabilities. However, HPSW availability does provide additional assurance of

being able to mitigate the effects of a Turbine Building flood event.

Although the HPSW Pumps are vulnerable to flood damage, the EWST is capable of providing a limited inventory of HPSW. (Ref. 2, Section 20.1.4.7).

. REFERENCES S

1. 10 CFR 50.65, " Maintenance Rule."

l 2. OSS-0254.00-00-1002 rev. 4, " Design Basis Specification for HPSW."

3. AP/1,2,3/A/1700/10, " Uncontrollable Flooding of Turbine Building,"

i ~

Approved 9/30/92. '

4. Letter dated 4/20/94 from J. W. Hampton (DPC) to NRC regarding supplemental information for revision to CTS 3.4.

16.9.8a-2 03/27/99

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

HPSW Requirements to Support. Loss of LPSW f 16.9.8a

5. Selected Licensee Commitments 16.9.1 and 16.9.8.

j 6. OSC 5945 rev. O, "HPSW Pump and Fire Protection Test Acceptance .

1 Criteria."

l 7. NRC Safety Evaluation Report, March 10, 1992, Supplemental December 3, i 1992 (Station Blackout).

8. UFSAR 8.3.2.2.4, Onsite Power Systems, Station Blackout Analysis.

9. UFSAR 9.2.2.1, Cooling Water Systems, Design Basis.

j 10. UFSAR 9.2.2.2.2, Cooling Water Systems, Description and Evaluation, HPSW.

j 11. NRC Safety Evaluation Report, Standby Shutdown Facility, 50-269, -270 &

-287.

j 12. OSC 5771, "PRA Risk-Significant SSC's for the Maintenance Rule."

13. Work Process Manual Section 607, " Maintenance Rule Assessment of

, Equipment Removed From Service."

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i 4

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e 4

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i 16.9.8a-3 03/27/99

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Auxiliary Service Water System and Main Steam Atmospheric Dump Valves l 16.9.9 l

16.9 AUXILIARY SYSTEMS l 16.9.9 Auxiliary Service Water (ASW) System and Main Steam Atmospheric Dump Valves COMMITMENT The ASW Pump and the associated piping and valves necessary to supply water to each Unit's Once Through Steam Generators (OTOGs) and to the HPI Pump motors shall be OPERABLE.

...................__.......N0TE----------------------------------

Included as part of the associated valves are the Main Steam Atmospheric Dump Valves.

APPLICABILITY: MODES 1, 2, and 3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME O A. ASW system inoperable. A.1 Restore ASW system to OPERABLE status.

30 days AND Standby Shutdown Facility (SSF) ASW System is OPERABLE.

B. ASW system B.1 Restore ASW system to 7 days inoperable. OPERABLE status.

_ANQ Standby Shutdown Facility (SSF) ASW System is inoperable.

l t C. Required Action and C.1 Submit eport to the NRC 30 day

[ associated Completion outliniri; plans and Time not met. procedures to be used to

provide for loss of the j system.

1 16.9.9-1 03/27/99

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Auxiliary Service Water System and Main Steam Atmospheric Dump Valves 16.9.9 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.9.9.1 Verify appropriate ASW pump discharge. 92 days capabilities.

SR 16.9.9.2 Stroke Main Steam Atmospheric Dump Valves. 18 months E85.ES The ASW system is designed to mitigate the consequences of a tornado event by providing emergency feedwater to one or more of the three units at Oconee.

This system shall be capable of supplying sdequate flow to all units simultaneously to remove core decay heat. The Main Steam atmospheric dump valves are required to be operable for the ASW system to be considered operable becrise the atmospheric dump valves must be opened to depressurize the OTSGs to allow the low-head ASW pump to supply water to the OTSGs.

Although it is desirable to maintain the ASW system operable to mitigate O design basis events, short periods of inoperability are necessary for testing and maintenance to assure a high degree of reliability for the ASW s'l stem.

Since the probability of any tornado striking the Oconee site is low, a seven day limiting enndition for operability (LCO) is reasonable for routine testing and maintenance.

The SSF ASW system is a redundant system and its availability reduces the need of the ASW system. The allowance of 30 days is deemed sufficient time for i extended maintenance to be performed on the system as long as the SSF ASW !

system is available. The testing requirements provide assurance that the 1 minimum OPERABILITY requirements of the ASW system are met.

REFERENCES

1. Design Basis Specification for the Emergency Feedwater and Auxiliary Service Water Systems (05S-0254.00-00-1000).

2. Oconee UFSAR, Section 9.2.3.

3. Oconee Probabilistic Risk Analysis, Section 3.4 :

4. Calculational File OSC-2262, " Tornado Protection Analysis."

5. Calculational File OSC-5771, "PRA Risk-Significant SSC's for the Maintenance Rule."

6. Work Process Manual Section 607, " Maintenance Rule Assessment of Equipment Removed From Service."

O 16.9.9-2 03/27/99

Comp:n:nt Cooling (CC) and HPI Seal Injection to Reactor Coolant Pumps 16.9.10

.16.9 ' AUXILIARY SYSTEMS 16.9.10 Component Cooling (CC)' and HPI Seal Injection to Reactor Coolant Pumps COMITMENT Both CC flow to Reactor Coolant Pump (RCP) thermal barrier / heat exchangers and High Pressure Injection (HPI) Seal Injection to RCP Seals should be maintained.

APPLICABILITY: Whenever the RCPs are' required to be OPERABLE.

ACTIONS

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

Station Abnormal Procedures (APs) specify appropriate actions to be taken in the event Component Cooling (CC) system flow and/or HP1 seal injection is lost to the RCPs.

CONDITION REQUIRED ACTION COMPLETION TIME A. CC flow to one or more A.1 Log unavailability NA RCP thermal barrier duration in the heat exchangers not Operations Log for maintained. Maintenance Rule O. Performance monitoring.

9.8 AND HPI seal injection to

-one or more RCP seals A.2 Perform a Risk Assessmer.t NA not maintained. considering equipment out of service.

SURVEILLANCE REQUIREMENTS !

SURVEILLANCE FREQUENCY l

SR 16.9.10.1 NA. NA !

BASES The intent of this Selected Licensee Commitment is to track Maintenance Rule unavailability and to ensure an acceptable level of risk associated with the j

O removal from service of the following system functions:

16.9.10-1 03/27/99 I

l i

l Component Cooling (CC) and HPI Seal Injection to Reactor Coolant Pumps ,

(~ 16.9.10 i C i

1. HPl seal injection to the RCPs. 4 1

2. Component Cooling flow to the RCP Thermal Barrier / Heat Exchangers.

)n the event that either or both of the listed functions is lost, instructions and limitations with respect to RCP operation are specified in existing plant procedures AP/1,2,3/A/1700/14, AP/1,2,3/A/1700/16, and AP/1,2,3/A/1700/20. !

l The interactions with other plant systems with respect to plant risk is managed via use of the ONS Risk Assessment Matrix found in WPM-607, in addition to applicable ITS specifications.

l Surveillance is by Operations Daily Rounas/ Checklists and normal Control Room monitoring.

REFERENCES ,

1. 10 CFR 50.65, ' Maintenance Rule."

2. OSS-0254.00-00-1001, " Design Basis Specification for the HP1 System." j

3. OSS-0254.00-00-1022, " Design Basis Specification for the CC System." l

4. AP/1,2,3/A/1700/14, revisions 3, 2, 1 respectively

5. AP/1,2,3/A/1700/16, revisions 4. 3, 4 respectively ,

6. AP/1,2,3/A/1700/20, revisions 2, 2, 2 respectively I O 7.

8.

Calculation No. 05C-5771, "PRA Risk-Significant SSC's for the Maintenance Rul e. "

WPM 607, " Maintenance Rule Assessment of Equipment Removed From Service."

O l

16.9.10-2 03/27/99

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l l Turbine Building Flood Protection Measures i p 16.9.11 v

16.9 AUXILIARY SYSTEMS 16.9.11 Turbine Building Flood Protection Measures .)

COPMITMENT Turbine Building Flood Protection Measures shall be OPERABLE as .

follows: I

a. CCW Pump Discharge Valves (1,2,3CCW-10 through -13) shall be ;

capable of being closed remotely unless one of the following conditions exists:

l

1. the unwatering blocks are installed for the associated i CCW inlet piping, l

1

2. the associated condensate coolers CCW flowpath is I isolated with locked closed valve (s), the associated !

waterbox inlet valves are locked closed, the crossover tie valves are locked closed, the CCW inlet piping is vented at the high point to disable the first siphon, and the CCW inlet piping is intact inside the Turbine Building, or

3. Keowee lake level is s 796.5 ft. absolute and the O

V associated CCW inlet piping is vented at the high point to disable the first siphon.

b. Condenser Outlet Valves (1,2,3CCW-20 through -25) shall be capable of closing automatically when all CCW pumps on the applicable unit are tripped to mitigate certain Turbine Building flood conditions unless one of the following conditions exists:

1. a condenser outlet valve is closed and air locked with air pressure vented and strongback installed,

2. a condenser outlet valve is closed with its operator removed and strongback installed,

3. the unwatering blocks are installed for the associated CCW discharge piping, or

4. Keowee lake level is s 791 ft. absolute and the associated CCW discharge piping is vented at the high point to prevent rave-se siphon flow.

c. Two flowpaths (one each from two different units) shall be available for reverse gravity flow through the Condensate coolers whenever Keowee lake level is greater than 791 ft. A i

p flowpath for reverse gravity flow consists of an open i

' d condenser discharge header, one failed-open condensate cooler CCW flow control valve, one open condensate cooler, and an open flowpath to the suction of the LPSW and SSF ASW Pumps.

16.9.11-1 03/27/99

Turbine Building Flood Protection Measures p 16.9.11 V

d. Prior to opening any condenser waterbox access hatch or creating any opening in the CCW, HPSW, or LPSW systems > 24 inches diameter (or multiple openings with equivalent diameter > 24 inches), an isolation boundary with single barriers shall be established to ' olate the opening from the lake using the following methodt 2s applicable:

1. Any manual valves > 24 inches diameter used for the isolation boundary shall be locked closed,

2. Any motor-operated valve > 24 inches diameter used for the isolation boundary shall be closed with its breaker locked open and the handwheel locked,

3. Any condenser outlet valve used for the isolation boundary shall be closed and air-locked with air pressure vented and strongback installed,

4. A physical barrier, such as unwatering blocks or blank flange, may be used for boundary isolation instead of valves,

e. The Turbine Building / Auxiliary Building boundary wall shall C be sealed below Elevation 795 ft with all water tight doors operable.

f. The Turbine Basement Water Emergency High Level alarm shall be operable.

g. The six foot diameter Turbine Building Flood drain shall be operable.

APPLICABILITY: At all times.

O O

16.9.11-2 03/27/99

1 Turbine Building Flood Protection Measures i 16.9.11 l

ACTIONS 1 CONDITION REQUIRED ACTION COMPLETION TIME I

A. Turbine Building Flood A.1 Initiate action to Immediately Protection Measures restore flood protection inoperable. measures to OPERABLE status.

AND A . 2 - - - - - - - - - - NOT E - - - - - - - - - - -

Entry into the associated Condition results in unavailability for all three units.

Log unavailability None duration in the Operations Log for l Maintenance Rule l O Performance monitoring.

AND A.3 Perform a Risk Assessment None using the PRA matrix considering CCW integrity '

not met for all three units.

O 16.9.11-3 03/27/99

l Turbine Building Flood Protection Measures 16.9.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY )

l L SR 16.9.11.1 Verify OPERABILITY of Turbine Basement 12 months )

Water Emergency High Level Alarm.

SR 16.9.11.2 Verify capability to close all four CCW 18 months pump discharge valves.

SR 16.9.11.3 Verify capability to automatically close 18 months condenser outlet valves when all CCW pumps are tripped.

BASES 1

One of the risk-significant Maintenance Rule functions for the CCW System is

._j to maintain system integrity to prevent or mitigate a Turbine Building flood.

The purpose of this Selected Licensee Commitment is to monitor the performance of the major design features associated with this function. To monitor .

performance of this function, any unavailability must be logged.

The Oconee UFSAR Section 3.4.1.1.1 describes the flood protection measures for the Turbine Building and Auxiliary Building. These measures are the basis for the commitments in SLC 16.9.11. The flood protection measures were implemented to reduce the overall risk of a Turbine Building flood, as determined by the Oconee Probabilistic Risk Assessment (PRA) study.

Upon detection of a Turbine Building flood, operators would trip the CCW pumps which would automatically close all condenser outlet valves. They would also close all CCW pump discharge valves. This may be done using a pushbutton in the control room that closes all four valves on that unit or by closing the valves individually using pushbuttons at the breaker compartment in the Equipment Room. This SLC is intended to ensure that the functional capability of the CCW pump discharge valves and condenser outlet valves will be maintained unless alternative actions have been taken.

In Commitment a, the CCW pump discharge valves shall be capable of being closed remotely. This precludes credit for manual operation of the valves during a flood since there may not be adequate time to take credit for manual operation. Additional options are provided in case the valves cannot be closed remotely. One option includes locking closed the condensate coolers O CCW flowpath, the waterbox inlet valves, and the crossover tie valves and venting the high point. Another option involves Keowee lake level s; 796.5 i 16.9.11-4 03/27/99 l

4 Turbine Building Flood Protection Measures 0 16.9.11 feet absolute with the CCW high point vented. The high point may be vented by opening valves or by other means, such as manways. These options provide additional flexibility to allow maintenance to be performed on the CCW pump discharge valves while preventing the possibility of CCW Siphoning into the i Turbine Building basement.

In commitment b of the SLC, additional options are provided to allow maintenance to be performed on the condenser outlet valves. Option "1" allows j a condenser outlet valve to be out of service if the valve is olocked closed with the air supply to the valve operator defeated. Option "2" is similar to "1" except that it allows the valve operator to be removed for maintenance if the strongback is installed. Option "3" involves installing the unwatering i blocks at the CCW discharge and venting the high point of the discharge piping. Option "4" allows the automatic valve operation to be out of service if the lake level is :s 791 feet absolute and the high point of the discharge piping is vented. Below this lake level, the CCW ' discharge pipe could not be refilled from the lake. Venting the high point may be accomplished by opening manways or by any available means. Credit cannot be taken for the normally open mid-point vent:: on the discharge piping, because these vents may not prevent reverse siphon flow.

Options "3" or "4" of Commitment b will make the affected flowpath incapable of applying towards the requirements in Commitment c, which requires two flowpaths for reverse gravity flow; however, Commitment c may still be met OT using other available flowpaths (e.g., other units).

Per UFSAR Section 3.4.1.1.1, the worst-case flood would involve failure of the expansion joint at the inlet to the condenser. There are other possible failures could lead to a Turbine Building flood. The flood consequences would vary depending upon the size of the opening and other factors. A flood that involved an opening greater than approximately 24 inches diameter may affect the Low Pressure Service Water (LPSW) pumps. Therefore, emphasis is placed on any activities that would create openings in the piping greater than 24 inches !

diameter.

Commitment d is provided to control activities that would create openings in the CCW, HPSW, or LPSW Systems. These activities are controlled to ensure that such openings are isolated from the lake using physical barriers (e.g.,

locks) and not just administrative barriers (e.g., valve tags). Commitment d requires that an isolation boundary be established on a case-by-case basis prior to opening a condenser waterbox access-hatch and for any openings > 24 inches, including multiple openings equivalent to 24 inches diameter. Single isolation is. acceptable, but the isolation boundary must include physical barriers, 'such as locked closed valves, and not just administrative barriers, such as valve tags. Physical barriers may include blocks or blank flanges. A stoppel plug or wet-tapping machine may also act as a physical barrier. This SLC is intended to address only the isolation of the opening from the lake.

If Keowee lake level is greater than' 791 ft., reverse gravity flow can be used O to provide suction to the LPSW and SSF ASW pumps. An analysis was performed to determine the optimum flowpath to supply suction to these pumps while minimizing.any excess flow that would contribute to additional flooding. This 16.9.11-5 03/27/99

Turbine Building Flood Protection Measures 16.9.11 analysis determined that flowpaths through one condensate cooler and one flow control valve on each of two units would be optimum. As a result of this analysis, Condensate Coolers CCW Flow Control Valves for Units 2 and 3 (2, 3CCW-84) have been permanently failed open by having their instrument air supplies removed. If either flowpath through Units 2 or 3 will be unavailable, an alternate flowpath should be provided on Unit I by failing open ICCW-84.

REFERENCES

1. UFSAR Sect'ons 3.4.1.1.1, 9.2.2, 9.6, and Figure 9-9,12/31/96 update.

2. Engineerii; Directives Manual EDM-210, " Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants or the Maintenance Rule." i

3. OSS-0754.00-00-1003, " Design Basis Specification for the Condenser '

Circulating. Water (CCW) System," Rev. 8.

4. 055-0254.00-00-3005, " Design Basis Specification for the Turbine Building l Structure," Rev. 1.

5. AP/1,2,3/A/1700/10, " Uncontrollable Flooding of Turbine Building,"

Approved 4/30/97. ,

6. Calculation No. C-0SA-SA-83-0002-0, Rev. O, 3/1/83, " Turbine Building '

Flood CCW Reverse Flow Analysis." l

7. Calculation NO. OSC-6522, Rev. O, 2/29/96, " Turbine Building Flood CCW Reverse Flow Analysis."

O- 8. Calculation No. OSC-6577, Rev. O, 6/7/96, "CCW Turbine Building Flood Analysis."

9. PT/1,2/A/0261/07, " Emergency CCW System Flow Test."

10. PT/3/A/0261/07, " Dam Failure Test."

11. IP/0/B/0235/03. " Turbine Basement Water Level Alarm System Check."

12. Calculation No. 05C-5771, PRA Ri.sk-Significant SSC's for the Maintenance l Rule."

13. Work Process Manual Section 607. " Maintenance Rule Assessment of Equipment Removed From Service".

14. OP/1,2,3/A/1104/12, " Condenser Circulating Water System."

15. Calculation 05C-6081, Rev. 2, CCW Seismic-LOOP Response."

16. Oconee Unit 3 Probabilistic Risk Assessment, Rev.1, November.1990.

l i

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! 16.9.11-6 03/27/99 l

t

l Additicnal Low Pressure Service Water (LPSW) Systea Operability Requirements 16.9.12 16.9 AUXILIARY SYSTEMS 16.9.12 Additional Low Pressure Service Water (LPSW) System OPERABILITY Requirements COMITMENT The following LPSW System Structures, Systems and Components (SSCs) shall be OPERABLE:

a. LPSW-4 ("A" LPI COOLER SHELL OUTLET)

b. LPSW-5 ("B" LPI COOLER SHELL OUTLET)

c. LPSW-139 (LPSW SUPPLY TO TB NON-ESSENTIAL HDR)

d. LPSW Pump Minimum Flow Recirculation Lines l

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

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l

A. LPSW flowpath though A.1 Declare associated Immediately an LPI cooler isolated LPI train inoperable, O by a manual valve.

B. LPSW-4 inoperable and B.1 Declare associated Immediately closed. LPI train inoperable.

9B LPSW-5 inoperable and closed.

O 16.9.12-1 03/27/99

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i l

Additional Low Pressure Service Water (LPSW) System Operability Requirements 16.9.12 CONDITION REQUIRED ACTION COMPLETION TIME C. LPSW flowpath through C.1 Enter applictble Imediately an LPI cooler not Condition of ITS isolated by manual 3.7.7 for 1 required valve. LPSW pump inoperable.

l 8!E LPSW-4 inoperable and l not closed. !

0.8 LPSW-5 inoperable and not closed.

D. One required LPSW pump D.1 Enter applicable Imediately es minimum recirculation Condition of V line inoperable. ITS 3.7.7 for one required LPSW pump

inoperable.

E. Two or more Unit 1 and E.1 Enter ITS LC0 3.0.3. Imediately 2 LPSW pump minimum recirculation lines inoperable when three l LPSW pumps are required to be OPERABLE.by ITS 3.7.7.

1

.............N0TE------------ !

F. LPSW-139 inoperable. Required Action is applicable to each ONS Unit supplied by ,

associated LPSW System.

F.1 Enter applicable Imediately Condition of ITS 3.7.7 for one required LPSW pump

( inoperable.

16.9.12-2 03/27/99

l i

Additional Low Pressure Service Water (LPSW) System Operability Requirements 16.9.12 I

O CONDITION REQUIRED ACTION COMPLETION TIME G. LPSW-139 inoperable on G.1 Enter ITS LC0 3.0.3 Immediately Unit 1. for DNS Unit 1 and !

ONS Unit 2. l 8!@

LPSW-139 inoperable on Unit 2.

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i O 16.9.12-3 03/27/99

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Additional Low Pressure Service Water (LPSW) System Operability Requirements 16.9.12 L

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

SR 16.9.12.1 Test LPSW-4, LPSW-5 and LPSW-139 in In accordance accordance with the Inservice testing with the

! Program. Inservice Testing Program i

S? 16.9.12.2 Test LPSW pump minimum recirculation lines. 18 months

[

BACKGROUND The Low Pressure Service Water (LPSW) System provides cooling water for normal and emergency services throughout the station. Safety related functions 3 served by this system include the Reactor Building cooling units (RBCUs), Low Pressure Injection (LPI) cooiers, and coolers for the High Pressure Injection (HPI) and Emergency Feedwater (EFW) motors.

APPLICABLE SAFETY ANALYSES

Sufficient LPSW System flow is required to meet the acceptance criteria of l containment heat removal safety analyses.

COMITMENT (S)

The following Low Pressure Service Water System SSC's shall be OPERABLE:

l a. LPSW-4 ("A" LPI COOLER SHELL OUTLET) l b. LPSW-5 ("B" LPI COOLER SHELL OUTLET)

c. LPSW-139 (LPSW SUPPLY TO TB NON-ESSENTIAL HDR)

d. LPSW Pump Minimum Flow Recirculation Lines O

V 16.9.12-4 03/27/99

Additional Low Pressure Service Water (LPSW) System Operability Requirements 16.9.12 APPLICABILITY This SLC applies in MODES 1, 2, 3, and 4. This applicability is consistent with the LPSW System operability requirements in Technical Specification 3.7.7. In MODES 5 and 6 the OPERABILITY requirements of the LPSW System are determined by the system it supports.

ACTIONS A.I. S.I. C.1 During normal operation, LPSW flow is isolated to the LPI coolers with block valves LPSW-4 and LPSW-5 in the closed position. If a LOCA occurs, LPSW-4 and LPSW-5 are required to be opened after Reactor Building Emergency Sump (RBES) recirculation is established. LPSW-251 and LPSW-252, for Units 1 and 2, and LPSW-404, and LPSW-405 for Unit 3 are the normal LPI cooler flow control valves and are normally in AUTO at a setpoint of 3,000 gpm. If a LOCA occurs, Instrument Air (IA) and Auxiliary Instrument Air (A1A) are assumed unavailable since they are not safety related. With LPSW-251 and LPSW-252, for Units 1 and 2, and LPSW-404 and LPSW-405, for Unit 3, failed open and unavailable, LPSW-4 and LPSW-5 are credited for throttling LPI cooler shell side flow to O maintain sufficient LPSW pump NPSH and adequate LPSW flow to the safety related loads. Therefore, as defined in this SLC, LPSW-4 and LPSW-5 operability is met by having the capability to throttle these valves from the control room.

If the LPSW flowpath through an LPI cooler is isolated due to a manual valve,

-then LPSW pump NPSH and LPSW flow to the other safety related loads would still be adequate. However, the LPSW flow to the affected LPI cooler would not be adequate. Thus, if the LPSW flowpath through an LPI cooler is isolated due to a manual valve, then the affected LPI train shall be declared

. inoperable.

If LPSW-4 or LPSW-5 is closed and not capable of throttling LPSW flow, then LPSW pump NPSH and LPSW flow to the other safety related loads would still be adequate. However, the LPSW flow to the affected LPI cooler would not be adequate. Thus, if LPSW-4 or LPSW-5 is closed and does not have throttle capability, then the affected LPI train shall be declared inoperable.

If one or both LPSW-4 and LPSW-5 are not closed but not capable of throttling LPSW flow, then LPSW pump NPSH and LPSW flow to the safety related loads may be inadequate. If a single failure of an LPSW pump is not assumed, then sufficient LPSW pump NPSH and LPSW flow to the safety related loads does exist. Thus, if one or both LPSW-4 and LPSW-5 are not closed and do not have throttle capability, then the LPSW system cannot withstand a single failure and the affected unit (s) shall enter the applicable Condition of ITS 3.7.7 for 16.9.12-5 03/27/99

Additional Low Pressure Service Water (LPSW) System Operability Requirements 16.9.12 l' one required LPSW pump inoperable. For Units 1 & 2, both units would be l affected if a valve on either unit is inoperable.

D.1. E.1 NSM ON-1,2,33001 removed LPSW-4 and LPSW-5 from ES actuation. By maintaining isolation of LPSW flow to the LPI Coolers during the initial phase of a LOCA, the potential exists for the LPSW pumps to be operated below the manufacturer's recommended minimum continuous flow rate of 4,250 gpa per pump.

If all LPSW pumps successfully start'and operate during the event, the potential exists for a stronger pump to deadhead a weaker pump during low flow conditions. To avoid damaging a pump due to minimum flow concerns, minimum flow recirculation piping exists for each LPSW pump. The minimum flow recirculation lines ensure the operability of a deadheaded pump until LPSW-4 or LPSW-5 are open on the LOCA unit after RBES recirculation is established.

If an LPSW pump's minimum flow recirculation line is inoperable, the LPSW system is not single failure proof and the associated unit shall enter the t applicable Condition of ITS 3.7.7 for one required LPSW pump inoperable. If both Unit 3 LPSW pump minimum flow recirculation lines are inoperable, the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ITS Completion Time is still appropriate because the stronger pump will always have sufficient flow. Likewise, if the Unit 1&2 LPSW system is in a condition that only requires two OPERABLE LPSW pumps per ITS 3.7.7, the O minimum flow recirculation lines associated with both OPERABLE pumps may be simultaneously inoperable for a duration of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months permitted by ITS i

l LCO 3.7.7. If the Unit 1&2 LPSW system is in a condition that requires three OPERABLE' LPSW pumps per ITS 3.7.7 and two or more minimum flow recirculation lines are out of service, both Unit 1 and Unit 2 shall enter ITS LCO 3.0.3.

F.1. G.I. H.l. 1.1 During normal operation, valve LPSW-139 is open to supply LPSW flow to the Main Turbine Oil Tank (MTOT) and other various non-essential loads on the applicable unit. In the event of a LOCA, LPSW-139 is credited to close after RBES Recirculation is established, but prior to opening valves LPSW-4 and LPSW-5. Since the Unit 1&2 LPSW system is shared, both LPSW-139 for Unit I and LPSW-139 for Unit 2 shall be closed if the non-LOCA unit has tripped due to a concurrent LOOP. Closing LPSW-139 maintains sufficient LPSW pump NPSH and adequate LPSW flow to the safety related loads. Remote closure capability for LPSW-139 shall exist from the control room. If LPSW-139 is not capable of closing, and a single failure of an LPSW pump occurs, LPSW pump NPSH and LPSW flow to the safety related loads may be inadequate. If LPSW-139 for Unit 1 or LPSW-139 for Unit 2 is closed or isolated by system block valves for maintenance, then the valve is still considered operable. Thus, if LPSW-139 is not capable of closing, the associated unit shall enter the applicable l Condition of ITS 3.7.7 for one required LPSW pump inoperable. Since the Unit

1&2 LPSW system is shared and LPSW-139 for Units 1 and 2 are normally open, l

the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time applies to both Unit 1 and Unit 2 if LPSW-139 for

. Unit 1 or LPSW-139 for Unit 2 is inoperable. If both ILPSW-139 and 2LPSW-139 l 16.9.12-6 03/27/99

l Additional Low Pressure Service Water (LPSW) System Operability Requirements 16.9.12 O

are inoperable, sufficient LPSW pump NPSH and LPSW flow to the safety related loads may not be available, even without a single failure. This scenario requires both Unit I and Unit 2 to entar ITS LC0 3.0.3.

SURVEILLANCE REQUIREMENTS SR 16.9.12.1

! This SR requires that LPSW-4, LPSW-5, and LPSW-139 be tested per Oconee's ASME

Section XI IST Program. Testing under this program is adequate to assure operability of these valves.

SR 16.9.12.2 l This SR requires that the LPSW pump minimum recirculation lines be tested

every 18 months. An 18 month frequency is adequate to ensure significant degradation has not occurred due to service water related fouling.

REFERENCES 1.- 05S-0254.00-00-1039, Design Basis Specification for the Low Pressure Service Water System, rev. 10.

2. OSC-2280, LPSW Pump NPSH and Minimum Required Lake Level, rev. 10.

l 3. 05C-4672, Unit 1&2 LPSW System Response to a large Break LOCA Using a Benchmarked Computer Hydraulic Model, rev. 7.

4. 05C-4489, Predicted Unit 3 LPSW System Response to a large Break LOCA Using a Benchmarked Computer Hydraulic Model, rev. 5.

5. PT/1/A/0251/023, LPSW System Flow Test, performed on 11/16/97.

l 6. PT/2/A/0251/023, LPSW System Flow Test, performed on 4/20/96. l l

7. PT/3/A/0251/023, LPSW System Flow Test, performed on 1/19/97. ;

8. PT/1,3/A/0251/01, LPSW Pump Test. l

9. ITS 3.5.3 and 3.7.7.

10. Oconee UFSAR Section 9.2.2,12/31/96 update. j

11. Letter from J. W. Hampton, (DPC), to USNRC, dated June 6, 1996, Proposed i Technical Specification amendment for LPSW-4, -5.

12. NRC Safety Evaluation Report, dated August 19, 1996, Technical Specification Amendment 217/217/214.

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l 16.9.12-7 03/27/99

Spent Fuel Cooling System i 16.9.13 16.9 AUXILIARY SYSTEMS 16.9.13 Spent Fuel Cooling System COMITMENT Perform specified SR. I APPLICABILITY: When irradiated fuel assemblies are stored in the spent fuel pool.

I ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A N/A 1 l

l l

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY O SR 16.9.13.1 Functionally Test the Spent Fuel Cooling Prior to each System. refueling Bases The requirement (s) of this SLC section were relocated from CTS Table 4.1-2, Item 9 during the conversion to ITS.

Functional testing of the Spent Fuel Cooling System is performed prior to refueling to assure proper system operation.

References

1. UFSAR 9.6.1.

2. DBD 055-0254.00-00-1006, Rev. 1.

O 16.9.13-1 03/27/99

r- 1 l

SSF Diesel Generator Inspection Requird.m nts 16.9.14

'16.9 AUXILIARY SYSTEMS 16.9.14 SSF Diesel Generator Inspection Requirements COMITMENT Perform specified SR.

APPLICABILITY: MODES 1, 2 and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME t

A. N/A A.1 N/A N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.9.14.1 Perform Diesel Generator inspection in 12 months accordance with procedures prepared in conjunction with its manufacturer's recommendations for this class of standby service.

EMU The requirement (s) of this SLC section were relocated from CTS 4.20.3.a.4 during the conversion to Improved Technical Specification.

The testing of the SSF electrical power systems are based upon a review of the surveillance requirements of other similar type of equipment contained within the technical specifications, manufacturer recommendations, and appropriate NRC guidelines.

References N/A O

16.9.14-1 03/27/99

l l Radioactive Material Sources 16.9.15 I i

1 16.9 AUXILIARY SYSTEMS l 16.9.15 Radioactive Material Sources !

COPMITMENT Leakage and/or contamination of sealed sources shall be

< 0.005 pCi of removable contamination.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l l

A. Leakage not within A.l.1 Remove source from Immediately- .

limit. use. l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.9.15.1 -----------------NOTES---------------------

1. Not required to be met for sources containing s 100 Ci of beta and/or gamma-emmitting material or s 10 Ci of alpha-emmitting material.

2. Not required to be met for sealed sources that are stored and not being used.

3. Not required to be performed for startup sources subject to core flux.

4. Only required to be met for sealed source containing radioactive material, other than tritium, with a half life greater than 30 days and in any form other than gas.

Verify leakage and/or contamination of each Once each 184 sealed source is within limit. days I

1 h

J (continued) 16.9.15-1 03/27/99

! Radioactive Material Sources 16.9.15

! SURVEILLANCE REQUIREMENTS (continued)

I SURVEILLANCE FREQUENCY l

SR 16.9.15.2 Verify leakage and/or contamination of each Prior to any i startup source is within limit. repair or l maintenance i AND 1

Following any I repair or maintenance i

AND Before being initially subjected to core flux 0 BASES The requirement (s) of this SLC section were relocated from Technical Specification 4.16 during the conversion to improved Technical Specification.

This specification assures that leakage from any byproduct, source, and special nuclear radioactive materials sources does not exceed allowable limits.

The leakage test shall be capable of detecting the presence of 0.005 microcuries of radioactive material on the test sample.

Leak testing of sources that are stored and not in use is not required. The sources excepted from this test shall be tested for leakage prior to any use or transfer to another user unless they have been leak tested within six months prior to the date or use or transfer. In the absence of a certificate from a transferor indicating that a test has been made within six months prior to transfer, sealed sources shall not be put into use until tested.

Sources removed from service due to leakage and/or contamination shall be decontaminated and repaired or disposed of in accordance with NRC regulations.

v) 16.9.15-2 03/27/99

1 l

Radioactive Material Sources I 16.9.15 Leak testing is not required for the check sources contained in extended range area monitors (1)(2)(3)RIA-3, 4, 6, 15, 16, and 17. SR 16.9.15.1 excludes testing sources which are " stored and not being used." Area monitors l (1)(2)(3)RIA-3, 4, 6, 15, 16, and 17 have high range detectors which are permanently installed and contain 200 Ci keep alive check sources which are not normally accessible and remain sealed within detector housings.

Therefore, these sources are considered to be " stored." These sources are considered as "not being used" because they remain stationary within the detector housing and are not subject to any friction due to contact with ,

moving parts.

4 REFERENCES N/A O

O 16.9.15-3 03/27/99

l Reactor Building Polar Crane and Auxiliary Hoist (RCS System Open) 16.9.16 16.9 AUXILIARY SYSTEMS 16.9.16 Reactor Building Polar Crane and Auxiliary Hoist (RCS System Open)

COMITMENT Operation of the Reactor Building Polar Crane and auxiliary hoist shall be . restricted as follows:

a. The Reactor Building Polar Crane shall not be operated over

! the fuel transfer canal when any fuel assembly is being moved,

b. The auxiliary hoist shall not be operated over the fuel transfer canal when any fuel assembly is being moved,

c. When irradiated fuel is in the reactor building and fuel is not being moved, the reactor building polar crane and auxiliary hoist shall be operated over the fuel transfer canal only where necessary and in accordance with approved operating procedures stating the purpose of such use, and

d. A flagman shall be located at the top of the secondary shield wall when the polar crane hook is above the elevation 1 of the fuel transfer canal when the polar crane is operated in areas away from the fuel transfer canal.

f

................____..----N0TE------------------------------

Commitment part b is not required to be met when the auxiliary hoist is being used to move the fuel assembly.

APPLICABILITY: Fuel in reactor building and reactor vessel head removed.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A 1

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16.9.16-1 03/27/99 j t

Reactor Building Polar Crane and Auxiliary Hoist (RCS System Open) 16.9.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.9.16.1 N/A. N/A EH1 4

~

The requirement (s) of this SLC section were relocated from CTS 3.12.1, 3.12.2, 3.12.3, and 3.12.4 and an associated Technical Specification Interpretation '

during the conversion to ITS.

Restriction of use of the reactor building polar crane and auxiliary hoist over the fuel transfer canal when the reactor vessel head is removed to those operations necessary for the fuel handling and core internals operations is to preclude the dropping of materials or equipment into the reactor vessel and possibly damaging the fuel to the extent that an escape of fission products would result. The fuel transfer canal will be delineated by readily visible markers at an elevation above which the reactor building polar crane would not normally handle loads.

The fuel transfer canal is the area bounded by the following:

Unit 1 - East and West by the secondary shield walls

- South by .the containment wall plate

- North by the 3rd floor handrail Unit 2 - East and West by the secondary shield walls

- North by the containment wall plate

- South by the 3rd floor handrail .

A fuel assembly is being moved when the Main or Auxiliary bridges are attached to a fuel assembly or a fuel assembly is within the transfer carriage while in the reactor building.

The polar crane is the trolley section, which contains the hooks, blocks and cable drums.

The purpose of having restrictions on use of the polar crane is to prevent the dropping of material or equipment and possibly damaging fuel to the extent that an escape of fission products would occur. The UFSAR, section 9.1.4.1.5 states that the fuel transfer canal is a passageway in the Reactor Building extending from the reactor vessel to the reactor building wall and formed by an upward extension of the primary shield wall. In order to form a boundary for polar crane operation, this area is modified slightly to conform to easily 3

identified structures which provide an extra margin of safety. Therefore the East and West side of the canal are denoted by the secondary shield wall 16.9.16-2 03/27/99

Reactor Building Polar Crane and Auxiliary Hoist (RCS System Open) 16.9.16 immediately adjacent to the primary shield wall and the 3rd floor handrail just outside of the primary shield wall of the canal shallow end. The reactor building wall plates determines the final side of the canal area.

It is permissible to operate the polar crane over the fuel transfer canal when absolutely necessary, except during times when any fuel assembly is being moved. Sir.ce fuel is in place in the reactor vessel, fuel movement takes place only when fuel is moved into or away from the vessel. Thus once a fuel assembly is attached to the Main or Auxiliary bridge it is considered to be in I the act of movement. Also while fuel is in the transfer carriage, being moved into or out of the reactor building, it is considered fuel movement. The reactor building polar crane consists of two connected steel beams on which a trolley assembly moves. The two beams stretch over the full length of the reactor building and are always positioned over the fuel transfer canal. The trolley is the component which actually move equipment up, down and around the reactor building. Therefore for purposes of this SLC, the polar crane consists of the trolley section, whether or not a load is attached.

?

1 REFERENCES

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16.9.16-3 03/27/99

Reactor Building Polar Crane (RCS at elevated temperature and pressure) 16.9.17 A

Q 16.9 AUXILIARY SYSTEMS 16.9.17 Reactor Building Polar Crane (RCS at elevated temperature and pressure)

COMITMENT The Reactor Building Polar Crane shall not be operated over the steam generator compartments.

APPLICABILITY: MODES 1, 2, MODES 3 and 4 with RCS pressure > 300 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.9.17.1 N/A. N/A BASES The requirement (s) of this SLC section were relocated from CTS 3.12.5 and an !

associated Technical Specification Interpretation during the conversion to i ITS.

Restriction of use of the reactor building polar crane over the steam genera-tor compartments during the time when steam could be formed from dropping a i load on the steam generator or reactor coolant piping resulting in rupture of !

the system is required to protect against a loss of coolant accident. The !

polar crane is the trolley section, which contains the hooks, blocks and cable drums.

The polar crane is the trolley section, which contains the hooks, blocks and cable drums.

16.9.17-1 03/27/99 i

i i

Reactor Building Polar Crane (RCS at elevated temperature and pressure) 16.9.17 The purpose of having restrictions on use of the polar crane is to prevent the dropping of material or equipment and possibly damaging fuel to the extent that an escape of fission products would occur.

The reactor building polar crane consists of two connected steel beams on which a trolley assembly moves. The two beams stretch over the full length of the reactor building and are always positioned over the fuel transfer canal.

The trolley is the component which actually move equipment up, down and around the reactor building. Therefore for purposes of this SLC, the polar crane consist of the trolley section, whether or not a load is attached.

Reference.

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16.9.17-2 03/27/99

l Snubbers

!. 16.9.18 16.9 AUXILIARY SYSTEMS L 16.9.18 Snubbers i

. COMMITMENT Hydraulic and Mechanical snubbers specified in the appropriate Station Procedure shall be OPERABLE.

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

ACTIONS

..............................__.......N0TE-----------------------------------

Separate Condition Entry is allowed for each snubber.

CONDITION REQUIRED ACTION COMPLETION TIME l

A. One or more snubbers A.1 Restore snubber (s) to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

B. Required Action and B.1 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months associated Completion Time not met SURVEILLANCE REQUIREMENTS SURVElLLANCE FREQUENCY SR 16.9.18.1 Perform visual inspections of each snubber ------NOTE-----

in accordance with Table 16.9.18-1. The provisions of SLC 16.2.7 do not apply.

In accordance with Table 16.9.18-1 g (continued) i t

\

16.9.18-1 03/27/99

Snubbers 16.9.18 O

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 16.9.18.2 ------------------NOTE---------------------

The maximum expected service life for the various seals, seal materials, and applications shall be estimated based on engineering information, and the seals shall be replaced so that the maximum expected service life is not exceeded by more than 10% during a period when the snubber is required to be OPERA 8LE. The seal replacements shall be documented and the documentation shall be retained in accordance with Quality Assurance Requirements.

Verify that the seal service life of N/A hydraulic snubbers is not exceeded by more than 10% between surveillance inspections.

SR 16.9.18.3 Perfnrm a functional test on a ------NOTE-----

representative sample of hydraulic snubbers The provisions and a representative sample of mechanical of SLC 16.2.7 snubbers in accordance with Table do not apply.

16.9.18-2. ---------------

In accordance with Table 16.9.18-2 O

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16.9.18-2 03/27/99 ;

I

1 Snubbers ;

/~T 16.9.18 l V

Table 16.9.18-1 (page 1 of 2)

Snubber Visual Inspections Visual inspections shall verify:

(1) that there are nc visible indications of damage or impaired OPERABILITY, (2) attachments to the fcundation or supporting structure are secure, and (3) in those locations where mechanical snubber movement can be manually induced, the snubbers shall be inspected as follows:

(a) Every 18 months +25%, the inaccessible snubbers shall be inspected near the beginning and the end of the outage.

(b) In the event of a severe aynamic event, snubbers in that system which experienced the event shall be inspected during '1e refueling outage to assure that the snubbers have frewom of ;

movement and are not frozen up. The inspection shall consist of

verifying freedom of motion. using one of the following: (i)

Manually induced snubber movement, (ii) evaluation of in place snubber piston setting; (iii) stroking the mechanical snubber Os through its full range of travel. If one or more mechanical snubbers are founa to be frozen up during this inspection, those 4 snubbers shall be replaced (or overhauled) before exceeding MODE l

5. Re-inspection shall subsequently be performed according to the .

schedule listed below.

i Snubbers which appear inoperable as a result of visual inspections may be determined OPERABLE for the purpose of establishing the next visual

inspection interval, providing that (1) the cause of the rejection is i clearly established and remedied for that particular snubber and for other snubbers that may be generically susceptible; and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE .

! However, when the fluid port of a hydraulic snubber is found to be uncovered, the snubber shall be tested by starting with the piston at the as found setting and. extending the piston rod in the tension mode direction.

G(^s 16.9.18-3 03/27/99

Snubbers 16.9.18

( Table 16.9.18-1 (page 2 of 2 Snubber Visual Inspections )

All snubbers be counted connected as inoperable to an inoperable common hydraulic fluid reservoi snubbers.

accordance with the following schedule: Snubber operability will be verified in No. Inoperable Snubbers

_per Inspection Period Subsequent Visual Insoection Period 0

1 18 months i 25%

2 12 mnnths i 25%

3,4 6 months i 25%

5,6,7 4 months i 25%

2B 2 months i 25%

1 month i 25%

Note: (1) The required inspection interval shall not be lengthened more than two steps per inspection.

(2)

Snubbers may be categorized in two groups, " accessible" or

" inaccessible," based on their accessibility during reactor operation.

These according to be twoschedule.

above groups may be inspected independently (3) q Hydraulic and mechanical snubber inspection schedules are independent.

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O 16.9.18 4 03/27/99

Snubbers 16.9.18 l l

Table 16.9.18-2 (page 1 of 2)

V Snubber Functional Testina At least once every 18 months +25%, a representative sample, a minimum of 10%

of the total of hydraulic and .a minimum of 10% of the total mechanical i snubbers in use in the plant, shall be functionally tested either in place or '

in a bench test. For each snubber that does not meet the functional test acceptance criteria, an additional minimum of 10% of the snubbers shall be functionally tested until none are found inoperative or all have been functionally tested. l 1

The representative sample selected for functional testing shall include the various configurations, operating environments and the range of size and capacity of hydraulic and mechanical snubbers. The representative sample shall be selected randomly from the total population of safety-related I hydraulic and mechanical snubbers. !

In addition to the regular sample, snubbers which failed the previous i functional test shall be retested during the next test period. If a spare i snubber has been installed in place of a failed snubber, then both the failed snubber (if it is repaired and installed in another position) and the spare snubber shall be retested. Test results of these snubbers may not be included for the re-sampling, and failures shall not require additional testing of other snubbers.

If any snubber selected for functional testing either fails to lockup or fails i l to move, i.e., frozen in place, an engineering evaluation will be performed to '

determine if the mode of failure could affect other snubbers of the same design. If this is det. ermined, then reporting requirements under 10CFR Part l 21 will be examined for applicability.

I When a snubber is found inoperable, an engineering evaluation will be performed in accordance with appropriate Station Procedure.

The hydraulic snubber functional test shall verify that:

1. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression. l

2. Snubber bleed, or release rate, where require is within the specified range in compression or tension. For hydraulic snubbers j specifically required not to displace under continuous load, the

! ability of the hydraulic snubber to withstand load without displacement shall be verified.

The mechanical snubber functional test shall verify that:

1. The force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.

/^N N]

16.9.18-5 03/27/99

r. - ._ - .. . . . _ _ ._ _ -.. _ _ _ _ . _ . . . _ . _ . . _

I Snubbers 16.9.18 Table 16.9.18-2 (page 2 of 2)

Snubber Functional Testina i

2. ,

Activation (restrain.ing action) is achieved within the specified range of velocity or acceleration in both tension and compression, l j

(Measuring the time required to travel a known distance, under load, is an acceptable method.) i 3.

Snubber release rate, where required, is within the specified range in compression or tension. For snubbers specifically r ' quired not to displace under continuous load, the ability of thc bber to withstand load without displacement shall be verifia.

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16.9.18-6 03/27/99

Snubbers 16.9.18 BASES The requirement s during the conve(rs) ion to ITS.of this SLC section were relocated from CTS 3.14 an Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup or shutdown. The consequence of an inoperable ,

snubber is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads. It is therefore i required that all snubbers required to protect the prim &ry coolant system or any other safety system or component be operable during reactor operation.

Since the snubber protection is required only during low probability events, a period of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed for repairs or replacements. In case a shutdown is required, the allowance of 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to reach MODE 5 will permit an orderly

' shutdown consistent with standard operating procedures.

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 failure or failure of the O system on which they are installed would have no adverse effect on any safety-related system.

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 tha$. interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections perfomed before the original required time interval has elapsed (nominal time less 25't.) may not be used to lengthen the required inspection interval unless so determined, by the engineer, from a previous window of a schedule. Any inspection whose results require a shorter inspection interval will override the previous schedule.

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by inservice functional testing, that snubber may be exempted from being counted as inoperable. Generically susceptible snubbers are those which are of a specific make or model and have the same design features directly related to rejection of the snubber by visual inspection, or are similarly located or exposed to the same environmental conditions such as temperature, radiation, and vibration.

When a snubber is found inoperable, an engineeriw avaluation is performed, in addition to the determination of the snebber mode of failure, in order to O '

16.9.18-7 03/27/99

Snubbers O 16.9.18 RAJf1 (continued) determine if any safety-related component or system has been adversely affected by the inoperability of the snubber.

To provide assurance of snubber functional reliability, a representative sample of the installed hydraulic snubbers will be functionally tested every 18 months. Observed failures of these sample snubbers shall require functional testing of additional units.

Hydraulic snubbers and mechanical snubbers may each be treated as a.different entity for the above surveillance programs.

Permanent or other exemptions from the surveillance program for individual snubbers may be granted L'y the Nuclear Regulatory 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. Snubbers so exempted shall be listed in a permanent record which references the exemption letter date.

REFERENCES N/A l N

16.9.18-8 03/27/99

Condensate Inventory Requirements for Emergency Feedwater 16.10.1

( 16.10 STEAM AND POWER CONVERSION SYSTEMS 16.10.1 Condensate Inventory Requirements for Emergency Feedwater C0ptilTMENT The combined inventory stored in the Upper Surge Tanks (UST) and the Hotweli shall be maintained greater than 145,000 gallons of water.

APPLICABILITY: MODE 1, MODE 2 with Thermal Power > 2% Rated Thermal Power ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Combined inventory in A.1 Restore inventory in the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ;

the UST and Hotwell UST and Hotwell to s 145,000 gallons of > 145,000 gallons of water. water.

O B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months I associated Completion Time not met.

1 i

SURVEILLANCE REQUIREMENTS 1 SURVEILLANCE FREQUENCY i SR 16.10.1.1 Verify combined inventcry in the UST and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ;

Hotwell is > 145,000 gallons of water (see Figure 16.10.1-1).

A b

16.10.1-1 03/27/99

. - . . . . . - - _. . . - - - - . - - . - . - . . . - - - - - . . . . . - . . . _ . _ = _ _ = . _ . - - .

Condensate Inventory Requirements for Emergency Feedwater l 16.10.1 Figure 16.10.1-1 Required Inventory For EFW REQUNtED INVENTCHtY FOR EPW i

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O 16.10.1 2 03/27/99

Condensate Inventory Requirements for Emergency Feedwater

,q 16.10.1 V

BASES The EFW design basis requires sufficient water supply be available to cool the Reactor Coolant System, to the point at which the Low Pressure Injection System can provide decay heat removal, after any of the design basis transients for the EFW system (Reference 3). The upper surge tanks are the assured, safety-related water source for the EFW system. The minimum ITS level of 6 feet ensures that adequate time is available to the operator to manually align alternate sources (Reference 2). The Hotwell, which is a non safety-related source, provides this alternate supply of water. Makeup may be available from the condensate storage tanks and the plant demineralized water system, but no credit will be taken for these additional makeup sources in this SLC.

UFSAR Section 10.4.7 states that an inventory of 145,000 gallons of water is required for a 50*F/hr cooldown to the point at which the Low Pressure Injection System can provide decay heat removal. This inventory is based on an initial power level of 102% prior to the loss of main feedwater. The reactor coolant pumps are assumed to be left on to maximize the heat input.

This inventory also assumes no recirculation via the turbine bypass valves.

REFERENCES

1. UFSAR Section 10.4.7.1

2. ITS 3.7.6

3. Design Basis Specification for the Emergency Feedwater and the Auxiliary Service Water Systems, Spec. OSS-0254.00-00-1000

4. 050-5964, EFW Combined Inventory sV 16.10.1-3 03/27/99

i Steam Generator Secondary Side P/T Limits 16.10.2 16.10 STEAM AND POWER CONVERSION SYSTEMS 16.10.2 Steam Generator Secondary Side Pressure and Temperature (P/T) Limits COMITMENT Tt.e secondary side of the steam generator shall not be pressurized above 237 psig if the temperature of the vessel i shell is below 110*F.

APPLICABILITY: At all times. i ACTIONS-CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A ERVEILLANCEREQUIREMENTS SURVEILLANCE FREQUENCY i

SR 16.10.2.1 N/A N/A BASES The requirement (s) of this SLC section were relocated from CTS 3.1.2.5 during the conversion to ITS.

The limitations on steam generator pressure and temperature provides protec-tion against nonductile failure of the secondary side of the steam generator.

At mtal temperatures lower than the RTuoi of +60'F, the protection against nonductile failure is achieved by limiting the secondary coolant pressure to 20 percent of the preoperational system hydrostatic test pressure.

The limitations of 110*F and 237 psig are based on the highest estimated RTuoy of +40*F and the preoperational system hydrostatic test pressure of 1312 psig.

The average metal temperature is assumed to be equal to or greater than the coolant temperature. The limitations include margins of 25 psi and 10*F for

, possible instrument error.

' REFERENCES 4

0

'C' N/A 4

16.10.2-1 03/27/99 ,

EFW Pump and Valve Testing :

'16.10.3 -

< 16.10 STEAM AND POWER CONVERSION SYSTEMS 16.10.3 Emergency Feedwater (EFW) Pump and Valve Testing COMITMENT Perform specified SRs..

APPLICABILITY: . When the associated EFW pump (s) ~and flow path (s) are required to be OPERABLE.

. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME' A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i

SR 16.10.3.1 Operate turbine-dr.iven and motor-driven EFW In accordance pumps on recirculation to the upper surge with Inservice tank for at least one hour. Testing Program SR 16.10.3.2 Verify EFW system flow to each steam Once prior to generator upon'an actual or simulated EFW exceeding 25%

actuation signal. RTP after any maintenance or modification which could degrade the EFW flow path I

l S.85.fd

! The requirement (s) of this SLC section were relocated from CTS 4.9.1, 4.9.2 l and 4.9.3 during the conversion to ITS.

16.10.3-1 03/27/99

' EFW Pump and Valve Testing 16.10.3 BASES (continued)

These tests shall.be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly. - In addition, during operation of the System Flow Test, flow to the steam generators shall be verified by control room indication.

Verification of correct operation is made both from the control room instrumentation and direct visual observation of the pumps. The parameters

, which are observed are detailed in the applicable edition of the ASME Boiler and Pressure Vessel Code,Section XI. The System Flow Test verifies correct total system operation following modifications or repairs.

~

REFERENCES i

UFSAR, Section 10.4.7.4 i

4

O i

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16.10.3-2 03/27/99

n LPSW Systea Testing 16.10.4 16.10 STEAM AND POWER CONVERSION SYSTEMS 16.10.4. Low Presssure Service Water (LPSW) System Testing COMITMENT Manually align valves LPSW-4 and LPSW-5 from the control

' room to demonstrate OPERABILITY of the Low Pressure

. Injection Coolers.

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

CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.10.4.1 Verify valves LPSW-4 and LPSW-5 actuate to ------NOTE-----

the correct position upon manual actuation The provisions from the control room. of SLC 16.2.7 do not apply.

18 months +25%

BASES The requirement (s) of this SLC section were relocated from CTS 4.5.1.1.2.a(2)

. during the conversion to ITS.

SR 16.10.4.1 verifies that LPSW-4 and (LPSW supply to LPI coolers) respond as required to manual alignment from the control room. The test will be considered satisfactory if valves LPSW-4 and LPSW-5 have completed their travel.

REFERENCES N/A 16.10.4-1 03/27/99

MSLB Feedwater Isolation Features 16.10.5 16.10 STEAM AND POWER CONVERSION SYSTEMS 16.10.5 Main Steam Line Break (MSLB) Feedwater Isolation Features COMITMENT Perform specified SR.

APPLICABILITY: -MODES 1 and 2, MODE 3 with main steam header pressure m 700 psig.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. . A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i

SR 16.10.5.1 Verify main feed pumps, main feedwater ------NOTC --

control valves and turbine driven emergency The provisions feedwater pump are appropriately of SLC 16.2.7

> actuated / inhibited by an actual or do not apply, simulated MSLB isolation signal. ---------------

18 months +25%

BMfJi The requirement (s) of this SLC section were relocated from CTS Table 4.1-2, Item 12 (TSC 96-09) during the conversion to ITS.

SR 16.10.5.1 verifies that equipment operated by the MSLB detection instrumentation functions properly upon reciept of a simulated or actual actuation signal. Main feedwater control valves include both the main feedwater and startup feedwater control valves j REFERENCES N/A-16.10.5-1 03/27/99

r l

l Emergency Feedwater Centrols 16.10.6 I

16.10 STEAM AND POWER CONVERSION SYSTEMS 16.10.6 Emergency Feedwater Controls COMMITMENT The controls of the emergency feedwater system shall be independent of the Integrated Control System.

i APPLICABILITY: MODES 1, 2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME j A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

SR 16.10.6.1 N/A N/A BASES The requirement (s) of this SLC section were relocated from CTS 3.4.6 during the comersion to ITS.

The EFW System is designed to start automatically in the event of loss of both main feedwater pumps as sensed by low hydraulic oil pressure. This specific automatic initiation logic is placed in service prior to criticality and may be bypassed when shutdown to present inadvertent actuation during startup and shutdown. All automatic initiation logic and control functions are indepen-dent from the Integrated Control System (ICS).

REFERENCES N/A O

16.10.6-1 03/27/99

Radioactive Liquid Effluents O 16.11.1 16.11 RADIOLOGICAL EFFLUENTS CONTROL 16.11.1 Radioactive Liquid Effluents COMITMENT Establish conditions for the controlled release of radioactive liquid effluents. Implement the requirements of 10 CFR 20,10 CFR 50.36a, Appendix A to 10 CFR 50, Appendix I to 10 CFR 50, 40 CFR 141 and 40 CFR 190.

P

a. Concentration The concentration of radioactive material released at anytime from the site boundary for liquid effluents to Unrestricted Areas [ denoted in Figure 2.1-4(a) of the Oconee Nuclear Station Updated Final Safety An:. lysis Report) shall be limited to 10 times the effluent concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2 for radionuclides other than dissolved or entrained noble gases. For dissolved or entrained noble gases the concentration shall be limited to 1 x 10 4 i Ci/mi total activity. l

b. Dose The dose or dose commitment to a Member Of The Public from radioactive materials in liquid effluents to Unrestricted Areas '

shall be limited to:

1. during any calendar quarter: '

s 4.5 mrem to the total body s 15 mrem to any organ, and;

2. during any calendar year:

s 9 mrem to the total body s 30 mrem to any organ.

c. Liquid Waste Treatment The appropriate subsystems of the liquid radwaste treatment system shall be used to reduce the radioactive materials in liquid waste prior to their discharge, if the projected dose due to liquid effluent releases to unrestricted areas, when averaged over 31 days would exceed 0.18 mrem to the total body or 0.6 mrem to any organ.

16.11.1-1 03/27/99

Radioactive Liquid Effluents 16.11.1

d. Chemical Treatment Ponds (CTP 1 and 2)

1. The quantity of radioactive material in the Chemical ,

Treatment Ponds (CTP) shall be limited so that, for all )

radionuclides identified, excluding noble gases and tritium, the sum of the ratios of activity (in curies) to the limits in 10 CFR 20, Appendix B, Table 2, column 2 shall not exceed 1.7 x 10 1

I Ai < 1.7 x 10' I j CJ Where Aj = pond inventory limit for single radionuclide "j" l (curies) l Cj = 10 CFR 20, Appendix B, Table 2, Column 2, concentration for single radionuclide "j" (curies)

2. No powdex resin shall be transferred to the CTPS unless the sum of the activity of the radionuclides identified is less than 0.1% of the limit identified in Commitment d.1.

< 1.0 x 103 where Qj = radionuclide activity in resin l

Aj = pond inventory limit for radionuclide "j"

3. The total radionuclide inventory of used powdex resin transferred to the Chemical Treatment Ponds over the previous 13 weeks, shall not excaed 0.4% of the pond I radionuclide inventory limit. Decay of radionuclides may be taken into account in determining inventory levels.

Oji + QJ2 + QJ3 + ------ + QJn s .004 x Aj where, Qj = Total inventory of radionuclide j in a transfer n - Nunber of transfers to the Chemical Treatment Ponds during the previous 13 - week period. l

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

Appendix I dose limits for radioactive liquid effluent releases are applicable only during normal operating conditions which include expected operational occurrences, and are not applicable during unusual operating conditions

( that result in activation of the Oconee Emergency Plan.

16.11.1-2 03/27/99

1 Radioactive Liquid Effluents 16.11.1 O '

APPLICABILITY: At all times I

I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of A.1 Restore concentration to Immediately radioactive material within the limit.

released in liquid effluents to Unrestricted Areas exceeds the limits specified-in Commitment a.

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16.11.1-3 03/27/99

Radioactive Liquid Effluents 16.11.1 i

CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose from B.1 ---------NOTE----------

the release of Not required during radioactive materials unusual operating in liquid effluents conditions that result exceeds any of the in activation of the limits in Commitment Oconee Emergency Plan,

b. -----------------------

Submit report to the 30 days from regional NRC Office the end of the which includes the quarter during following: which the release

a. Cause(s) for occurred exceeding the limit (s).

b. A description of the program of corrective action initiated to: reduce the releases of Os radioactive materials in liquid effluents, and to keep these levels of radioactive materials in liquid effluents in compliance with the above limits, or as low as reasonably achievable.

c. Results of radiological analyses of the drinking water source and the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141.

O 16.11.1-4 03/27/99

I Radioactive Liquid Effluents 16.11.1 CONDITION REQUIRED ACTION COMPLETION TIME C. Radiosctive liquid C.1 Submit report to the 30 days waste is discharged regional NRC Office without treatment and which includes the -

in excens of the following:

specified limit,

a. Cause of equipment or subsystem inoperability.

b. Corrective action to restore equipment and prevent recurrence.

D. Total radioactive D.1 Submit report to the 30 days i

inventory of used regional NRC Office f- s powdex resins describing the reason (s)

( ,,) transferred to the for exceeding the limit l Chemical Treatment and plans for future l Ponds over previous operation.

13 weeks greater than 0.4% of the pond radionuclide inventory limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

, SR 16.11.1.1 N/A N/A

-m k

16.11.1-5 03/27/99

I l

Radioactive Liquid Effluents 16.11.1 mm The concentration commitment is provided to ensure that the concentration of radioactive materials released in liquid waste effluents from the site to unrestricted areas will be less than 10 times the effluent concentration levels- specified in 10 CFR Part 20, Appendix B, Table 2, Column 2. The concentration limit for noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its EC in air (submersion) was converted to an equivalent concentration in water using the methods described in International. Commission on Radiological Protection (ICRP) Publication 2.

The basic' requirements for Selected Licensee Commitments concerning effluent 1 from nuclear power reactors are stated in 10 CFR 50.36a. Compliance with i effluent Selected Licensee Commitments will ensure that average annual j releases of radioactive material in effluents will be small percentages of the l limits specified in the old 10 CFR 20.106 (new 10 CFR 20.1302). The i requirements contained in 10 CFR 50.36a further indicate that operational i flexibility is allowed,. compatible with considerations of health and safety, I which may temporarily result in releases higher than such small percentages, but still within the limits specified in the old 10 CFR 20.106 which references Appendix B, Table II concentrations (MPCs). These referenced concentrations are specific values which relate to an annual dose of 500 mrem.

It is further indicated in 10 CFR 50.36a that when using operational 1 O flexibility, best efforts shall be exerted to keep levels of radioactive materials in. effluents as low as reasonably achievable (ALARA) as set forth in 10 CFR 50 Appendix 1. Also, for fresh water sites with drinking water supplies which can be potentially affected by plant operations, there is reasonable assurance that the operation of the facility will not result in radionuclide concentrations in the finished drinking water that are in excess of the requirements of 40 CFR 141. Therefore, to accommodate operational flexibility needed for effluent releases, the limits associated with this SLC are based on ten times the instantaneous dose rate value of 50 mrem / year to apply at all times. Compliance.with the limits of the new 10 CFR 20.1001 vill be demonstrated by operating within the limits of 10 CFR 50, A?pendix I, 40 CFR 141 and 40 CFR 190.

Section I of Appendix I of 10 CFR 50 states that this appendix provides specific numerical guides for design objectives and limiting conditions for operation, to assist holders of licmes for light water cooled nuclear power reactors in meeting the requirements to keep releases of radioactive material

~

to unrestricted areas as low as practical. and reasonably achievable, during normal reactor operations, including expected operational occurrences. Using the flexibility granted during unusual operating conditions, and the stated applicability of the design objectives for the Oconee Nuclear Station, Appendix I dose limits for radioactive liquid effluent releases are concluded

to be not applicable during unusual operating conditions that result in the activation of the Oconee Emergency Plan.

For units with shared radwaste treatment systems, the liquid effluents from j the shared system are proportioned among the units sharing that system.

16.11.1-6 03/27/99

~ . .

1 f

Radioactive Liquid Effluents 16.11.1 I p]

The requirements that the appropriate portions of this system be used when specified provides assurance that the releases of radioactive materials in ;

liquid effluents will be kept "as low as is reasonably achievable." This SLC implements the requirements of 10 CFR Part 50.36a. General Design Criterion 60 of Appendix A to 10 CFR Part 50 and design objective Section II.D of Appendix A to 10 CFR Part 50.

)

The inventory limits of the chemical treatment ponds are based on limiting the consequences of an uncontrolled release of the pond inventory. The short term rate limit (2 mrem /hr) of 10 CFR 20.1301 is applied to 10 CFR 20.1302 in the following expression:

A.i x 10' uCi x cal s 2 mrem /hr x 8760 hr 1.3 x 10 gal Curie 3785 ml 500 mrem /yr yr 10 x CJ 8,1 s 1.7 x.10e CJ Where Aj = pond inventory limit for radionuclide "j" (curies)

Cj = 10 CFR 20. Appendix B. Table 2, Column 2, concentration radionuclide "j" 1.3 x 10' gal = estimated volume of smaller chemical treatment pond The transfer limits provide assurance that activity input to the CTP will be minimized.

REFERENCES:

1. 10 CFR Part 20, Appendix B

2. 40 CFR Part 141

3. 10 CFR Part 50, Appendix A and I

4. 40 CFR Part 190

5. Offsite Dose Calculation Manual

6. Regulatory Guide 1.109 0 .

16.11.1-7 03/27/99

Radioactive Gasetus Effluents 16.11.2 16.11 RADIOLOGICAL EFFLUENTS' CONTROL 16.11.2 Radioactive Gaseous Effluents COMITMENT Establish conditions for the controlled release of radioactive gaseous effluents. Implement the requirements of 10 CFR 20,10 CFR 50.36a, Appendix A to 10 CFR 50, Appendix I to 10 CFR 50, and 40 CFR 190.

a. Dose Rate The instantaneous dose rate at the site (exclusion area) boundary for gaseous effluents [ Figure 2.1-4(a) of the Oconee Nuclear Station Updated Final Safety Analysis Report due to radioactive raterials released in gaseous effluents from the site shall be limited to the following values:

1. The dose rate limit for noble gases shall be:

]

s 500 mrem /yr to the total body I s 3000 mrem /yr to the skin and;

2. The dose rate limit for all radiciodines and for all radioactive materials in particulate form and radionuclides

, other than noble gases with half-lives greater than 8 days shall be s 1500 mrem /yr to any organ.

b. Dose

1. The air dose due to noble gases released in gaseous effluent i from the site shall be limited to the following: l

i. During any calendar quarter:

s 15 mrad for gamma radiation s 30 mrad for beta radiation ii. During any calendar year:

s 30 mrad for gamma radiation s 60 mrad for Deta radiation

2. The dose to a Member Of The Public from radioiodines, tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents released from the site, shall be limited to the following:

O v

16.11.2-1 03/27/99

A Radioactive Gaseous Effluents 16.11.2 O

i. During any calendar quarter:

, s 22.5 mrem to any organ ii. During any calendar year:

s 45 mrem to any organ,

c. Gaseous Radwaste Treatment

1. The Gaseous Radwaste Treatment System shall be used to 4

reduce the noble gases in gaseous wastes prior to their discharge, if the projected gaseous effluent air dose due to gaseous effluent release from the site, when averaged over 31 days exceeds 0.6 mrad for gamma radiation and 1.2 mrad for beta radiation.

2. The Ventilation Treatment. Exhaust System shall be used to reduce radioactive materials other than noble gases in gaseous waste prior to their discharge when the projected doses due to effluent releases to unrestricted areas when averaged over 31 days would exceed 0.9 mrem to any organ.

d. Used Oil Incineration During incineration of used oil contaminated by radioactive material in the Station Auxiliary Boiler, the dose to a Member Of The Public from radiciodines, tritium and radioactive materials in particulate form with half-lives greater than 8 days in gaseous effluents released from the Station Auxiliary Boiler shall be s 0.045 mrem to any organ in any calendar year.

NOTE--------------------------------- 1 The requirement of c.2 does not apply to the Auxiliary Building j Exhaust System since it is not " treated" prior to release. l APPLICABILITY: At all times O

G 16.11.2-2 03/27/99

Radioactive Gaseous Effluents 16.11.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Dose rate exceeds the A.1 Restore release rate to Immediately limits specified in within limits.

commitment a.

B. Calculated dose B.1 Submit report to the 30 days from exceeds specified regional NRC Office the end of the limits. which includes the quarter during following: which the release

a. Cause(s) for occurred exceeding the limit (s), and

b. A description of the i program of

' corrective action

( initiated to: reduce the releases of radioactive materials in gaseous effluents, and to keep these levels of radioactive

' materials in gaseous effluents in compliance with the specified limits or as low as reasonably achievable.

i 1

O 16.11.2-3 03/27/99

Radioactive Gaseous Effluents 16.11.2 CONDITION REQUIRED ACTION COMPLETION TIME C. Radioactive gaseous C.1 Submit a report to the 30 days waste is discharged regional NRC Office greater than limits which includes the specified in following:

Commitment c.1 or c.2. a. Cause of equipment or subsystems M inoperability, and Radioactive gaseous b. Corrective action to waste is discharged restore equipment without treatment for and prevent more than 31 days. recurrence.

j SURVEILLANCE REQUIREMENTS

, SURVEILLANCE FREQUENCY

O v

l

{ SR 16.11.2.1 N/A N/A i

I 4

4 N

i

!O 16.11.2-4 03/27/99

1 Radioactive Gaseous Effluents 16.11.2 I E The basic requirements for Selected Licensee Commitments concerning effluent from nuclear power reactors are stated in 10CFR50.36. Compliance with effluent Selected Licensee Commitments will ensure that average annual releases of radioactive material in effluents will be small percentages of the limits specified in the old 10CFR20.106 (new 10CFR20.1302). The requirements contained in'10CFR50.36a further indicate that operational flexibility is allowed, compatible with considerations of health and safety, which may tem >orarily result in releases higher than such small percentages, but still wittin the limits specified in the old 10CFR20.106 which references Appendix

8. Table II concentrations (MPCs). These referenced concentrations are specific values which relate to an annual dose of 500 mrem to the total body,

3000 mrom to the skin, and 1500 mrem to an infant via the milk animal-milk-infant pathway. It is further indicated in 10CFR50.36a that when using operational flexibility, best efforts shall be exerted to keep levels of radioactive materials in effluents as low is reasonably achievable (ALARA) as set forth in 10CFR50 Appendix 1. Therefore, to accommodate operational flexibility needed for effluent releases, the limits associated with gaseous release rate SLCs will be maintained at the current instantaneous dose rate limit for noble gases of 500 mrem / year to the total body and 3000 mrem / year to the skin; and for Iodine-131, for Iodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days. an

/N instantaneous-dose rate limit of 1500 mrem / year.

d The ODCM calculational methods for calculating the doses due to the actual release rates of the subject materials will be consistent with the methodology provided in Regulatory Guide 1.109, " Calculating of Annual Doses to man from i Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance I with 10 CFR Part 50, Appendix I, " Revision I. October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of i Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors."

Equations in the ODCM are provided for determining the actual doses based upon the historical average atmospheric conditions. The release rate commitments for radioiodines, radioactive material in particulate form and radionuclides other than noble gases are dependent on the existing radionuclide pathways to man, in the unrestricted area. The pathways which are examined in the development Of these calculations are: 1) individual inhalation of airborne radionuclides, 2) deposition of radionuclides into green leafy vegetation with subsequent consumption by man, 3) deposition onto grassy areas where milk j animals and meat producing animals graze with consumption of the milk and meat

. by man, and 4) deposition on the ground with subsequent evposure of man.

The requirement that the appropriate portions of these systems be used when specified provides reasonable assurance that the release of radioactive !

materials in gaseous effluents will be kept "as low as is reasonably achievable." This commitment implements the requirements of 10 CFR Part i 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50, and design objective Section IID of Appendix I to 10 CFR Part 50.

16.11.2-5 03/27/99

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

1 Radioactive Gaseous Effluents I 16.

11.2 REFERENCES

1 10 CFR Part 20. Appendix 8

2. 10 CFR Part 50. Appendix A and I

3. Regulatory Guide 1.109

4. 40 CFR Part 190

5. Offsite Dose Calculation Manual O

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16.11.2-6 03/27/99 1

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l Radioactive Effluent Monitoring Instrumentation 16.11.3 i 16.11 RADIOLOGICAL EFFLUENTS CONTROL 16.11.3 Radioactive Effluent Monitoring Instrumentation COMITNENT Radioactive Effluent Monitoring Instrumentation shall be OPERABLE as follows:

a. Liquid Effluents The radioactive liquid effluent monitoring instrumentation i channels shown in Table 16.11.3-1 shall be operable with l their alarm / trip setpoints set to ensure that the limits of '

SLC 16.11.1.a are not exceeded. I

b. Gaseous Process and Effluents The radioactive gareous process and effluent monitoring l instrumentation channels shown in Table 16.11.3-2 shall be operable with their alarm / trip setpoints set to ensure that the limits of SLC 16.11.2.a are not exceeded. .

1

c. The setpoints shall be determined in accordance with the '

methodology described in the ODCM and shall be recorded.

......---NOTE-------------------------------

Correction to setpoints determined in accordance with Commitment c may be permitted without declaring the channel inoperable. ;

j APPLICABILITY: According to Table 16.11.3-1 and Table 16.11.3-2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Alarm / trip setpoint A.1 Declare channel Immediately less conservative than ' inoperable.

required for one or more Effluent OR monitoring instrument channels. A.2 Suspend release of Immediately effluent monitored by the channe'. .

16.11.3-1 03/27/99

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

Radioactive Effluent Monitoring Instrumentation 16.11.3 C. CONDITION REQUIRED ACTION COMPLETION TIME B. One or more required B.1 Enter the Condition Immediately l Liquid Effluent referenced in Table monitoring instrument 16.11.3-1 for the channels inoperable. function.

AND B.2 Restore the 30 davs instrument (s) to OPERABLE status. I 1

C. One or more required C.1 Enter the Condition Immediately Gaseous Effluent referenced in Table monitoring instrument 16.11.3-2 for the channels inoperable. function.

AND C.2 Restore the 30 days instrument (s) to O- OPERABLE status.

l I

D. Required Action and D.) Explain in next Annual April 30 of associated Completion Radiological Effluent following '

Time of Required Release Report why calendar year Action B.2 or C.2 not inoperability was not met. corrected in a timely manner.

O 16.11.3-2 03/27/99 i

i , , . _ ,

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

Radioactive Effluent Monitoring Instrumentation n 16.11.3 U

CONDITION REQUIRED ACTION COMPLETION TIME i E. As required by E.1.1 Analyze two Prior to i Required' Action B.1 independent samples in initiating !

and referenced in accordance with subsequent i Table 16.11.3-1. SLC 16.11.4. release l (RIA-33)

AND E.1.2 Conduct two Prior to independent data entry initiating checks for release subsequent rate calculations release AND E.1.3 Conduct two Prior to independent valve initiating lineups of the subsequent effluent pathway. release O =

E.2 Suspend release of Immediately 1

radioactive effluents by this pathway.

F. As required by F.1 Suspend release of Immediately Required Action B.1 radioactive effluents and referenced in by this pathway.

Table 15 11.3-1.

(RIA-5s, E F.2 Collect and analyze Prior to each grab samples for gross discrete radioactivity (beta release of the and/or gamma) at a sump lower limit of detection 4

of at least 10 pCi/ml.

O 16.11.3-3 03/27/99

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

l Radioactive Effluent Monitoring Instrumentation 16.11.3 l

CONDITION REQUIRED ACTION COMPLETION TIME G. As required by ------------NOTE-------------

Required Action B.1 Not required during short, and referenced in controlled outages of liquid Table 16.11.3 . . effluent monitoring (Liquid Radwasu instrumentation. Short Effluent Line Flow controlled outages arc Rate Monitor) defined as planned removals from service for durations not to exceed I hour, for purposes of sample filter changeouts, setpoint adjustments, service checks, and/or routine maintenance procedures. This guidan e may be applied successively, provided that time between

' successive short, controlled outages is always at least equal to duration of

! !.(_b!!! .$!!!_

G.1 Suspend release of Immediately radioactive effluents by this pathway.

OB G.2 Estimate flow rate Immediately during actual releases. AND Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter l

O 16.11.3-4 03/27/99

l Radioactive Effluent Monitoring Instrumentation l- 16.11.3 l

l CONDITION REQUIRED ACTION COMPLETION TIME l

l H. As required by ------------NOTE------------- 1 i

Required Action B.1 Not required during short, I and referenced in controlled outages of liquid Table 16.11.3-1. effluent monitoring (RIA-35,' #3 Chemical instrumentation. Short Treatment Pond controlled outages are Composite Sampler and defined as planed removals Sampler Flow Monitor from service for durations (Turbine Building not to exceed I hour, for Sumps Effluent)) purposes of sample filter changeouts, setpoint adjustments. service checks, and/or routine maintenance procedures. This guidance may be applied successively, provided that time between successive short, controlled

, outages is always at least l equal to duration of

. . _ $$$ _S!_$ !I I

H.1 Suspend release of Immediately radioactive effluents by this pathway.

08 H.2 Collected and analyze Immediately grab samples for gross radioactivity (beta AND and/or gamma) at a lower limit of Once per 12 detection of at least hours 10-7 pCi/ml. thereafter I

l O.

e U

, 16.11.3-5 03/27/99 L ,

l Radioactive Effluent Monitoring Instrumentation q 16.11.3 b

CONDITION REQUIRED ACTION COMPLETION TIME I. As required by ------------NOTE-------------

Required Action C.1 Not required during short, and referenced in controlled outages of gaseous Table 16.11.3-2 for effluent monitoring effluent releases from instrumentation. Short waste gas tanks controlled outages are (RIA-37, RIA-38) or defined as planned removals containment purges from service for durations (RIA-45). not to exceed I hour, for purposes of sample filter changeouts, setpoint adjustments. service checks, and/or routine maintenance procedures. This guidance may be applied successively, provided that time between successive short, controlled outages is always at least equal to duration of

') immediately preceding outage.

I.1.1 Analyze two Prior to independent samples. initiating subsequent release AND I.1.2 Conduct two Prior to independent data entry initiating checks for release subsequent rate calculations release AND I.1.3 Conduct two Prior to independent valve initiating lineups of the subsequent effluent pathway. release 93 1.2 Suspend release of Immediately radioactive effluents by this pathway.

O 16.11.3-6 03/27/99

k

~

Radioactive Effluent Monitoring Instrumentation

. 16.11.3 CONDITION REQUIRED ACTION COMPLETION TIME l

J. As required by ----------NOTE--------------- !

Required Action C.1 Not required during short, and referenced in controlled outages of gaseous Table 16.11.3-2. effluent monitoring (Effluent Flow Rate instrumentation. Short Monitor (Unit Vent , controlled outages are Containment Purge, defined as planned removals Interin Radwaste from service for durations Exhaust, Hot Machine not to exceed 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , for Shop Exhaust, Radwaste purposes of sample filter Facility Exhaust, changeouts, setpoint Waste Gas Discharge )) adjustments. service checks, and/or routine maintenance procedures. This guidance may ,

be applied successively, provided that time between successive short, controlled outages is always at least i equal to duration of I immediately preceding outage.

( ...................... l J.1 Suspend release of Immediately radioactive effluents by this pathway.

9.3 J.2 Estimate flow rate Immediately AND Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months thereafter O

16.11.3-7 03/27/99

Radioactive Effluent Monitoring Instrumentation 16.11.3 l

! CONDITION REQUIRED ACTION COMPLETION TIME I

K. As required by ------------NOTE-------------

Required Action C.1 Not required during short, and referenced in controlled outages of gaseous Table 16.11.3-2. effluent monitoring (4RIA-45,RIA-53) instrumentation. Short controlled outages are defined as planned removals ;

from service for durations l not to exceed I hour, for purposes of sample filter <

changeouts, setpoint adjustments. service checks, and/or routine maintenance procedures. This guidance may be applied successively, provided that time between l successive short, controlled outages is. always at least equal to duration of O immediately preceding outage.

K.1 Suspend release of Immediately radioactive effluents by this pathway.

0R K.2.1 Collect grab sample. Immediately AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months AND K.2.2 Analyze grab samples 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from for gross activity collection of (beta and/or gamma). sample O

16.11.3-8 03/27/99

Radioactive Effluent Monitoring Instrumentation 16.11.3

, CONDITION REQUIRED ACTION COMPLETION TIME I

L. As required by ------------NOTE-------------

Re. quired Action C.1 Not required during short, j and referenced in controlled outages of gaseous Table 16.11.3-2. effluent monitoring (Unit Vent Monitoring instrumentation. Short

Iodine Sampler, Unit controlled outages are 1 Vent Monitoring defined as planned removals l Particulate Sampler, from service for durations Interim Radwate not to exceed I hour, for

Building Ventilation purposes of sample filter i Monitoring Iodine changeouts, setpoint

. Sampler, Interim adjustments. service checks,

, Radwate Building and/or routine maintenance i Ventilation Monitoring procedures. This guidance may l Particulate Sampler, be applied successively,

Hot Machine Shop provided that time between i Iodine Sampler, Hot successive short, controlled 2

Machine Shop outages is always at least i Particulate Sampler, equal to duration of

Radwaste Facility immediately preceding outage.

j -

Iodine Sampler, -----------------------------

i Radwaste Facility

! Particulate Sampler L.1 Suspend release of Immediately l radioactive effluents l by this pathway.

OR L.2.1 ---------NOTES--------

The collection time of j

each sample shall not j exceed 7 days.

l Collect samples Immediately i continuously using auxiliary sampling i equipment.

1 AND L.2.2 Analyze each sample. 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from end of each sample collection

O j

16.11.3-9 03/27/99

l Radioactive Effluent Monitoring Instrumentation 16.11.3 CONDITION REQUIRED ACTION COMPLETION TIME M. As required by ------------NOTE-------------

Required Action C.1 Not required durir.g short, and referenced in controlled outages of gaseous ,

Table 16.11.3-2 for effluent monitoring effluent releases from instrumentation. Short ventilation system or controlled outages arc condensor air defined as planned removals ejectors. (RIA-40) from service for durations 1 not to exceed I hour, for -

purposes of sample filter changeouts, setpoint adjustments. service checks, and/or routine maintenance procedures. This guidance may be applied successively, provided that time between successive short, controlled outages is always at least equal to duration of

.5! ! $.f. $$$ !.S$ I!!.

M.1 Continuously monitor Immediately release through the unit vent.

M M.2 Suspend release of Immediately radioactive effluents .

l by this pathway.

l M

M.3.1 Collect grab sample. Immcdiately AND '

Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months AND M.3.2 Analyze grab sample 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from for gross activity collection of O

D (beta and/or gamma). grab sample 16.11.3-10 03/27/99

Radicactive Effluent Monitoring Instrumentation 16.11.3 1

l' SURVEILLANCE REQUIREMENTS I l

SURVEILLANCE FREQUENCY 1

i SR 16.11.3.1 ------------------NOTE--------------------- l The Channel Response check shall consist of l verifying indications during periods of release. Channel response checks shall be made at least once per calendar day on days in which continuous, periodic or batch i releases are made. '

Perform Channel Response Check. During each release via this pathway SR 16.11.3.2 ------------------NOTE--------------------- 1 The Channel. Response check shall consist of g verifying indications during periods of Pg release. Channel response checks shall be made at least once per calendar day on days in which continuous, periodic or batch releases are made.

Perform Channel Response Check. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months l

l l SR 16.11.3.3 Perform Source Check, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months SR 16.11.3.4 Perform Source Check. 31 days l

l l

SR 16.11.3.5 Perform Source Check. 92 days O

16.11.3-11 03/27/99

l 1

i Radioactive Effluent Monitoring Instrumentation l

O 16.11.3 SURVEILLANCE FREQUENCY SR 16.11.3.6 ------------------NOTE---------------------

The Channel Functional Test shall also demonstrate that automatic isolation of this pathway and control room annunciation occurs if any of the following conditions exist:

1. Instrument indicates measured levels above the alarm / trip setpoint.

2. Circuit failure (downscale only).

Perform CHANNEL FUNCTIONAL TEST. 92 days' SR 16.11.3.7 ------------------NOTE---------------------

The Channel Functional Test shall also O demonstrate that control room annunciation occurs if any of the following conditions exist:

1. Instrument indicates measured levels above the alarm / trip setpoint.

2. Circuit failure (downscale only).

Perform CHANNEL FUNCTIONAL TEST. 9? days SR 16.11.3.8 Perform CHANNEL FUNCTIONAL TEST. 92 days lO 16.11.3-12 03/27/99

l Radioactive Effluent Monitoring Instrumentation 16.11.3 Ot .

SURVEILLANCE FREQUENCY SR 16.11.3.9 ------------------NOTE---------------------

The initial Channel Calibration shall be

, performed using one or more of the reference standards certified by the National Bureau of Standards or using standards that have been obtained from suppliers that participate in measurement assurance activities with the National Institute of Standards and Technology I (NIST). The standards shall permit !

calibrating the system over its intended I range of energy and measurement, For subsequent Channel Calibration sources that have been related to the initial calibration shall be used. (Operating plants may substitute previously established calibration procedures for these requirements.) ;

Perform CHANNEL CALIBRATION. 12 months SR 16.11.3.10 Perform CHANNEL CALIBRATION. 12 months SR 16.11.3.11 Perform leak test. When cylinder gates or wicket gates are reworked SR 16.11.3.12 Perform Source Check. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to each release via associated pathway 16.11.3 13 03/27/99 e m a e , , -

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

Radiological Effluent Monitoring Instrumentation O Table 16.11.3-1

( LIQUID EFFLUENT MONITORING INSTRUMENTATION OPERATING CONDITIONS AND SURVE1LLANCE REQUIREMENTS CONDITION MINIMUM REFERENCED OPERABLE SURVEILLANCE FROM REQUIRED INSTRLMENT CHANNELS APPLICABILITY REQUIREMENTS ACTION B.1

1. Monitors Providing Automatic Termination of Release

a. Liquid Radweste 1 At all times SR 16.11.3.1 E Effluent Line SR 16.11.3.3 Monitor, RIA-33 SR 16.11.3.6 SR 16.11.3.9

b. Turbine Butiding 1 At all times SR 16.11.3.2 F Sump, RIA-54 SR 16.11.3.4 SR 16.11.3.7 SR 16.11.3.9

2. Monitors not Providing Automatic Termination of Release Low Pressure Service 1 At all times SR 16.11.3.2 M Water RIA-35 SR 16.11.3.4 SR 16.11.3.7 SR 16.11.3.9

3. Flow Rate Measuring Devices

a. Liquid Rachvaste 1 At all times SR 16.11.3.1 G Effluent Line Flow SR 16.11.3.10 Rate Monitor (0LVCR0725or OLW 550920)

b. Liquid Radwaste NA NA SR 16.11.3.1 NA Effluent Line SR 16.11.3.10 Mintrun Flow Device

c. Turbine Building NA NA SR 16.11.3.1 NA Sump Minimum Flow SR 16.11.3.13 Device

d. Low Pressure Service NA NA SR 16.11.3.1 NA Water Minimum Flow SR 16.11.3.10 Device l

%/

16.11.3-14 03/27/99

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

Radiological Effluent Monitoring Instrumentation 16.11.3 Table 16.11.3-1 LIQUID EFFLUENT MON!TORING INSTRUMENTATION OPERATING CONDITIONS AND SURVEILLANCE PrQUIREMENTS Y/ . .

CONDITION MINIMUM REFERENCED OPERABLE SURVEILLANCE .FRDM REQUIRED INSTRUMENT CHANNELS APPLICABILITY REQUIREMENTS ACTION B.1

e. Keouse Hydroelectric NA NA SR 16.11.3.11 NA Tallrace Discharge "'

4 Continuous Composite Sampler f3 Chemical Treatment 1 At all times SR 16.11.3.2 H I Pond Composite Sampler SR 16.11.3.10 and Sampler Flow Mut.itor (Turbine Building Sunps i

Effluent)

(a) Flow is determined from the number of hydro units operating. If no hydro units are operating, leakage flow will be assumed to be 38 cfs based on historical data.

O V

[

16.11.3-15 03/27/99

_ . _ . _.._. _ _ _ _ . _ . _ . . _ _ _ . _ . . __ _ - . . . _ _ _ _ _ . _ . . . _ . ,__m .m_. _ _ _ _ _ _

Radir.ogical Effluent Monitoring Instrumentation r

f 16.11.3 Table 16.11.3-2 GASEOUS EFFLUENT MONITORING INSTRUMENTATION-OPERATING CONDITIONS AND SURVE!LLANCE REQUIREMENTS MINIMUM CONDITION OPERABLE REFERENCED CHANNELS (PER SURVEILLANCE FROM REQUIRED INSTRtMENT RELEASE PATH) APPLICABILITY REQUIREMENTS ACTION C.1

1. Unit Vent Monitoring System

a. Noble Gas Activity 1 At All Times SR 16.11.3.2 1 Monitor Providing SR 16.11.3.4 Alam and Automatic SR 16.11.3.7 Temination of SR 16.11.3.9 Contalrunent Purge Release (RIA Purge Isolation Function)

b. Noble Gas Activity 1 At all times SR 16.11.3.2 K Monitor Providing SR 16.11.3.4 Alam. (RIA SR 16.11.3.7 Vent Stack Monitor SR 16.11.3.9 Function)

c. lodine Sampler 1 At All Times SR 16.11.3.2 L

d. Particulate Sampler 1 At All Times SR 16.11.3.2 L

e. Effluent Flow Rate 1 At All Times SR 16.11.3.2 J Monitor (Unit Vent SR 16.11.3.10 Flow) (GWD CR0037) j

f. Sampler Flow Rate 1 At All Times SR 16.11.3.2 NA Monitor *** l SR 16.11.3.10 '

(Annunciator) g .' Effluent Flow Rate 1 During Containment SR 16.11.3.2 J Monitor (Containment Purge Operation SR 16.11.3.10 l Purge) (PR CR0082) '

h. CSAE Off Gas Monitor 1 During Operation SR 16.11.3.2 M (RIA-40) of CSAE SR 16.11.3.5 SR 16.11.3.8 SR 16.11.3.9

2. Interim Radweste Building Ventilation Monitoring System

a. Noble 6as Activity 1 At All Times SR 16.11.3.2 K Monitor (RIA - $3) SR 16.11.3.4 SR 16.11.3.7 I SR 16.11.3.9 l

b. lodine Sampler 1 At All Times SR 16.11.3.2 L

c. Particulate Sampler- 1 At All Times SR 16.11.3.2 L

d. Effluent Flow Rate 1 At All Times SR 16.11.3.2 J j Monitor (Interim SR 16.11.3.10 Radweste Exhaust) '

(GWO FT0082) l 16.I1.3-16 03/27/99 1

4 - - - - -- - ,.,-na u,

. . _ _ . ._ __ _, m. _ __ . _ - __. _ . . _ . . _

p !

I Radiological Effluent Monitoring Instrumentation 16.11.3 i Table 16.11.3-2 l

Q GASEOUS EFFLUENT MONITORING INSTRUMENTATION {

i OPERATING CONDITIONS AND SURVElt. LANCE REQUIREMENTS l MINIMUM CONDITION l OPERABLE REFERENCED CHANNELS (PER SURVE!LLANCE FROM REQUIRED INSTRUMENT RELEASE PATH) APPLICABILITY REQUIREMENTS ACTION C.1

e. Sampler Flow Rate 1 At All Times SR 16.11.3.2 NA Monitor" SR 16.11.3.10 (Annunciator)

3. Hot Machine Shop Ventilation Sampiing System

a. todine Sampler 1 At All Times SR 16.11.3.2 L I 1

b. Particulate Sampler 1 At All Times SR 16.11.3.2 L i

c. Effluent Flow Rate 1 At All Times SR 16.11.3.2 J Monitor (Hot Machine SR 16.11.3.10 ShopExhaust)

(Totalizer) l

d. Sampler Flow Rate 1 At All Times SR 16.11.3.2 NA Monitor

SR 16.11.3.10 (Annunciator)

4. Radwaste Facility p l Ventilation Monitoring

( System

a. Noble Gas Activity 1 At All Times SR 16.11.3.2 K Monitor (4-RIA-45) SR 16.11.3.4 SR 16.11.3.7 SR 16.11.3.9

b. lodine Sampler 1 At All Times SR 16.11.3.2 L

c. Particulate Sampler 1 At All Times SR 16.11.3.2 L

d. Effluent Flow Rate 1 At All Times SR 16.11.3.2 J Monitor (Radwaste SR 16.11.3.10 {

Facility Exhaust) '

(0VSCR2060)

e. Sampler Flow Rate 1 At All Times SR 16.11.3.2 NA Monitor

SR 16.11.3.10 (Annunciator)

5. Waste Gas Holdup Tanks

a. Noble Gas Activity 1 During Waste Gas SR 16.11.3.1 I Monitor - Providing Holdup Tank Releases SR 16.11.3.6 Alarm and Automatic SR 16.11.3.9 Termination of SR 16.11.3.12 Release (RIA-37.-

38)*

l b. Effluent Flow Rate 1 During Waste Gas SR 16.11.3.1 J l Monitor (WasteGas Holdup Tank Releases SR 16.11.3.10 l l Olscharge Flow) (GVD i

! CR033) p

, k)Alarmsindicatinglowflowmaybesubstitutedforflow .. .suring devices.

l (b)Either Normal or High Range monitor is required deper 4r. upon activity in tank being released.

l l

1 16.11.3-17 03/27/99 l

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

Radiological Effluent Monitoring Instrumentation l 16.11.3 EMLE.1 The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or potential releases. The alarm / trip.setpoints for these instruments shall be calculated in accordance with NRC approved methods in the 00CM to assure that the alarm / trip will occur prior to exceeding 10 times the limits of 10 CFR Part 20. The operability and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

The radioactive gaseous effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in gaseous effluents during actual or potential releases. The alarm / trip setpoints for these instruments shall be calculated in accordance with NRC approved methods in the ODCM to assure that the alarm / trip will occur prior to exceeding applicable dose limits in SLC 16.11.2. The operability end use of this instrumentation is consistent with the requirements of General Design criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50.

For certain applicable cases, grab samples or flow estimates are required at frequencies between every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months end every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months upon RIA removal from service. SLC 16.11.3 does not explicitly require Action (grab samples or O flow estimates) to be initiated immediately upon RIA removal. from service, V when removal is for the purposes of sample filter changeouts, setpoint adjustments, service checks, or routine maintenance. Therefore, during the defined short, controlled outages, Action is not required.

For the cases in which Action is defined as continuous sampling by auxiliary equipment (Action L) initiation of continuous sampling by auxiliary sampling equipment requires approximately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . One hour is the accepted reasonable time to initiate collect and change samples. Therefore, for the defined short, controlled outages (not to exceed I hour), Action is not required.

Failures such as blown instrument fuses, defective indicators, and faulted amplifiers are, in many cases, revealed by alarm or annunciator action.

Comparison of output and/or state of independent channels measuring the same variable supplements this type of built-in surveillance. Based on experience in operation of both conventional and nuclear systems, when the unit is in operation. the minimum checking frequency stated is deemed adequate.

EEERENCES:

1. 10 CFR Part 20

2. 10 CFR Part 50. Appendix A l 3. Offsite Dose Calculation Manual

! 4. UFSAR, Section 7.2.3.4 O

l l 16.11.3-18 03/27/99

Operational Safety Review 16.11.4 16.11 RADIOLOGICAL EFFLUENTS CONTR0;.

16.11.4 Operational Safety Review ComITMENT' Required sampling should be performed as detailed in Table 16.11.4-1.

APPLICABILITY: At all times AC7 IONS CONDITION REQUIRED ACTION COMPLETION TIME A. NA A! NA NA 1

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.11.4.1 N/A N/A l

i I

I O

16.11.4-1 03/27/99

1 Operational Safety Review l p 16.11.4

) ' Table 16.11.4-1 l

i Minimum Sampling Frequency and Analysis Program Lower Limit of Detection (b) of Lab Analysis Item Check Frequency for Waste

1. Condensate Test a. Principal Ganna Composite Grab Sample <5E-06 pCl/ml (Ce-144 and Mo-99)

Tank. Condensate Emitters (c) prior to release of <5E-07 pC1/ml (Other Ganna huclioes)

Monitoring Tank. including each batch (h) <1E-05 pCi/ml (Dissolved Gases)

Laundry-Hot Dissolved Noble <1E-06 pC1/ml (1-131) shower Tank. Gases Weste and Recycle Monitor Tanks

b. Radiochemical Quarterly from all <5E-08 pCl/ml Analysis Sr-89 composited batches (f) and Sr-90

c. Tritium Monthly Composite <1E 35 pCi/mi

d. Gross Alpha Monthly Composite <1E-07 pC1/ml Activity

2. Unit Vent a. Iodine Continuous monitor. <1E-10 pCl/cc (I-133)

Sampilng Spectrum (a) weekly sample (e) <1E-12 pCi/cc (1-131)

(IncludesWaste t

~

Gas Decay Tanks. '

Reactor Building Purges, b. Particulates(a)

Auxiliary Building 1. Ce-144 & Mo-99 Weekly Composite (e) <5E-09 pCi/cc Ventilation.

Spent Fuel Pool Ventilation. Air Ejectors)

11. Other Principle Weekly Composite (e) <1E-10 pC1/cc Ganna Emitters (d) 111. Gross Alpha Monthly, using <!E-11 pct /cc Activity composite samples of one week tv. Radiochemical Quarterly Composite <!E-11 pct /cc Analysis Sr-89

$r-90

c. Gases by Weekly Grab Sample <1E-04 pCi/cc Principle Gamma Emitters (d)

d. Tritium Weekly Grab Sample <1E-06 pCi/cc

3. Weste Gas Decay a. Principle Ganna Grab Sample prior to <1E-04 pCi/cc (gases)

Tank Emitters (d) release of each batch <1E-10 pCi/cc (particulates and S todines) l D) b. Tritium Grab Sample prior to release of each batch

<1E-06 pCi/cc l

16.11.4-2 03/27/99

Operational Safety Review 16.11.4 O Table 16.11.4-1 Minimum Sampling Frequency and Analysis Program I

l Lower Limit of Detection (b) of Lab Analysis item Check Frequency for Waste 3

4. Reactor Building. a. Principle Gama Grab sample each purge <1E-04 pC1/cc (gases)

Emitters (d) <1E-10 pC1/cc (particulates and iodines) -

b. Tritium Grab sample each purge <1E-06 pCi/cc l S. Backwash Principle Gama Grab Sample prior to NA ,

Receiving Tanks Emitters including release of each batch dissolved Noble G4ses

6. f3 Chemical a. Principle Gama Weekly Continuous <5E-07 pC1/ml Treatment Pond Emitters (c) Composite (g)

Ef fluent'"

b. 1-131 Weekly Continuous <!E-06 pCi/ml Composite (g)

c. Irltium Monthly Continuous <1E-05 pct /ml Composite (g)

d. Gross Alpha. Monthly Continuous <!E-07 pCi/ml Activity Composite (g) i

e. Sr-89 & Sr-90 Quarterly Continuous

-} Composite (g)

<5E-08 Ci/mi

f. Disolved and Monthly Grab <1E-05 C1/ml Entrained gases (Game Emitters)

7. Radweste a. Iodine Continuous monitor, (1-133) <1E-09 pC1/cc

, Facility Spectrum (a) weekly sample (e) (I-131) <1E-11 pCi/cc Ventilation

b. Particulate (a)

1. Ce-144 and Mo- Weekly Composite (e) <5E-09 pCi/cc 99

11. Other Principle Weekly Composite (e) <1E-10 pct /cc Gamma Emitters (d) 111. Gross Alpha Monthly, using <!E-11 pCi/cc Activity composite samples of one week tv. Radiochemical Quarterly Composite <1E-11 pct /cc Analysis Sr-89,

$r-90

c. Gases by Weekly Grab Sample <1E-04 pC1/cc Principle Gama(d)

Emitters

, d. Tritium Weekly Grab Sample <!E-06 pct /cc 6~dvag 16.11.4-3 03/27/99

Operational Safety Review

(~~T 16.11.4 V Table 16.11.4-1 Minimum Sampling Franuency and Analysis Program (a) Samples shall be changed at least once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analysis sh.ll be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing (on or after removal from sampler).

(b) The LLD is defined for purposes of these comitments as the smallest concentration of radioactive material in a sample that would be detected with 95% probability with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation) .

LLD = 4.66 sb E x V x 2.22E06 x Y x exp (46 t)

Where:

LLD is the "a priori" lower limit of detection as defined above (as micro Curles per unit mass or volume),

sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (ascountsperminute).

E is the counting efficiency (as counts per disintegration),

V is the sample size (in units of mass or volume).

2.22E06 is the number of disintegrations per minute per micro Curie, f Y is the fractional radiochemical yield (when applicable),

A is the radioactive decay constant for the particular nuclide A t is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples,). NOTE: This assumes decay correction is applied (at the time of analysis) for the duration of sample collection, for the time between collection and analysis, and for the duration of the counting. Additionally, it does not apply to isolated systems such as Waste Gas Decay Tanks and Waste Monitor Tanks.

Typical values of E. V, Y and A t should be used in the calculation.

It should be recognized that the LLD is an a priori (before the fact) limit representing the capability of a measurement system and not an a costeriori (after the fact) limit for a particular measurement.

(c) The principal gama emitters for which the LLD control applies include the following radionuclides: Mn-54. e-59, Co-58 Co-60. Zn-65. No-99, Cs-134 Cs-137. and Ce-141. Ce-144 shall also be meassured, but with a LLD of SE-06 C1/ml. This list does not mean that only these nuclides are to be considered. Other gama peaks that are identifiable. together with the above nuclides shall also be analyzed and reported in the Annual Radioactive Effluent Release Report.

(d) The principal ganma emitters for which the LLD comitment appites exclusively are the following radionuclides: Kr-87.

Kr-88, Xe-133. Xe-133m, Xe-135. and Xe-138 for gaseous emissions and Mn-54, Fe-59. Co-58, Co-60, Zn-65. Mo-99, Cs-134 Cs-137, Ce-141, and Ce-144 for particulates. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides shall also be identified and reported.

(e) The ratio of the sample flow rate to the sampled stream flow rate shall be known for the time period covered by each dose or dose rate calculation made in accordance with SLC 16.11.2.a. SLC 16.11.2.b.1, SLC 16.11.2.b.2.

(f) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids r i released.

(g) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously in proportion to the rate of low of the effluent stream. Prior to analysis, all samples taken for the composite shall be thoroughly mixed in order for the composite sample to be representative of the effluent release.

16,11.4-4 03/27/99

Operational Safety Review 16.11.4 i

Table 16.11.4-1 Minimum Sampling Frequency and Analysis Program (h) A 'atch release is the discharge of liquid wastes cf a discrete volume. Prior to sampling for analysis, each batch i sMll be isolated, and then thoroughly mixed, to assure representative sampling.

(1) A conteuous release is the discharge of liquid wastes of a non-discrete volume, e.g., from a volume of a system that has an in M ilow durig the continuous release.

4 e

4 d

1 4

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16.11.4-5 03/27/99

Operational Safety Review 16.11,4 BASES N/A

REFERENCES:

N/A 1

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16.11.4-6 03/27/99

Solid Radioactive Waste 16.11.5 16.11 RADIOLOGICAL EFFLUENTS CONTROL 16.11.5 Solid Radioactive Waste i

C0fMITMENT The Solid Radwaste System shall be used in accordance with a Process Control Program, for the solidification of wet radioactive .

wastes. Prior to the shipment of containers of radioactive wastes l from the site, radioactive wastes shall be processed and packaged ;

to ensure meeting the requirements of 10 CFR Part 20,10 CFR Part '

71, and Federal and State regulations governing the disposal of radioactive wastes.

The PROCESS CONTROL PROGRAM (PCP) shall contain the current 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 10 CFR Parts 20, 61, and 71, State regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

The Process Control Program shall be used to verify the Solidification of at least one representative test specimens from at least every tenth batch of each type of wet radioactive waste O to be solidified. l APPLICABILITY: At all times ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of 10CFR A.1 Suspend shipments of Immediately Part 20 are not defectively packaged '

satisfied. solid radioactive wastes from the site.

E Requirements of 10CFR Part 71 are not satisfied.

O 16.11.5-1 03/27/99

l Solid Radioactive Waste g3 16.11.5 G'

CONDITION REQUIRED ACTION COMPLETION TIME B. Any test specimen B.1 Suspend solidification Immediately fails to verify of the batch under Solidification. test until such time as additional test specimens can be obtained, alternative Solidification l parameters can be '

determined in accordance with the l Process Control l Program, and a .

subsequent test l verifies 1 Solidification. i Solidification of the i batch may then be J resumed using the alternative

' Solidification ,

parameters determined l by the Process Control Program.

l 3

C. Initial test specimen C.1 Process Control NA l from a batch of waste Program shall provide fails to verify for the collection and Solidification. testing of representative test specimens from each consecutive batch of the same type of wet waste until at least 3 consecutive initial j

test specimens I demonstrate l Solidification. The l Process Control l Program shall be modified as required to assure Solidification of

, subsequent batches of waste.

16.11.5-2 03/27/99

Solid Radioactive Waste 16.11.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.11.5.1 NA NA

, E8}U The solid radwaste system will be used whenever radwastes require processing and packaging prior to being shipped offsite. This commitment implements the requirements of 10 CFR Part 50.36a and General Design Criterion 60 of Appendix A to 10 CFR Part 50. The process parameters included in establishing the

, Process Control Program may include, but are not limited to waste type, waste d

pH, waste / liquid / solidification agent / catalyst ratios, waste oil content,

' waste principal chemical constituents, mixing and curing times.

(V3

REFERENCES:

1. 10 CFR Part 50, Appendix A

2. Process Control Program Manual iO 16.11.5-3 03/27/99

,w -

i

) Radiological Environmental Monitoring l 16.11.6 1

16.11 RADIOLOGICAL EFFLUENTS CONTROL J

16.11.6 Radiological Environmental Monitoring

! ColWITMENT a. The radiological environmental monitoring samples shall be f

collected in accordance with Table 16.11.6-1 and shall be

' analyzed pursuant to the requirements of rables 16.11.6-1, i 16.11.6-2 and 16.11.6-3.

i b. A land use census shall be conducted and shall identify the 4

' location of the nearest milk animal and the nearest residence in each of the 16 meteorological sectors within a

! distance of five miles. Broad leaf vegetation sampling shall

, be performed at the site boundary in the direction sector 1

with the highest D/Q in lieu of the garden census.

! c. Analyses shall be performed on radioactive materials supplied as part of an Interlaboratory Comparison Program.

A summary of the results obtained as part of the i Interlaboratory Comparison Program and in accordance with j the methodology and parameters in the 00CM shall be included in' the Annual Radiological Environmental Operating Report.

d. The results of the land use census shall be included in the i Annual Radiological Environmental Operating Report.

1

...........................N0TE-----------------------------------

If samples required by Commitment part a, become permanently l unavailable from any of the required sample locations, the j locations from which samples were unavailable may then be deleted i from the program provided replacement samples were obtained and T added to the environmental monitoring program, if available. These i new locations will be identified in the Annual Radioactive 1 Effluent Release report.

a i

1 j APPLICABILITY: At all times s

O 4

16.11.6-1 03/27/99

I l

Radiological Environmental Monitoring 16.11.6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Radiological A.1 Submit a description May 15 of environmental of the reason for not following monitoring program is conducting the program calendar year not conducted as as required and plans required. to prevent a recurrence shall be included in the Annual Radiological Environmental Operating Report.

l l

B. Land use census B.1 ---------NOTE---------

identifies a Location The sampling location which yields a having the lowest calculated dose or calculated dose or dose comitment (via dose commitment (via ;

Os the same exposure the same exposure l pathway) greater than pathway) may be '

a location from which deleted from this i samples are currently monitoring program being obtained. after October 31 of the year in which this land use census was conducted.

Add new location to 30 days the radiological environmental monitoring program.

AND B.2 Identify new locations April 30 of in the next Annual following Radioactive Effluent calendar year Release report.

/

i 16.11.6-2 03/27/99

Radiological Environmental Monitoring 16.11.6 CONDITION REQUIRED ACTION COMPLETION TIME C. Interlaboratory C.1 Report corrective May 15 of Comparison Program actions in the Annual following analyses not performed Radiological calendar year as required. Environmental Operating Report.

SURVEILLANCE REQUIREMENTS g

SURVEILLANCE FREQUENCY SR 16.11.6.1 Conduct land use census during growing 12 months season using that information that will provide the best results, such as by a door-to-door survey, aerial survey, or by O .

consulting local agriculture authorities.

O 16.11.6-3 03/27/99

4 Radiological Environmental Monitoring Program C 16.11.6 Table 16.11.6-1 Radiological Environmental Monitoring Program

. Number of Sampling and

. Exposure Pathway Sample Collection and/or Sample Locations Frequency (d) Time and Frequency of (b) Analysis

1. AIRBORNE Radioiodine 5 Continuous Radiciodine canister:

and sampler I-131 analysis weekly.

Particulates operation with Particulate sampler:

sample Gross beta radioactivity collection analysis following filter j weekly, or more change; and gamma isotopic frequ'ently if analysis of composite (by required by dust location) quarterly. (c)

] loading.

O 2. DIRECT 40 Quarterly. Gama dose quarterly.

RADIATION !

3. MTERBORNE

a. Surface 2 Composite (a) 3ima isotopic analysis I sample over a rsonthly.

. 1-month period.

Composite for tritium analysis quarterly.

b. Drinking 3 Composite (a) Composite for gross beta sample over a and gama isotopic 1-month period. analyses monthly.

Composite for tritium analysis quarterly.

c. Sediment 2 Semiannually. Gama isotopic analysis

, from semiannually.

\ Shoreline 16.11.6-4 03/27/99

I i

Radiological Environmental Monitoring Program i

' 16.11.6 ;

Table 16.11.6-1 Radiological Environantal Monitoring Program j Number of Sampling and Exposure Pathway Sample Collection and/or Sample Locations Frequency (d) Time and Frequency of i (b) Analysis

? l I

4. INGESTION l

a. Milk 3 Semimonthly when Gama isotopic and I-131 i animals are on analysis semimonthly when l pasture; monthly animals are on pasture; at other times. monthly at other times. l

b. Fish 2 Semiannually. Gama isotopic analysis One sample of semiannually on edible each of the portion.

following species:

1. Bass l

2. Catfish !

c. Broad-leaf 2 Monthly. Gama isotopic analysis Vegetation monthly, i (a) Composite samples shall be collected by collecting an aliquot at intervals not exceeding 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

(b) Sample locations are identified in the ODCM.

(c) Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than 10 times the yearly mean of control samples, gama isotopic analysis shall be performed on the individual samples.

(d) Deviations are permitted from the required sampling schedule if specimens

, are unobtainable due to hazardous conditions, seasonal unavailability, or l to malfunction of automatic sampling equipment. If the latter, every

effort shall be made to complete corrective action prior to the end of l

\ the next sampling period.

i

! 16.11.6-5 03/27/99

Radiological Environmental Monitoring Program 16.11.6 x Table 16.11.6-2 Maximum Values for the 1.ower Limits of Detection (LLD) (a)(c)

Airborne Broad-leaf Particulate Fish Vegetation Sediment Water (pCi/kg, Milk (pCi/kg, (pCi/kg, Analysis (pCi/1) er Gasp)s wet)

(pci/m (pci/1) wet) dry)

Gross 4 lE-02 Beta H3 2,000 Mn-54 15 130 Fe-59 30 260 Co-58 15 130 Co-60 15 130 Zn-65 30 260 Zr-95 30 Nb-95 15 I-131 15(b) 7E-02 1 60 Cs-134 15 5E-02 130 15 60 150 Cs-137 18 6E-02 150 18 80 180 Ba-140 60 60 La-140 15 15 (a) The LLD is defined, for purposes of these commitments, as the smallest concentration of radioactive material in a sample with 95% probability of detection and with 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which n'ay include radiochemical separation):

LLD = 4.66 Sb E x V x 2.22 x Y x exp (-Aa t)

Where:

16.11.6-6 03/27/99

l '

Radiological Environmental Monitoring Program 16.11.6 1

Table 16.11.6-2 Maximum Values for the Lower Limits of Detection (LLD) (a)(c) l LLD is the lower limit of detection as defined above (as pCi per unit l mass or volume)

Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute)

E is the counting efficiency (as counts per disintegration)

V is the sample size (in units of. mass or volume) 2.72 is the number of disintegrations per minute per picocurie l

Y is the fractional radiochemical yield (when applicable) l A is the radioactive decay constant for the particular radionuclide l

A t is the elapsed time between sample collection (or end of the sample ,

collection period) and time of counting Typical values of E, V, Y and A t should be used in the calculation.

O The LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as a posteriori (after the fact) limit for a particular measurement.

Analyses.shall be performed in such a manner that the stated LLDs Wil be achieved under routine conditions. Occasionally background fluctuations, unavoidably small sample sizes, the presence of interfering nuclides, or other uncontrollable circumstances, may render these LLDs unachievable.

In such cases, the contributing factors will be identified and described in the Annual Radiological Environmental Operating Report.

(b) LLD for gamma isotopic analysis for I-131 in drinking water samples. Low l 1evel I-131 analysis on drinking water will not be routinely performed because the calculated dose from I-131 in drinking water at all locations is less than 1 mrem per year. Low level I-131 analyses will be performed if abnormal releases occur which could reasonably result in > 1 pCi/ liter of I-131 in drinking water. For low level analyses of I-131 an LLD of 1 pCi/ liter will be achieved.

(c) Other peaks which are measurable and identifiable, together with the radionuclides in Table 16.11.6-2, shall be identified and reported.

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16.11.6-7 03/27/99

l 1

Radiological Environmental Monitoring Program l O Table 16.11.6-3 Reporting Levels for Radioactivity Concentrations in Environmental Samples 16.11.6 l i

Airborne

, Particulite Fish Broad-leaf I

Water or Gases (pCi/kg, Milk Vegetation !

l Analysis (pCi/1) (pci/m ) wet) (pCi/1) (pCi/kg, wet) i H-3 2E04(a)

Mn-54 IE03 3E04 Fe-59 4E02 1E04 Co-58 IE03 3E04 Co-60 3E02 lE04 Zn-65 3E02 2E04 Zr-Nb-95 4E02 I-131 2(b) 1.0 3 1E02 Cs-134 30 10 lE03 60 1E03

-Cs-137 50 20 2E03 70 2E03 Ba-La-140 2E02 3E02 (a) For drinking water samples. This is 40 CFR Part 141 value.

(b) If low level I-131 analyses are performed.

lO 16.11.6-6 03/27/99

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

l

Radiological Environmental Monitoring Program 16.11.6 E81El The environmental monitoring program required by this commitment provides measurements of radiation and of radioactive materials in those exposure pathways and for whose radionuclides which lead to the highest potential radiation exposures of individuals resulting from the station operation. This monitoring program thereby supplements the radiological effluent monitoring program by verifying that the measurable concentrations of radioactive materials and levels of radiation are not higher than expected on the basis of the effluent measurements and modeling of exposure pathways. The initially specified monitoring program will be effective for at least the first three )

years of commercial operation. Following this period, program changes may be '

l initiated based on operational experience.

The detection capabilities required by Table 16.11.6-2 are considered optimum for routine environmental measurements in industrial laboratories. The specified lower limits of detection correspond to less than the 10 CFR 50.

Appendix I, design objective dose-equivalent of 45 mrem / year for atmospheric

! releases to the most sensitive organ and individual. The land use census l commitment is provided to assure that changes in the use of unrestricted areas are identified and that modifications to the monitoring program are provided ,

if required by the results of this census. '

The requirements for participation in an Interlaboratory Comparison Program is b provided to assure that independent checks on the precision and accuracy of l the measurements of radioactive material in environmental sample matrices are performed as part of a quality assurance program for environmental monitoring in order to demonstrate that the results are reasonably valid.

l The following requirement (s) were relocated from the CTS 6.4.4.f during the conversion to ITS.

The station shall have L program to monitor the radiation and radionuclides in i the environs of the plant. The program shall provide (1) representative l 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 l

conteined in UFSAR Chapter 16, (2) conform to the guidance of Appendix I to 10 CFR Part 50, and (3) include the following:

l l 1. Monitoring, sampling, analysis, and reporting of radiation and

! radionuclides in the environment in accordance with the methodology and l 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,

3. Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of 16.11.6-9 03/27/99 l

l l

Radiological Environmental Monitoring Program i 16.11.6 O !

I 1

radioactive materials in environmental ~ sample matrices are performed as I part of the quality assurance program for environmental monitoring. I

REFERENCES:

1. 10 CFR Part 50. Appendix I

2. Offsite Dose Calculation Manual 1

O !

l I

1 a

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16.11.6-10 03/27/99

S l

Dose Calculations 16.11.7 ,

i l 16.11 RADIOLOGICAL EFFLUENTS CONTROL "

16.11.7 Dose Calculations COMITMENT The annual (calendar year) dose or dose commitment, to any Member

. of The Public due to releases of radioactivity and to radiation from uraatum fuel cycle sources shall be limited to s 25 mrems to the total body or to any organ, except the thyroid, which shall be limited to s 75 mrems.

APPLICABILITY: . At all times ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated doses from A.1 Determine by None l the release of calculation, including )

radioactive materials direct radiation '

O V in liquid or gaseous contributions from the effluents exceeding reactor units and from twice the limits of outside storage tanks, SLC 16.ll.1.b, SLC whether the limits of 16.ll.2.b.1, or SLC Commitment 16.11.7 16.11.2.b.2 have been exceeded.

1 l

~

y i V

1 16.11.7-1 03/27/99

Dose Calculations 16.11.7 CONDITION REQUIRED ACTION COMPLETION TIME B. Calculated dose --------------NOTE----------- ,

exceeds limits of This Special Report, as l Commitment 16.11.7. defined in 10 CFR Part 20.2203(a), shall include an analysis that estimates the radiation exposure (dose) to ;

a Member Of The Public from i uranium fuel cycle sources, (including all effluent pathways and direct l radiation), for the calendar year that includes the l release (s) covered by this l report. It shall also describe the levels of radiation and concentration of radioactive material involved, and the cause of the exposure levels or concentrations.

B.1 Prepare and submit to 30 days the Commission a Special Report that defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the specified limits and includes the schedule for achieving conformance with the specified limits.

O 16.11.7-2 03/27/99

Dose Calculations 16.11.7 O CONDITION REQUIRED ACTION COMPLETION TIME C. Calculated dose C.1 - - - - - - - - - 140T E - - - - - - - - -

exceeds limit of Submittal of the Commitraent 16.11.7. report is considered a t'aely request, and a ANQ variance is granted until staff action on Release condition the request is resulting in violation complete.

of 40 CFR 190 not ----------------------

corrected at time of report submittal. Include a request for 30 days from a variance in . exceeding the accordance with the limit provisions of 40 CFR Part 190. j l

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.11.7.1 Determine cumulative dose contributions 31 days from liquid effluents in accordance with Offsite Dose Calculation Manual.

SR 16.11.7.2 Determine cumulative dose contributions 31 days from gaseous effluents in accordance with Offsite Dose Calculation Manual.

BASES The dose commitment is provided to assure that the release of radioactive material in liquid and gaseous effluents will be kept "as low as is reasonably achievable." The dose calculations in the ODCM implement the requirements in Section III.A of Appendix I in that conformance with the guides of Appendix I is to be shown by calculations and procedures based on models and data such that the actual exposure of an individual through appropriate pathways is O- unlikely to be substantially underestimated.

16.11.7-3 03/27/99 ey +

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

Dose Calculations 16.11.7 that the actual exposure of an individual through appropriate pathways is unlikely to be substantially underestimated.

REFERENCES:

1. 10 CFR Part 20

2. 40 CFR Part 190

3. Offsite Doss Calculation Manual

4. 10 CFR Part 50, Appendix I I

O l

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O 16.11.7-4 03/27/99

Reports I 16.11.8 1

16.11 RADIOLOGICAL EFFLUENTS CONTROL I 16.11.8 Reports i

COMITMENT- Special reports shall be submitted to the Regional Administrator, J Region II, within the time period specified for each report. These reports shall be submitted covering the activities identified below pursuant to the requirements of the applicable SLC:

a. Radioactive Liquid Effluents, ,

Dose, SLC 16.11.1.b '

Liquid Waste Treatment, SLC 16.11.1.c Chemical Treatment Ponds, SLC 16.11.1.d

b. Radioactive Gaseous Effluents.

Dose, SLC 16.11.2.b Gaseous Radwaste Treatment, SLC 16.11.2.c

c. Radiological Environmental Monitoring Program, SLC 16.11.6.a, b, and c

d. Land Use Census, SLC 16.11.6.d

e. Dose Calculations, SLC 16.11.7 APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Individual milk A.1 Submit plan advising 7 days samples show I-131 the NRC of the

, concentrations of 10 proposed action to i picocuries per liter ensure the plant 1 or greater. related annual doses will be within the design objective of 45 mrem /yr to the thyroid i of any individual. l 1

i O

16.11.8-1 03/27/99 1

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

l l

l Reports 16.11.8 O

CONDITION REQUIRED ACTION COMPLETION TIME B. Milk samples collected B.1 Submit a plan 30 days over a calendar advising the NRC of l quarter show average the proposed action to concentrations of ensure the plant 2: 4.8 picoCuries per related annual doses liter will be within the design objective of 45 mrem /yr to the thyroid of any individual.

I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.11.8.1 NA NA BASES Reference applicable commitments.

REFERENCES:

1. 10 CFR Part 20

2. 40 CFR Part 190

3. Offsite Dose Calculation Manual 1

16.11.8-2 03/27/99

l Radioactive Effluent Release Report !

q 15.11.9 l

V 16.11 RADIOLOGICAL EFFLUENTS CONTROL 16.11.9 Radioactive Effluent Release Report COP 91ITHENT The Annual Radioactive Effluent Release Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year.

A single submittal may be made for a multiple unit station. The I submittal shall combine those sections that are common to all l units at the station; however, for units with radwaste systems, the submittal shall specify the release of radioactive material 4

from each unit. '

i The Annual Radioactive Effluent Release Report shall include a j summary of the quantities of radioactive liquid and gaseous a

effluents and solid waste released from the station during the

],

reporting period.

The annual Radioactive Effluent Release Report shall include a

! summary of the meteorological conditions concurrent with the release of gaseous effluents during each quarter.

p The Annual Radioactive Effluent Release Report shall include an j

4 Q assessment of the radiation dose from radioactive effluents to members of the public due to their activities inside the

l unrestricted area boundary during the reporting period. All assumptions used in making these assessments (e.g., specific i activity, exposure time and location) shall be included in these reports.

1 4

The Annual Radioactive Effluent Release Report shall include the

following information for all unplanned releases to unrestricted areas of radioactive materials in gaseous and liquid effluents

a. A description of the event and equipment involved;

b. Cause(s) for the unplanned release;

c. Actions taken to prevent recurrence; and,

d. Consequences of the unplanned release.

The Annual 11adioactive Effluent Release Report shall include an assessment of radiation doses from the radioactive liquid and gaseous effluents released from the station during each calendar quarter. In addition, the unrestricted area boundary maximum noble gas gamma air and beta air doses shall be evaluated. The annual average meteorological conditions shall be used for yS determining the gaseous pathway doses. Approximate and conservative approximate methods are acceptable. The assessment 16.11.9-1 03/27/99

Radioactive Effluent Release Report 16.11.9

%.)

of radiation doses shall be performed in accordance with the Offsite Dose Calculation Manual. I 1

The Annual Radioactive Effluent Release Report shall include an I explanation of why the inoperability of liquid or gaseous effluent i

monitoring instrumentation out of service for greater than 30 days was not corrected in a timely manner per SLC 16.11.3.

The Annual Radioactive Effluent Release Report shall include the l

following information for each type of solid waste shipped offsite

during the report period

a. Total container volume (cubic meters);

b. Total curie quantity (determined by measurement or estimate);- l

c. Principal radionuclides (determined by measurement or estimate);

d. Type of waste, (e.g., spent resin. compacted dry waste evaporator bottoms);

e. Number of shipments; and,

f. Solidification agent (e.g., cement, or other approved agents (media)).

The Annual Radioactive Effluent Release Report shall include a list and description of unplanned releases from the site to Unrestricted Areas of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Annual Radioactive Effluent Release Report shall include any changes made during the reporting period to the Offsite Dose Calculation Manual (ODCM), as well as a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census.

The Annual Radioactive Effluent Re' lease Report shall also include an assessment of radiation doses to the likely most exposed Member Of The Public from reactor releases and other nearby uranium fuel cycle sources (including doses from primary effluent pathways and i direct radiation) for the previous calendar year to show l conformance with 40 CFR 190, Environmental Radiation Protection i Standards for Nuclear Power Operation. Methods for calculating the dose contribution from liquid and gaseous effluents are given l in the ODCM.

O l APPLICABILITY: At all times.

l 16.11.9-2 03/27/99

Radioactive Effluent Release Report 16.11.9 ACTIONS CONDITION- REQUIRED ACTION COMPLETION TIME A. NA A.1 NA NA SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I

SR 16.11.9.1 NA NA 1

BASES i O NA

REFERENCES:

i

1. Oconee ITS

2. Offsite Dose Calculation Manual O

16.11.9-3 03/27/99

1 l

i Radiological Environmental Operating Report 16.11.10 16.11 RADIOLOGICAL EFFLUENTS CONTROL 16.11.10 Radiological' Environmental Operating Report COMITMENT Routine Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May I of each year.

1 The Annual Radiological Environmental Operating Report shall include sumaries, interpretations. and statistical evaluation of the results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of .

the observed iapacts of the plant operation on the environment. I The reports shall also include the results of the land use censuses. If harmful effects are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

I The Annual Radiological Environment Operating Report shall include a sumary of the results obtained as part of the required i Interlaboratory Comparison Program and in accordance with the O ODCM. Alternatively, participants in the EPA cross-check program shall provide the EPA program code designation for the unit.

The Annual Radiological Environmental Operating Report shall include sumarized and tabulated results of the radiological environmental samples required by SLCs taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted as soon as practical in a supplementary report.

The initial report shall also include the following: a sumary description of the radiological environmental monitoring program including sampling methods for each sample type, size and physical characteristics of each sample type, sample preparation methods, analytical methods, and measuring equipment used; a map of all sampling locations keyed to a table giving distances and directions from one reactor; and, the result of land use censuses.

Subsequent reports shall describe all substantial changes in these aspects.

APPLICABILITY: At all times.

O 16.11.10-1 03/27/99

j Radiologica.' Environmental Operating Report 16.11.10

. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. NA A.1 NA NA i

1 I

l SURVEILLANCE REQUIREMENTS I

SURVEILLANCE FREQUENCY SR 16.11.10.1 NA NA BASES l

REFERENCES:

1. Oconee ITS l

2. Offsite Dose Calculation Manual l

l l

I lO 16.I1.10-2 03/27/99 l

Iodine Radiation Monitoring Filters 16.11.11 16.11_ RADIOLOGICAL EFFLUENTS CONTROL 16.11.11 Iodine Radiation Monitoring Filters COMITMENT Assure that the iodine radiation monitoring filters perform their intended function.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. NA A.1 NA NA SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY O SR 16.11.11.1 Remove and replace iodine radiation 30 days of monitoring filters in RIA-44. operation SR 16.11.11.2 Discard spare iodine radiation monitoring After 24 months filters. of shelf life.

EA_SES The purpose of this commitment is to assure the reliability of the iodine radiation monitoring charcoal filters.

REFERENCES:

1. Oconee CTS Amendment No. 3/3 SER date July,1974.

V 16.11.I1-1 03/27/99 i

Radioactive Material in Outside Temporary Tanks Exceeding Licit 16.11.12 O

Q 16.11 RADIOLOGICAL EFFLUENTS CONTROL 16.11.12 Radioactive Material in Outside Temporary Tanks Exceeding Limit C0ffilTMENT The quantity of radioactive material in outside temporary storage tanks shall not exceed the limit specified in ITS 5.5.13.c. .

APPLICABILITY: At all times. I l

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l

A. The quantity of A.1 Suspend addition of Immediately radioactive material radioactive material in outside temporary to tank.

storage tank not within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR'16.11.12.1 Verify the quantity of radioactive material within 7 days contained in each of the outside temporary after addition tanks is within the limit by analyzing a of radioactive representative sample of the tanks' materials to contents. the tank BASES The requirement (s) of this SLC section were relocated from CTS 3.9.1.c during the conversion to ITS.

The tanks included in this specification are all those outdoor radwaste liquid storage tanks that 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 cor.nected to the liquid radwaste treatment system. Restricting the quantify of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of a tank's contents, p the resulting concentrations would be irss than the limits of 10CFR Part 20,

'Q Appendix B, Table II, Column 2, at thn nearest potable water supply and the nearest surface water supply in an UNhESTRICTED AREA.

16.11.12-1 03/27/99

Radioactive Material in Outside Temporary Tanks Exceeding Limit

p 16.11.12

\}g

REFERENCES N/A i

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l 16.11.12-2 03/27/99 f I

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l Radioac+.ive Material in Waste Gas Holdup Tank Exceeding Limit 3 16.11.13

q i

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16.11 RADIOLOGICAL EFFLUENTS CONTROL l

16.11.13 Radioactive Material in Waste Gas Holdup Tank Exceeding Limit )

COMITMENT The quantity of radioactive material in the Waste Gas Holdup

1 tanks shall not exceed the limit specified in ITS 5.5.13.b.

?!i ~

)

X APPLICABILITY: At all times.

ACTIONS

.......................................N0TE----------------------------------- l Separate Condition Entry is allowed .for each tank.

................................................................ ............. 1 COMPLETION TIME CONDITION REQUIRED ACTION Suspend addition of Immediately A. The quantity of A.1 radioactive material radioactive material in the Waste Gas to tank.

Holdup tank not within j AND limit.

A.2 Reduce tank contents 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months J

to within limit.

i SURVEILLANCE REQUIREMENTS

. FREQUENCY

" SURVEILLANCE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when SR 16.11.13.1 Verify quantity of radioactive materials in tank is being each tank is within limit, filled 4W [i'4

- BASES The requirement (s) of this SLC section were relocated from CTS 3.10.1.

3.10.1.c daring the conversion to ITS.

4 iO 16.11.13-1 03/27/99 2

Radioactive Material in Waste Gas Holdup Tank Exceeding Limit 16.11.13 O Restricting the quantity of radioactivity contained in each waste gas holdup tank provides assurance that in the event of an uncontrolled release of the tank contents, the resulting total body exposure to an individual at the exclusion area boundary will not exceed 0.5 rem.

REFERENCE UFSAR, Section 15.10 O

O 16.11.13-2 03/27/99

l Explosive Gas Mixture 16.11.14 l

() 16.11 RADIOLOGICAL EFFLUENTS CONTROL 16.11.14 Explosive Gas Mixture COMMITMENT The concentration of Hydrogen in the Waste Gas Holdup Tanks shall be s 3% by volume.

APPLICABILITY: At all times.

ACTIONS

................................-------NOTE-----------------------------------

l Separate Condition Entry is allowed for each tank.

m CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of A.1 Reduce Concentration 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Hydrogen in Waste Gas of Hydrogen to within Holdup tank is > 3% limit, and s 4% by volume.

O B. Concentration of B.1 Suspend addition of Immediately Hydrogen in Waste Gas waste gases'to tank.

Holdup tank is > 4% by volume. AND B.2 Reduce Concentration 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of Hydrogen to within limit.

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l 16.11.14 1 03/27/99

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Explosive Gas Mixture 16.11.14 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR'16.ll.14.1 Verify Hydrogen concentration in Waste Gas 5 times / week on Holdup Tank is s 3% by volume. each tank when in service bND once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after isolation of the tank E8KS.

The requirement (s) of this SLC section were relocated from CTS 3.10.2 and Table 4.1-3, Item 13 during the conversion to ITS.

This Commitment is provided to ensure that the concentration of potentially explosive gas mixtures contained in the Waste Gas Holdup Tanks is maintained below the flamability limits of hydrogen. (Administrative controls are used to prevent the hydrogen concentrations from reaching the flammability limit.)

These controls include sampling each tank 5 times a week while in service, and/or once in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after isolation of the tank; injection of dilutants to reduce the concentration of hydrogen below its flammability limits provides assurance that the releases of radioactive material will be controlled in conformance with the requirements of GDC 60 of Appendix A to CFR Part 50.

REFERENCES N/A O

16.11.14-2 03/27/99

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Decay Time for Movement of Irradiattd Fuel 16.12.1 ,

16.12 REFUELING OPERATIONS 16.12.1 Decay Time for Movement of Irradiated Fuel COMITMENT Irradiated fuel shall not be moved from the reactor until the unit has been subcritical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 1

APPLICABILITY: MODE 6.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY J G

l SR 16.12.1.1 N/A. N/A BASES I

The requirement (s) of this SLC section were relocated from CTS 3.8.11 during i the conversion to ITS.

The safety analysis for the fuel handling accident is based on the assumption that the reactor has been shutdown for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . j I

REFERENCES UFSAR, Section 15.11.2.1 a

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16.12.1-1 03/27/99

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

Area Radiation Monitoring for Fuel Loading and Refueling 16.12.2 16.12 REFUELING OPERATIONS 16.12.2 Area Radiation Monitoring for Fuel Loading and Refueling COMITMENT Radiation levels in the reactor building refueling area shall be monitored by RIA-3 and by a portable bridge monitor for each bridge which is being used for fuel handling.

Radiation, levels in the spent fuel storage area shall be monitored by RIA-6 and by a portable bridge monitor.

. APPLICABILITY: During fuel handling.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required Area A.1 Use portable survey Immediately Radiation Monitor not. instrumentation, 4 monitoring radiation having the i levels, appropriate ranges and sensitivity to fully protecc individuals involved O- in refueling operation.

SURVEILLANCE REQUIREMENTS SURVEILLAliCE FREQUENCY SR 16.12.2.1 N/A, N/A E8.lf.S The requirement (s) of this SLC section were relocated from CTS 3.8.1 during the conversion to ITS.

O 16.12.2-1 03/27/99

l Arca Radiation Monitoring for Fuel Loading anc' Refueling 16.12.2 l Continuous monitoring of radiation levels provides imediate indication of an unsafe condition.

1 REFERENCES N/A O

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i Communication Between Control Room and Refueling Personnel 16.12.3 16.12 REFUELING OPERATIONS.

i 16.12.3 Communication Between Control Room and Refueling Personnel C0milTMENT Direct communications between the control room and the refueling personnel in the reactor building sha'l exist. I l

APPLICABILITY: During C6RE ALTERATIONS.

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ACTIONS I CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE O -

FREQUENCY SR 16.12.3.1 N/A.. N/A BASES The requirement (s) of this SLC section were relocated from CTS 3.8.5 during the conversion to ITS.

This Commitment allows the control room operator to inform the reactor building prsonnel of any impending unsafe condition detected from the main control board indicators during fuel movement.

REFERENCES l

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Handling of Irradiated Fuel Asstablies 16.12.4 l

16.12 REFUELING OPERATIONS 16.12.4 Handling of Irradiated Fuel Assemblies 1

L CONIITNENT When two irradiated fuel assemblies are being handled !

simultaneously within the fuel transfer canal, a minimum of l 10 feet separation shall be maintained between the ,

assemblies at all times. 1

..............-N0TE----------------------------

The Auxiliary Hoist may be used provided no other irradiated fuel assembly is being handled in the fuel transfer canal.

l APPLICABILITY: During movement of irradiated fuel assemblies inside containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 1 SR 16.12.4.1 N/A. N/A E83.E.1 The requirement (s) of this SLC section were relocated from CTS 3.8.8 during the conversion to ITS.

REFERENCES N/A L

16.12.4-1 03/27/99 l

Loads Suspended Over Spsnt Fuel in Spent Fual Pool 16.12.5 4

16.12 REFUELING OPERATIONS j 16.12.5 Loads Suspended Over Spent Fuel in Spent Fuel Pool

CONI!~ENT' No suspended loads of more than 3000 lb, shall be i

. transported over spent fuel stored in the spent fuel pool.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME j A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS l

SURVEILLANCE FREQUENCY

O SR 16.12.5.1 N/A. N/A BASES The requirement (s) of this SLC section were relocated from CTS 3.8.14 during the conversion to ITS.

This commitment is required to prohibit transport of. loads greater than a fuel assembly with a control rod and the associated fuel handling tool (s).

REFERENCES N/A 4

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16.12.5-1 03/27/99

Fira Brigade 1 16.13.1 1 l

0 1- >> co ouc' or oetaa'ious -

16.13.1 Fire Brigade I

l COMITNENT A Fire Brigade of five members shall be maintained onsite at all l l times.

....................__..__..-N0TE--------------------------------- !

.This excludes 3 members of the minimum operating shift l requirements that are required to be present in the control rooms. ;

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APPLICABILITY: At all times.

ACIIONS CONDITION REQUIRED ACTION COMPLETION TIME i

A. NA A.1 NA NA i

O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.13.1.1 NA NA BASES The primary purpose of the Fire Protection Program is to minimize both the probability and consequences of postulated fires. Despite designed active and passive Fire Protection Systems installed throughout the plant, a properly trained and equipped Fire Brigade organization of at least five members is needed to provide immediate response to fires that may occur at the site.

Fire Brigade equipment and training conform to Oconee's commitments to

! Appendix A to Branch Technical Position 9.5-1 and supplemental NRC Staff guidelines including Nuclear P1 ant Functional Responsibilities, Administrative Controls and Quality Assurance.

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!O 16.13.1-1 03/27/99

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Fire Brigade l 16.13.1 v

This !alected Licensee Commitmer,t is part of the Oconee Fire Protection ;

Program and therefore subject to the provisions of Oconee Facility Operating l License Conditions. l The following requirement was relocated from the CTS 6.1.1.8 during the conversion to ITS.

A training program for the fire brigade shall meet or exceed the requirements of Section 27 of the NFPA Code-1975, except that training sessions may be held quarterly.

REFERENCES:

1. Oconee UFSAR, Chapter 9.5.1. l

2. Oconee Fire Protection SER dated August 11, 1978.

3. Oconee Fire Protection Review, as revised.

4. Duke letter of January 16, 1978 to NRC in response to " Nuclear Plant Functional Responsibilities, Administrative Controls, and Quality Assurance".

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O 16.13.1-2 03/27/99

Technical Review and Control g 16.13.2 V

16.13 CON 00CT OF OPERATIONS 16.13.2 Technical Review and Control COMITMENT A Technical Review and Control Program covering the preparation, j review, and approval of documents important to station operation shall be established and maintained for the site.

Personnel performing the preparation, review, and approval activities covered by this commitment shall meet or exceed the qualifications of ANSI N18.1-1971 (the conformance status for this standard is as listed in Table 17-1 of the Duke Pcwer Topical Report, Quality Assurance Program, Duke-1-A).

a. The preparation, review, and approval of station procedures 4

shall be done in accordance with station Technical Specifications. Individuals responsible for these reviews shall be members of the supervisory staff assigned to the i site, be previously designated by the Site Vice President as a Qualified Reviewer, and successfully complete the site l Qualified Reviewer training program. Review of ;

environmental radiological analysis procedures, shall be i performed by the General Manager, Environmental Services or a designee. 'Each such review shall include a determination O- of whether or not additional, cross-disciplinary review shall be performed by the appropriately designated site review personnel.

b. Proposed modifications shall be designed and the design reviewed'in accordance with station Technical Specifications. The proposed modification design, the design review, and design approval shall be in accordance with ANSI N45.2.ll as described in Table 17-1 of the Duke Power Topical Report, Quality Assurance Program, Duke 1-A.

Proposed modifications to nuclear safety related structures, systems, and components shall be approved prior to implementation by the Station Manager or the Manager of Engineering; or for the Station Manager, by a Maintenance Superintendent, the Operations Superintendent, or the Work Control Superintendent, as previously designated by the Station Manager. Upon implementation approval, the modification shall be implemented in accordance with the Duke Power Nuclear Station Modification Program and approved procedures (as discussed in Item a above).

c. Proposed changes to the station Technical Specifications shall be prepared in accordance with station Technical Specifications. Each proposed Technical Specification 4 change shall be reviewed by the Plant Operations Review Committee (PORC) and the Nuclear Safety Review Board (NSRB) prior to submittal to the Nuclear Regulatory Commission.

Proposed changes to the Technical Specifications shall be 16.13.2-1 03/27/99

d i

Technical Review and Control 16.13.2 approved by the Station Manager, or for the Station Manager by a designated manager or company officer. Technical l Specifications submittal cover letters shall be signed by an l officer of Duke Power Company. j l d. Proposed tests and experiments which affect station nuclear i safety and are not addressed in the UFSAR or Technical Specifications shall be reviewed by the Plant Operations Review Committee (PORC).

e. Incidents reportable pursuant to station Technical Specifications and all violations of Technical Specifications shall be investigated and a report prepared which evaluates the occurrence and which provides recommendations to prevent recurrence. Such reports shall i be approved by the Manager, Safety Assurance and provided to l the Site Vice President and the Plant Operations Review Comittee (PORC).

f. The Manager, Safety Assurance shall assure the performance of special reviews and investigations, and the preparation and submittal of reports thereon, as requested by the Site Vice President. Such reports shall be provided to the Plant Operations Review Committee (PORC). ;

g. The Manager, Safety Assurance shall assure the performance of a review by a knowledgeable individual / organization of every unplanned onsite release of radioactive material to tie environs, including the preparation and forwarding of re.oorts covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence to the Site Vice President. and to tha Plant Operations Review Committee (PORC),

h. The Manager, Safety Assurance shall assure the performance of a review by.a knowledgeable individual / organization of changes to the Process Control Program, Offsite Dose Calculation Manual (0DCM), and Radwaste Treatment Systems.

1. The Manager, Safety Assurance shall ensure the performance of a review by a knowledgeable individual / organization of the Fire Protection program and implementing procedures and submittal of recommended changes to the Director, Organization Effectiveness Services.

i j. Reports documenting each of the activities performed under

! this commitment shall be ,e4intained. Copies shall be j provided to the NSRB.

> (3 U APPL.ICABILITY: At all times.

I 16.13.2-2 03/27/99

t Technical Review and Control 16.13.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l

A. NA A.1 NA NA l l .

l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.13.2.1 NA NA 1

BASES The requirements contained in this selected licensee commitment were relocated from the Oconee Technical Specifications with the approval of the U. S.

Nuclear Regulatory Commission. Changes to this SLC shall be considered a Q

V change in an NRC commitment and shall be made only in accordance with the approved Compliance Manual Procedure fer the Control of Selected Licensee Commitments and by use of the 10 CFR 50.59 evaluation process.

This SLC implements the review requirements of ANSI N18.7-1976/ANS-3.2 and ANSI N45.2.11-1974 as referenced in the Duke Power Company Topical Report, Quality Assurance Program, Duke-1-A.

REFERENCES:

1. ANSI N18.1-1971, Selection and Training of Nuclear Power Plant Personnel

2. ANSI N18.7-1976/ANS-3.2, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants

3. ANSI N45.2.11-1974, Quality Assurance Requirements for the Design of Nuclear Power Plants

4. Compliance Manual Procedure for the Control of Selected Licensee Commitments

5. Nuclear System Directive 209,10 CFR 50.59 Evaluations

6. 10 CFR 50.59

7. Nuclear System Directive 703, Administrative Instructions for Station Procedures l

O 16.13.2-3 03/27/99

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' Plant Operations Review Committee 16.13.3 j

16.13 CONDUCT OF OPERATIONS 16.13.3 Plant Operations Review Committee j

j C0f04ITMENT A Plant Operations Review Committee (PORC) shall be established and maintained for the site. The PORC shall be composed of the Manager of Safety Assurance, the Station Manager and his/her direct reports most responsible for station operation and maintenance, the Manager of Engineering and his/her direct reports i most responsible for engineering support of station operation and maintenance, or designated alternates. The PORC Chairperson, members, and alternate members shall be qualified in accordance with ANSI N18.1-1971 and be appointed by the Site Vice President.

3 The quorum necessary for conducting the PORC functions shall i consist of the Chairperson, or his/her designated alternate, and at least three other PORC members including alternates. .

i Reports of reviews encompassed by this Selected Licensee Commitment shall be prepared and forwarded to the Site Vice President and the Nuclear Safety Review Board.

1

a. The PORC shall be responsible for reviewing the following

prior to final approval:

1. All proposed tests and experiments which affect stat;on nuclear safety and are not addressed in the UFSAR or

]

(_, Technical Specifications;

2. OPERABILITY evaluations resulting in a Justification for l Continued Operation and a proposal for discretionary enforcement;

3. OPERABILITY evaluations resulting in the decision that affected systems, structure or components are OPERABLE but degraded; and

4. All proposed changes to the' station Technical i Specifications, Bases, or Facility Operating License. i

b. The PORC shall be responsible for reviewing the effectiveness of corrective actions for:

1. Licensee Event Reports and Special Reports made to the NRC;

2. Violations of Technical Specifications;

3. Special reviews and investigations as requested by the Site Vice President; and

{ 4. Reports on unplanned onsite releases of radioactive J

material to the environs.

}

16.13.3-1 03/27/99

Plant Operations Review Committee l 16.13.3 C. The PORC shall review additional programs, procedures and plant activities as directed by the Site Vice President.

l APPLICABILITY: At all times.

ACTIONS L CONDITION REQUIRED ACTION COMPLETION TIME l

A. NA A.1 NA hA SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.13.3.1 NA NA l

BASES

, The PORC shall be established to recommend to the Station Manager approval or L disapproval of the items listed under this comicitment prior to their final approval.

The PORC shall report to the Site Vice President on the areas of responsibility specified in this selected licensee commitment.

REFERENCES:

1. ANSI N18.1-1971, Selection and Training of Nuclear Power Plant Personnel

2. ANSI N18.7-1976/ANS-3.2, Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants

3. Nuclear System Directive 308, Plant Operations Review Committee LO 16.13.3-2 03/27/99

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Reactivity Anomaly 16.13.4 16.13 CONDUCT OF OPERATIONS 16.13.4 REACTIVITY ANOMALY CCMITMENT ' Report the cause of reactivity anomalies to the Nuclear

- Regulatory Commission.

APPLICABILITY: At all times.

! ACTIONS CONDITION REQUIRED ACTION COMPLET'ON TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.13.4.1 N/A. N/A BASES The requirement (s) of this SLC section were relocated from CTS 4.10 during the conversion to ITS.

REFERENCES t

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16.13.4 1 03/27/99

Additional Operating Shift Requirements ,

16.13.5 j 16.13 CONDUCT OF OPERATIONS 16.13.5 Additional Operating Shift Requirements C0WilTMENT a. Minimum operating shift staffing shall include the following additional personnel in excess of ITS requirements:

1. One Shift Work Manager when any unit is in MODES 1, 2, 3 or 4

2 One additional Non-Licensed Operator when Unit 1 and 2 are i in MODES 1, 2, 3, or 4 and Unit 3 is not in MODES 1, 2, 3, !

or 4.

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3. Two Non-Licensed operator when no units are in MODES 1, 2, l 3 or 4.

4. At least one licensed operator shall be in the reactor building when fuel handling operations in the reactor building are in progress.

5. If the computer for a reactor is inoperable for more than ,

eight hours, an operator, in addition those specified in '

O ITS 5.5.2.b and 10 CFR 50.54(m) shall supplement the control room staff.

b. The Shift Work Manager shall be an exper';enced SRO.

APPLICABILITY: With fuel in any of the three reactor vessels.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.13.5.1 N/A N/A O

16.13.5-1 03/27/99

. Additional Operating Shift Requirements 16.13.5 mm The requirement (s) o'f this SLC section were relocated from CTS 6.1.1.9 and I Table 6.1-1 during the conversion to ITS.

The requirements of this commitment are in addition to the requiremements specified in ITS 5.2.2. For example, when Units 1 and 2 are in MODES 1, 2, 3, or 4 and Unit 3 is not in MODES 1, 2, 3, or 4 (i.e., two units are operating from one control room), ITS 5.2.2.a requires 4 non-licensed operators. SLC 16.13.5.1.b requires an additional non-licensed operator for a total of five non-licensed operators.

REFERENCES

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Retraining and Replacement of Station Persennnel 16.13.6 ,

16.13 CONDUCT OF OPERATIONS 16.13.6 Retraining and Replacement of Station Personnnel C0fMITMENT Retraining and replacement of station personnel shall be in accordance with Section 5.5 of the ANSI /ANS-3.1-1978, t

" Selection and Training of Nuclear Power Plant Personnel." 1 APPLICABILITY: At all times.

ACTIONS j CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE O FREQUENCY SR 16.13.6.1 N/A N/A l f4 BA The requirement (s) of this SLC section were relocated from CTS 6.1.1.7 during the conversion to ITS.

REFERENCES N/A 1

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O 16.13.6-1 03/27/99

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Precedures for Control of pH in Recirculated Coolant After LOCA

& Long-Term Emergency Core Cooling Systems 16.13.7 O 16.13 CONDUCT OF OPERATIONS 16.13.7 Procedures for Control of pH in Recirculated Coolant After Loss-of-Coolant Accident (LOCA) & Long-Term Emergency Core Cooling Systems COMITMENT a. Procedures shall state that pH will be measured and the addition of appropriate caustic to coolant will commence within 30 minutes after switchover to recirculation mode of core cooling to adjust the pH to a range of 7.0 to 8.0 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

b. Procedures shall include provision for remote or local operation of system components necessary to establish high and low pressure injection within 15 minutes after a line break.

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A N/A SURVEILLANCE REQUIREMENTS SVRVEILLANCE FREQUENCY SR 16.13.7.1 N/A N/A BJA fji The requirement (s) of this SLC section were relocated from CTS 6.4.1.i and 6.4.1.k during the conversion to ITS.

REFERENCES N/A O

16.13.7-1 03/27/99

l Respiratory Protective Progran 1 16.13.8 ;

O g 16.13 CON 00CT OF OPERATIONS 16.13.8 Respiratory Protective Program l COMITMENT A respiratory protective program approved by the Commission I shall be in force. l APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A N/A l

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY I O SR 16.13.8.1 N/A N/A BASES The requirement (s) of this SLC section were relocated from CTS 6.4.4.a during the conversion to ITS.

REFERENCES N/A O

16.13.8-1 03/27/99 f

c Startup Report 16.13.9 16.13 CONDUCT OF OPERATIONS 16.13.9 Startup Report C0f#ilTMENT A summary report of unit startup and power escalation testing shall be submitted following (1) receipt of an operating license, (2) amendment to the facility 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 unit.

Startup reports shall be submitted (1) within 90 days following completion of the startup test program, (2) 90 days following resumption of commencement of commercial power operation, or (3) nine months following initial criticality, whichever occurs first. If a starty, report does not cover all three events, i.e., initial criticality, completion of the startup test program and resumption or commencement of commercial power operation supplementary reports shall be submitted at least every three months until all three events are completed.

APPLICABILITY: At all times.

{

}

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A N/A SURVEILLANCE REQUIREMENTS SURVE!LLANCE FREQUENCY i

SR 16.13.9.1 N/A N/A i

O 16.13.9-1 03/27/99 i

l

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

Startup Report 16.13.9 mu The requirement (s) of this SLC section were relocated from CTS 6.6.1.1 during the conversion to ITS.

REFERENCES N/A O

i c

O 16.13.9-2 03/27/99

Core Operating Limits Report 16.13.10 16.13 CONDUCT OF OPERATIONS 16.13.10 Core Operating Limits Report '

COPMITMENT Concentrated Boric Acid Storage Tank volume and boron concentration limits shall be established prior to each reload cycle, or prior to any remaining part of a reload !

cycle and shall be documented in the CORE OPERATING LIMITS '

REPORTS.

l APPLICABILITY:. At all times. l l

ACTIONS I CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A N/A l

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.13.10.1 N/A N/A BASfd The requirement (s) of this SLC section were relocated from CTS 6.9.1.4 during the conversion to ITS.

REFERENCES ,

N/A O

a 16.13.10-1 03/27/99

Procedure for Station Survey Fo11cwing an Earthquake 16.13.11 16.13 CONDUCT OF OPERATIONS 16.13.11 Procedure for Station Survey Following an Earthquake 1 COMITMENT Written procedures with appropriate check-off lists and '

instructions shall be provided for a station survey following an earthquake:

APPLICABILITY: At all times.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A N/A ,

l I

1 SURVEILLANCE REQUIREMENTS

]

SURVEILLANCE FREQUENCY O SR 16.13.11.1 N/A N/A l

E85.fl The requirement (s) of this SLC section were relocated from CTS 6.4.1.f during the conversion to ITS. .

l REFERENCES N/A O

16.13.11-1 03/27/99

APSR Movement 16.14.1 16.14 CONTROL RODS AND POWER DISTRIBUTION 16.14.1 APSR Movement

C00MITMENT Perform specified SR.

APPLICABILITY:- MODES 1 and 2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A. A.1 N/A. N/A SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.14.1.1 Verify that loss of power will not cause --- --NOTE-----

APSR movement. The provisions of SLC 16.2.7 do not apply.

18 months +25%

BASES The requirement (s) of this SLC section were relocated from CTS 4.7.1 during the conversion to ITS.

I REFERENCES N/A lO 16.14.1-1 03/27/99 L _ - . - - - _ .

Control Rod Prcgram Verification 16.14.2 16.14 CONTROL RODS AND POWER DISTRIBUTION 16.14.2 Control Rod Program Verification 1

COMi!TMENT CONTROL RODS shall be operated in their programmed functional position and group.

APPLICA8ILITY: MODES I and 2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. CONTROL ROD not A.1 Declare CONTROL R00 Immediately ;

operating in its inoperable. !

programmed functional position and group.

O SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY l

SR 16.14.2.1 Each control rod drive mechanism shall be Whenever the selected from the control room and control rod exercised by a movement of approximately drive patch two inches to verify that the proper rod panel is locked has responded as shown en the unit computer (after printout of that rod. inspection, test, reprogramming, or maintenance)

(continued) !

l O

16.14.2-1 03/27/99 I

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

l l

Control Rod Program Verification 16.14.2 SURVEILLANCE REQUIREMENTS (continued) i SURVEILLANCE FREQUENCY SR 16.14.2.2 Independently verify that the power or Upon instrumentation cables to control rod drive reconnecting assemblies atop the reactor are connected. after the cables have been disconnected or removed BASES The requirement (s) of this SLC section were relocated from CTS 3.5.2.2.b.6 and CTS 4.7.2 during the conversion to ITS.

Each control rod has a relative and an absolute position indicator system.

One set of outputs goes to the plant computer, identified by a unique number (1 through 69) associated with only one core position. The other set of outputs goes to a programmable bank of 69 edgewise meters in the control room.

O In the event that a patching error is made in the patch panel or connectors in the cables leading to the control rod drive assemblies or to the control room meter bank are improperly transposed upon reconnection, these errors and transpositions will be discovered by a comparative check by: (1) selecting a specific rod from one group (e.g., Rod 1 in Regulating Group 6), (2) noting that the program-approved core position for this rod of the group (assume the approved core position is No. 53), (3) exercising of the selected rod and (4) noting that the computer prints out both absolute and relative position response for the approval core position (assumed to be position No. 53) and that the proper meter responds in the control room display bank (assumed to be Rod 1 in Group 6) for both absolute and relative meter positions. This type of comparative check will not assure detection of improperly connected cables inside the reactor building. For these, it is necessary for a responsible person, other than the one doing the work, to verify by appropriate means that each cable has been matched to the proper control rod drive assembly.

REFERENCES

1. UFSAR, Section 7.6.

O 16.14.2-2 03/27/99 i -

l i Power Mapping l 16.14.3 O 16.14 CONTROL RODS AND POWER DISTRIBUTION 16.14.3 Power Mapping COMITMENT Perform specified SRs.

APPLICABILITY: MODE 1 with THERMAL POWER > 20% RTP.

l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME l

l l A. N/A. A.1 N/A. N/A !

l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 16.14.3.1 Using the Incore Instrumentation System, a -----NOTE------ l l power map shall be made to verify expected SR 3.0.2 is not power distribution, applicable '

10 EFPD E8511 The requirement (s) of this SLC section were relocated from CTS 4.1.5 during the conversion to ITS.

Periodic use of the Incore Instrumentation System for power mapping is sufficient to assure that axial and radial power peaks and the peak locations are controlled in accordance with the provisions of the Technical

- Specifications.

REFERENCES I N/A L

16.14.3-1 03/27/99 l

w , e ,- -,.-m w - - , -,y - - - -., , , , -

Control Rod Drive Patch Panals 16.14.4 16.14 CONTROL RODS AND POWER DISTRIBUTION 16.14.4 Control Rod Drive Patch Panels l

I COMITMENT The control rod, drive 9atch panels shall be locked with ;

limited access to be aithorized by the manager or his designated alternate. l l

APPLICABILITY: At all times.

ACTIONS 1

CONDITION REQUIRED ACTION COMPLETION TIME l

A. N/A. A.1 N/A. N/A g SURVEILLANCE REQUIREMENTS U SURVEILLANCE FREQUENCY SR 16.14.4.1 N/A. N/A BASES The requirement (s) of this SLC section were relocated from CTS 3.5.2.7 during the conversion to ITS.

REFERENCES N/A i

i

[

16.14.4-1 03/27/99 1

l

Penetration Room Ventilation Room System Testing 16.15.1 ,

16.15 VENTILATION FILTER TESTING PROGRAM 16.15.1 Penetration Room Ventilation Room System Testing i COMITNENT Perform specified Surveillance Requirements. i 1

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

i I

ACTIONS 1 i

CONDITION REQUIRED ACTION COMPLETION TIME l l

A. N/A. A.1 N/A. N/A i l

1 l

SURVEILLANCE REQUIREMENTS i

SURVEILLANCE FREQUENCY SR 16.15.1.1- Verify carbon sample removed from the 31 days after Penetration Room Ventilation System Filters removal of provide a: 90% radioactive methyl iodide carbon sample removal when tested in accordance with ASTM D3803-1989 (30*C, and 95% R.H.).

SR 16.15.1.2 Verify each Penetration Room Ventilation ------NOTE-----

System fan when tested in accordance with The provisions ANSI N510-1975 operates at system design of SLC 16.2.7 flow (i 10%). do not apply.

I 18 months +25%

l (continued) 4

!O 16.15.1-1 03/E7/99

1 1

Penetration Room Ventilation Room System Testing 16.15.1 i

SURVEILLANCE FREQUENCY SR 16.15.1.3 Verify Penetration Room Ventilation System ------NOTE-----

HEPA filters provide a: 99% DOP removal when The provisions tested in accordance with ANSI N510-1975 at of SLC 16.2.7 system design flow (i 10%). do not apply.

18 months +25%

AND J i

Once after each !

complete or i partial-replacement of a HEPA filter bank AND

(T Once after any

\- / structural maintenance on .

I the system housing AND i Once after painting, fire, )

or chemical release in any I ventilation zone communicating with the system (continued) 16.15.1-2 03/27/99

l Penetration Room Ventilation Room System Testing 16.15.1 SURVEILLANCE FREQUENCY SR 16.15.1.4 Verify Penetration Room Ventilation System ------NOTE-----

charcoal adsorber filters provide a: 99% The provisions halogenated hydrocarbon removal when tested of ^LC 16.2.7 in accordance with ANSI N510-1975 at system do not apply.

design flow (i 10%). ---------------

18 months +25%

AND Once after each complete or partial replacement of a charc.;al adsorber bank I AND O Once after any structural maintenance on the system housing AND Once after painting, fire, cr chemical release in any ventilation zone communicating with the system (continued) r d

v 16.15.1-3 03/27/99

1

( '

Penetration Room Ventilation Room System Testing 16.15.1 i' r

SURVEILLANCE FREQUENCY SR 16.15.1.5 Remove carbon saaples from Penetration Room ------NOTE-----

Ventilation System for laboratory analysis. The provisions ,

of SLC 16.2.7 )

do not apply.

18 months +25% l l

ANQ Once after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation AND ;

i Once after !

painting, fire !

or chemical l release in any O ventilation zone i

communicating with the system SR 16.15.1.6 Verify pressure drop across the combined ------NOTE-----

HEPA filters and carbon adsorber banks is The provisions

< 6 inches of water when tested in of SLC 16.2.7 accordance with ANSI N510-1975 at system do not apply, design flow (i 10%). ---------------

18 months +25%

BASES The requirement (s) of a portion of this SLC section were relocated from CTS 4.5.4.1.b.1, 4.5.4.1.c and 4.5.4.1.e during the conversion to ITS. This SLC 1 e.lso includes the Ventilation Filter Testing Program requirements for the Penetration Room Ventilation System specified in ITS 5.5.12. Ventilation O Filter Testing Program.

1 l

16.15.1-4 03/27/99 l l' - _ . - . . __ _ _ _ _ - __ _.!

._ _ ____- _.. _ _ _.-_ __ ._... _ _._ _ _ . _ ..._ _ ._ ____ _. _._ - . _ .m.

I l

i Penetration Room Ventilation Room System Testing 16.15.1 l Operation of. the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers. If the l performance is as specified, the calculated doses should be less than the guidelines stated in 10 CFR 100 for the accidents analyzed.

! (HEPA) filters are installed before the charcoal adrorbers to prevent clogging of the iodine adsorbers. The charcoal adsorbers are installed to reduce the potential release of radiciodine. Bypass leakage for the charcoal adsorbers and particulate removal efficiency for HEPA filters are determined by .

l halogenated hydrocarbon and DOP respectively. The laboratory carbon sample l test results indicate a radioactive methyl iodide removal efficiency for expected accident conditions.

The frequency of tests and sarole analysis are necessary to show that the HEPA l filters and charcoal adsorbers can perform as evaluated. Replacement adsor- i

, bent should be qualified according to the guidelines of Regulstory Guide 1.52.

The charcoal adsorber efficiency test procedures should allow for the removal i l of one adsorber tray, emptying of one bed from the tray, mixing the adsorbent '

L thoroughly and obtaining at least two samples. Each sample should be replaced.

Any HEPA filters found defective should be replaced with filters qualified pursuant to Regulatory Position C.3.d 'of Regulatory Guide 1.52.

, If painting, fire or chemical release occurs during system operation such that I the HEPA filter or charcoal adsorber could become contaminated from the fumes, l chemicals or foreign materials, the same tests and sample analysis should be l performod as required for operational use.

! REFERENCES

1. Regulatory Guide 1.52, Rev. 2.

2. ITS 5.5.12, Ventilation Filter Testing Program.

L i

!O 1

16.15.1-5 03/27/99

Control Room Pressurizatien and Filtering System '

16.15.2 16.15 VENTILATION FILTER TESTING PROGRAM 16.15.2 Control Room Pressurization and Filtering System )

COMITMENT Perform specified Surveillance Requirements.

APPLICABILITY: MODES 1, 2, 3, and 4. I ACTIOhS CONDITION REQUIRED ACTION COMPLETION TIME l

A. N/A. A.1.1 N/A. N/A.

l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

/3 U SR 16.15.2.1 Verify carbon sample removed from the 31 days after Control Room Booster Fan train Filters removal of provide a 90% radioactive methyl iodide carbon sample reri. oval when tested in accordance with ASTM D3803-1989 (30*C, and 95% R.H.).

SR 16.15.2.2 Verify pressure drop across pre-filters is 92 days s 1 inch of water and pressure drop across HEPA filters is s 2 inches of water when tested in accordance with ANSI N510-1975 at system design flow ( 10%).

(continued)

~

O 16.15.2 1 03/27/99

r ntrol Room Pressurization and Filtering System 16.15.2 SURVEILLANCE REQUIRENENTS (continued)

SURVEILLANCE FREQUENCY SR 16.15.2.3 Verify Control Room Booster Fan train HEPA ------NOTE-----

filters provide a: 99.5% DOP removal when The provisions tested in accordance with ANSI N510-1975 at of SLC 16.2.7 system design flow i (10%). do not apply.

18 months +25%

AND Once after each complete or partial replacement of a HEPA filter bank AND Once after any ,

structural maintenance on the system housing AND Once after painti.ng, fire, or chemical release in any ventilation zone communicating with the system (continued)

. O 16.15.2-2 03/27/99

i l

l Control Room Pressurization and Filtering System

16.15.2 l . l SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY l

SR 16.15.2.4 Verify Control Room Booster Fan train ------NOTE-----

charcoal adsorber filters provide 2 99% The provisions halogenated hydrocarbon removal when tested of SLC 16.7. 7 in accordance with ANSI N510-1975 at system do not apply, design flow i (10%). ---------------

i 18 months +25% i l

AND l

Once after each 1 complete or i partial replacement of a charcoal adsorber bank AND O Once after any structural i

maintenance on the system housing l

AND Once after painting, fire, or chemical release in any ventilation zone communicating with the system i

i

!O t

16.15.2-3 03/27/99

1 I

, Control Room Pressurization and Filtering System l 16.15.2 O

4 4

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY ;

1 SR 16.15.2.5 Remove carbon samples from Control Room ------NOTE-----

Booster Fan trains for laboratory analysis The provisions of SLC 16.2.7 do not apply.

18 months +25% 1 AND Once after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation AND Once after painting, fire O or chemical relcase in any ventilation zone communicating with the system BASES The requirement (s) of this SLC section were ralocated from CTS 4.12 during the conversion to ITS. This SLC also includes the Ventilation Filter Testing Program requirements for the Control Room Booster Fan train filters specified in ITS 5.5.12, Ventilation Filter Testing Program.

The purpose of the control room pressurization filtering system is to protect the control room operators from the effects of accidental release of radioactive effluents or toxic gases in the Turbine Building or Auxiliary Building only. The system is designed with two 50 percent capacity filter trains each of which consists of a prefilter, high efficiency particulate filters, carbon filters, booster fans, air handling unit fans, and essociated ductwork to pressurize the control room with outside air.

Since these systems are not normally operated, a periodic test is required to

( insure their operability when needed. Quarterly testing of this system will show that the system is available.

16.15.2-4 03/27/99

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

Control Room Pressurization and Filtering System 16.15.2 O

Refueling frequency testing of the installed carbon adsorber stage and absolute filters will verify the leak integrity of the cleanup system.

REFERENCES N/A O

i l

O '

16.15.2-5 03/27/99

Spent Fuel Pool Ventilation System 16.15.3

( 16.15 VENTILATION FILTER TESTING PROGRAM 16.15.3 Spent Fuel Pool Ventilation System COMITHENT Perform specified Surveillanca Requirements APPLICABILITY: During movement of fuel within the spent fuel storage pool.

During crane operations with loads over the spent fuel storage pools.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. N/A A.1 N/A N/A ,

l SURVEILLANCE REQUIREMENTS :

SURVEILLANCE FREQUENCY

'V SR 16.15.3.1 Verify carbon sample removed from the 31 days after Reactor Building Purge Filters provide removal of ,

a 90% radioactive methyl iodide removal carbon sample l when tested in accordance with ASTM ;

D3803-1989 (30*C, and 95% R.H.).

1 SR 16.15.3.2 Verify each Spent Fuel Ventilation fan ------NOTE-----

operates at design flow (i 10%) when tested The provisions in accordance with ANSI N510-1975. of SLC 16.2.7 do not apply.

18 months +25%

(continued)

[v 16.15.3-1 03/27/99 t

Spent Fuel Pool Ventilation System 16.15.3 l

SURVEILLANCE FREQUENCY SR 16.15.3.3 Remove carbon sample from Reactor Building ------NOTE-----

Purge Filters for laboratory analysis. The provisions of SLC 16.2.7 do not apply.

18 months +25%

/,ND Once after 720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months of system operation i AND Once after painting, fire, or chemical release in any ventilation O

zone communicating with the System (continued)

)

O 16.15.3-2 03/27/99

. - . . . - . - .- .- . - . . . . . = . - - - - _ . - - - . - . . - - _ _ . ~- - .-.- . -.

i l

Spent Fuel Pool Ventilation System n 16.15.3

'%A SURVEILLANCE FREQUENCY 3R 16.15.3.4 Verify Reactor Building purge HEPA filters ------NOTE----- )

provide a 99% D0P removal when tested in The provisions

, acccrdance with ANSI N510-1975 at design of SLC 16.2.7 flow (i 10%). do not apply.

18 months +25%

AND Once after each complete or partial replacement of HEPA filter bank j AND !

l N Once after any l -

structural maintenance on the system

! housing l AND 1 Once after

! painting, fire, or chemical release in any vertilation zone com1unicating with the system (continued) e i

l iO

U f

j

. 16.15.3-3 03/27/99

l Spent Fuel Pool Veatilation System 16.15.3 v

SURVEILLANCE FREQUENCY SR 16.15.3.5 Verify Reactor Building purge charcoal ------NOTE----- l adsorber filters provide 2: 99% halogenated The provisions hydrocarbon removal when tested in of SLC 16.2.7 accordance with ANSI N510-1975 at design do not apply.

flow (i 10%). ---------------

18 months +25%

AND Once after each complete or i partial replacement of charcoal adsorber bank ;

l AND !

l (s Once after any U structural maintenance on l

the system i housing AND t

Once after

! painting, fire, l

or chemical

( release in any ventilation zone communicating with the system i BASES The requirement (s) of this SLC section were relocated from CTS 4.14 during the conversion to ITS. This SLC also includes the Ventilation Filter Testing

Program requirements for the Spent Fuel Pool Ventilation System specified in k ITS 5.5.12, Ventilation Filter Testing Program.

16.15.3-4 03/27/99

J-Spent fuel Pool Ventilation System 16.15.3 1

} The offsite doses for the fuel handling accident are within the guidelines of

10 CFR 100; however, to further reduce the doses resulting from this accident, j it is required that the spent fuel pool ventilation system be operable when-ever the possibility of a fuel handling accident could exist.

l The Unit 2 Reactor Building purge filter is used in the ventilation system for i the common spent fuel pool for Units 1 and 2. The Unit 3 Reactor Building

! purge filter is used in the Unit 3 spent fuel pool ventilation system. Each j filter is constructed with a prefilter, an absolute filter and a charcoal filter in series. The high efficiency particulate air (HEPA) filters are

! installed before the charcoal adsorbers to prevent clogging of the iodine l adsorbers. The charcoal adsorbers are installed to reduce the potential i release of radiciodine.

! Bypass leakage for the charcoal adsorbers and particulate removal efficiency

for HEPA filtars are determined by halogenated hydrocarbon and DOP respec-

i tively. The laboratory carbon sample test results indicate a radioactive i methyl iodide removal efficiency for expected accident conditions. Operation of the fans significantly different from the design flow will change the removal efficiency of the HEPA filters and charcoal adsorbers. If the per-j formances are as specified, the doses for a fuel handling accident would be ,

j minimized.

4

! The frequency of tests and sample analysis are necessary to show that the HEPA i . filters and charcoal adsorbers can perform as evaluated. Replacement

adsorbent should be qualified according to the guidelines of Regulatory Guide

! 1.52. The charcoal adsorber efficiency test procedures should allow for the l removal of one adsorber tray, emptying of one bed from the tray, mixing the d

adsorbent thoroughly and obtaining at least two samples. Each sample should

. be replaced. Any HEPA filters found defective should be replaced with filters j

qualified pursuant to Regulatory Position C.3.d of Regulatory Guide 1.52. .

Operation of the spent fuel pool ventilation System every month will demon-strate aperability of the fans, filters and adsorber system.

i If painting, fire or chemical release occurs during system operation such that the HEPA filter or charcoal adsorber could become contaminated from the fumes,

chemicals or foreign materials, the same tests and sample analysis should be

< performed as required for operational use.

l l REFERENCES

1. Regulatory Guide 1.52, Rev. 2.

2. ITS 5.5.12, Ventilation Filter Testing Program O

16.15.3-5 03/27/99 Q .,

ML20204B414

Contents

Text

16.1 INTRODUCTION

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6.2 REFERENCES

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7.7 REFERENCES

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11.2 REFERENCES

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ML20204B414

Text

16.1 INTRODUCTION

REFERENCES:

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6.2 REFERENCES

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7.7 REFERENCES

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11.2 REFERENCES

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