Text

Revised CNS Selected Licensee Commitments Manual
	ML20199H406
Person / Time
Site:	Catawba
Issue date:	01/16/1999
From:	; DUKE POWER CO.
To:	;
Shared Package
ML20199H404	List: ML20199H399; ML20199H406;
References
	PROC-990116, NUDOCS 9901250201
	Download: ML20199H406 (300)
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CATAWCA NUCLEAR STATION SELECTED LICENSEE COMMITMENTS MANUAL Tab 16.9 (continued) 16.9-6 Page 4 of 12 01/16/99 16.9-6 Page 5 of 12 01/16/99 16.9-6 Page 6 of 12 01/16/99 16.9-6 Page 7 of 12 01/16/99 16.9-6 Page 8 of 12 01/16/99 16.9-6 Page 9 of 12 01/16/99 16.9-6 Page 10 of 12 01/16/99 16.9-6 Page 11 of 12 01/16/99 16.9-6 Page 12 of 12 01/16/99 16.9-7 Page 1 of 2 01/16/99 16.9-7 Page 2 of 2 01/16/99 16.9-8 Page 1 of 2 01/16/99 16.9-8 Page 2 of 2 01/16/99 16.9-9 Pege 1 of 2 01/16/99 16.9-9 Page 2 of 2 01/16/99 16.9-10 Page 1 of 1 01/16/99 16.911 Page 1 of 3 01/16/99 16.9-11 Page 2 of 3 01/16/99 16.9-11 Page 3 of 3 01/16/99 16.9-12 Page 1 of 3 01/16/99 2 16.9-12 Page 2 of 3 01/16/99 16.9-12 Page 3 of 3 01/16/99 16.9-13 Page 1 of 10 01/16/99 16.9-13 Page 2 of 10 01/16/99 3 16.9-13 Page 3 of 10 01/16/99 1 16.9-13 Page 4 of 10 01/16/99 16.9-13 Page 5 of 10 01/16/99 16.9-13 Page 6 of 10 01/16/99 16.9-13 Page 7 of 10 01/16/99 16.9-13 Page 8 of 10 01/16/99 16.9-13 Page 9 of 10 01/16/99 l 16.9-13 Page 10 of 10 01/16/99 !

16.9-14 Page 1 of 2 01/16/99 l 16.9-14 Page 2 of 2 01/16/99 l 16.9-15 Page 1 of 1 01/16/99 16.9-16 Page 1 of 1 01/16/99 16.9-17 Page 1 of 1 01/16/99 16.9-18 Page 1 of 1 01/16/99 16.9-19 Page 1 of 2 01/16/99 4 16.9-19 Page 2 of 2 01/16/99 16.9-20 Page 1 of 1 01/16/99 Q 16.9-21 Page 1 of 1 01/16/99 b) 16.9-22 Page 1 of 2 01/16/99 )

l Page 5 of 8 01/16/99 l

CATAW2A NUCLEAR STATION SELECTED LICENSEE COMMITMENTS MANUAL Tab 16.9 (continued) 16.9-22 Page 2 of 2 01/16/99 16.9-23 Page 1 of 3 05/04/98 16.9-23 Page 2 of 3 05/04/98 l 16.9-23 Page 3 of 3 05/04/98 Tab 16.10 i

16.10-1 Page 1 of 2 01/16/99 16.10-1 Page 2 of 2 01/16/99 16.10-2 Page 1 of 2 01/15/97 !

16.10-2 Page 2 of 2 01/15/97 Tab 16.11 16.11-1 Page 1 of 7 01/16/99 16.11-1 Page 2 of 7 01/16/99 16.11-1 Page 3 of 7 01/16/99 16.11-1 Page 4 of 7 01/16/99 16.11-1 Page 5 of 7 01/16/99 l h

V 16.11-1 Page 6 of 7 16.11-1 Page 7 of 7 01/16/99 01/16/99 I

16.11-2 Page 1 of 6 01/16/99 16.11-2 Page 2 of 6 01/16/99 16.11-2 Page 3 of 6 01/16/99 16.11-2 Page 4 of 6 01/16/99 16.11-2 Page 5 of 6 01/16/99 16.11-2 Page 6 of 6 01/16/99 16.11-3 Page 1 of 2 01/16/99 16.11-3 Page 2 of 2 01/16/99 16.11-4 Page 1 of 2 01/16/99 16.11-4 Page 2 of 2 01/16/99 16.11-5 Page 1 of 3 08/01/94 16.11-5 Page 2 of 3 08/01/94 16.11-5 Page 3 of 3 08/01/94 16.11-6 Page 1 of 7 01/16/99 16.11-6 Page 2 of 7 01/16/99 16.11-6 Page 3 of 7 01/16/99 16.11-6 Page 4 of 7 01/16/99 16.11-6 Page 5 of 7 01/16/99 16.11-6 Page 6 of 7 01/16/99 16.11-6 Page 7 of 7 01/16/99 16.11-7 Page 1 of 9 01/16/99

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CATAWBA NUCLEAR STATION SELECTED LICENSEE COMMITMENTS MANUAL g

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) .

List of Effective Pages Tab 16.11 (continued) 16.11-7 Page 3 of 9 01/16/99 16.11-7 Page 4 of 9 01/16/99 16.11-7 Page 5 of 9 01/16/99 16.11-7 Page 6 of 9 01/16/99 16.11-7 Page 7 of 9 01/16/99 16.11-7 Page 8 of 9 01/16/99 16.11-7 Page 9 of 9 01/16/99 16.11-8 Page 1 of 2 01/16/99 16.11-8 Page 2 of 2 01/16/99 16.11-9 Page 1 of 2 01/16/99 16.11-9 Page 2 of 2 01/16/99 16.11-10 Page 1 of 2 01/16/99 16.11-10 Page 2 of 2 01/16/99 16.11-11 Page 1 of 2 01/16/99 16.11-11 Page 2 of 2 01/16/99 16.11-12 Page 1 of 2 01/16/99 16.11-12 Page 2 of 2 01/16/99 16.11-13 Page 1 of 14 01/16/99

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CATAWBA NUCLEAR STATION SELECTED LICENSEE COMMITMENTS MANUAL Tab 16.11 (continued) 16.11-19 Page 2 of 2 01/16/99 16.11-20 Page 1 of 3 01/16/99 16,11-20 Page 2 of 3 01/16/99 16.11-20 Page 3 of 3 01/16/99 16.11-21 Page 1 of 2 01/16/99 16.11-21 Page 2 of 2 01/16/99 Tab 16.12 16.12-1 Page 1 of 2 01/16/99 16.12-1 Page 2 of 2 01/16/99 Tab 16.13 16.13-1 Page 1 of 2 12/90 16.13-1 Page 2 of 2 12/90 16.13-2 Page 1 of 3 01/16/99 !

16.13-2 Page 2 of 3 01/16/99

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16.13-2 Page 3 of 3 01/16/99 16.13-3 Page 1 of 2 01/16/99 l 16.13-3 Page E af 2 01/16/99 l O :

Page 8 of 8 01/16/99 I

CATAWBA NUCLEAR STATION I FINAL SAFETY ANALYSIS REPORT q SELECTED LICENSEE COMMITMENTS I

J Table of Contents 16.0 Selected Licensee Commitments 16.1 introduction 16.2 Applicability 16.3 Definitions 16.4 Commitments Related to Reactor Components 16.5 Commitments Related to Reactor Coolant System i

16.5-1 Mid-Loop Operation with irradiated Fuel in the Core 16.5-2 Safety Valves - Shutdown l

16.5-3 Chemistry l

16.5-4 Pressurizer l

16.5-5 Structuralintegrity l

16.5-6 Reactor Coolant System Vents l

16.5-7 Steam Generator Pressure / Temperature Limitation l 16.6 Commitments Related to Engineered Safety Features 16.6-1 Containment Sump 16.6-2 Ice Bed Temperature Monitoring System l 16.6-3 Inlet Door Position Monitoring System l 16.6-4 Chlorine Detection Systems l 16.7 Commitments Related to Instrumentation t i

16.7-1 ATWS Mitigation System Actuation Circuitry (AMSAC) 16.7-2 Seismic Instrumentation 16.7-3 MeteorologicalInstrumentation 16.7-4 Loose-Part Detection System 16.7-5 Turbine Overspeed Protection 16.7-6 RN Discharge instrumentation

!O 16.7-7 Movable incore Detectors O l Chapter 16.0 Page 1 of 4 01/16/99

CATAW2A NUCLEAR STATION FINAL SAFETY ANALYSIS REPORT SELECTED LICENSEE COMMITMENTS Table of Contents 16.7-8 Groundwater Level l

16.7-9 Standby Shutdown System l

16.7-10 Radiation Monitoring for Plant Operations l

16.7 11 Position Indication System - Shutdown l

16.7-12 Position Indication System - Test Exceptions l

16.8 Commitments Related to Electrical Power Systems 16.8-1 Containment Penetration Conductor Overcurrent Protective Devc' es 16.8-2 230 kV Switchyard Systems 16.8 3 230 kV Switchyard 125 VDC Power System 16.8-4 6900 V Shared Transformers 16.8-5 Diesel Generator Supplemental Testing Requirements l 16.9 Commitments Related to Auxiliary Systems 16.9-1 Fire Suppression Water Systems 16.9-2 Sprayand/or Sprinkler Systems 16.9-3 CO2 Systems 16.9-4 Fire Hose Stations 16.9-5 Fire Barrier Penetrations 16.9-6 Fire Detection Instrumentation 16.9-7 Boration Systems Flow Path - Shutdown l

16.9-8 Boration Systems Flow Paths - Operating l

16.9-9 Boration Systems Charging Pump - Shutdown l

16.9-10 Boration Systems Charging Pumps - Operating l <

16.9-11 Boration Systems Borated Water Source - Shutdown l

16.9-12 Boration Systems Borated Water Sources - Operating l 16.9-13 Snubbers j 16.9-14 Lake Wylie Water Temperature l

Chapter 16.0 Page 2 of 4 01/16/99

CATAWCA NUCLEAR STATION FINAL SAFETY ANALYSIS REPORT SELECTED LICENSEE COMMITMENTS Table of Contents 1 C io Auxiliary Building Filtered Exhaust System Filter Cooling Bypass Valves I f 16.9-16 Fuel Handling Ventilation Exhaust System Filter Cooling Bypass Valves l 16.9-17 Refueling Operations - Decay Time l f 16.9-18 Refueling Operations - Communications '

l 16.9-19 Refueling Operations - Manipulator Crane l ,

16.9-20 Refueling Operations - Crane Travel - Spent Fuel Storage Pool Building l 16.9-21 Refueling Operations - Storage Pool Water Level l 16.9-22 Control Room Area Ventilation System - Intake Alarms l 16.9-23 Fire Hydrants 16.10 Commitments Related to Steam and Power Conversion Systems 16.10-1 Steam Vent to Atmosphere 16.10-2 Condenser Circulating Water System ,

(/ 16.11 Commitments Related to Radioactive Waste Management ,

16.11-1 Liquid Effluents l 16.11 2 Radioactive Liquid Effluent Monitoring instrumentation 16.11-3 Dose i 16.11-4 Liquid Radwaste Treatment System -

16.11 5 ChemicalTreatment Ponds 16.11-6 Gaseous Effluents 16.11-7 Radioactive Gaseous Effluent Monitoring instrumentation 16.11-8 Dose - Noble Gases 16.11-9 Dose - lodine-131, lodine-133, Tritium, and Radioactive Material in Particulate Form 16.11-10 Gaseous Radwaste Treatment System 16.11 11 Solid Radioactive Wastes '

16.11-12 Total Dose 16.11-13 Monitoring Program Chapter 16.0 Page 3 of 4 01/16/99

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

CATAWCA NUCLEAR STATION FINAL SAFETY ANA1 VSIS REPORT

'f] SELECTED LICENSEE COMMITMENTS l NJ Table of Contents 16.11-14 Land Use Census 16.11-15 Interlaboratory Comparison Program l

16.11-16 Annual Radiological Environmental Operating Report and Annual Radioactive i

Effluent Release Report

-16.11 17 Liquid Holdup Tanks l

16.11-16 Explosive Gas Mixture l I 16.11 19 Gas Storage Tanks l

16.11-20 Explosive Gas Monitoring Instrumentation l

16.11-21 Major Changes to Liquid, Gaseous, and Solid Radwaste Treatment Systems l 16.12 Commitments Related to Radiation Protection 16.12-1 Sealed Source Contamination l 16.13 Commitments Related to Conduct of Operations 16.13-1 Fire Brigade 16.13-2 Technical Reviewand Control 16.13-3 Plant Operations Review Committee O

Chapter 16.0 Page 4 of 4 01/16/99

1.6A SELECTED LICENSEE COMMITMENTS

16.1 INTRODUCTION

i l

1 l

. COMMITMENT:

3 This chapter provides a single location in the UFSAR where certain selected

' l licensee commitments (SLC) are presented. The content of this chapter is based on the results of application of a set of criteria to detemline the content of technical specifications. For purposes of administrative ease, this chapter is !

maintained in a separate manual, The Catawba Nuclear Station Selected Licensee Commitment Manual. Those previous technical specification l requirements which did not meet the criteria are relocated in this chapter. 1 Catawba Technical Specification 5.4 (Procedures) requires written procedures to l be established, implemented, and maintained on those selected licensee ,

commitments. '

I The control of the Catawba Nuclear Station SLC program and manual shall be accordance with Section 3.8 of the Compliance Functional Area Manual, l

" Selected Licensee Commitments", and is included in this manual as the '

Appendix. The manualis officially designated as Chapter 16 of the Catawba i UFSAR. The originalissue and subsequent revisions of the manual are l

( approved by the station manager. Administrative requirements of the manual l

are the responsibility of the site Regulatory Compliance section.

Changes to these SLC may be made, pursuant to 10 CFR 50.59, only after the j bases for the requirement have been clearly established and after a multi- !

disciplinary review by qualified reviewers, including onsite operation's personnel l

(52FR3788, February 6,1987, Interim Policy Statement on Technical Specification improvements for Nuclear Power Plants).

Additional operational related 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 FSAR, but are n.c.to of such a level of importance that they need to be under the rigorous control provided by technical specifications.

This chapter includes testing requirements for certain systems, and remedial actions to be taken in the event the system is not fully capable of performing its design function. A bases for the commitment is also provided. Reference is also provided to specific sections of the UFSAR where the information relative to the l commitment is further described.

Chapter 16.1 Page 1 of 1 01/16/99 I

l

16.2 APPLICABILITY This section provides the general requirements applicable to each of the COMMITMENTS and Testing requirements within Section 16.0, Selected Licensee Commitments (SLCs).

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 COMMITMENT, the associated l REMEDIAL ACTION (S) shall be met. If the COMMITMENT is met or is !

no longer applicable prior to expiration of the specified time interval, completion of the REMEDIAL ACTION (S)is not required, unless l 1

otherwise stated.

16.2.3 When a COMMITMENT is not met, entry into an OPERATIONAL MODE or other specified condition in the Applicability shall not be made ;

except when the associated REMEDIAL ACTIONS to be entered permit '

continued operation in the MODE or other specified condition in the Applicability for an unlimited period of time. This COMMITMENT shall not prevent changes in OPERATIONAL MODES or other specified O conditions in the Applicability that are required to comply with REMEDIAL ACTIONS. Exceptions to this COMMITMENT are stated in the individual COMMITMENTS.

16.2.4 COMMITMENTS including the associated REMEDIAL ACTIONS shall apply to each unit individually unless otherwise indicated as follows:

a. Whenever the COMMITMENT refers to systems or components which are shared by both units, the REMEDIAL ACTIONS will apply to both units simultaneously. This will be indicated in the REMEDIAL ACTIONS;

b. Whenever the COMMITMENT applies to only one unit, this will be identified in the APPLICABILITY section of the COMMITMENT; and

c. Whenever certain portions of a COMMITMENT contain operating parameters, setpoints, etc., which are different for each unit, this will be identified in parentheses or footnotes. (for example, "... flow rate of 54,000 cfm (Unit 1) or 43,000 cfm (Unit

2) . . .") .

16.2.5 O Testing Requirements shall be met during the OPERATIONAL MODES or other specified conditions in the Applicability for individual COMMITMENTS unless otherwise stated in an individual Testing Requirement or Reference. Failure to meet a Testing Requirement, Chapter 16.2 Page 1 of 2 01/16/99

f)

V whether such failure is experienced during the performance of the Testing Requirement or between performances of the Testing Requirement, shall be failure to meet the COMMITMENT. Failure to perform a Testing Requirement within the specified Frequency shall be failure to meet the COMMITMENT except as provided in COMMITMENT 16.2.7. Testing Requirements do not have to be performed on inoperable equipment or variables outside specified limits.

16.2.6 Each Testing Requirement shall be performed on its specified frequency with a maximum allowable extension not to exceed 25% of the test frequency. The phrase "at least" associated with a testing frequency does not negate this tolerance value, and permits the performance of more frequent testing activities. This tolerance is i

necessary to provide operational flexibility because of scheduling and performance considerations.

16.2.7 If it is discovered that a Testing Requirement was not performed within its specified Frequency, then compliance with the requirement to declare the COMMITMENT not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or up to the limit of the specified Frequency, whichever is less. This delay period is permitted to allow performance r of the Testing Requirement.

C]'

If the Testing Requirement is not performed within the delay period, the COMMITMENT must immediately be declared not met, and the applicable REMEDIAL ACTIONS must be entered.

When the Testing Requirement is performed within the delay period and the Testing Requirement is not met, the COMMITMENT must immediately be declared not met, and the applicable REMEDIAL ACTIONS must be entered.

16.2.8 Entry into an OPERATIONAL MODE or other specified condition in the Applicability of a COMMITMENT shall not be made unless the COMMITMENT'S Testing Requirement (s) have been met within their specified frequency. This provision shall not prevent entry into OPERATIONAL MODES or other specified conditions in the Applicability that are required to comply with REMEDIAL ACTIONS.

, 16.2.9 Testing Requirements shall apply to each unit individually ur. ess otherwise indicated as stated in Section 16.2.4 for individual COMMITMENTS or whenever certain portions of a COMMITMENT contain testing parameters different for each unit, which will be

,e identified in parentheses or footnotes.

O Chapter 16.2 Page 2 of 2 01/16/99

l 16.3 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Selected Licensee Commitments. The definitions in the Catawba Technical Specifications apply to the defined terms used herein. The following l

additional defined terms appear in capitalized type and are applicable throughout t this Selected Licensee Commitment document:

ATWS MITIGATION SYSTEM ACTIVATION CIRCUlTRY (AMSAC)

AMSAC is the Westinghouse System for mitigating ATWS events.

ANTICIPATED TRANSIENT WITHOUT SCRAM (ATWS)

An ATWS is an expected operational transient (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power) which is accompanied by a failure of the reactor trip system to shut down the reactor.

COMMITMENT A COMMITMENT is a method of ensuring the lowest functional capability or O, performance levels of equipment which are important to the safety of the facility but are not of such a level of importance that they need to be under the rigorous control !

provided by Technical Specifications.

FREQUENCY NOTATION ,

The FREQUENCY NOTATION specified for the performance of Surveillance Requirements shall correspond to the intervals defined in Table 16.3-1.

i MEMBER (S) OF THE PUBLIC MEMBER (S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the licensee, its contractors, or vendors. Also excluded from this category are persons who enter the site to service equipment or to make deliveries. This category does include persons who use portions of the site for recreational, occupational, or other purposes not associated with the plant.

PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM (PCP) shall contain the current formulas, sampling, analyses, tests, and determinations to be made to ensure that processing O

v 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 Chapter 16.3 Page 1 of 4 01/16/99

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

r assure compliance with 10 CFR Parts 20,61, and 71 and Federal and State !

regulations, burial ground requirements, and other requirements goveming the disposal of radioactive waste.

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

REMEDIAL ACTION REMEDIAL ACTION shall be that part of a Selected Licensee Commitment which prescribes remedial measures required under designated conditions.

i SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned, nor leased, nor otherwise controlled by licensee. 1 i

SOURCE CHECK 0%

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

UNRESTRICTED AREA 1 i

An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY '

access to which is not controlled by the licensee for purposes of protection of )

individuals from exposure to radiation and radioactive materials, or any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, and/or recreational purposes.

VENTILATION EXHAUST TREATMENT SYSTEM A VENTILATION EXHAUST TREATMENT SYSTEM shall be any system designed and installed to reduce gaseous radioiodine or radioactive material in particulate form in effluents by passing ventilation or vent exhaust gases through activated carbon adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents. Engineered Safety Features (ESF) Atmospheric Cleanup Systems are not considered to be VENTILATION EXHAUST TREATMENT SYSTEM components.

I Chapter 16.3 Page 2 of 4 01/16/99

_ A

VENTING VENTING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that replacement air or gas is not provided or required during VENTING. Vent, used in system names, does not imply a )

VENTING process. '

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

O 1

Chapter 16.3 Page 3 of 4 01/16/99 1

I

TABLE 16.3-1 1

FREQUENCY NOTATION 1

NOTADON FREQUENCY S At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

W At least once per 7 days.

1 M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 184 days.

R At least once per 18 months.

S/U Prior to each reactor startup.

O N.A. Not applicable.

l P Completed prior to each release.

O Chapter 16.3 Page 4 of 4 01/16/99 ,

I 16.5 REACTOR COOLANT SYSTEM 16.5-1 MID-LOOP OPERATION WITH IRRADIATED FUEL IN THE CORE COMMITMENT _:

Operations with Reactor Coolant (NC) system levels less than or equal to 16% with fuelin the core shall be conducted under the following conditions:

1) At least one hot leg will be maintained with no S/G nozzle dam installed until the reactor vessel head has been removed. I

2) If S/G nozzle dams are to be used, one hot leg dam and a hot or cold leg manway on the associated S/G shall remain out anytime the reactor vessel head is in place. If a cold leg manway is being used, then all co!d leg nozzle dams must be installed.

3) Two independent trains of NC levelinstruments are required. These instruments shall have independent transmitters and shall not include the NC System sightglass (NCLG-6450) or tygon tubing.

O O 4) Two core exit thermocouples shall be maintained operating with temporary high alarms set at 140 F and monitored except as noted below: l

. Final disconnection of the last two core exit thermocouples shall occur no sooner than two hours prior to reactor vessel head removal.

. Reconnection of at least two thermocouples within two hours after reinstalling the reactor vessel head.

. The total time without thermocouple indication shall not exceed 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

5) Three power sources shall be available as follows:

. Two off-site power sources and one D/G, or

. One off-site power source and two D/Gs.

6) Two independent makeup paths of borated water shall be available, during each of the following conditions a) Reactor Coolant System intact:

e l b

Chapter 16.5-1 Page 1 of 4 01/16/99

COMMITMENT (con't)

[] '

One Centrifugal Charging Pump (NV) as required per SLC 16.9-7 and 16.9-9.

One Safety injection Pump (NI) having its breaker installed in its ,

associated cubicle and flow path available from the FWST to the NC System.

b) Reactor Coolant System open to Containment atmosphere via a hot leg vent path: !

One Centrifugal Charging Pump (NV) as required per SLC 16.9-7 and 16.9-9.

. One of the following gravity flowpaths:

1. FWST through ND-33 to the cold legs via NI-173A and/or NI- ,

1788. '

2. FWST through the ND suction lines to the hotlegs.

3. FWST through ND-33 to the hotlegs via Nl-183B.

/ NOTE: The number of open containment penetrations is limited such that the k penetrations can be closed within two hours of losing ND.

7) Containment Closure must be established. Containment Closure is verified by the performance of PT/1/(2)/A/4200/02C-1, Containment Closure Verification, with penetrations not verified acceptable, administratively controlled per OP/0/A/6100/14, Penetration Control During Modes 6 and 6. .

8) The reactcr has been subcritical for at least 7 days; or Design Engineering '

has provided a required subcritical time based on plant operating history and actual reduced NCS level.

APPLICABILITY:

Whenever irradiated fuel is in the reactor vessel and NC System wide range level is less than or equal to 16%.

REMEDIAL ACTION: 1 a) If the primary method of monitoring core exit thermocouples is unavailable then the backup means shall be used to check core exit temperatures. The i backup means is the use of the incore Instrumental Pane!. ;

O ,

Chapter 16.5-1 Page 2 of 4 01/16/99

___ _ _ _ . , _ . ._ -. . ___ .___ m. _ _

1

( REMEDIAL ACTION (con't)

The thermocouple temperatures on the incore Instrument Panel are to be

periodically checked and recorded by an operator in the control room at no

{ greater that 15 minutes.

a j b) If any of the above commitments cannot be met during the time that the j reactor vessel is in a reduced inventory condition, take immediate corrective actions to bring the plant into compliance with the COMMITMENT and i contact the Station Manager and/or responsible Group Superintendent for additional guidance.

t

TESTING REQUIREMENTS

None

REFERENCES:

1) Generic Letter 88-17 (Loss of Decay Heat Removal)

2) NUREG 1410 (Loss of Vital AC Power and Residual Heat Removal during Mid-Loop Operation at Vogtle) .

/'

3) Catawba Nuclear Station Directive 3.1.30 (Mid-Loop Operation) r i

4) OP/1(2)/A/6150/06 (Draining the Reactor Coolant System) i

, 5) Catawba Nuclear Station Technical Specifications 2

6) Catawba Nuclear Station Technical Specification Interpretations

, 7) Oconee Nuclear Station Selected Licensee Commitment 16.5.3

8) Integrated Scheduling Management Procedure 3.1 (Outage Planning and

Execution Responsibilities)

9) Catawba Nuclear Station responses to Generic Letter 88-17 dated January 3, 1989 BASES:

Generic Letter 88-17 and NUREG 1410 involve concems associated with a loss of Residual Heat Removal during NC System reduced inventory. Numerous events have occurred in the industry that resulted in a loss of residual heat removal during O reduced inventory operation. This is of great concern due to the potential for substantial core damage occurring in a relatively short time period.

Chapter 16.5-1 Page 3 of 4 01/16/99

BASES (con't)

This Selected Licensee Commitment depicts those commitments which are !

extremely important to nuclear safety, however, are not presently covered by i' Technical Specifications.

P h

i O :

I

)

l O

Chapter 16.5-1 Page 4 of 4 01/16/99

[

jjd REACTOR COOLANT SYSTEM

$ 313-2

. SAFETY VALVES - SHUTDOWN l

COMMITMENT:

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

)

APPLICABILITY:

MODES 4 and 5. I REMEDIAL ACTION:

I With no pressurizer Code safety valve OPERABLE, immediately suspend all i operations involving positive reactivity changes and place an OPERABLE residual !

heat removal loop into operation in the shutdown cooling mode. J l

TESTING REQUIREMENTS-Verify each pressurizer code safety valve is OPERABLE in accordance with the Inservice Testing Program. :

REFERENCES:

1) Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998. l BASES:

1 The pressurizer Code safety valves operate to prevent the Reactor Coolant System from being pressurized above its Safety Limit of 2735 psig. Each safety valve is l

L designed to relieve 420,000 lbs per hour of saturated steam at the valve setpoint. '

The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown. In the event that no safety valves are OPERABLE, an operating residual heat removal loop, connected to the Reactor ,

Coolant System, provides overpressure relief capability and will prevent j overpressurization in addition, the Overpressure Protection System provides a

~ diverse means of protection against overpressurization at low temperatures.

/ ' The lift setting pressure shall correspond to ambient conditions of the valve at nominal operating

' temperature and pressure.

Chapter 16.5-2 Page 1 of 2 01/16/99

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BASES (con't) f Demonstration of the safety valves' lift settings will occur only during shutdown and will be pedormed in accordance with the provisions of Section XI of the ASME Boiler

and Pressure Vessel Code.

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O Chapter 16.5-2 Page 2 of 2 01/16/99

( ,14 REACTOR COOLANT SYSTE.M 16.5-3 CHEMISTRY COMMITMENT:

[

The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 16.5-3A.

APPLICABILITY: '

At all times.

REMEDIAL ACTION: '

MODES 1,2,3 od 4:

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

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

At all Other Times:

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

TESTING REQUIREMENTS:

The Reactor Coolant System chemistry shall be determined to be within the limits by analysis of those parameters at the frequencies specified in Table 16.5-3B. '

O Chapter 16.5-3 Page 1 of 4 01/16/99

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady-State Limits provides adequate corrosion protection to ensure the structuralintegrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady-State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.

The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady-State Limits.

The Testing Requirements provide adequate assurance that concentrations in

(]

G excess of the limits will be detected in sufficient time to take corrective action.

O Chapter 16.5-3 Page 2 of 4 01/16/99

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

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e TABLE 16.5-3A I

REACTOR COOLANT SYSTEM

. CHEMISTRY LIMITS i

1

. STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT

' ~

Dissolved Oxygen

s 0.10 ppm s 1.00 ppm i

Chloride s 0.15 ppm s 1.50 ppm l

Fluoride s 0.15 ppm s 1.50 ppm l

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

Chapter 16.5-3 Page 3 of 4 01/16/99

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

i TABLE 16.5-3B REACTOR COOLANT SYSTEM CHEMISTRY LIMITS TESTING REQUIREMENTS SAMPLE AND PARAMETER ANALYSIS FREQUENCY Dissolved Oxygen At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months i Chloride At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Fluoride At least once per 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months O i l

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Not required with T.vgl ess than or equal to 250' F.

Chapter 16.5-3 Page 4 of 4 01/16/99

16.5 REACTOR COOLANT SYSTEM 16.5-4 PRESSURIZER COMMITMENT:

The pressurizer temperature shall be limited to:

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

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

APPLICABILITY:

At all times.

REMEDIAL ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within the limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural i t integrity of the pressurizer; determine that the pressurizer remains acceptable for

\ i continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and '

reduce the pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

TESTING REQUIREMENTS:

The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutec during system heatup or cooldown.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The temperature and pressure changes during heatup and cooldown are limited to I be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code, Section lil, Appendix G. Although the pressurizer operates in temperature ranges above those for which there is reason for concem of nonductile failure, 1 operating limits are provided to assure compatibility of operation with the fatigue

- analysis performed in accordance with the ASME Code requirements.

Chapter 16.5-4 Page 1 of 1 01/16/99

19.4 REACTOR COOLANT SYSTEM 19.4-1 STRUCTURAL INTEGRITY COMMITMENT:

, The structural integrity of ASME Code Class 1,2, and 3 components shall be l maintained. '

APPLICABILITY:

All Modes.

REMEDlAL ACTION:

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

b. With the structural integrity of any ASME Code Class 2 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) {

prior to increasing the Reactor Coolant System temperature above 200 F. l

c. With the structural integrity of any ASME Code Class 3 component (s) not conforming to the above requirements, restore the structural integrity of the affected component (s) to within its limit or isolate the affected component (s) from service. j TESTING REQUIREMENTS: l 1

The structural integrity of ASME Code Class 1,2, and 3 components shall be ;

maintained in accordance with the Inservice Inspection and Testing Program.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

O l

Chapter 16.5-5 Page 1 of 2 01/16/99

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

1 BASES:

The inservice inspection and testing programs for ASME Code Class 1,2, and 3 l

components ensure that the structural integrity and operational readiness of these i

components will be maintained at an acceptable level throughout the life of the i plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific written relief has been granted by the Commission pursuant to l 10 CFR 50.55a(g)(6)(i).

Components of the Reactor Coolant System were designed to provided access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, and applicable Addenda as required by 10 CFR 50.55a(g) ;

except where specific written relief has been granted by the Commission pursuant to l 10 CFR 50.55a(g)(6)(i).

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Chapter 16.5-5 Page 2 of 2 01/16/99

16.5 REACTOR COOLANT SYSTEM 1911 REACTOR COOLANT SYSTEM VENTS COMMITMENT:

At least one Reactor Coolant System vent path consisting of at least two, valves in series powered from emergency buses shall be OPERABLE and closed at each of '

the following locations: '

a. Reactor Vessel Head

b. Pressurizer steam space t APPLICABILITY:

MODES 1,2,3 and 4.

REMEDIAL ACTION:

i

a. With one of the above Reactor Coolant System vent paths inoperable, STARTUP and/or POWER OPERATION may continue provided the :

inoperable vent path is maintained closed with power removed from the valve ,

actuator of all the valves in the inoperable vent path; restore the inoperable vent path to OPERABLE status within 30 days, or, be in HOT STANDBY ,

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within se following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With both of the above Reactor Coolant System vent paths inoperable, ,

maintain the inoperable vent paths closed with power removed from the valve :

actuators of all the valves in the inoperable vent paths, and restore at least '

one of the vent paths to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

TESTING REQUIREMENTS:

Each Reactor Coolant System vent path shall be demonstrated OPERABLE at least once per 18 months by:

a. Verifying all manual isolation valves in each vent path are locked in the open l position, and I l

O

For the plants using power operated relief valve (PORV) as a vent path, PORV block is not required to be closed if the PORV is operable. i Chapter 16.5-6 Page 1 of 2 01/16/99

(u TESTING REQUIREMENTS (con't) '

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

REFERENCES:

i

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical '

Specifications Amendments for Catawba, September 30,1998. !

BASES: '

Reactor Coolant System Vents are provided to exhaust noncondensible gases and/or steam from the primary system that could inhibit natural circulation core l cooling. The OPERABILITY of at least one Reactor Coolant System vent path from the reactor vessel head, and the pressurizer steam space ensures the capability ;

exists to perform this function. ;

The valve redundancy of the Reactor Coolant System vent paths serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, power supply or control system does not prevent isciation of the vent path.

O '

V The function, capabilities, and testing reouirements of the Reactor Coolant System !

vent systems are consistent with the requirements of item II.B.1 of NUREG-0737, l

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

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O Chapter 16.5-6 Page 2 of 2 01/16/99

11 4 REACTOR COOLANT SYSTEM ,

16.5-7 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION ,

COMMITMENT:

The temperatures of both the reactor and secondary coolants in the steam .

generators shall be greater than 70 F when the pressure of either coolant 17. the steam generator is greater than 200 psig.

APPLICABILITY: '

At all times.

REMEDIAL ACTION:

With the requirements of the above COMMITMENT not satisfied:

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

b. Perform an engineering evaluation to determine the effect of the :

overpressurization on the structuralintegrity of the steam generator. :

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

TESTING REQUIREMENTS:

The pressure in each side of the steam generator shall be determined to be less than 200 psig at least once per hour when the temperature of either the reactor or I secondary coolant is less than 70 F. I

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The limitation on steam generator pressure and temperature ensures that the pressure-induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 70 F and 200 psig are based on a steam generator RTum of 60*F and are sufficient to prevent brittle O fracture. l Chapter 16.5-7 Page 1 of 1 01/16/99 ,

16.6 ENGINEERED SAFETY FEATURES 16.6-1 CONTAINMENT SUMP COMMITMENT:

The containment sump sha!! be maintained free of loose debris.

APPLICABILITY:

MODES 1,2,3, and 4.

REMEDIAL ACTION:

Remove all debris which is found within containment which could be transported to the containment sump.

TESTING REQUIREMENTS:

Perform a visual inspection which verifies that no loose debris (rags, trash, clothing, etc.) is present in the containment which could be transported to the containment sump and cause restriction of the pump suctions during LOCA conditions. This visual inspection shall be performed:

1. For all accessible areas of the containment prior to establishing containment integrity, and

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

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

Periodic inspections of the containment for loose debris ensures that the sump suction inlet will remain unrestricted in the event of a LOCA and subsequent ECCS recirculation injection operation.

/

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Chapter 16.6-1 Page 1 of 1 01/16/99

16.6 ENGINEERED SAFETY FEATURES 16.6-2 ICE BED TEMPERATURE MONITORING SYSTEM COMMITMENT:

i The Ice Bed Temperature Monitoring System shall be OPERABLE with at least two OPERABLE RTD channels in the ice bed at each of three basic elevations (< 11',

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

APPLICABILITY:

MODES 1,2,3, and 4.

REMEDIAL ACTION:

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

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

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

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

a) Twenty-one OPERABLE air handling units, b) Two OPERABLE glycol circulating pumps, and l

c) Three OPERABLE refrigerant units.

Otherwise, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

b. With the Ice Bed Temperature Monitoring System inoperable and with the Ice Condenser Cooling System not satisfying the minimum components OPERABILITY requirements of REMEDIAL ACTION a.3 above, POWER OPERATION may continue for up to 6 days provided the ice compartment lower inlet doors, intermediate deck doors, and top deck doors are closed and the last recorded mean ice bed temperature was less than or equal to Chapter 16.6-2 Page 1 of 2 01/16/99 1

REManlAL ACTION fcon't) 15*F and steady; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

TESTING REQUIREMENTS:

The Ice Bed Temperature Monitoring System shall be detemiined OPERABLE by performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

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

O O

Chapter 16.6-2 Page 2 of 2 01/16/99

9 ,1gj (V ENGINEERED SAFETY FEATURES '

16.6-3 INLET DOOR POSITION MONITORING SYSTEM COMMITMENT:

The inlet Door Position Monitoring System shall be OPERABLE.

APPLICABILITY:

MODES 1,2,3, and 4.

REMEDIAL ACTION:

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

O TESTING REQUIREMENTS: i The inlet Door Position Monitoring System shall be determined OPERABLE by:

a. Perfomling a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , i

b. Performing a TRIP ACTUATING DEVICE OPERATIONAL TEST at least >

once per 18 months, and

c. Verifying that the Monitoring System correctly indicates the status of each inlet door as the door is opened and reclosed during its testing per Technical Specification 3.6.13.

l

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

l Chapter 16.6-3 Page 1 of 2 01/16/99 l l

BASES:

The OPERABILITY of the inlet Door Positioning Monitoring System ensures that the capability is available for monitoring the individual inlet door position. In the event the system is inoperable, the REMEDIAL ACTION requirements provide assurance that the ice bed heat removal capacity will be maintained.

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, Chapter 16.6-3 Page 2 of 2 01/16/99

i. 1E& ENGINEERED SAFETY FEATURES 16.6-4 CHLORINE DETECTION SYSTEMS COMMITMENT:

Two independent Chlorine Detection Systems, with their Alarm / Trip Setpoints adjusted to actuate at a chlorine concentration of less than or equal to 5 ppm, shall be OPERABLE.

APPLICABILITY:

i All MODES.

REMEDIAL ACTION: '

a. With one Chlorine Detection System inoperable, restore the inoperable ;

system to OPERABLE status within 7 days or within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months initiate ;

and maintain 0;eration of the Control Room Area Ventilation System in high chlorine protection mode with flow through the HEPA filters and activated carbon adsorbers.

E

b. With both Chlorine Detection Systems inoperable, within 1 hourinitiate and maintain operation of the Control Room Area Ventilation System in high chlorine protection mode with flow through the HEPA filters and activated '

carbon adsorbers.

TESTING REQUIREMENTS:

1. Each Chlorine Detection System shall be demonstrated OPERABLE by performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a CHANNEL OPERATIONAL TEST at least once per 31 days and a CHANNEL CALIBRATION at least once per 18 months.

2. Verify once per 18 months that on a High Chlorine / Toxic Gas test signal, the system automatically isolates the affected intake from outside air with recirculating flow through the HEPA filters and activated carbon adsorbers banks within 10 seconds (plus air travel time between the detectors and the isolation dampers).

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

Chapter 16.6-4 Page 1 of 2 01/16/99

s

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[ BASES:

The OPERABILITY of the Chlorine Detection Systems ensures that sufficient capability is available to promptly detect and initiate protective action in the event of a an accidental chlorine release. This capability is required to protect control room personnel and is consistent with the recommendations of Regulatory Guide 1.95, J

Revision 1," Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release", January 1977.

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O Chapter 16.6-4 Page 2 of 2 01/16/99 i

D 16.7 INSTRUMENTATION (O

16.7-1 ATWS MITIGATION SYSTEM ACTUATION CIRCUlTRY (mSAC)

COMMITMENT:

The AMSAC System shall be OPERABLE.

APPLICABILITY:

MODE 1, above 40% of RATED THERMAL POWER (based on Turbine impulse Pressure).

REMEDIAL ACTION:

With the AMSAC System inoperable, restore it to OPERABLE status within 7 days or submit a Special Report to the Nuclear Regulatory Commission within the following 30 days. This report shall outline the cause of the malfunction and the plans for restoring the system to OPERABLE status.

TESTING REQUIREMENTS:

Perform a CHANNEL CALIBRATION on the AMSAC System instruments at least once per 18 months.

REFERENCES:

1. 10 CFR 50.62, Requirements for Reduction of Risk From Anticipated Transients Without Scram (ATWS) Events for Light-Water Cooled Nuclear Power Plants.

2. Generic Letter 85-06, " Quality Assurance Guidance for ATWS Equipment That is Not Safety-Related".

3. Information Notice 92-06, " Reliability of ATWS Mitigation System and Other NRC Required Equipment Not Controlled by Plant Technical Specifications".

O Chapter 16.7-1 Page 1 of 2 05/04/98

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BASES:

Per 10 CFR 50.62, "Each pressurized water reactor must have equipment from sensor output to final actuation device, that is diverse from the reactor trip system, to automatically initiate the auxiliary feedwater system and initiate a turbine trip under conditions indicative of an ATWS. This equipment must be designed to perform its function in a reliable manner and be independent from the existing reactor trip system". When this rule was issued, the NRC did not require licensees to address the OPERABILITY of this equipment in the plant Technical Specifications nor require that this equipment be designated as safety-related.

On January 15,1992, the NRC issued Information Notice 92-06 which discussed the reliability of AMSAC and other equipment not controlled by plant Technic &l l

Specifications. This notice described two separate incidents where violations were :

cited because the licensees failed to adequately maintain the reliability of their AMSAC systems. The NRC is concemed that Licensees may not place an i

appropriate level of priority on resolving problems with the AMSAC System because !

it is not a safety-related system and because the plant's Technical Specifications do '

not govem its operability. The NRC considers the failure of licensees to adequately l ensure the reliable operation of AMSAC equipment to be a significant regulatory concem.

rN This Selected Licensee Commitment was developed to ensure that approp:iate C) attention is given to maintaining the AMSAC System in a reliable conditica and that prompt action will be taken to repair and restore any AMSAC equipmer.t that is discovered ir a condition where it is incapable of performing its intended function.

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Chapter 16.7-1 Page 2 of 2 05/04/98

4 1

.G O 16.7 INSTRUMENTATION 16.7-2 SEISMIC INSTRUMENTATION 4

COMMITMENT:

a. The seismic monitoring instrumentation shown in Table 16.7-2A shall be OPERABLE.

APPLICABILITY: j i

At all times.

l ftEMEDIAL ACTION:

a. With one or more of the above required seismic monitoring instruments

inoperable for more than 30 days, prepare and submit a Special Report to the !

Commission within the next 10 days outlining the cause of the malfunction l and the plans for restoring the instrument (s) to OPERABLE status.

TESTING REQUIREMENTS: l

.O

. U a. Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CALIBRATION and CHANNEL OPERATIONAL TEST operations l at the frequencies shown in Table 16.7-2B.

b. Each of the above accessible seismic monitoring instruments actuated during a seismic event greater than or equal to 0.01 g shall be restored to j OPERABLE status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the seismic event. Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion. Data retrieved from the triaxial time-history accelerograph shallinclude a post-event CHANNEL CALIBRATION obtained by actuation of the intemal test and calibrate !

function immediately prior to removing data. CHANNEL CAllBRATION shall be performed immediately after insertion of the new recording media in the triaxial time-history accelerograph recorder. A Special Report shall be prepared and submitted to the Commission within 10 days describing the l magnitude, frequency spectrum, and resultant effect upon facility features important to safety.

O Chapter 16.7-2 Page 1 of 4 01/16/99

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REFERENCES:

N/A BASES: ,

l The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to '

permit comparison of the measured response to that used in the design basis for the i

! facility to determine if plant shutdown is required pursuant to Appendix A of 10 CFR Part 100. The instrumentation is consistent with the recommendations of Regulatory Guide 1.12, " Instrumentation ter Earthquakes", April 1974. i l

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O Chapter 16.7-2 Page 2 of 4 01/16/99 r

4 TABLE 16.7-2A SElSMIC MONITORING INSTRUMENTATION j Minimum

' Measurement instruments i instruments And Sensor Locations Range Operable e

i 1. Triaxial Time-History Accelerographs

a. 1MIMT 5070 (Remote Sensor A) -1 g to + 1 g 1 Containment Base Slab

b. -1 g to + 1 g 1MIMT 5080 (Remote Sensor B) 1

, Containment Vessel Elev 619'5" 4

c. 0.005 g to 0.05 g 1MIMT 5090 (Starter Unit) 1 t

, Containment Base Slab

2. Triaxial Peak Accelerographs

a. 1MIMT 5010- Containment Bldg. O g to + 2 g 1 Elev 588' + 61/8" -

b. 1MIMT 5020- Containment Bldg. O g to + 2 g 1 Elev 567' 2 %" '

c. 1MIMT 5030- Auxiliary Bldg. O g to + 2 g 1 Elev 543'

3. Triaxial Seismic Switch '

1 MIMT 5000- Containment 0.025 g to 0.25 g 1* t Base Slab

4. Triaxial Response - Spectrum Recorders

a. 1MIMT 5040 - Containment 0 to 34 g at 1' Base Slab 2 to 25 Hz ,

b. 1MIMT 5050 - Containment Bldg. O to 34 g at 1 Elev 579' 3 %" 2 to 25 Hz

c. 1MIMT 5060 - Auxiliary Bldg. O to 34 g at 1 Elev 577' 2 to 25 Hz 1 With reactor control room indication.

Chapter 16.7-2 Page 3 of 4 01/16/99

TABLE 16.7-2B (A .

SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Channel Channel Channel Operational Instruments and Sensor Locations Check Calibration Test

1. Triaxial Time-History Accelerographs

a. 1MIMT 5070 (Remoto Sensor A) M* R SA Containment Base Slab

b. 1MIMT 5080 (Remote Sensor B) M* R SA Containment Vessel Elev 619'5"

c. 1MIMT 5090 (Starter Unit) N.A. R SA Containment Base Slab

2. Triaxial Peak Accelerographs

a. 1MIMT 5010 - Containment Bldg. N.A. R N.A.

Elev 588' + 61/8" O ]

b b. 1MIMT 5020-Containment Bldg. N.A. R N.A.

i Elev 567' 2 %"

c. 1MIMT 5030- Auxiliary Bldg. N.A. R N.A.

Elev 543'

3. Triaxial Seismic Switch i

1MIMT 5000 - Containment M R SA '

Base Slab "

4. Triaxial Response - Spectrum Recorders

a. 1MIMT 5040- Containment Base M R SA Slab"

b. 1MIMT 5050 - Containment Bldg. N.A. R N.A.

Elev 579' 3 %"

c. 1MIMT 5060 - Auxiliary Bldg. N.A.

R N.A.

Elev 577'

Except seismic trigger.

- With reactor control room indications.

Chapter 16.7-2 Page 4 of 4 01/16/99 l

4

16.7 INSTRUMENTATION 16.7-3 METEOROLOGICAL INSTRUMENTATION COMMITMENT:

a. The meteorological monitoring instrumentation channels shown in Table 16.7-3A shall be OPERABLE.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

a. With one or more required meteorological monitoring channels inoperable for more than 7 days, prepare and submit a Special report to the Commission l within the next 10 days outlining the cause of the malfunction and the plans for restoring the channel (s) to OPERABLE status.

TESTING REQUIREMENTS:

O a. Each of the above meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of a CHANNEL CHECK and instrument Calibration operations at the frequencies shown in Table 16.7-38.

REFERENCS:

' N/A BASES:

I The OPERABILITY of the meteorological instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public and is consistent with the recommendations of Regulatory Guide 1.23, "Onsite Meteorological Programs,"

February 1972.

An Instrument Calibration will consist of the following test:

1) A bench based test and calibration of the tower mounted sensors for:

I Chapter 16.7-3 Page 1 of 4 01/16/99

. Wind Speed

. Wind Direction Temperature RTD's and Temperature Processing Module (Ambient and Delta T)

2) An instrument Loop Calibration from the input of the signal processors to the end devices.

3) For Wind Direction a Line Phase Differential Compensation will be :

performed, which includes the tower signal cable.

4) A CHANNEL CHECK, subsequent to any work performed. This will verify continuity of the signal cable between the sensor and signal processors. ,

5) The Wind Speed Sensors and cup-sets or Wind Direction Sensors and Vanes do not require wind tunnel testing as an assembly.

6) Replacement of cup-sets or vanes does not require an Instrument Calibration '

of the affected channel.

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Chapter 16.7-3 Page 2 of 4 01/16/99

l TABLE 16.7-3A METEOROLOGICAL MONITORING INSTRUMENTATION MINIMUM INSTRUMENT LOCATION OPERABLE

1. Wind Speed

a. MeteorologicalTower Nominal Elev. 663.5 ft. 1

b. MeteorologicalTower Nominal Elev. 830.5 ft. 1

2. Wind Direction

a. Meteorological Tower - Nominal Elev. 663.5 ft. 1

b. MeteorologicalTower Nominal Elev. 830.5 ft. 1

3. Air Temperature - AT Meteorological Tower Nominal Elev. 827.25 - 660.25 ft. 1 Note: Elevations are feet above Mean Sea Level O

Chapter 16.7-3 Page 3 of 4 01/16/99 t

( TABLE 16.7-3B

! METEOROLOGICAL MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS l l CHANNEL CHANNEL INSTRUMENT CHECK CAllBRATION

1. Wind Speed

a. Nominal Elev. 663.5 ft D SA j b. Nominal Elev. 830.5 ft. D SA l

i 2. Wind Direction 4

a. Nominal Elev. 663.5 ft D SA

b. Nominal Elev. 830.5 ft. D SA

3. AirTemperature- AT l Nominal Elev. 827.25 - 660.25 ft. D SA i

l l Note: Elevations are feet above Mean Sea Levei 1

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', Chapter 16.7-3 Page 4 of 4 01/16/99

10 I INSTRUMENTATION If,Z-i. LOOSE-PART DETECTION SYSTEM -

COMMITMENT:

The Loose-Part Detection System shall be OPERABLE.

APPUCABlWTY:

MODES 1 and 2.-

REMEDIAL ACTION:

a. With all channels of one or more Loose-Part Collection Region (s) inoperable for more than 30 days:

1. Restore at least one channel per Loose Part Collection Region to OPERABLE status, or l

2. Prepare and submit a Special Report to the Commission within the l next 10 days outlining the cause of the malfunction and the plans for O- restoring the channals to OPERABLE status.

TESTING REQUIREMENTS:

a. Each channel of the Loose-Part Detection Systems shall be demonstrated OPERABLE by performance of:

1. A CHANNEL CHECK at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , i

)

2. An Channel Functional Test at least once per 31 days as follows:

1. Verify each analog channel by listening to the audio signal, ii. Verify each digital channel by confirming stored data is within i expected limits, and ;

iii. Inject a simulated signalinto one or more channels and verify j the alarm function. j i

3. A CHANNEL CALIBRATION at least once per 18 months.

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Chapter 16.7-4 Page 1 of 2 01/16/99 i

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REFERENCES:

N/A BASES: 1 1

The OPERABILITY of the loose-part detection instrumentation ensures that i sufficient capability is available to detect loose metallic parts in the Reactor System '

and avoid or mitigate damage to Reactor System components. The allowable out-of-service times and surveillance requirements are consistent with the ;

recommendations of Regulatory Guide 1.133, " Loose-Part Detection Program for the Primary System of Light-Water-Cooled Reactors," May 1981..

t A Loose Part Collection Region is an area within the Reactor Coolant System where I loose parts can possibly collect and which is monitored by the Loose Part Detection {

System. Collection Regions are: '

1) Upper Reactor Vessel

2) Lower Reactor Vessel

3) Primary side of Steam Generator A

4) Primary side of Steam Generator B

5) Primary side of Steam Generator C

6) Primary side of Steam Generator D l

Chapter 16.7-4 Page 2 of 2 01/16/99

16.7 INSTRUMENTATION 16.7-5 TURBINE OVERSPEED PROTECTION COMMITMENT:

a. At least one Turbine Overspeed Protection System shall be OPERABLE.

APPLICABILITY:

MODES 1,2, and 3 i

REMEDIAL ACTION:

a. With one stop valve or one control valve per high pressure turbine steam line inoperable and/or with one intemiediate stop valve or one intercept valve per low pressure turbine steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , or close at least one valve in the affected steam line(s) or isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

C b. With the above required Turbine Overspeed Protection System otherwise inoperable, within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the turbine from the steam rupply.

TESTING REQUIREMENTS:

a. The above required Turbine Overspeed Protection System shall be !

demonstrated OPERABLE

1. At least once per 31 days while in MODE 1 and while in MODE 2 with the turbine operating, by cycling each of the following valves through at least one complete cycle from the running position:

a) Four high pressure turbine stop valves, b) Six low pressure turbine intermediate stop valves, and c) Six low pressure turbine intercept valves.

2. At least once per 92 days while in MODE 1 and while in MODE 2 with the turbine operating, by direct observation of the movement of each of the above valves and the four high pressure turbine control valves, through one complete cycle from the running position, Chapter 16.7-5 Page 1 of 2

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05/04/98

[ TESTING REQUIREMENTS (con't)

] 3. At least once per 18 months by performance of a CHANNEL '

CALIBRATION on the Turbine Overspeed Protection Systems, and

4. At least once per 40 months by disassembling at least one of each of the above valves (including the four high pressure turbine control valves) and performing a visual and surface inspection of valve seats,

, disks and stems and verifying no unacceptable flaws or corrosion.

1

REFERENCES:

N/A BASES:

t This commitment is provided to ensure that the turbine overspeed protection ;

instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related i components, equipment, or structures.

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Chapter 16.7-5 Page 2 of 2 05/04/98 l 1

f) v 16.7 INSTRUMENTATION

igJ-f RN DISCHARGE INSTRUMENTATION COMMITMENT:

The following shall be OPERABLE:

Instrument Components:

Loops:

1RNFE7520 RN Pump A Discharge Flow Annubar 1RNLP7520 1RNFT7520 RN Pump A Discharge Flow 1RNP7520 RN Pump A Discharge Flow (Control Room Indication) 1RNFE7510 RN Pump B Discharge Flow Annubar i 1RNLP7510 1RNFT7510 RN Pump B Discharge Flow 1RNP7510 RN Pump B Discharge Flow (Control Room Indication) 2RNFE7520 RN Pump A Discharge Flow Annubar '

2RNLP7520 2RNFT7520 RN Pump A Discharge Flow 2RNP7520 RN Pump A Discharge Flow (Control Room Indication)

/9 2RNFE7510 RN Pump B Discharge Flow Annubar V 2RNLP7510 2RNFT7510 RN Pump B Discharge Flow 2RNP7510 RN Pump B Discharge Flow (Control Room Indication)

Annunicators:

1 1AD12-A/1 RN Pump A Flow Hi/ Low ;

1AD12-A/4 RN Pump B Flow Hi/ Low I 2AD12-A/1 RN Pump A Flow Hi/ Low 2AD12 A/4 RN Pump B Flow Hi/ Low APPLICABILITY:

Any time the associated pump is required to be OPERABLE.

REMEDIAL ACTION:

a. With any component on both trains of the above listed instrument loops associated with the same unit inoperable OR both annunciators associated with the same unit inoperable OR any component of one instrument loop and one annunciator from different trains associated with the same unit inceerable:

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t For Example: 1RNFT7520 and 1RNFT7510 QR 1 AD12-A/1 and i 1 AD12-A/4 OR 1RNFT7520 and 1 AD12-A/4.

Chapter 16.7-6 Page 1 of 2 01/16/99

REMEDIAL ACTION (con't)

Open 1RN-58B(RN Hdr B Rtn to SNSWP) and 1RN-63A (RN Hdr A Rtn to SNSWP), OR enter a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months action statement per Technical Specification 3.7.8 on RN for the affected unit.

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b. With the conditions specified in remedial action a. met on both units:

Open 1RN-58B(RN Hdr B Rtn to SNSWP) and 1RN-63A (RN Hdr A Rtn to SNSWP), OR enter Technical Specification 3.0.3 for both units.

TESTING REQUIREMENTS:

Perform channel calibrations per IP/1(2)/B/3112/04 as required.

REFERENCES:

1) NRC Inspection Report,50-413,414/94-17, September 9,1994

2) Catawba Nuclear Station PIP 0-C94-1555 BASES:

The bases for this Selected License Commitment is stated in reference 1, page 11, "Part of the licensing bases of the RN system included the ability for operators to manually change the discharge path for the RN system from the non-safety related Lake Wylie to the safety related SNSWP."

This Selected License Commitment was developed to address the NRC concerns about actions to be taken in the event instrumentation / alarms were out of service which would impair the ability of Operations to recognize loss of the non-safety discharge. This SLC shall ensure the proper actions are taken based on plant equipment conditions.

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Chapter 16.7-6 Page 2 of 2 01/16/99

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( 16.7 INSTRUMENTATION 16.7-7 MOVABLE INCORE DETECTORS COMMITMENT:

The Movable incore Detection syste shall be OPERABLE with:

a. At least 75% of the detector thimbles,

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

)

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

APPLICABILITY:

i When the Movable incore Detection System is used for:

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

b. Monitoring the QUADRANT POWER TILT RATIO, or l

c. Measurement of F,s", and Fo(Z).

l REMEDIAL ACTION:

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

TESTING REQUIREMENTS:

The Movable incore Detection System shall be demonstrated OPERABLE at lease once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by irradiating each detector used and determining the I acceptability of its voltage curve for:

a. Recalibration of the Excore Neutron Flux Detection system, or

b. Monitoring the QUADRANT POWER TILT RATIO, or N

c. Measurement of F,s , ggg p0(7).

i Chapter 16.7-7 Page 1 of 2 01/16/99 1

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The OPERABILITY of the movable 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 core. The !

OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptability of its voltage curve.

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For the purpose of measuring Fo(Z) or F,s a full inCore flux map is used. Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and fullincore flux maps or symmetric ;

incore thimbles may be used for monitoring the OUADRANT POWER TILT RATIO when one Power Range channel is inoperable.

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l 16.7

[O INSTRUMENTATION 16.7-8 GROUNDWATER LEVEL COMMITMENT:

The groundwater level shall be maintained at or below the top of the adjacent floor slabs of the Reactor Containment Building and the Auxiliary Building.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

a. With the groundwater level above the top of the adjacent floor slab by less than or equal to 5 feet, reduce the groundwater level to or below the top of the affected adjacent floor slab within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

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b. With the groundwater level above the top of the adjacent floor slab by greater than 5 feet but less than 15 feet, reduce the groundwater level to less than or equal to 5 feet above the top of the affected adjacent floor slab within 24 l hours and to or below the top of the affected adjacent floor slab within 7 days i of initially exceeding the above limits or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . j

c. With the groundwater level above the top of the adjacent floor slab by greater j than or equal to 15 feet, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . Perform an engineering ,

evaluation to determine the effects of this higher groundwater level on the affected building (s) and submit the results of this evaluation and any '

corrective action determined necessary to the Commission as a Special Report prior to increasing T.vg above P0012

d. Determine the rate of rise of groundwater when the level reaches the top of the floor slab. If the rate of rise of the groundwaterlevelis greaterthan or i equal to 0.3 foot per hour, determine the rate of rise at least once per 30 minutes. If the rate of rise exceeds 0.5 foot per hour for more than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , be in at least HOT STANDBY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and in COLD SHUTDOWN within 3

(v Chapter 16.7-8 Page 1 of 2 01/16/99

REMEDIAL ACTION (con't) the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . If the rate of rise is less than 0.5 feet per hour, comply with the requirements of REMEDIAL ACTIONS a., b., and c. above.

TESTING REQUIREMENTS:

The groundwater level shall be determined at the following frequencies by monitoring the water level and by verifying the absence of alarm in the six grcundwater monitor wells as shown in UFSAR Figure 2-60 installed around the perimeter of the Reactor and Auxiliary Buildings:

a. At least once per 7 days when the groundwater level is at or below the top of the adjacent floor slab, and

b. At lea,se once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the groundwater level is above the top of the adjacent floor slab.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

V BASES:

This COMMITMENT is provided to ensure that groundwater levels will be monitored and prevented from rising to unacceptable levels. High groundwater levels could result in unacceptable structural stresses in the Containment and/or Auxiliary Building due to uplitt and hydrostatic forces during design basis events. Although these buildings have been statically analyzed to withstand soil pressure along with the uplift and hydrostatic forces resulting from groundwater rebound to yard elevation (593'6"), this analysis did not include any other loadings and was not a design condition for these buildings.

Chapter 16.7-8 Page 2 of 2 01/16/99

if 7 INSTRUMENTATION 16.7-9 STANDBY SHUTDOWN SYSTEM 1 COMMITMENT:

The Standby Shutdown System (SSS) shall be OPERABLE.

4 APPLICABILITY:

MODES 1,2, and 3.

4 j REMEDIAL ACTION: (Units 1 and 2)

a. With the Standby Shutdown System inoperable, restore the inoperable equipment to OPERABLE status within 7 days or be in at least HOT

! STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in at least HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ,

b. With the total leakage from UNIDENTIFIED LEAKAGE, IDENTIFIED LEAKAGE and reactor coolant pump seal leakage greater than 26, gpm,

,O declare the Standby Makeup Pump inoperable and take REMEDIAL ACTION i V a., above.

1

c. The provisions of SLC 16.2.3 are not applicable.

4 TESTING REQUIREMENTS:

! 1. The Standby Shutdown System diesel generator shall be demonstrated

OPERABLE:

4

a. At least once per 31 days by verifying

1) The fuel level in the fuel storage tank is greater than or equal to 67 inches, and

2) The diesel starts from ambient conditions and operates for at least 30 minutes at greater than or equal to 700 kW.

b. At least once per 92 days by verifying that a sample of diesel fuel from the fuel storage tank, obtained in accordance with ASTM-D270-1975, is within the acceptable limits specified in Table 1 of ASTM-D975-1977 when checked for viscosity and water and sediment; and O

Chapter 16.7-9 Page 1 of 4 01/16/99

_ __.__-____.____..____._______.______.-____.__.m_. .

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TESTING REQUIREMENTS (con't)

c. At least once per 18 months by subjecting the diesel to an inspection !

in accordance with procedures prepared in conjunction with its manufacturer's recommendations for the class of service.

2. The Standby Shutdown System diesel starting 24-volt battery bank and charger shall be demonstrated OPERABLE

i

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

1) The electrolyte level of each battery is at or above the low mark :

and at or below the high mark; and i

2) The overall battery voltage is greater than or equal to 24 volts on float charge.

b. At least once per 92 days by verifying that the individual cell voltage is greater than or equal to 1.36 volts on float charge, and

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

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

2) The battery-to-battery and terminal connections are clean, tight, and free of corrosion.

3. The Standby Makeup Pump water supply shall be demonstrated OPERABLE by:

a. Verifying at least once per 7 days that the requirements of SLC 16.9-21 are met and the boron concentration in the storage pool is greater -

than or equal to the minimum specified in the Core Operating Limits Report.

b. Verifying at least once per 92 days that the Standby Makeup Pump develops a flow of greater than or equal to 26 gpm at a pressure greater than or equalto 2488 psig.

4. The Standby Shutdown System 250/125-Volt Battery Bank and its associated charger shall be demonstrated OPERABLE:

a. At least once per 31 days by verifying:

O Chapter 16.7-9 Page 2 of 4 01/16/99

(')

v TESTING REQUIREMENTS (con't)

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

2) The total battery terminal voltage is greater than or equal to i 258/129 volts on float charge.

b. At least once per 92 days by verifying that the specific gravity is appropriate for continued service of the battery, and

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

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

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

5. The Steam Turbine Driven Auxiliary Feedwater Pump and associated components snail be demonstrated OPERABLE at least once per 18 months by verifying that the system functions as designed from the Standby fa Shutdown System.

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6. Each Standby Shutdown System instrumentation device shall be demonstrated OPERABLE by performance of a CHANNEL CHECK at least once per 31 days and a CHANNEL CAllBRATION at least once per 18 months.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The Standby Shutdown System (SSS) is designed to mitigate the consequences of certain postulated fire, security, and station blackout incidents by providing capability to maintain HOT STANDBY conditions and by controlling and monitoring vital systems from locations extemal to the main control room. This capability is consistent with the requirements of 10 CFR Part 50, Appendix R, NUREG 0800 Section 9.5-1 and Appendix A to Branch Technical Position APSCB 9.5-1.

The Testing Requirements ensure that the SSS systems and components are e capable of performing their intended functions. The required level in the SSS (x

Chapter 16.7-9 Page 3 of 4 01/16/99

1 BASES (con't) diesel generator fuel storage tank ensures sufficient fuel for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months uninterrupted operation. It is assumed that, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , either offsite power can be restored or additional fuel can be added to the storage tank.

Although the Standby Makeup Pump is not nuclear safety-related and was not designed according to ASME code requirements, it is tested quarterly to ensure its OPERABILITY. The Testing Requirement conceming the Standby Makeup Pump water supply ensures that an adequate water volume is available to supply the pump continuously for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

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16.7 INSTRUMENTATION 16.7-10 RADIATION MONITORING FOR PLANT OPERATIONS COMMITMENT:

i The radiation monitoring instrumentation channels for plant operations shown in Table 16.7-10A shall be OPERABLE with their Alarm / Trip Setpoints within the specified limits.

APPLICABILITY:

As shown in Table 16.7-10A REMEDIAL ACTION:

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a. With a radiation monitoring channel Alarm / Trip Setpoint for plant operations

, exceeding the value shown in table 16.7-10A, adjust the Setpoint to within the l limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or declare the channel inoperable. j

b. With one or more radiation monitoring channels for plant operations inoperable, take the REMEDIAL ACTION shown in Table 16.7-10A.

TESTING REQUIREMENTS:

Each radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL CAllBRATION and CHANNEL OPERATIONAL TEST operations for the MODES and at i

the frequencies shown in Table 16.7-10B

REFERENCES:

i 1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

J

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

(d Chapter 16.7-10 Page 1 of 6 01/16/99

BASES (con't) maintenance. The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded. The radiation monitors send actuation signals to initiate alarms or automatic isolation action and actuation of emergency exhaust or ventilation systems. Some of the final actuations are dependent on plant condition in addition to the actuation signals from the radiation monitors.

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l TABLE 16.7-10A j RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS j FUNCTIONAL UNIT CHANNELS MINIMUM APPLICABLE ALARM / TRIP REMEDIAL TO CHANNELS MODES SETPOINT ACTION

! TRIP / ALARM OPERABLE

1. Containment Atmosphere- 1 1 All ***

C

{ High Gaseous Radioactivity l (Low Range - EMF-39) 2.~ Fuel Storage Pool Areas

, a. High Gaseous 1 1 s 1.7 x 10" pCi/mi F Radioactivity (Low Range

- EMF-42)

b. Criticality-Hadiation Level 1 1 515 mR/h E (Fuel Bridge- Low Range

- 1 EMF-15, 2 EMF-4)

3. Control Room Air intake- All s 1.7 x 10" pCi/mi D Radiation Level- High 1/ intake 2 (1/ intake)

Gaseous Radioactivity (Low Range - EMF-43 A & B)

4. Auxiliary Building Ventilation 1 1 1,2,3,4 51.7 x 10" Ci/mi G High Gaseous Radioactivity (Low Range- EMF-41)

5. Component Cooling Water s 1 x 104 pCi/ml H System (EMF-46 A & B) 1 1 All r

Chapter 16.7-10 Page 3 of 6 01/16/99

(V9 TABLE 16.7-10A TABLE NOTATIONS

With fuel in the fuel storage pool areas.

- With irradiated fuel in the fuel storage pool areas.

"* When venting or purging from containment to the atmosphere, the trip setpoint shall not exceed the equivalent lirnits of SLC 16.11-18 in accordance with the methodology and parameters in the ODCM. When not venting or purging in Modes 5 or 6, the alarm setpoint concentration ( Ci/ml) shall be such that the actual submersion dose rate would not exceed SmR/hr without alarm. When not venting or purging in Modes 1 through 4 the alarm setpoint shall be no more than 3 times the containment atmosphere activity as indicated by the radiation monitor.

REMEDIAL ACTION STATEMENTS ACTION C - With less than the Minimum Channels OPERABLE requirement, operation may continue provided the containment purge and exhaust valves are maintained closed.

(

U) ACTION D - With the number of operable channels one less than the Minimum Channels OPERABLE requirement, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months initiate and maintain operation of the Control Room Area Ventilation System with flow through the HEPA filters and activated carbon adsorbers.

ACTION E - With less than the Minimum Channels OPERABLE requirement, operation may continue for up to 30 days provided an appropriate portable continuous monitor with the same Alarm Setpoint is provided in the fuel storage pool area. Restore the inoperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel building.

ACTION F - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, operation may continue provided ;

the Fuel Handling Ventilation Exhaust System is operating and l discharging through the HEPA filters and activated carbon adsorbers.

Otherwise, suspend all operations involving fuel movement in the fuel !

building.

ACTION G - With the number of OPERABLE channels less than the Minimum I Channels OPERABLE requirement, operation may continue provided p the Auxiliary Building Filtered Exhaust System is operating and b discharging through the HEPA filter and activated carbon adsorbers. {

Chapter 16.7-10 Page 4 of 6 01/16/99 1

TABLE 16.7-10A REMEDIAL ACTION STATEMENTS (con't)

ACTION H - With the number of OPERABLE channels less than the Minimum Channels OPERABLE requirement, operation may continue for up to 30 days provided that, at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , grab samples are collected and analyzed for radioactivity (gross gamma) at a lower limit ,

of detection of no more than 10 Ci/ml.

O i

i L

O Chapter 16.7-10 Page 5 of 6 01/16/99

O O O TABLE 16.7-10B RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS TESTING REQUIREMENTS FUNCTIONAL UNIT CHANNEL CHANNEL CHANNEL MODES FOR CHECK CAllBRATION OPERATIONAL WHICH TEST SURVEILLANCE IS REQUIRED

1. Containment Atmosphere-High 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 18 months 92 days All Gaseous Radioactivity (Low Range -

EMF-39)

2. Fuel Storage Pool Areas

a. High Gaseous Radioactivity (Low 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 18 months 92 days "

Range - EMF-42)

b. Criticality-Radiation Level (Fuel 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 18 months 92 days

Bridge - Low Range - 1 EMF-15, 2 EMF-4)

3. Control Room Air intake Radiation Level- High Gaseous Radioactivity- 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 18 months 92 days All (Low Range - EMF-43 A & B)

4. Auxiliary Building Ventilation High Gaseous Radioactivity (Low Range - 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 18 months 92 days 1,2,3,4 EMF-41)

5. Component Cooling Water 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 18 months 92 days All System (EMF-46 A & B)

TABLE NOTATIONS

With fuel in the fuel storage pool area.

" With irradiated fuelin the fuel storage pool areas.

Chapter 16.7-10 Page 6 of 6 01/16/99

16.7 INSTRUMENTATION 16.7-11 POSITION INDICATION SYSTEM - SHUTDOWN COMMITMENT:

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

APPLICABILITY:

MODES 3',4', and 5'.

REMEDIAL ACTION:

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

l TESTING REQUIREMENTS:

Testing Requirements are specified in Technical Specification Surveillance Requirement 3.1.7.1.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical I Specifications Amendments for Catawba, September 30,1998. j BASES:

OPERABILITY of the Digital Rod Position Indicators is required to determine control rod positions and thereby ensure compliance with the control rod alignment and insertion limits of the Technical Specifications.

O

With the Reactor Trip System breakers .in the closed position.

Chapter 16.7-11 Page 1 of 1 01/16/99

f 16.7 INSTRUMENTATION 16.7-12 POSITION INDICATION SYSTEM -TEST EXCEPTION COMMITMENT:

The limitations of SLC 16.7-11 may be suspended during the performance of individual full-length shutdown and control rod drop time measurements provided;

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

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

APPLICABILITY:

MODES 3,4, and 5 during performance of rod drop time measurements.

REMEDIAL ACTION:

With the Position Indication System inoperable or with more than one bank of rods withdrawn, immediately open the Reactor trip breakers.

TESTING REQUIREMENTS:

The above required Position Indication Systems shall be determined to be OPERABLE within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to the start of and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter during rod drop time measurements by verifying the Demand Position Indication System and the Digital Rod Position Indication System agree:

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

b. Within 24 steps during rod motion.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendrnents for Catawba, September 30,1998.

This requirement is not applicable during the initial calibration of the Position Indication System provided: (1) Keff si maintained less than or equal to 0.95, O

V and (2) only one shutdown or control rod bank is withdrawn from the fully inserted position at one time.

Chapter 16.7-12 Page 1 of 2 01/16/99

o l

BASES:

J This special tes' exception permits the Position Indication Systems to be inoperable during rod drop time measurements. The exception is required since the data '

necessary to determine the rod drop time are derived from the induced voltage in ,

the position indicator coils as the rod is dropped. This induced voltage is small !

compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE. l 1

i l

1 O !

O Chapter 16.7-12 Page 2 of 2 01/16/99

p) 16.8 ELECTRICAL POWER SYSTEMS 16.8-1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES COMMITMENT:

)

+

Primary and backup containment penetration conduction overcurrent protective devices show in Table 16.8-1 A and 16.8-1B shall be operable.

APPLICABILITY: i

^

MODES 1,2,3, and 4.

REMEDIAL ACTION:

With one or more of the primary or backup containment penetration conductor overcurrent protective device (s) shown in Table 16.8-1 A and 16.8-1B inoperable:

a. Restore the protective device (s) to OPERABLE status or de-energize the circuit (s) by tripping the associated backup circuit breaker or racking out or removing the inoperable circuit breaker within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , declare the affected

(( system or component inoperable, and verify the backup circuit breaker to be l j tripped or the inoperable circuit breaker racked out or removed at least once per 7 days thereafter; the provisions of SLC 16.2.3 are not applicable to overcurrent devices in circuits which have their backup circuit breakers tripped, their inoperable circuit breakers racked out, or removed, or

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

TESTING REQUIREMENTS:

The above noted primary and backup containment penetration conductor overcurrent protective devices shall be demonstrated OPERABLE:

a. At least once per 18 months:

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

O Chapter 16.8-1 Page 1 of 49 01/16/99

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

TESTING REQUIREMENTS (con't) a) A CHANNEL CALIBRATION of the associated protective relays, b) An integrated protective system functional test which includes simulated automatic actuation of the system and verifying that '

each relay and associated circuit breakers function as designed, and c) For each circuit breaker found inoperable during these functional tests, an additional representative sample of at least

.10% of all the circuit breakers of the inoperable type shall also be functionally tested until no more failures are found or all ,

circuit breakers of that type have been functionally tested.

2) By selecting and functionaliy testing a representative sample of at !

least 10% of each type of lower voltage circuit breakers. Circuit breakers selected for functional testing shall be selected on a rotating .

basis. Testing of these circuit breakers shall consist of injecting a l current in excess of the breakers nominal Setpoint and measuring the response time. The measured response time will be compared to the manufacturer's data to ensure that it is less than or equal to a value !

specified by the manufacturer. Circuit breakers found inoperable (n) during functional testing shall be restored to OPERABLE status prior to ;

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

3) By selecting and functionally testing a representative sample of each type of fuse on a rotating basis. Each representative sample of fuses i shallinclude at least 10% of all fuses of that type. The functional test shall consist of a nondestructive resistance measurement test which demonstrates that the fuse meets its manufacturer's design criteria.

Fuses found inoperable during these functional tests shall be replaced with OPERABLE fuses prior to resuming operation. For each fuse found inoperable during these functional tests, an additional representative sample of at least 10% of all fuses of that type shall be functionally tested until no more failures are found or all fuses of that type have been functionally tested.

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

O Chapter 16.8-1 Page 2 of 49 01/16/99

l

REFERENCES:

1

1. i.etter from NRC to Gary R. Peterson, Duke, issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

Containment electrical penetrations and penetration conductors are protected by .

either deenergizing circuits not required during reactor operation or by demonstrating the OPERABILITY of primary and backup overcurrent protection circuit breakers during periodic testing.

The Testing Requirements applicable to lower voltage circuit breakers and fuses provide assurance of breaker and fuse reliability by testing at least one representative sample of each manufacturer's brand of circuit breaker and/or fuse.

Each manufacturer's molded case circuit breakers and/or fuses are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers and/or fuses are tested. If a wide variety exists within any manufacturer's brand of circuit breakers and/or fuses, it is necessary to divide that manufacturer's breakers and/or fuses into groups and treat each group as a separate type of breaker or fuse for testing purposes.

]

G The lists of components for which this COMMITMENT is applicable exclude those circuits for which credible fault currents would not exceed the electrical penetration design rating.

O Chapter 16.8-1 Page 3 of 49 01/16/99

4 o

p TABLE 16.8-1 A t

UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED 1, 6900 VAC Swgr Primary Bkr RCP1 A Reactor Coolant Pump 1 A Backup Bkr 1TA-3 Primary Bkr RCP1B Reactor Coolant Pump 1B Backup Bkr 1TB-3 Primary BKR RCP1C Reactor Coolant Pump 1C

, Backup Bkr 1TC-3

Primary BKR RCP1D Reactor Coolant Pump 1D Backup Bkr 1TD-3

2. 600 VAC MCC 1EMXC-F01B Primary Bkr Accumulator 10 Discharge Backup Fuse Isol Viv 1NI76A 1EMXC-F01C Primary Bkr Check Valve Test Header l Backup Fuse Cont isol Viv 1N!95A 1EMXC-F02A Primary Bkr Train A Attemate Power Backup Fuse To ND LTDN Viv 1ND1B 1EMXC-F02B 1 Primary Bkr Hot Leg inj. Check Viv Backup Fuse ' Test isol Viv 1N1153A 1EMXC-F02C Primary Bkr Cont isol at 134 Deg Backup Fuse Annulus Area Viv 1Vl312A 1

1EMXC-FO3A Primary Bkr NC Pump 1C Thermal Barrier Outlet Backup Fuse Isol Viv 1KC345A 1EMXC-FO3B Primary Bkr N2 to Prt Cont Isol inside Backup Fuse Viv 1NC54A 1EMXC-FO3C Primary Bkr Pressurizer Power-Operated Backup Fuse Relief Isol Viv 1NC33A O

l 4

Chapter 16.8-1 Page 4 of 49 05/04/98 l

-( TABLE 16.8-1 A UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVicFR DEVICE NUf1P5fi & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continu'ed) 1EMXC-FO5A Primary Bkr NCDT Vent inside Cont isol Backup Fuse Viv 1WL450A 1EMXC-F05B Primary Bkr Cont Sump Pumps Discharge inside Backup Fuse Cont Isot Viv 1WL825A 1EMXC-F05C Primary Bkr Vent Unit Cond Dm Tank Backup Fuse Outside Cont isol Viv 1WL867A 1EMXC-F06A Primary Bkr NCDT Pumps Disch inside Cont isol -

Backup Fuse Viv 1WL805A 1EMXC-F07B Primary Bkr Cont H2 Purge Outlet Cont isol Backup Fuse Viv 1VY17A P 1EMXD-F01 A Primary Bkr ND Pump 1 A Suction From NC Backup Fuse Loop B Viv 1ND1B 1EMXD-F01B Primary Bkr Accumulator 1B Discharge Backup Fuse Isol Viv 1N165B l

1EMXD-F010 Primary Bkr Ni Pump A to Hot Leg Check j Backup Fuse Viv Test isol Viv 1N1122B 1 1EMXD-F02A Primary Bkr ND Pump 18 Suction from NC Backup Fuse Loop C Viv 1ND368 1 EMXD-F02B Primary Bkr ND to Hot Legs Chk 1N1125,1N1129 Backup Fuse Test isol Viv 1N1154B D

LJ Chapter 16.8-1 Page 5 of 49 05/04/98 i I

l

' TADLE 16.8-1 A UNIT 1 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVicFC ,

DEVICE NUMBER & LOCATION SYSTEM POWERED i

2. 600 VAC MCC (Continued) 1EMXD-F02C Primary Bkr Pressurizer Power-Operated Relief Backup Fuse isol Viv 1NC31B 1EMXD-F05A Primary Bkr Pressurizer Power-Operated Relief Backup Fuse isol Viv 1NC35B 1EMXD-F05B Primary Bkr Rx Bldg Drain Hdr inside Cont

Backup Fuse Isol Viv 1KC429B 1EMXD-F05C Primary Bkr NCDT Hx Cing Water Retum inside Backup Fuse Isol Viv 1KC332B i 4

1EMXD-FCSA Primary Bkr NC Pump 18 Thermal Barrier Outlet '

Backup Fuse isol Viv 1KC364B 1EMXD-F06B (Q) Prir.1ary Bkr Backup Fuse NC Pumps Rtn Hdr inside Cont Isol Viv 1KC424B 1 EMXK-F01 A Drimary Bkr UHI Check Viv Test Line Inside Note 1 t Backup Fuse Cont isol Viv 1N1266A 1EMXK-F010 Primary Bkr Backup N2 to PORV 1NC34A From Backup Fuse Accum Tnk 1 A Viv 1NI438A 1EMXK-F02A '

Primary Bkr NC Pump 1 A Thermal Barrier Backup Fuse Outlet isol Viv 1KC394A 1EMXK-F02B Primary Bkr Lower Cont Vent Units Retum Backup Fuse Cont Isol Viv 1RN484A Note 1: Upon removal of cable from power source associated with the deletion of UHI, this specification is no longer applicable.

N Chapter 16.8-1 Page 6 of 49 05/04/98

4 p

TABLE 16.8-1 A UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE l

\

DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Contirtued) !

^

1EMXK-F02C Primary Bkr NV Supply to Pressurizer Viv Backup Fuse 1NV037A

^ !

1EMXK-F03A Primary Bkr S/G C Blowdown Line Sample Backup Fuse l Inside Cont Isol Viv 1NM210A '

1EMXK F04A Primary Bkr S/G A Upper Shell Sample Inside Backup Fuse )

Cont isol Viv 1NM187A i 1EMXK-F04B Primary Bkr S/G A Blowdown Line Sample Backup Fuse inside Cont Isol Viv 1NM190A 4

1EMXK-F04C i Primary Bkr S/G C Upper Shell Sample 2

Backup Fuse Inside Cont isolViv 1NM207A i O 1EMXK-F06A Primary Bkr Backup Fuse Hydrogen Skimmer Fan 1 A inlet Viv 1VX1 A 1EMXK-F07C Primary Bkr Electric Hydrogen Recombiner Backup Fuse Power Supply Panel 1 A 1EMXK-F09A Primary Bkr Accumulator 1 A Discharge isol Backup Fuse Viv 1N154A 1EMXK-F09B Primary Bkr UHI Ck Viv Test Line Inside Note 1 Backup Fuse Cont isol Viv 1N1267A i

1EMXK-F090 l Primary Bkr NC Pump Oil Fill Header !

Backup Fuse Cont Isol Viv 1NC196A l Note 1: Upon removal of cable from power source associated with the deletion of UHl, this specification is no longer applicable.

i O !

Chapter 16.8-1 Page 7 of 49 05/04/98

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

I !

i TABLE 16.8-1 A 1.INIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVicFR DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) l l 1EMXK-F10A i

! Primary Bkr Containment Air Retum Damper !

Backup Fuse 1 ARF-D-2

)

1EMXK-F108 Primary Bkr VO Fans Suction From Containment !

Backup Fuse isol Viv 1VQ2A 1EMXK-F10C 1 Primary Bkr Cont Air Addition Containment I Backup Fuse isol Viv 1VO16A 1EMXK-F11 A Primary Bkr Containment Air Retum Fan Backup Fuse Motor 1 A 1EMXK-F11B Primary Bkr Hydrogen Skimmer Fan Motor 1 A Backup Fuse 1EMXL-F01B

'g Primary Bkr Backup Fuse Tm B Altemate Power to ND Letdn Viv 1ND37A 1EMXL-F01C Primary Bkr Ni Accum D Sample Line inside

_l Backup Fuse Cont Isol Viv 1NM81B l l

1EMXL-F02A I Primary Bkr NC Pump 1D Thermal Barrier Backup Fuse Outlet isol Viv 1KC413B 1EMXL-F02B Primary Bkr Air Handling units Glycol Retum <

Backup Fuse Cont Isol Viv 1NF233B 1EMXL-F02C Primary Bkr NI Accum C Sample Line inside Backup Fuse Cont IsolViv 1NM78B 1EMXL-F03A Primary Skr S/G D Blowdown Sample Line i Backup Fuse inside Cont isol Viv 1NM2208 i

l iO Chapter 16.8-1 Page 8 of 49 05/04/98

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

l TABLE 16.8-1 A UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 1EMXL-F03B Primary Bkr Ni Accum A Sample Line Inside Backup Fuse Cont isol Viv 1NM728 1EMXL-F030 i Primary Bkr NI Accum B Sample Line inside I Backup Fuse Cont isol Viv 1NM75B i

1EMXL-F04A Primary Bkr S/G B Upper Shell Sample inside i Backup Fuse Cont isol Viv 1NM1978 l 1EMXL-F04B I Prime.y Bkr S/G B Blowdown Sample Line inside Backup Fuse Cont isol Viv 1NM2008 ;

i 1EMXL-F04C Primary Bkr S/G D Upper Shell Sample inside l Backup Fuse Cont isol Viv 1NM217B

,q 1EMXL-F06A

(~) Primary B%r Backup Fuse Hydrogen Skimmer Fan 1B inlet Viv 1VX2B 1EMXL-F06B Primary Bkr Backup N2 to PORV 1NC32B Backup Fuse from Accum Tnk 1B Viv 1NT439B 1EMXL-F07C Primary Bkr Electric Hydrogen Recombiner Backup Fuse Power Supply Panel 1B 1EMXL-F09A Primary Bkr Accumulator 1D Discharge Backup Fuse isol Viv 1N188B 1EMXL-F10A Primary Bkr Containment Air Retum Damper Backup Fuse 1 ARF-D-4 O

Chapter 16.8-1 Page 9 of 49 05/04/98

A

/ TABLE 16.8-1 A v UNIT 1 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVicFR <

1 DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 1EMXL-F10B Primary Bkr Reactor Vessel Head Vent Backup Fuse Viv 1NC251B 1 1EMXL-F100 Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse 1NC252B 4

1EMXL-F11 A Primary Bkr Containment Air Retum i-Backup Fuse Fan Motor 1B 1EMXL-F11B i Primary Bkr Hydrogen Skimmer Fan Motor 1B l Backup Fuse 1EMXS-F01B ;

Primary Bkr NC Pumps Seal Rtn '

Backup Fuse Inside Cont iso! Viv 1NV89A 1EMXS-F02A j Primary Bkr ND Pump 1B Suction from NC

+

Backup Fuse Loop C Viv 1ND37A l 1EMXS-F02B t Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse 1NC250A

1EMXS-F03D Primary Bkr ND Pump 1 A Suction from NC Backup Fuse Loop B Viv 1ND2A i

4 1EMXS-F03E Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse 1NC253A 1

- 1EMXS-F04B Primary Bkr S/G 1D Blowdown inside Cont Backup Fuse Isol Viv 1BB8A 1EMXS-F04C Primary Bkr S/G 1B Blowdown inside Cont Backup Fuse Isol Viv 1BB19A O

Chapter 16.8-1 Page 10 of 49 05/04/98

i TABLE 16.8-1 A rO UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES !

DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) i 1EMXS-F05A Primary Bkr S/G 1 A Blowdown inside Cont >

Backup Fuse isol Viv 1BB56A 1EMXS-F05B l

Primary Bkr S/G 10 Blowdown inside Cont .

Backup Fuse isolViv 1BB60A 1EMXS-F05C I Primary Bkr Pn Liquid Sample Line inside Backup Fuse Cont isol Viv 1NM3A 1EMXS-F06A i

Primary Bkr Pzr Steam Sample Line inside l Backup Fuse Cont IsolViv 1NM6A 3 1EMXS-F06B Primary Bkr NC Hot Leg A Sample Une i Backup Fuse inside Cont isol Viv 1NM22A '

l 1EMXS-F06C O Primary Bkr Backup Fuse NC Hot Leg C Sample Line inside Cont isol Viv 1NM25A i

1MXM-F01 A Primary Bkr Reactor Coolant Pump Motor Backup Fuse Drain Tank Pump Motor 1MXM-F02A Primary Bkr NC Pump 1B Oil Lift Backup Fuse Pump Motor 1 1MXM-F02B Primary Bkr NC Pump 1C Oil Lift Backup Fuse Pump Motor 1 1MXM-F03A Primary Bkr Ice Condenser Power Backup Fuse Transformer ICT1 A iMXM-F03B Primary Bkr Ice Condenser Air Handling Unit Backup Fuse 186 Fan Motor A & B O ;

Chapter 16.8-1 Page 11 of 49 05/04/98 l

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

t

,. g TABLE 16.8-1 A UNIT 1 CONTA"""""MT PEb i NATION CONDUCTOR OVERCURRENT PROTEC n/E DEVmER '

DEVICE NUMBER & LOCATION SYSTEM POWERED !

r

2. 600 VAC MCC (Continued) !

i 1MXM-F03C i Primary Bkr ice Condenser Equipment Access Backup Fuse Door Hoist Motor 1 A 1MXM-F04D Primary Bkr Lighting Transformer 1LR10 Backup Fuse 1MXM-F04E Primary Bkr Lighting Transformer 1LR13 ;

Backup Fuse ,

1MXM-F05A !

Primary Bkr 175 Ton Polar Crane and 25 !

Backup Fuse Ton Aux Crane No. RO13 and RO1S ;

1MXM-F05C i Primary Bkr Upper Containment Welding Feeder !

Backup Fuse 1MXM F06A O Primary Bkr Backup Fuse Ice Condenser Air Handling Unit 1 A7 Fan Motor A & B i

1MXM-F06B ,

Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 188 Fan Motor A & B !

1MXM-F06C Primary Bkr Ice Condenser Air Handling ;

Backup Fuse Unit 1 A9 Fan Motor A & B l 1MXM-F06D I Primary Bkr Ice Condenser Air Handling i Backup Fuse Unit 1B10 Fan Motor A & B !

1MXM-F07B

{

Primary Bkr ice Condenser Air Handhng Backup Fuse Unit 1 A13 Fan Motor A & B 1MXM-F07C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1B14 Fan Motor A & B Chapter 16.8-1 Page 12 of 49 05/04/98

l l

(-

TABLE 16.8-1 A UNIT 1 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE <

DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 1MXM-F08D Primary Bkr ice Condenser Refrigeration Backup Fuse Floor Cool Defrost Heater 1 A 1MXM-F09A Primary Bkr Ice Condenser Air Handling Backup Fuse {

Unit 1 A1 Fan Motor A & B i 1MXM-F09B ,

Primary Bkr Ice Condenser Air Handling Backup Fuse . Unit 1B2 Fan Motor A & B

)

~

1MXM-F09C Primary Bkr Ice Condenser Air Handling Backup Fuee Unit 1 A3 Fan Motor A & B 1MXM-F09D Primary Bkr Ice Condenser Air Handling l Backup Fuse Unit 1B4 Fan Motor A & B 1MXM-F10A I

Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 1 A1 1MXM-F10B Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 1B1 1

1MXN-F01F l Primary Bkr Stud Tensioner Backup Fuse Hoist 1B i i

1MXN-F02A Primary Bkr NC Pump 1B Oil Lift Pump Motor 2 Backup Fuse i

1MXN-F02B '

Primary Bkr NC Pump 1C Oil Lift Pump Motor 2 Backup Fuse !

1MXN-F02E !

Primary Bkr Stud Tensioner Hoist 1C Backup Fuse l

O l

Chapter 16.8-1 Page13 of 49 05/04/98

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

l TABLE 16.8-1 A

%.J UNIT 1 CONTAINaaFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE DEVICE NUMBER & LOCATION GYSTEM POWEREQ

2. 600 VAC MCC (Continued) 1MXN-F03A Primary Bkr Ice Condenser Power Transformer Backup Fuse ICT1B j 1MXN-F03B Primary Bkr Ice Condenser Bridge Crane 1 !

Backup Fuse Crane No. R011 l

1MXN-F03E Primary Bkr Stud Tensioner Hoist 1 A j Backup Fuse i l

1MXN-F04D

{

Primary Bkr Lighting Transformer 1LR5 ;

Backup Fuse 1MXN-F04E Primary Bkr Lighting Transformer 1LR6 Backup Fuse j

/^g 1MXN-F05A U Primary Bkr Backup Fuse Ice Condenser Refrigeration Floor Cool Defrost Heater 1B 1MXN-F05B Primary Bkr Ice Condenser Refrigeration Floor Backup Fuse Cool Pump Motor 1B 1MXN-F05C Primary Bkr Ice Condenser Equipment Access Backup Fuse Door Hoist Motor 1B 1MXN-F06A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 181 Fan Motor A & B 1MXN-F06B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A2 Fan Motor A & B 1MXN-F06C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 183 Fan Motor A & B tm U

Chapter 16.8-1 Page 14 of 49 05/04/98

i l

l I TABLE 16.8-1A UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICER DEVICE NUMBER & LOCATION SYSTEM POWERED l 1

2. 600 VAC MCC (Continued)

' 1MXN-F06D Primary Bkr Ice Condenser Air Handling

, Backup Fuse Unit 1 A4 Fan Motor A & B 1MXN-F078 Primary Bkr Ice Condenser Air Handling ,

' Backup Fuse Unit 1B5 Fan Motor A & B 1

l 1

1MXN-F07C Primary Bkr Ice Condenser Air Handling !

j Backup Fuse Unit 1 A6 Fan Motor A & B j

1MXN-F08A

Primary Bkr Ice Condenser Air Handling j Backup Fuse Unit 187 Fan Motor A & B

)

1MXN-F08B i Primary Bkr ice Condenser Air Handling Backup Fuse l Unit 1 A8 Fan Motor A & B !

g 1MXN-F08C

{Q Primary Bkr Backup Fuse Ice Condenser Air Handling Unit 189 Fan Motor A & B j

1MXN-F08D f

Primary Bkr ice Condenser Air Handling 3 Backup Fuse Unit 1 A10 Fan Motor A & B !

1MXN-F09A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1811 Fan Motor A & B 1MXN-F09B i Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1 A12 Fan Motor A & B 1MXN-F09C Primary Bkr- Ice Condenser Air Handling Backup Fuse Unit 1813 Fan Motor A & B 1MXN-F09D Primary Bkr Ice Condenser Air Hanoling Backup Fuse Unit 1 A14 Fan Motor A & B l O

Chapter 16.8-1 Page 15 of 49 05/04/98

(m TABLE 16.8-1 A UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 1MXN-F10A Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 1 A2 j

1MXN-F10B Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 182 1MXN-F100 Primary Bkr incore Instrumentation Backup Fuse . Sump Pump Motor 1 1MXN-F10D '

Primaiy Bkr Ice Condenser Air Handling Backup Fuse Unit 1815 Fan Motor A & B 1MXO-F01 A i Primary Bkr Upper Containment Air Retum Backup Fuse Fan Motor 1C 1MXO-F01B (Qj Primary Bkr Backup Fuse incore instrument Tunnel Booster Fan Motor 1A

]

l 1MXO-F02B Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 1 A 1MXO-F03A l l

Primary Bkr Lower Containment Ventilation Backup Fuse Unit 1C Fan Motor 1MXO-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 1C Fan Motor 1MXO-F05C Primary Bkr Containment Pipe Tunnel Backup Fuse Booster Fan Motor 1 A 1MXP-F01 A Primary Bkr Upper Containment Retum Backup Fuse j Air Fan 1B O

O j i

Chapter 16.8-1 Page 16 of 49 05/04/98 i

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

l i

s TABLE 16.8-1 A l.INIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVicFR DEVICE NUMBER & LOCATION SYSTEM POWERED

, 2. 600 VAC MCC (Continued) 1MXP-F01B l

Primary Bkr incore Instrument Tunnel Backup Fuse Booster Fan Motor 1B i 1MXP F02B Primary Bkr Control Rod Drive Vent .

Backup Fuse Fan Motor 1B 1MXP-F03A !

Primary Bkr Lower Containment Ventilation Backup Fuse Unit 1B Fan Motor i

1MXP-F04D i

Primary Bkr Upper Containment Ventilation '

Backup Fuse Unit 1B Fan Motor 1MXP-F05C Primary Bkr Containment Pipe Tunnel Backup Fuse Booster Fan Motor 1B p 1MXQ-F01 A Q Primary Bkr Backup Fuse Upper Containment Retum i Air Fan Motor 1 A 1MXQ-F01B Primary Bkr incore Instrument Room Ventilation Unit 1 A Fan Motor Unit 1 A Fan Motor 1MXQ-F028 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 1C l 1MXQ-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit 1 A Fan Motor 1MXQ-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 1 A Fan Motor 1MXR-F01 A Primary Bkr Upper Containment Retum Air Backup Fuse Fan Motor 1D O

Chapter 16.8-1 Page 17 of 49 05/04/98

t

't TABLE 16.8-1 A C) UNIT 1 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 1MXR-F01B

, Primary Bkr Incore Instrument Room Ventila "

Backup Fuse tion Unit 1B Fan Motor 1MXR-F02B Primary Bkr Control Rod Drive Vent Backup Fuse Fan Motor 1D l l

1MXR-F03A Primary Bkr Lower Containment Ventilation i Backup Fuse Unit 1D Fan Motor ;

1MXR-F04C i Primary Bkr Upper Containment Ventilation i Backup Fuse Unit 1D Fan Motor 1MXY-F02A l Primary Bkr NC Pump 1 A Oil Lift Pump Motor 1 Backup Fuse )

1MXY-F02B i O Primary Bkr Backup Fuse NC Pump 1D Oil Lift Pump Motor 1 i 1MXY-F02C Primary Bkr Reactor Building Lower Containment Backup Fuse Welding Machine Receptacle 1RCPLO185 1MXY FO2D Primary Bkr Upper Containment Reactor Building Backup Fuse Welding Receptacle 1RCPLO193 1MXY-F03A Primary Bkr Reactor Coolant Drain Tank Pump Backup Fuse Motor 1 A 1MXY-F03D Primary Bkr Ice Condenser Refrigeration Backup Fuse Floor Cool Pump Motor 1 A 1MXY F05A 4

Primary Bkr Lighting Transformer Backup Fuse 1LR8 1MXY-F05B Primary Bkr Lighting Transformer Backup Fuse O 1LR11 Chapter 16.8-1 Page 18 of 49 05/04/98

_. . _ . _ _ . - _ _ . . _ _ _ _ _ m._.. _.__. .. . . ~ _ _ .. __. _ .. _. .

l TABLE 16.8-1 A ,

C) UNIT 1 CONTAINaazNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE DEVICE NUMBER & LOCATION SYSTEM POWERED l

2. 600 VAC MCC (Continued) I 1MXY-F05C Primary Bkr Lighting Transformer Backup Fuse 1LR14 1MXY-F06A Primary Bkr Ice Condenser Air Handling l Backup Fuse Unit 1 A5 Fan Motor A & B 1MXY-F06B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1 A11 Fan Motor A & B 1MXY-F06C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1812 Fan Motor A & B 1MXY-F06D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 1A15 Fan Motor A & B 1MXY-F08A Primary Bkr incore Drive Assembly )

Backup Fuse Motor 1 A l 1MXY-F08B Primary Bkr incore Drive Assembly Backup Fuse Motor 10 1MXY-F08C Primary Bkr incore Drive Assembly Backup Fuse Motor 1E 1MXY-F08D Primary Bkr Lower Containment Auxiliary l Backup Fuse Charcoal Filter Unit Fan Motor 1A 1MXZ-F02A Primary Bkr NC Pump 1 A Oil Lift Pump Backup Fuse Motor 2 1MXZ-F02B Primary Bkr NC Pump 1D Oil Lift Pump Backup Fuse Motor 2 i

O i Chapter 16.8-1 Page 19 of 49 05/04/98

TABLE 16.8-1 A !

UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES '

DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Conta?ued) 1MXZ-F03A Primary Bkr Reactor Coolant Drain Tank i Backup Fuse Pump Motor 1B 1MXZ-F04B Primary Bkr Lighting Transformer 1LR1 ;

Backup Fuse i

1MXZ-F04C Primary Bkr Lighting Transformer 1LR2 Backup Fuse 1MXZ-F04D Primary Bkr Lighting Transformer 1LR3 Backup Fuse 1MXZ-F05A Primary Bkr Reactor Coolant Pump Jib Backup Fuse Hoist No. RO19 TH R022 3 1MXZ-F05C (Q Primary Bkr Backup Fuse Lower Containment Auxiliary Charcoal Filter Unit Fan Motor 1B J 1MXZ-F06A Primary Bkr incore Drive Assembly Motor 1B l Backup Fuse j 1MXZ-F068 '

Primary Bkr incore Drive Assembly Motor 1D Backup Fuse ;

1MXZ-F06C Primary Bkr incore Drive Assembly Motor 1F Backup Fuse 1MXZ-FO6D Primary Bkr Lower Containment Reactor Building Backup Fuse Welding Receptacle 1RCPLO194 1MXZ F078 Primary Bkr Lighting Transformer 1LR4 Backup Fuse 1MXZ-F07C Primary Bkr 5 Ton Jib Crane in Containment Backup Fuse Crane No. R005 O

Chapter 16.8-1 Page 20 of 49 05/04/98

i TABLE 16.81A

[m)%./ UNIT 1 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 1MXZ-F07D Primary Bkr Reactor Cavity Manipulator Backup Fuse Crane No R007 & R027 1MXZ-F08A Primary Bkr Steam Generator Drain Pump Backup Fuse Motor 1 1MXZ-F080 Primary Bkr 15 Ton Equipment Access Hatch Backup Fuse Hoist Crar's Nc,. R009 1MXZ-F08D Primary Bkr Control Rod Drive 2 Ton Jib Backup Fuse Hoist Crane No. RO17 1MXZ-F08E Primary Bkr Reactor Side Fuel Handling Backup Fuse Control Console

()

f"% SMXG-F010 Primary Bkr Backup Fuse Standby Makeup Pump Drain isol Viv 1NV876 SMXG F05C Primary Bkr Pressurizer Heaters 28,55 & 56 Backup Fuse SMXG-F06A Primary Bkr Standby Makeup Pump to Seal Backup Fuse Water Line iso! Viv 1NV877

3. 600 VAC Pressurizer Heater Power Panels PHP1A-F01A Primary Bkr Pressurizer Heaters Backup Fuse 1,2, & 22 PHP1A-F01B Primary Bkr Pressurizer Heaters Backup Fuse 5,6,& 27 Chapter 16.8-1 Page 21 of 49 05/04/98

f i

TABLE 16.8-1A UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTfVE DEVICEE DEVICE NUMBER & LOCATION SYSTEM POWERED

3. !

600 VAC Pressurizer Heater Power Panels (Continued)

PHP1A-F01C 3 Primary Bkr Pressurizer Heaters Backup Fuse 9,10 & 32 !

PHP1A-F02C Primary Bkr Pressurizer Heaters ,

Backup Fuse 11,12 & 35 i PHP1A-F02D Primary Bkr Pressurizer Heaters '

Backup Fuse 13,14 & 37 PHP1A-F02E i

Primary Bkr Pressurizer Heaters !

Backup Fuse 17,18 & 42 l

)

PHP1B-F01A Primary Bkr )

Pressurizer Heaters Backup Fuse j 21,47 & 48 i G PHP1B-F01B Q Primary Bkr Backup Fuse Pressurizer Heaters 26,53 & 54 PHP1B-F01C Primary Skr Pressurizer Heaters Backup Fuse 31,59 & 60 PHP18-F02C Primary Bkr Pressurizer Heaters Backup Fuse 36,65 & 66 PHP1B-F02D Primary Bkr Pressurizer Heaters Backup Fuse 41,71 & 72 PHP1B-F02E Primary Bkr Pressurizer Heaters Backup Fuse 46,77 & 78 PHP10-F01 A Primary Bkr Pressurizer Heaters Backup Fuse 7,8 & 30 I

Chapter 16.8-1 Page 22 of 49 05/04/98

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

I O TABLE 16.81 A '

m .BAIlOhLCQHfEl.CIQB DEVICE NUMBER & LOCATION gy,gTEM POWERED s 3.

600 VAC Pressurizer Heater Power Panels (Continued)

PHP10-F01B Primary Bkr Pressurizer Heaters Backup Fuse 19,20 & 45 PHPIC-F01C Primary Bkr Pressurizer Heaters Backup Fuse !

24, 51 & 52 PHP1C-F01D Primary Bkr Pressurizer Heaters Backup Fuse 29, 57 & 58 PHP10-F02C Primary Bkr Pressurizer Heaters Backup Fuse 34,63 & 64 PHP10-F02D Primary Bkr Pressurizer Heaters '

Backup Fuse 39,69 & 70 t

f PHP1C-F02E Primary Bkr Pressurizer Heaters Backup Fuse 44, 75 & 76 PHP1D-F01 A Primary Bkr Pressurizer Heaters Backup Fuse 3,4 & 25 ,

PHP1D-F01B Primary Bkr Pressurizer Heaters Backup Fuse 15,16 & 40 '

PHP1D-F01C Primary Bkr Pressurizer Heaters Backup Fuse 23,49 & 50 PHP1D-F02C Primary Bkr Pressurizer Heaters ,

Backup Fuse 33,61 & 62 l PHP1D-F02D Primary Bkr Pressurizer Heaters Backup Fuse 38,67 & 68 i

L O ;

Chapter 16.8-1 Page 23 of 49 05/04/98

TABLE 16.8-1 A p)g UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED 3.

600 VAC Pressurizer Heater Power Panels (Continued)

PHP1D-F02E Primary Bkr Pressurizer Heaters Backup Fuse 43,73 & 74

4. 250 VDC Reactor Building Deadlight Panelboard 1DLD-2 1

Primary Bkr Lighting Panelboard No.1LR1, Backup Fuse 1LR2,1LR3,1LR4 1 1DLD-3 Primary Bkr Lighting Panelboard No.1LR13, Backup Fuse 1LR14 1 DLD-4 Primary Bkr Lighting Panelboard No.1LRS, Backup Fuse 1LR6 1DLD-5 Primary Bkr Lighting Panelboard No.1LR10, (N Backup Fuse 1LR11 1DLD-10 Primary Bkr Lighting Panelboard No.1LR8 Backup Fuse

5. 120 VAC Panelboards 1ELB1-5 Primary Bkr Emergency A.C. Lighting Backup Fuse 1ELB1-7 Primary Bkr Emergency A.C. Lighting Backup Fuse i

1ELB1-13 !

Primary Bkr Emergency A.C. Lighting Backup Fuse 1ELB1 15 Primary Bkr Emergency A.C. Lighting '

Backup Fuse R.)

Chapter 16.8-1 Page 24 of 49 05/04/98 l

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

TABLE 16.8-1 A O UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEViccm DEVICE NUMBER & LOCATION SYSTEM POWERED P

5. 120 VAC Panelboards (Continued) 1ELB1-17 Primary Bkr Emergency A.C. Lighting !

Backup Fuse 1KPM-1 Primary Bkr NC Pump Motor 1 A Space Heater l Backup Fuse

1KPM-2 Primary Bkr NC Pump Motor 1C Space Heater Backup Fuse 1KPM-7 t Primary Bkr Lower Containment Vent Unit Backup Fuse 1 A Fan Motor Space Heater :

1KPM-8-1 Primary Bkr Lower Containment Vent Unit I Backup Fuse 1C Fan Motor Space Heater

(~ 1KPM-24-1

{

Primary Bkr Control Rod Drive Vent Fan l Backup Fuse Motor 1 A Space Heater ,

l 1KPM-24-2 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 1B Space Heater 1 KPM-24-3 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 10 Space Heater 1KPM-24 4 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 1D Space Heater 1KPM-33 4,5,6,7 Primary Bkr Safety injection System Backup Fuse Temperature Transmitters 1KPN-1 Primary Bkr NC Pump Motor 1B Space Heater

,. Backup Fuse O l 1

Chapter 16.8-1 Page 25 of 49 05/04/98

l 7N TABLE 16.8-1 A

) UNIT 1 CONTA!NM8:NT PENETRATION CONDUC DEVICE NUMBER & LOCATION SYSTEM POWERED i S. 120 VAC Panelboards (Continued) 1KPN-2 Primary Bkr NC Pump Motor 1D Space Heater Backup Fuse 1KPN-7-1

Primary Bkr Lower Containment Vent Unit Backup Fuse 1B Fan Motor Space Heater 1KPN-8-1 Primary Bkr Lower Containment Vent Unit Backup Fuse 1D Fan Motor Space Heater i 1 KPN-8-2,3,4,5 NC Pump Seal Stand Pipe Vent and Drain Valves !

i 1KPN-11 Primary Bkr Misc Control Power Backup Fuse for 1 ATC 24

6. DC Welding Circuits e3 !

1EOCB0001 ;

Primary Bkr-AA Lower Containment DC Backup Bkr-AB Welding Circuit 1EOCB0002 Primary Bkr-AA Upper Containment DC Backup Bkr-AB Welding Circuit l

l l

L O

Chapter 16.8-1 Page 26 of 49 05/04/98 i

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

r TABLE 16.8-1B

( UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE DEVICE NUMBER & LOCATION SYSTEM POWERED

1. 6900 VAC Swgr Primary Bkr RCP2A Reactor Coolant Pump 2A Backup Bkr 2TA-3 Primary Bkr RCP2B Reactor Coolant Pump 2B Backup Bkr 2TB-3 Primary BKR RCP2C Reactor Coolant Pump 2C Backup Bkr 2TC 3 Primary BKR RCP2D Reactor Coolant Pump 2D Backup Bkr 2TD 3 2.- 600 VAC MCC 2EMXC-F01B Primary Bkr Accumulator 20 Discharge Backup Fuse Iso! Viv 2NI76A 2EMXC-F01C Primary Bkr Check Valve Test Header Backup Fuse Cont Iso! Viv 2N195A 2EMXC-F02A Primary Bkr Train A Altemate Power Backup Fuse To ND LTDN Viv 2ND1B 2EMXC-F028 Primary Bkr Hot Leg inj. Check Viv Backup Fuse ,

Test Isol Viv 2N1153A 2EMXC-FO2C Primary Bkr Cont isol at 134 Deg Backup Fuse Annulus Area Viv 2Vl312A 2EMXC-F03A Primary Bkr NC Pump 2C Thermal Barrier Outlet Backup Fuse Isol Viv 2KC345A 2EMXC-FO3B Primary Bkr N2 to Pit Cont isolinside Backup Fuse Viv 2NC54A 2EMXC-FO3C Primary Bkr Pressurizer Power-Operated Backup Fuse Relief Isol Viv 2NC33A O

i Chapter 16.8-1 Page 27 of 49 05/04/98

{

TABLE 16,8-1B

[~ UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES I

1 DEVICE NUMBER & LOCATION SYSTEM POWERED

2. !

600 VAC MCC (Continued) 2EMXC-FO5A Primary Bkr NCDT Vent inside Cont isol Backup Fuse Viv 2WL450A

(

i 2EMXC-F05B Primary Bkr Cont Sump Pumps Discharge Inside Backup Fuse Cont isol Viv 2WL825A

[

2EMXC-F05C Primary Bkr Vent Unit Cond Dm Tank Backup Fuse l Outside Cont isol Viv 2WL867A 2EMXC-F06A '

Primary Bkr NCDT Pumps Disch inside Cont isol Backup Fuse Viv 2WL805A 2EMXC-F07B Primary Bkr Cont H2 Purge Outlet Cont Isol ,

Backup Fuse Viv2VY17A (g 2EMXD-F01A i t

Primary Bkr ND Pump 2A Suction From NC

(]' Backup Fuse Loop B Viv 2ND1B :

2EMXD-F01B Primary Bkr Accumulator 2B Discharge Backup Fuse Isol Viv 2NI65B 2EMXD-F010 Primary Bkr NI Pump A to Hot Leg Check !

Backup Fuse Viv Test isol Viv 2NI122B 2EMXD-F02A Primary Bkr ND Pump 2B Suction from NC Backup Fuse Loop C Viv 2ND36B 2EMXD-F02B Primary Bkr ND to Hot Legs Chk 2N1125,2N1129 Backup Fuse Test isol Viv 2N1154B O

l Chapter 16.8-1 Page 28 of 49 05/04/98 '

4 TABLE 16.8-1B UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PPOTECTIVE DEVICEE DEVICE NUMBER & LOCATION SYSTEM POWERED i

2. 600 VAC MCC (Continued) 2EMXD-F02C Primary Bkr Pressurizer Power-Operated Relief Backup Fuse Isol Viv 2NC31B 2EMXD-F05A Primary Bkr Pressurizer Power-Operated Relief ,

., Backup Fuse Isol Viv 2NC35B 2EMXD-F05B i Primary Bkr Rx Bldg Drain Hdr inside Cont Backup Fuse isol Viv 2KC429B 2EMXD-F05C Primary Bkr NCDT Hx Cing Water Retum inside Backup Fuse Isol Viv 2KC332B 2EMXD-F06A Primary Bkr NC Pump 2B Thermal Barrier Outlet Backup Fuse isol Viv 2KC364B n 2EMXD-F06B '

Primary Bkr NC Pumps Rtn Hdr inside Cont Backup Fuse Isol Viv 2KC424B

, 2EMXK-F01A !

Primary Bkr UHI Check Viv Test Line inside

Note 1 i Backup Fuse Cont isol Viv 2N1266A 2EMXK-F01C

, Primary Bkr Backup N2 to PORV 2NC34A From Backup Fuse Accum Tnk 2A Viv 2N1438A 2EMXK-F02A Primary Bkr NC Pump 2A Thermal Barrier a

Backup Fuse Outlet isol Viv 2KC394A 1

2EMXK-F02B Primary Bkr Lower Cont Vent bnits Retum Backup Fuse Cont isol Viv 2RN484A 4

Note 1: Upon removal of cable from power source associated with the deletion of UHI, this specification is no longer applicable.

!O Chapter 16.8-1 Page 29 of 49 05/04/98 ,

N a w---lLa- i-,---es.2 ,aa = 4L .M4 rk iu-a+ * --ws--A a-uos-u--,- 4e- A bs-J- m A a,a n"4ss-l l

TABLE 16.8-1B

~) UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVicFA J

%/

DEVICE NUMBER & LOCATION SYSTEM POWERED l

2. 600 VAC MCC (Continued) l 1

2EMXK-F02C 1 Primary Bkr NV Supply to Pressurizer Viv i

Backup Fuse 2NV037A 2EMXK-F03A i

Primary Bkr S/G C Blowdown Line Sample Backup Fuse inside Cont isol Viv 2NM210A l

2EMXK-F04A j

Primary Bkr S/G A Upper Shell Sample inside i Backup Fuse Cont iso! Viv 2NM187A 1 2EMXK-F04B i

Primary Bkr S/G A Blowdown Line Sample l Backup Fuse inside Cont 1801 Viv 2NM190A 2EMXK-F04C i Primary Bkr S/G C Upper Shell Sample :

Backup Fuse inside Cont Isol Viv 2NM207A ,

J 2EMXK-F06A

(()N Primary Bkr Hydrogen Skimmer Fan 2A Backup Fuse inlet Viv 2VX1 A 2EMXK-F07C Primary Bkr Electric Hydrogen Recombiner Backup Fuse Power Supply Panel 2A 2EMXK-F09A Primary Bkr Accumulator 2A Discharge isol Backup Fuse Viv 2N154A 2EMXK-F09B j

Primary Bkr UHI Check Viv Test Line inside

Note 1 Backup Fuse Cont Isol Viv 2N1267A 2EMXK-F09C Primary Bkr NC Pump Oil Fill Header Backup Fuse Cont isol Viv 2NC196A 1

Note 1: Upon removal of cable from power source associated with the deletior, of UHI, this specification is no longer applicable. i l

l Chapter 16.8-1 Page 30 of 49 05/04/98 l

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

TABLE 16.8-1B tJNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCf]RRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 2EMXK-F10A Primary Bkr Containment Air Retum Damper Backup Fuse 2 ARF-D-2 2EMXK-F10B Primary Bkr VO Fans Suction From Containment 1

~

Backup Fuse Isol Viv 2VO2A ,

2EMXK F10C Primary Bkr Cont Air Addition Containment Backup Fuse Isol Viv 2VO16A 2EMXK-F11 A Primary Bkr Containment Air Retum Fan Backup Fuse Motor 2A '

2EMXK-F11B Primary Bkr Hydrogen Skimmer Fan Motor 2A Backup Fuse

(

t 2EMXL-F01B Primary Bkr Tm B Altt;nate Power to ND Backup Fuse Letdn Viv 2ND37A 2EMXL-F010 Primary Bkr NI Accum D Sample Line inside Backup Fuse Cont isol Viv 2NM81B 2EMXL-F02A Primary Bkr NC Pump 2D Thermal Barrier Backup Fuse Outlet isol Viv 2KC413B 2EMXL-F02B Primary Bkr Air Handling units Glycol Retum Backup Fuse Cont isol Viv 2NF233B 2EMXL-F02C Primary Bkr Ni Accum C Sample Line inside Backup Fuse Cont isolViv 2NM78B 2EMXL-F03A Primary Bkr S/G D Blowdown Sample Line inside Backup Fuse Cont Iso! Viv 2NM220B O

Chapter 16.8-1 Page 31 of 49 05/04/98

( TABLE 16.8-1B UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 2EMXL-F03B Primary Bkr NI Accum A Sample Line inside Backup Fuse Cont isol Viv 2NM72B 2EMXL-F03C Primary Bkr NI Accum B Sample Line inside Backup Fuse Cont isol Viv 2NM758 2EMXL F04A Primary Bkr S/G B Upper Shell Sample inside Backup Fuse Cont isol Viv 2NM1978 2EMXL-F04B Primary Bkr S/G B Blowdown Sample Line inside Backup Fuse Cont isol Viv 2NM2008 2EMXL F04C Primary Bkr S/G D Upper Shell Sample inside Backup Fuse Cont Isol Viv 2NM2178

/7 2EMXL-F06A d Primary Bkr Backup Fuse Hydrogen Skimmer Fan 2B inlet Viv 2VX2B 2EMXL-F06B Primary Bkr Backup N2 to PORV 2NC32B Backup Fuse from Accum Tnk 2B Viv 2N1439B 2EMXL-F07C Primary Bkr Electric Hydrogen Recombiner Backup Fuse Power Supply Panel 2B 2EMXL-F09A Primary Bkr Accumulator 2D Discharge Backup Fuse Isol Viv 2N188B 2EMXL F10A Primary Bkr Containment Air Retum Damper Backup Fuse 2 ARF-D-4 O

Chapter 16.8-1 Page 32 of 49 05/04/98

p) s TABLE 16.8-1B UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICER v

DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 2EMXL F10B Primary Bkr Reactor Vessel Head Vent Backup Fuse Viv 2NC251B 2EMXL-F10C Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse 2NC252B 2EMXL-F11 A '

Primary Bkr Containment Air Retum Backup Fuse Fan Motor 2B j 2EMXL-F11B 4 Primary Bkr Hydrogen Skimmer Fan Motor 2B !

Backup Fuse l

2EMXS-F01B i Primary Bkr NC Pumps Seal Rtn Backup Fuse Inside Ccnt isol Viv 2NV89A

[~]

(/

2EMXS-F02A Primary Bkr ND Pump 2B Suction from NC l

Backup Fuse Loop C Viv 2ND37A '

2EMXS-F02B Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse 2NC250A 2EMXS-F03D Primary Skr ND Pump 2A Suction from NC Backup Fuse Loop B Viv 2ND2A

~

2EMXS-F03E -

Primary Bkr Reactor Vessel Head Vent Viv Backup Fuse 2NC253A 2EMXS-F04B Primary Bkr S/G 2D Blowdown inside Cont Backup Fuse isol Viv 2BB8A 2EMXS-F04C Primary Bkr S/G 2B Blowdown inside Cont ,

Backup Fuse Isol Viv 2BB19A l

O V

~

Chapter 16.8-1 Page 33 of 49 05/04/98

p. s TABLE 16.81B UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) !

2EMXS-F05A Primary Bkr S/G 2A Blowdown inside Cont Backup Fuse isol Viv 2BB56A 2EMXS-F05B Primary Bkr S/G 2C Blowdown inside Cont Backup Fuse isol Viv 2BB60A 2EMXS-F05C Primary Bkr Pzr Liquid Sample Line inside Backup Fuse Cont Isol Viv 2NM3A 2EMXS-F06A Primary Bkr Pzr Steam Sample Line inside Backup Fuse Cont isol Viv 2NM6A 2EMXS-F06B Primary Bkr NC Hot Leg A Sample Line Backup Fuse inside Cont isol Viv 2NM22A Q

v 2EMXS-F06C Primary Bkr NC Hot Leg C Sample Line Backup Fuse inside Cont isol Viv 2NM25A 2MXM-F01A Primary Bkr Reactor Coolant Pump Motor Backup Fuse Drain Tank Pump Motor 2MXM-F02A Primary Bkr NC Pump 28 Oil Lift Backup Fuse Pump Motor 1 2MXM-F02B Primary Bkr NC Pump 2C Oil Lift Backup Fuse Pump Motor 1 2MXM-F03A Primary Bkr Ice Condenser Power Be& 9 Fuse Transformer ICT2A 2MXM-F03B Primary Bkr Ice Condenser Air Handling Unit Backup Fuse 2B6 Fan Motor A & B O

Chapter 16.8-1 Page 34 of 49 05/04/98

I l

l TABLE 16.8-1B i UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE DEYlCE NUMBER & LOCATION SYSTEM POWERED l

2. 600 VAC MCC (Continued) 2MXM-F03C Primary Bkr Ice Condenser Equipment Access Backup Fuse Door Hoist Motor 2A 2MXM-F04D I Primary Bkr Lighting Transformer 2LR10 Backup Fuse l

i 2MXM-F04E Primary Bkr Light)ng Tramformer 2LR13 Backup Fuse 2MXM-F05A Primary Bkr 175 Ton Polar Crane and 25 Backup Fuse Ton Aux Crane No. R014 and RO16 2MXM-F050 Primary BKr Upper Containment Welding Feeder Backup Fuse I

p 2MXM-F06A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A7 Fan Motor A & B 2MXM-F06B Primary Bkr Ice Condenser Ai Handling Backup Fuse Unit 2B8 Fan Motor A & B 2MXM-F060 1 Primary Bkr Ice Condenser Air Handling l Backup Fuse Unit 2A9 Fan Motor A & B 2MXM-F06D j Primary Bkr Ice Condenser Air Handling '

Backup Fuse Unit 2B10 Fan Motor A & B 2MXM-F07B Primary Bkr ice Condenser Air Handling Backup Fuse Unit 2A13 Fan Motor A & B l

2MXM-F070 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B14 Fan Motor A & B I

O Chapter 16.8-1 Page 35 of 49 05/04/98

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

?

i e

4 r~ TABLE 16.8 ,

inwT 2 cGidTe--^^ WT PEb mATEN nrMLicTOft CVSN ^ .._ZidT pie 6TE6 udE DEMEos

DEVICE NUMBER & LOCATION SYSTEM POWERED 2.' 600 VAC MCC (Contint.ed)

! 2MXM-F080 -

! Primary Bkr Ice Condenser Refrigeration

- Backup Fuse Floor Cool Defrost Heater 2A 2MXM-F09A Primary Bkr Ice Condenser Air Handling j_ Backup Fuse Unit 2A1 Fan Motor A & B 2MXM-F09B t-Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B2 Fan Motor A & B 2MXM-F09C Primary Bkr ice Condenser Air Handling Backup Fuse Unit 2A3 Fan Motor A & B 2MXM-F09D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B4 Fan Motor A & B 2MXM-F10A Primary Bkr Containment Floor and Equipment Backup Fuse ' Sump Pump Motor 2A1 2MXM-F108 Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 2B1 '

2MXN-F01F ,

Primary Bkr Stud Tensioner Backup Fuse Hoist 2B 2MXN-F02A Primary Bkr NC Pump 2B Oil Lift Pump Motor 2 Backup Fuse 2MXN-F02B Primary Bkr NC Pump 2C Oil Lift Pump Motor 2 Backup Fuse 2MXN-F02E Primary Bkr Stud Tensioner Hoist 20 Backup Fuse O

Chapter 16.8-1 Page 36 of 49 05/04/98

i /] TABLE 16.8-1B UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued)

. 2MXN-F03A i

Primary Bkr Ice Condenser Power Transformer

. Backup Fose ICT2B 2MXN-F03B Primary Bkr Ice Condenser Bridge Crane 2 Backup Fuse Crane No. R012 2MXN-F03E Primary Bkr Stud Tensioner Hoist 2A Backup Fuse 2MXN-F04D Primary Bkr Lighting Transformer 2LRS Backup Fuse i 2MXN-F04E Primary Bkr Lighting Transformer 2LR6 Backup Fuse D 2MXN-F05A

( Primary Bkr Ice Condenser Refrigeration l Backup Fuse Floor Cool Defrost Heater 28 2MXN-F05B 1

Primary Bkr Ice Condenser Refrigeration Floor Backup Fuse Cool Pump Motor 2B 2MXN-F05C Primary Bkr Ice Condenser Equipment Access Backup Fuse Door Hoist Motor 2B 2MXN-F06A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B1 Fan Motor A & B 2MXN-F06B Primary Bkr ice Condenser Air Handling Backup Fuse Unit 2A2 Fan Motor A & B 2MXN-F06C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B3 Fan Motor A & B ,

i l

O ;

Chapter 16.8-1 Page 37 of 49 05/04/98

TABLE 16.8-1B O 11tilI.1CQHIAltiMEliJ PENETRATION CONDUCTOR OVEqCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 2MXN-F06D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A4 Fan Motor A & B 2MXN-F07B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 285 Fan Motor A & B 2MXN-F07C Primary Bkr Ice Condenser Air Handling !

Backup Fuse Unit 2A6 Fan Motor A & B ;

2MXN-F08A l Primary Bkr Ice Condenser Air Handling Backup Fuse l Unit 2B7 Fan Motor A & B 2MXN-F08B ,

Primary Bkr Ice Condenser Air Handling l Backup Fuse Unit 2A8 Fan Motor A & B l l

{N s")

2MXN-F08C Primary Bkr Ice Condenser Air Handling i l

Backup Fuse Unit 2B9 Fan Motor A & B 2MXN-F08D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A10 Fan Motor A & B 2MXN-F09A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B11 Fan Motor A & B 2MXN-F09B Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A12 Fan Motor A & B ,

2MXN-F09C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B13 Fan Motor A & B 2MXN-F09D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A14 Fan Motor A & B Chapter 16.8-1 Page 38 of 49 05/04/98

I TABLE 16.8-1B

(

C UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICER DEVICE NUMBER & LOCATION SYSTEM POWERED l

l

2. 600 VAC MCC (Continued) l 2MXN-F10A i

Primary Bkr Containment Floor and Equipment Backup Fuse Sump Pump Motor 2A2 2MXN-F108 Primary Bkr Containment Floor and Equipment l

l Backup Fuse Sump Pump Motor 2B2 '

2MXN-F10C l Primary Bkr incore Instrumentation Backup Fuse Sump Pump Motor 2 2MXN-F10D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B15 Fan Motor A & B 2MXO-F01A Primary Bkr Upper Containment Air Retum i Backup Fuse Fan Motor 2C 1

,c 2MXO-F02B Q) Primary Bkr Backup Fuse Control Rod Drive Vent Fan Motor 2A 2MXO-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit 2C Fan Motor 2MXO-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 20 Fan Motor 2MXO-F05C Primary Bkr Containment Pipe Tunne!

Backup Free Booster Fan Motor 2A i

2MXP.r A Prirrary Bkr Upper Containment Retum Backup Fuse Air Fan 2B Chapter 16.8-1 Page 39 of 49 05/04/98

(q TABLE 16.8-1B UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE

- DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued)

-l 2MXP-F02B Primary Bkr Control Rod Drive Vent Backup Fuse Fan Motor 2B 2MXP-F03A Primary Bkr Lower Containment Ventilation Backup Fuse Unit 28 Fan Motor i

2MXP-F04D Primary Bkr Upper Containment Ventilation Backup Fuse Unit 2B Fan Motor j

2MXP-F05C Primary Bkr Containment Pipe Tunnel Backup Fuse Booster Fan Motor 2B 2MXO-F01A Primary Bkr Upper Containment Retum Backup Fuse Air Fan Motor 2A 2MXQ-F01B !

q,, Primary Bkr - Incore instrument Room Venti-Backup Fuse lation Unit 2A Fan Motor 2MXO-F02B Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 20 2MXO-F03A Primary Bkr Lower Containment Ventilation ,

Backup Fuse Unit 2A Fan Motor 2MXO-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 2A Fan Motor 2MXR-F01A Primary Bkr Upper Containment Retum Air Backup Fuse Fan Motor 2D O

Chapter 16.8-1 Page 40 of 49 05/04/98

b o TABLE 16.8-1B 1 UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 2MXR-F01B Primary Bkr Incore instrument Room Ventila-Backup Fuse tion Unit 2B Fan Motor '

2MXR-F02B Primary Bkr Control Rod Drive Vent =

Backup Fuse Fan Motor 2D 2MXR-F03A I

Primary Bkr Lower Containment Ventilation Backup Fuse Un't- 2D Fan Motor 2MXR-F04C Primary Bkr Upper Containment Ventilation Backup Fuse Unit 2D Fan Motor 2MXY-F02A i

Primary Bkr NC Pump 2A Oil Lift Pump Motor 1 l Backup Fuse i

O 2MXY-F028 Primary Bkr Backup Fuse NC Pump 2D Oil Lift Pump Motor 1 2MXY-F02C Primary Bkr Reactor Building Lower Containment Backup Fuse Welding Machine Receptacle 2RCPLO185 2MXY-FO2D Upoer Containment Reactor Building Primary Bkr Welding Receptacle 2RCPLO193 Backup Fuse 2MXY-F03A Primary Bkr Reactor Coolant Drain Tank Pump Backup Fuse Motor 2A 2MXY F03D Primary Bkr lee Condenser Refrigeration Backup Fuse Floor Cool Pump Motor 2A 2MXY-F05A Primary Bkr Lighting Transformer Backup Fuse 2LR8 2MXY F05B Primary Bkr Lighting Transformer Backup Fuse 2LR11 Chapter 16.8-1 Page 41 of 49 05/04/98

TABLE 16.8-1B

,O UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE M DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 2MXY-F05C Primary Bkr Lighting Transformer Backup Fuse 2LR14 .

2MXY-F06A Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A5 ran Motor A & B 2MXY-F068 Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A11 Fan Motor A & B 2MXY-F06C Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2B12 Fan Motor A & B 2MXY-F06D Primary Bkr Ice Condenser Air Handling Backup Fuse Unit 2A15 Fan Motor A & B

^ 2MXY F07C Primary Bkr EXH Reactor Building Equipment Backup Fuse Hatch Jib Cranes R035 & R036 2MXY-F08A Primary Bkr incore Drive Assembly Backup Fuse Motor 2A 2MXY F08B Primary Bkr incore Drive Assembly Backup Fuse Motor 20 2MXY F08C Primary Bkr incore Drive Assembly Backup Fuse Motor 2E 2MXY-F08D Primary Bkr Lower Containment Auxiliary Backup Fuse Charcoa! Filter Unit Fan Motor 2A 2MXZ-F02A Primary Bkr NC Pump 2A Oil Lift Pump Backup Fuse Motor 2 O

Chapter 16.8-1 Page 42 of 49 05/04/98

l TABLE 16.8-1B UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICER DEVICE NUMBER & LOCATION SYSTEM POWERED

2. 600 VAC MCC (Continued) 2MXZ-F02B

, Primary Bkr NC Pump 2D Oil Lift Pump Backup Fuse l Motor 2 !

2MXZ-F03A Primary Bkr Reactor Coolant Drain Tank {

Backup Fuse Pump Motor 2B '

2MXZ-F04B Primary Bkr Lighting Transformer 2LR1

, Backup Fuse

, 2MXZ-F04C i i Primary Bkr Lighting Transformer 2LR2 Backup Fuse 2MXZ-F04D Primary Bkr Lighting Transformer 2LR3 Backup Fuse f 2MXZ-F05A

I Primary Bkr Reactor Coolant Pump Jib Backup Fuse Hoist No. R023 TH R026 j

2MXZ-F05C l l

1 Primary Bkr Lower Containment Auxiliary Backup Fuse Charcoal Filter Unit Fan Motor 2B 2MXZ-F06A Primary Bkr !

incore Drive Assembly Motor 2B Backup Fuse 2MXZ-F06B Primary Bkr incore Drive Assembly Motor 2D Backup Fuse

2MXZ-F06C i

Primary Bkr incore Drive Assembly Motor 2F Backup Fuse 2MXZ-F06D Lower Containment Reactor Building Primary Bkr Welding Receptacle 2RCPLO194 Backup Fuse i

2MXZ-F07B !

Primary Bkr Lighting Transformer 2LR4 Backup Fuse Chapter 16.8-1 Page 43 of 49 05/04/98

(

t TABLE 16.8-1B

\s UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE DEVICE NUMBER & LOCATION SYSTEM POWERED l

2. 600 VAC MCC (Continued) 2MXZ-F07C j

Primary Bkr 5 Ton Jib Crane in Containment i Backup Fuse Crane No. R006 !

2MXZ-F07D Primary Bkr Reactor Cavity Manipulator Backup Fuse Crane No. R008 & R028 2MXZ-F08A Primary Bkr Steam Generator Drain Pump Backup Fuse Motor 2 2MXZ-F08C Primary Bkr 15 Ton Equipment Access Hatch Backup Fuse Hoist Crane No. RO10 2MXZ-F08D Primary Bkr Control Rod Drive 2 Ton Jib Backup Fuse Hoist Crane No. R018 C~)' Primary Bkr Backup Fuse Reactor Side Fuel Handling Control Console SMXG-F06B Primary Bkr Standby Makeup Pump Drain isol Backup Fuse Viv 2NV876 SMXG-R05B Primary Bkr Pressurizer Heaters 28,55 & 56 Backup Fuse SMXG-F06C Primary Bkr Standby Makeup Pump to Seal Backup Fuse Water Line Isol Viv 2NV877

3. 600 VAC Pressurizer Heater Power Panels PHP2A-F01A Primary Bkr Pressurizer Heaters Backup Fuse 1,2, & 22 PHP2A-F01B Primary Bkr Pressurizer Heaters Backup Fuse 5,6, & 27 m

U Chapter 16.8-1 Page 44 of 49 05/04/98

p TABLE 16.8-1B UMfT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED l

3. 600 VAC Pressurizer Heater Power Panels (Continued)

PHP2A F01C Primary Bkr - Pressurizer Heaters Backup Fuse 9,10 & 32 PHP2A-F02C Primary Bkr Pressurizer Heaters Backup Fuse 11,12 & 35 PHP2A-F02D i Primary Bkr Pressurizer Heaters Backup Fuse 13,14 & 37 PHP2A-F02E Primary Bkr Pressurizer Heaters Backup Fuse 17,18 & 42 PHP2B-F01B Primary Bkr Pressurizer Heaters Backup Fuse 26, 53 & 54

]

PHP28-F01C !

q Primary Bkr Pressurizer Heaters j Backup Fuse 31,59 & 60 j PHP2B-F02C Primary Bkr Pressurizer Heaters Backup Fuse 36,65 & 66 PHP2B-F02D Primary Bkr Pressurizer Heaters Backup Fuse 41, 71 & 21 PHP2B-F02E Primary Bkr Pressurizer Heaters j Backup Fuse 46, 77 & 78 ,

l PHP2C-F01A Primary Bkr Pressurizer Heaters Backup Fuse 7,8 & 30 0

Chapter 16.8-1 Page 45 of 49 05/04/98 L - - - - - - - - _ -- - . _ . _ . . - - - _ _ _

TABLE 16.8-1B !

UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICEE

. DEVICE NUMBER & LOCATION SYSTEM POWERED 3, I 600 VAC Pressurizer Heater Power Panels (Continued)

PHP2C-F01B 1 Primary Bkr Pressurizer Heaters i Backup Fuse 19,20 & 45 PHP2C-F01C Primary Bkr Pressurizer Heaters Backup Fuse 24, 51 & 52 PHP2C-F01D Primary Bkr Pressurizer Heaters Backup Fuse {

29,57 & 58 PHP2C-F02C Primary Bkr Pressurizer Heaters Backup Fuse 34,63 & 64 PHP20-F02D i Primary Bkr Pressurizer Heaters Backup Fuse 39,69 & 70 i

p PHP20-F02E Primary Bkr Pressurizer Heaters Backup Fuse 44,75 & 76 PHP2D-F01 A Primary Bkr Pressurizer Heaters Backup Fuse 3,4 & 25 PHP20-F01B Primary Bkr Pressurizer Heaters Backup Fuse 15,16 & 40 PHP2D-F01C Primary Bkr Pressurizer Heaters Backup Fuse 23,49 & 50 PHP2D-F02C Primary Bkr Pressurizer Heaters Backup Fuse 33,61 & 62 PHP2D-F02D Primary Bkr Pressurizer Heaters Backup Fuse 38,67 & 68 O

Chapter 16.8-1 Page 46 of 49 05/04/98

. .. m . __ _. _ _ . __ _ _ .

I r] TABLE 16.81B UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED 4

3.

600 VAC Pressurizer Heater Power Panels (Continued)

PHP2D-F02E Primary Bkr Pressurizer Heaters Backup Fuse 43,73 & 74

4. 250 VDC Reactor Building Deadlight Panelboard I 2DLD-2 1

Primary Bkr Lighting Panelboard No. 2LR1, '

Backup Fuse 2LR2,2LR3,2LR4 2DLD-3 Primary Bkr Lighting Panelboard No. 2LR13,

Backup Fuse 2LR14 i 2DLD-4 Primary Bkr Lighting Panelboard No. 2LRS, !

Backup Fuse 2LR6 i l

2DLD-5 Primary Bkr Lighting Panelboard No. P.LR10, !

Backup Fuse 2LR11 '

2DLD-10 Primary Bkr Lighting Panelboard No. 2LR8 Backup Fuse i

, 5. 120 VAC Panelboards 2ELB1-5 Primary Bkr Emergency A.C. Lighting Backup Fuse ,

l 2ELB17 Primary Bkr Emergency A.C. Lighting '

Backup Fuse 2ELB1-13 '

Primary Bkr Emergency A.C. Lighting Backup Fuse 2ELB1-15 Primary Bkr Emergency A.C. Lighting i

Backup Fuse G

Chapter 16.8-1 Page 47 of 49 05/04/98

'q TABLE 16.8-1B UNIT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED

5. 120 VAC Panelboards (Continued) 2ELB1 17 Primary Bkr Emergency A.C. Lighting Backup Fuse 2KPM-1 i Primary Bkr NC Pump Motor 2A Space Heater !

Backup Fuse

)

2KPM-2 i Primary Bkr NC Pump Motor 20 Space Heater !

Backup Fuse 2KPM-7-1 Primary Bkr

{

Lower Containment Vent Unit '

Backup Fuse 2A Fan Motor Space Heater 2KPM-8-1 Primary Bkr Lower Containment Vent Unit Backup Fuse 2C Fan Motor Space Heater j C 2KPM-241 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 2A Space Heater 2KPM-24-2 ;

Primary Bkr Control Rod Drive Vent Fan !

Backup Fuse Motor 2B Space Heater j I

2KPM-24-3 '

Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 2C Space Heater j i

2KPM-24-4 Primary Bkr Control Rod Drive Vent Fan Backup Fuse Motor 2D Space Heater 2KPM 6, 7,8,9 Primary Bkr Safety injection System Backup Fuse Temperature Transmitters 2KPN-1 Primary Bkr NC Pump Motor 2B Space Heater Backup Fuse 4

Chapter 16.8-1 Page 48 of 49 05/04/98

i TABLE 16.8-1B

( UNrT 2 CONTAINMFNT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICER DEVICE NUMBER & LOCATION SYSTEM POWERED

5. 120 VAC Panelboards (Continued) 2KPN-2 Primary Bkr NC Pump Motor 2D Space Heater Backup Fuse 2KPN-7-1 PrimaryBkr Lower Containment Vent Unit Backup Fuse 2B Fan Motor Space Heater 2KPN-8-1 Primary Bkr Lower Containment Vent Unit Backup Fuse 2D Fan Motor Space Heater 2KPN-8-2,3,4,5 NC Pump Seal Stand Pipe Vent and Drain Valves 2KPN-11 Primary Bkr Misc Control Power Backup Fuse for 2ATC 24

6. DC Welding Circuits !

O h 2EOCB0001 Primary Bkr- AA Lower Containment Backup Bkr- AB DC Welding Circuit 2EOCB0002 Primary Bkr- AA Upper Containment '

Backup Bkr- AB DC Welding Circuit O

Chapter 16.8-1 Page 49 of 49 05/04/98

16.8 ELECTRICAL POWER SYSTEMS 1E8.-g 230 KV SWITCHYARD SYSTEMS COMMITMENH The following Switchyard equipment shall be in its normal alignment:

a. Switchyard Unit 1 PCBS 14,15,17, and 18 including their associated Manual Disconnects, Current Transformers, Interconnecting Bus, and Support Structures (EBA system).

b. Switchyard Unit 2 PCBS 20,21,23, and 24 including their associated Manual Disconnects, Current Transformers, Interconnecting Bus, and Support Structures (EBA system).

c. Buslines 1 A,18 (from Main Stepup Transformers to Switchyard Unit PCBs), including their associated Motor Operated Disconnects, Coupling Capacitor Voltage Transformers, Interconnecting Bus, and Support Structures (EBA system).

d.

O' Buslines 2A,2B (from Main Stepup Transformers to Switchyard Unit PCBs), including their associated Motor Operated Disconnects, Coupling Capacitor Voltage Transformers, Interconnecting Bus, and Support Structures (EBA system).

e. Controls associated with the equipment above (EBE, ERE systems).

f. Protective Relaying associated with the equipment above (EBD, ERD systems).

g. 480 VAC Aux Power Load Centers STA for both Units Train A, STB for both Units Train B (EBI system).

h. 125 VDC Aux Power (EBH system). See SLC 16.8-3.

APPLICABILITY:

At all times in accordance with Technical Specifications (all Modes) and Site Directive 3.1.30 (Modes 4,5, & 6).

O Chapter 16.8-2 Page 1 of 2 01/16/99

i O REMEDIAL ACTIONS:

Retum Switchyard equipment to the normal commitment alignment in accordance with Risk Assessment Matrix priorities.

TESTING REQUIREMENTS:

l None l

l

REFERENCES:

l 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at i Nuclear Power Plants l

WPM 607, Maintenance Rule Assessment of Equipment out of Service CNC-1535.00-00-0008, Severe Accident Analysis Report, CNS PRA Risk )

Significant SSCs for the Maintenance Rule.

, CNS-010.01-EB-0001, Switchyard Design Basis Specification TECH SPEC sections 3.8.1 and 3.8.2, LCOs for AC Power Sources during l Operating and Shutdown modes SITE DIRECTIVE 3.1.30, Unit Shutdown Configuration Control BASES:

i Effective implementation of the Maintenance Rule,10 CFR 50.65, requires the i continuous assessment of systems determined to be Risk Significant in the )

protection against Core Damage or Radiation Release. It has been determined l through PRA numerical methods that Switchyard Systems are Risk Significant from the stanopoint of causing or being able to recover from Loss of Offsite Power Events. This SLC serves two purposes. It defines the Risk Significant portions of the Switchyard. It also provides a method of tracking the Switchyard Systems for the purposes of supporting 10 CFR 50.65 and WPM 607.

Chapter 16.8-2 Page 2 of 2 01/16/99 m

. i O 16.8 ELECTRICAL POWER SYSTEMS 4.

V 114-1 230 KV SWITCHYARD 125 VDC POWER SYSTEM COMMITMENT: '

With the Switchyard in service, providing a power exchange between the site ;

and the transmission grid, the 230 KV Switchyard 125 VDC Power System (EBH) !

shall be AVAILABLE, with a minimum of one battery (SYB-1 or SYB-2) and one ,

, charger (SYBC-1, SYBC-S, or SYBC-2) aligned to each distribution bus. This I will provide an adequate, uninterruptable power source for relaying, control, and l

associated equipment requirements for normal switchyard operations.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

Restore to the normal commitment alignment in accordance with the Risk Assessment Matrix priorities.

TESTING REQUIREMENTS:

Periodic Tests performed in accordance with the IEEE Std 450-1980, "lEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations".

REFERENCES:

1 IEEE Std 450-1980, "lEEE Recommended Practice for Maintenance, Testing, and Replacement of Large Lead Storage Batteries for Generating Stations and Substations".

WPM 607, Maintenance Rule Assessment of Equipment out of Service 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.

SAAG File: 160 Severe Accident Analysis Report, CNS Probability Risk Assessment (PRA) Risk Significant SSCs for the Maintenance Rule.

O Chapter 16.8-3 Page 1 of 2 01/15/97

i O BASES:

The effective implementation of the Maintenance Rule,10 CFR 50.65, requires the continuous assessment of systems determined to be Risk Significant in the protection against Core Damage or Radiation Release. It has been determined through PRA numerical methods that portions of the Switchyard Systems are Risk Significant from the standpoint of being able to recover from the Loss of Offsite Power Events. This SLC serves two purposes. It defines the Risk Significant portion of the Relaying and Power Control System of the Switchyard through acceptable EBH system configuration alignments. It also provides a method of tracking the Relaying and Power Control System for the purposes of supporting 10 CFR 50.65 and WPM 607.

a O

O Chapter 16.8-3 Page 2 of 2 01/15/97

( j,.6, .8 Electrical Power Systems 16.8-4 6900V Shared Transformers COMMITMENT:

The 6900V Shared Transformers (SATA and SATB) will be available to be energized by the appropriate power source at all times of maintenance in which the indicated equipment is tagged and isolated. During the activities listed below (or their equivalent affects), ensure the supply power source is available, as directed:

Power Source / SATA from Bkr SATB from Brk SATA from Brk SATB from Brk Maintenance l'IE4 ITB4 2TC4 2TB4 Activity Unit 1 Zone A Main Power X

Unit 1 Zone B Main Power X

Unit 1 A EDG X

Unit 1 B EM X yg Unit 2 Zone A X i ) Main Power Unit 2 Zone B X

Main Power Unit 2 A X EDG Unit 2 B EDG X 1 APPLtCABILITY:

At all times REMEDIAL ACTION: 1 Restore the appropriate power source to the shared transformers to available in accordance with the Risk Assessment Matrix priorities.

TESTING REQUIREMENTS:

None

/3 k)

Chapter 16.8-4 Page 1 of 2 05/29/97

REFERENCES:

Maintenance Rule,10CFR50.65.

WPM 607 BASIS:

Effective implementation of the Maintenance Rule,10CFR50.65, requires the continuous assessment of systems determined to be risk significant in the protection against Core Damage or Radiation Release. It has been determined through PRA numerical methods that the 6900V Shared Transformers (SATA and SATB) are risk ,

significant from the standpoint of being able to recover from the Loss of Offsite Power Events. This SLC provides a method tracking the availlability of the Shared Transformers for the purpose of supporting 10 CFR 50.65 and WPM 607.

l O !

l i

O Chapter 16.8-4 Page 2 of 2 05/29/97

19,& ELECTRICAL POWER SYSTEMS 16.8-5 DIESEL GENERATOR SUPPLEMENTAL TESTING REQUIREMENTS COMMITMENT:

4

a. The diesel generator shall be operated at less than or equal to 5750 kW. I

b. The Cathodic Protection System shall be OPERABLE. !

c. The diesel generator Testing Requirements specified below shall be met.

APPLICABILITY:

1 MODES 1, 2, 3, 4, 5, and 6.

REMEDIAL ACTION: i

a. . With a diesel generator operating at greater than 5750 kW, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months reduce the diesel generator output to less than or equal to 5750 kW.

b. With the Cathodic Protection System inoperable, restore the System to OPERABLE status within 10 days or prepare and submit a Special Report outlining the cause of the inoperability and the plans for restoring the Syst( .

to OPERABLE.

c. With the specified Testing Requirements, except Testing Requirement b, not met, declare the affected diesel generator inoperable and take the ACTIONS of Technical Specification 3.8.1 or 3.8.2.

TESTING REQUIREMENTS:

Each diesel generator shall be demonstrated OPERABLE:

a. After each operation of the diesel where the period of operation was greater than or equal to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months by checking for and removing accumulated water from the day tank; !

b. By verifying that the Cathodic Protection System is OPERABLE by verifying:

1) At least once per 60 days that cathodic protection rectifiers are OPERABLE and have been inspected in accordance with the

/

manufacturer's inspection procedures, and Chapter 16.8-5 Page 1 of 2 01/16/99

[ TESTING REQUIREMENTS (con't)

2) At least once per 12 months that adequate protection from corrosion is provided in accordance with manufacturer's inspection procedures. )

c. At least once per 18 months by:

1) Subjecting the diesel to an inspection, during shutdown, in accordance with procedures prepared in conjunction with its manufacturer's ,

recommendations for this class of standby service 1

2) Verifying that the fuel transfer valve transfers fuel from each fuel !

3 storage tank to the day tank of each diesel via the installed cross- j connection lines; and 1

3) Verifying that the following diesel generator lockout features prevent diesel generator starting only when required:

a) Tuming gear engaged, or b) Maintenance mode.

]

C

d. At least once per 10 years by performing tank wall thickness measurements.

The resulting data shall be evaluated and any abnormal degradation will be justified or corrected. Any abnormal degradation wi!! be documented in a report to the Commission.

d

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The supplemental Testing Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guide 1.108, " Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants", Revision 1, August 1977, Regulatory Guide 1.137, " Fuel-Oil Systems for Standby Diesel Generators", Revision 1, October 4

1979, and the NRC Staff Evaluation Report conceming the Reliability of Diesel Generators at Catawba, August 14,1984. If any other metallic structures (building, new or modified piping systems, conduits) are placed in the ground near the Fuel Oil Storage System or if the original system is modified, the adequacy and frequency of inspections for the Cathodic Protection System shall be reevaluated and adjusted in J

O accordance with the manufacturer's recommendations.

Chapter 16.8-5 Page 2 of 2 01/16/99 i

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iga AUXILIARY SYSTEMS - FIRE PROTECTION SYSTEMS 16.9-1 FIRE SUPPRESSION WATER SYSTEM COMMITMENT:

The Fire Suppression Water System shall be OPERABLE with:

a. At least two fire suppression pumps, each with a capacity of 2500 gpm, with I their discharge aligned to the fire suppression header, and j

b. An OPERABLE flow path capable of taking suction from Lake Wylie and transferring the water through distribution piping with OPERABLE sectionalizing control valves and isolation valves for each sprinkler, hose standpipe, or Spray System nser required to be OPERABLE per Commitments 16.9-2 and 16.9-4.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

a. With one of the above required pumps and/or one Water Supply / Distribution System inoperable, restore the inoperable equipment to OPERABLE status within 7 days or provide an attemate backup pump or supply,

b. With the Fire Suppression Water System otherwise inoperable establish a backup Fire Suppression Water System within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

TESTING REQUIREMENTS:

a. The Fire Suppression Water System shall be demonstrated OPERABLE:

1. At least once per 31 days on a STAGGERED TEST BASIS by starting each electric motor- driven pump and operating it for at least 15 minutes on recirculation flow, ii. By visual verification that each valve (manual, power-operated, or automatic) in the flow path, which is accessible during plant operations, is in the correct position. The frequency of the verification shall be determined by the performance based criteria stated in the Bases Section.

Chapter 16.9-1 Page 1 of 4 01/16/99

a d

4 TESTING REQUIREMENTS (con't) iii. At least once per 6 months by performance of a system flush of the

' outside distribution loop to v9rify no flow blockage by fully opening the hydraulically most remote hydrant, '

iv. At least once per 12 months by cycling each testable valve in the flow i path through at least one complete cycle of full travel,

v. At least once per 18 months by verifying that each valve (manual, power-operated, or automatic) in the flow path which is inaccessible during plant operations is in its correct position, vii. At least once per 18 months by performing a system functional test which includes simulated automatic actuation of the system throughout its operating sequence, and:

1) Verifying that each automatic valve in the flow path actuates to its correct position,

2) Verifying that each pump develops at least 2500 gpm at a net pressure of 144 psig by testing at three points on the pump G performance curve, V

3)- Cycling each valve in the flow path which is not testable during plant operation through at least one complete cycle of full travel, and

4) Verifying that each fire suppression pump starts within 10 psig of its intended starting pressure (A pump, primary switch-95 psig; B pump, primary switch -90 psig; and C pump, primary switch-85 psig).

viii. At least once per 3 years by performing a flow test of the system in accordance with Chapter 8, Section 16 of the Fire Protection Handbook,15th Edition, published by the National Fire Protection Association.

REFERENCES:

1) Catawba FSAR, Section 9.5.1

2) Catawba SER, Section 9.5.1

3) Catawba SER, Supplement 2, Section 9.5.1

4) Catawba SER, Supplement 3, Section 9.5.1 Chapter 16.9-1 Page 2 of 4 01/16/99

1 1

REFERENCES (cont'd)

5) Catawba Fire Protection Review, as revised

6) Catawba Fire Protection Commitment Index

7) Startup and Normal Operation of Fire Protection System - OP/1/A/6400/02A BASES:

The OPERABILITY of the Fire Suppression Systems 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 system, spray, and/or sprinklers, CO2 ,

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.

The proper positioning of RF/RY valves is critical to delivering fire suppression water at the fire source as quickly as possible. The option of increasing or decreasing the frequency of valve position verification allows the ability to optimize plant operational resources. Should an adverse trend develop with RF/RY valve positions, the T frequency of verification shall be increased. Similarly if the RF/RY valve position trends are positive, the frequency of verification could be decreased. Through programmed trending of RF/RY as found valve positions, the RF/RY System will be maintained at predetermined reliability standards. The RF/RY System Engineer is responsible for trending and determining verification frequencies based on the following: :

Initially the frequency will be monthly.

Annually review the results of the completed valve position verification procedures.

If the results demonstrate that the valves are found in the correct position at least 99% of the time, the frequency of conducting the valve position verification may be decreased from - monthly to quarterly or - quarterly to semiannually or -

semiannually to annually - as applicable. The frequency shall not be extended beyond annually (plus grace period).

e if the results demonstrate that the valves are not found in the correct position at least 99% of the time, the frequency of conducting the valve position verification shall be increased from - annually to semiannually or - semiannually to quarterly or - quarterly to monthly - as applicable. The valve position verification need not be conducted more often that monthly.

Chapter 16.9-1 Page 3 of 4 01/16/99

4 5

In the event that portions of the Fire Suppression Systems are inoperable, altemate backup fire-fighting equipment is required to be made available in the affected areas

until the inoperable equipment is restored to service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a

longer period of time is allowed to provide an attemate means of fire fighting than if

the inoperable equipment is the primary means of fire suppression.

! In the event the Fire Suppression Water System becomes inoperable, immediate corrective measures must be taken since this system provides the major fire i j suppression capability of the plant.

This Selected l'icensee Commitment is part of the Catawba Fire Protection Program j and therefore subject to the provisions of the Catawba Facility Operating License

Conditions #6 for NPF-52 and #8 for NPF-35.

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Chapter 16.9-1 ' Page 4 of 4 01/16/99

}

l Iga AUXILIARY SYSTEJAS - FIRE PROTECTION SYSTEMS 16.9-2 SPRINKLER SYSTEMS l )

l COMMITMENT: !

Sprinkler systems in Table 16.9-1 shall be OPERABLE:

l APPLICABILITY:

Whenever equipment protected by the Sprinkler System is required to be l l OPERABLE.

)

REMEDIAL ACTION:

a. With one or more of the above required Sprinkler Systems inoperable, within l l 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , in accordance with the " Fire Watch Code" given in Table 16.9-1, :

established a continuous fire watch or an hourly fire watch. !

i

b. Verify backup fire suppression (fire extinguisher, nearby fire hose station) is available, and if not, establish backup fire supprossion equipment for the affected area. This must be accomplished within the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months given above.

TESTING REQUIREMENTS:

a. Each of the above required Sprinkler Systems shall be demonstrated l OPERABLE:

1. By verifying that each valve (manual or power-operated) in the flow l path, which is accessible during plant operations, is in the correct position. The frequency of the verification shall be determined by the performance based criteria stated in the Bases Section.

ii. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle of full travel, iii. At least once per 18 months by verifying that each valve (manual or l power-operated) in the flow path which is inaccessible during plant operations is in its correct position and O

Chapte- 16.9-2 Page 1 of 5 01/16/99

l TESTING REQUIREMENTS: (cont'd) ;

iv. At least once per 18 months:

1) By performing a system functional test which includes an 4

inspector's test connection flow test and cycling each valve in the flow path that is not testable during plant operation through at least one complete cycle of full travel.

2) By a visual inspection of each Sprinkler System starting at the system isolation valve to verify the system's integrity; and

3) By a visual inspection of each nozzle's spray area to verify the l spray pattem is not obstructed.

REi:ERENCEk

1) Catawba FSAR, Section 9.5.1

2) Catawba SER, Section 9.5.1

3) Catawba SER, Supplement 2, Section 9.5.1

4) Catawba SER, Supplement 3. Section 9.5.1

5) Catawba Fire Projection Review, as revised Q 6) Catawba Fire Protection Commltment index 2

b 2 BASES:

i The OPERABILITY of the Fire Suppression Systems 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 !ocated. The Fire i Suppression System consists of the water system, sprinklers, CO2, and fire hose l l 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.

The proper positioning of RF/RY valves is critical to delivering fire suppression water at the fire source as quickly as possible. The option of increasing or decreasing the frequency of valve position verification allows the ability to optimize plant operational resources. Should an adverse trend develop with RF/RY ;

valve positions, the frequency of verification shall be increased. Similarly if the RF/RY valve position trends are positive, the frequency of verification could be decreased. Through programmed trending of RF/RY as found valve positions, the RF/RY System will be maintained at predetermined reliability standards. The RF/RY System Engineer is responsible for trending and determining verification frequencies based on the following:

O Chapter 16.9-2 Page 2 of 5 01/16/99 i i

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

i BASES (cont'd)

Initially the frequency will be monthly.

Annually review the results of the completed valve position verification procedures.

if the results demonstrate that the valves are found in the correct position at least 99% of the time, the frequency of conducting the valve position verification may be decreased from - monthly to quarterly or - quarterly to semiannually or - '

semiannually to annually - as applicable. The frequency shall not be extended beyond annually (plus grace period).

e if the results demonstrate that the valves are not found in the correct position at !

least 99% of the time, the frequency of conducting the valve position verification shall be increased from - annually to semiannually or - semiannually to quarterly or - quarterly to monthly - as applicable. The valve position verification need not be conducted more often that monthly.

In the event that portions of the Fire Suppression Systems are inoperable, attemate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a O longer period of time is allowed to provide an attemate means of fire fighting than if the inoperable equipment is the primary means of fire suppression.

This Selected Licensee Commitment is part of the Catawba Fire Protection Program and therefore subject to the provisions of Section 2.C. of the Catawba Facility Operating Licenses.

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Chapter 16.9-2 Page 3 of 5 01/16/99

TABLE 16.9-1 SPRINKLER SYSTEMS l

( Room No. Eauipment Fire Watch Code

a. Elevation 522+0 - Auxiliary Building 100,101,106 ND & NS Connecting (2) 111,112 Corridor l 104 ND Pump 1B (3) 105 ND Pump 1 A (3) 109 ND Pump 2B (3) 110 ND Pump 2A (3)

b. Elevation 543+0 - Auxiliary Building i 230 NV Pump 1A (3) 231 NV Pump 1B (3) 240 NV Pump 2A (3)

, 241 NV Pump 2B (3) 250 Unit 1 CA Pump Room (1) 260 Unit 2 CA Pump Room (1)

c. Elevation 554+0 - Auxiliary Building

, O, -

340

, U2 Battery Room Corridor (DD-EE) (2) 350 U1 Battery Room Corridor (DD-EE) (2)

d. Elevation 560+0 - Auxiliary Building 300 KC Pumps 1 A1,1 A2, (3) 300 KC Pumps 1B1,1B2 (3) ;

l

e. Elevation 574+0 - Auxiliary Building !

480 U2 Cable Room Corridor (DD-EE) (2) 490 U1 Cable Room Corridor (DD-EE) (2)

f. Elevation 577+0 - Auxiliary Building 400 KC Pumps 2A1,2A2, (3) 400 KC Pumps 281,2B2 (3)

g. Peactor Buildings Annulus (1)

O Chapter 16.9-2 Page 4 of 5 01/16/99 I

TABLE 16.9-1 l SPRINKLER SYSTEMS l ;

O Fire Watch Codes for Table 16.9-1 (Sprinkler Systems):

l (1) Continuous. ,

(2) Hourly unless Standby Shutdown System (SSS) is inoperable. If SSS is !

inoperable - continuous watch is required.

(3) Hourly unless opposite train component is inoperable, OR sprinkler system for opposite train component is inoperable, OR Standby Shutdown System (SSS) is inoperable. If opposite train component, or sprinkler for opposite train component, or SSS is inoperable - continuous watch is required. i i

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Chapter 16.9-2 Page 5 of 5 01/16/99

1ga AUXILIARY SYSTEMS - FIRE PROTECTION SYSTEMS 16&-3 CO2 SYSTEMS COMMITMENT:

The following High Pressure and Low Pressure CO2 Systems shall be OPERABLE:

a. Low Pressure CO2 Syster: - Diesel generator rooms, and

b. High Pressure CO2 System - Auxiliary feedwater pump rooms APPLICABILITY:

Whenever equipment protected by the Systems is required to be OPERABLE.

REMEDIAL ACTION:

a. With one or more of the Low Pressure CO2 System (Diesel Generator Rooms) inoperable, within one hour establish an houriy fire watch with backup fire suppression equipment as long as the fire barrier between the affected A and B train D/G rooms is intact. If the fire barrier between the affected A and B train D/G rooms is not intact or backup fire suppression equipment is not available, establish a continuous fire watch

b. With one or more of the High Pressure CO2 Systems (Aux. Feedwater Pump Rooms) inoperable, within one hour establish a continuous fire watch.

TESTING REQUIREMENTS:

a. Each of the above required CO2 Systems shall be demonstrated OPCRABLE by visual verification that each valve (manual, power operated, or automatic) in the flow path is in the correct position.

' The frequency of the verification shall be determined by the performance based criteria in the Bases Section.

b. Each of the above required Low Pressure CO2 Systems shall be demonstrated OPERABLE:

1. At least once per 7 days by verifying the CO2 storage tank level to be greater than 44% of full capacity, and O

CHAPTER 16.9-3 Page 1 of 3 01/16/99

iO 4

TESTING REQUIREMENTS (cont'd)

11. At least once per 18 months by verifying: .

)

i

{

] 1) Each system actuates manJally and automatically, upon receipt '

of a simulated actuation signal, ,

4

2) Normal and Emergency Ventilation System Fans receive an j "off" signal upon system operation, and 1

3) By a visual inspection of discharge nozzles to assure no I

) blockage. :

4

c. Each of the aoove required High Pressure CO2 Systems shall be '

, demonstrated OPERABLE:

i

i. At least once per 6 months by verifying the weight of each CO2 storage >

cylinder to be at least 90% of full charge weight, and '

ii. At least once per 18 months by: i

1) Verifying each system actuates manually and automatically upon receipt of a simulated actuation signal,

2) Verifying that damper closure devices receive an actuation i signal upon system operation, and j l

3) A visualinspection of the discharge nozzles to assure no ;

blockage. i REFERENCES

1) - Catawba FSAR, Section 9.5.1

2) Catawba SER, Section 9.5.1

3) Catawba SER, Supplement 3, Section 9.5.1 SASES:

The OPERABILITY of the Fire Suppression Systems 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 CHAPTER 16.9-3 Page 2 of 3 01/16/99

BASES: (cont'd)

Suppression System consists of the water system, spray, and/or sprinklers, CO2 ,

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.

The proper positioning of RF/RY valves is critical to delivering fire suppression CO2 at the fire source as quickly as possible. The option of increasing or decreasing the frequency of valve position verification allows the ability to optimize plant operational resources. Should an adverse trend develop with CO2 Systems valve positions, the frequency of verification shall be increased. Similarly if the CO2 Systems valve position trends are positive, the frequency of verification could be decreased. Through programmed trending of CO 2 Systems as found valve positions, the CO2 fire protection systems will be maintained at predetermined reliability standards. The RF/RY System Engineer is responsible for trending and determining verification frequencies based on the following:

Initially the frequency will be monthly.

Annually review the results of the completed valve position verification procedures.

(3 .

If the results demonstrates that the valves are found in the correct position at U least 99% of the time, the frequency of conducting the valve position verification may be decreased from - monthly to quarterly or - quarterly to semiannually or -

semiannually to annually - as applicable. The frequency shall not be extended beyond annually (plus grace period).

if the results demonstrates that the valves are not found in the correct position at least 99% of the time, the frequency of conducting the valve position verification shall be increased from - annually to semiannually or - semiannually to quarterly or - quarterly to monthly - as applicable. The valve position verification need not be conducted more often than monthly.

In the event that portions of the Fire Suppression Systems are inoperable, attemate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an attemate means of fire fighting than if the inoperable equipment is the primary means of fire suppression.

This Selected Licensee Commitment is part of the Catawba Fire Protection Program and therefore subject to the provisions of the Catawba Facility Operating License Conditions #6 for NPF-52 and #8 for NPF-35.

CHAPTER 16.9-3 Page 3 of 3 01/16/99 I

__. ~ - - - - - . ... . . -. -

,1.ja AUXILIARY SYSTEMS - FIRE PROTECTION SYSTEMS M FIRE HOSE STATIONS L

COMMITMENT: 1 The fire hose stations given in Table 16.9-2 shall be OPERABLE:

APPLICABILITY: .

Whenever equipment in the areas protected by the fire hose stations is required to

be OPERABLE.

REMEDIAL ACTION: !

With one or more of the fire hose stations given in Table 16.9-2 inoperable, provide gated wye (s) on the nearest OPERABLE hose station (s). One outlet of the wye shall be connected to the standard length of hose provided for the hose station (s).

The second outlet of the wye shall be connected to a length of hose sufficient to provide coverage to the unprotected area resulting from the inoperable hose station. 1 To prevent a hazard to station personnel, plant equipment, or the hose itself, the fire hose shall be stored in a roll at the outlet of the OPERABLE hose station. Signs !

shall be mounted above the gated wye (s) to identify the proper hose to use. The l

above REMEDIAL ACTION requirement shall be accomplished within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months if the i inoperable fire hose is the primary means of fire suppression; otherwise route the l additional hose within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

h TESTING REQUIREMENTS:

a. Each of the fire hose stations given in Table 16.9-2 shall be demonstrated OPERABLE:

1. By a visual inspection of the fire hose stations, accessible during plant operations, to assure all required equipment is at the station and the fire hose shows no physical damage. The frequency of the inspection I shall be determined by the performance based criteria stated in the Bases Section.

ii. At least once per 18 months, by:

1

1) Visualinspection of the stations not accessible during plant !

('

operations to assure all required equipment is at the station, and the fire hose shows no physical damage. !

CHAPTER 16.9-4 Page 1 of 5 01/16/99

TESTING REQUIREMENTS: (cont'd)

2) Inspecting all gaskets and replacing any degraded gaskets in the couplings. !

iii. At least once per 3 years, by:

1) Partially opening each hose station valve to verify valve OPERABILITY and no flow blockage, and

2) Conducting a hose hydrostatic test at a pressure of 200 psig or ;

at least 50 psig above maximum fire main operating pressure, '

whichever is greater, and-l t

3) Removing the hose from fire hose stations for inspection and reracking.

I REFERENCES; i

1) Catawba FSAR, Section 9.5.1

2) Catawba SER, Section 9.5.1 _ ;

O 3) 4)

Catawba SER, Supplement 3, Section 9.5.1 Catawba Fire Protection Review, as revised

5) Catawba Fire Protection Commitment index :

i R& Egg:

The OPERABILITY of the lv Suppression Systems ensures that adequate fire suppression capability is avikbla to confine and extinguish fires occurring in any portion of the faci!!!y e Sere saica -related equipment is located. The Fire i Suppression Systew visists oMhe water system, spray, and/or sprinklers, CO 2, and fire hose stattres The collective capability of the Fire Suppression Systems is adequate to minin a potential damage to safety-related equipment and is a major ;

element in the fMhy Fire Protection Program. l l

The location of the required er:uipment at the fire hose station and the physical condition of fire hose is cri'.ical to fire brigade operations. The option of increasing or decreasing the frequency of the firo hose inspections allows the ability to optimize plant operational resource,s. Should an adverse trend develop with fire hose station equipment or fire hose condition, the frequency of the inspection shall be increased.

Similariy if the fire hose station equipment or fire hose condition trends are positive,

- the frequency of verification could be decreased. Through programmed trending of O l CHAPTER 16.9-4 Page 2 of 5 01/16/99

[ BASES (cont'd) fire hose station inspections, fire hose stations will be maintained at predetermined reliability standards. The RF/RY System Engineer is responsible for trending and determining inspection frequencies based on the following:

Initially the frequency will be monthly.

Annually review the results of the completed fire hose station inspection procedure.

If the results demonstrate that the fire hose stations are found acceptable at least 99% of the time, the frequency of conducting the fire hose station inspection may be decreased from - monthly to quarterly or - quarterly to semiannually or - semiannually to annually - as applicable. The frequency shall not be extended beyond annually (plus grace period).

If the results demonstrates that the fire hose stations are not found acceptable at least 99% of the time, the frequency of conducting the fire hose station inspections shall be increased from - annually to semiannually or - semiannually to quarter 1y or- quarterly to monthly- as applicable. The verification need not be conducted more often than monthly.

In the event that portions of the Fire Suppression Systems are inoperable, attemate U backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an altemate means of fire fighting than if the inoperable equipment is the primary means of fire suppression.

This Selected Licensee Commitment is part of the Catawba Fire Protection Program and therefore subject to the provisions of Section 2.0. the Catawba Facility Operating Licenses.

O CHAPTER 16.9-4 Page 3 of 5 01/16/99 l

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TABLE 16.9-2 I FIRE HOSE STATIONS !

l LOCATION ELEVATION HOSE RACK #

1. AUXILIARY BUILDING I

i. 59. FF 522+0 1RF235 I i 55, FF 522+0 1RF248 ,

63-64, KK 543+0 1RF210

); 63, MM 543+0 l

1RF211 (

[ 60, MM 543+0 1RF212 !

! 58, PP 543+0 1RF218 ;

i 59, . GG-HH 543+0 1RF236 i j 60-61, FF-GG 543+0 1RF237 l

61-62, CC-DD 543+0 2RFA64 l l l 57, JJ 543+0 1RF242 54-55, . GG 543+0 1RF249

) 57, FF 543+0 1RF250 !

t 52-53, GG 543+0 1RF255 i j 52-53, CC-DD 543+0 1RFA64 l l i 50-51, JJ-KK 543+0 1RF262 3 53, MM 543+0 1RF268 50-51, NN i

62, MM-NN 63, JJ-KK 543+0 560+0 560+0 1RF271 1RF203 !

1RF213 (

58, PP 560+0 1RF219 '

! 56, NN 560+0 1RF220 t l 59, HH 560+0 1RF239 l 57, KK 560+0 1RF243 !

li 54-55, FF-GG 560+0 1RF251 l 51, KK 560+0 1RF263 52, MM-NN 560+0 1RF269 ;

58, BB 554+0 1RF484 t

65, BB-CC 560+0 1RF485

, 62, . AA-BB 560+0 1RF486 j

i. 56, BB 554+0 1RF487 52, AA-BB 560+0 1RF488 49, BB-CC - 560+0 1RF489 68-69, BB 560+0 1RF996 (

4 45-46, BB 560+0 1RF997 !

p 63, NN 577+0 1RF204 l l- 61, LL 577+0 1RF214 1

63, KK-LL 577+0 1RF215 ;

{~ 58, PP 577+0 1RF221

59, JJ 577+0 1RF230 i

CHAPTER 16.9-4 Page 4 of 5 01/16/99

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

4 TABLE 16.9-2 FIRE HOSE STATIONS I

, LOCATION ELEVATION HOSE RACK #

58 GG 577+0 1RF240 !

56, KK 577+0 1RF244 l 54, GG 577+0 1RF252 52-53, KK 577+0 1RF258

51, KK 577+0 1RF264 l 51-52, NN 577+0 1RF272 56, PP 577+0 1RF278 68-69, BB 577+0 1RF478 I 3 65, BB-CC 577+0 1RF479 59, DD 574+0 1RF480 60, AA 574+0 1RF481

~1 49, BB-CC 577+0 1RF490 l

. 45, BB 577+0 1RF491 l 55, DD 574+0 1RF492 54, AA 574+0 1RF493

'. 63, AA 577+0 1RF993 51, AA 577+0 1RF998 62, NN 594+0 1RF205 O' 57, MM 594+0 1RF222 63, JJ 594+0 1RF231

! 57, HH 594+0 1RF245

, 57, EE 594+0 1RF253 51, JJ 594+0 1RF259 53, NN 594+0 1RF275 l

64, BB 594+0 1RF984 50, BB 594+0 1RF985 51, KK 605+10 1RF265 63, JJ 605+10 1RF233 63-64, MM 631+6 1RF483 50-51, MM 631+6 1RF495

2. FUEL POOLS l 65, TT-UU 605+10 1RF208 48, TT-UU 605+10 1RF276 I 63-64, MM 605+10 1RF482 50-51, MM 605+10 1RF822 CHAPTER 16.9-4 Page 5 of 5 01/16/99

Igj AUXILIARY SYSTEMS - FIRE PROTECTION SYSTEMS 1gj-5, FIRE BARRIER PENETRATIONS COMMITMENTS:

All fire barriers (walls, floor / ceilings, cable tray enclosures and other fire barriers) and all sealing devices in fire barrier penetrations (fire doors, fire dampers, cable, pipe and ventilation duct penetration seals) separating:

. Safety from non-safety related areas or, e

Redundant analyzed Post Fire Safe Shutdown Equipment or, e

Control Complex (Control Room, Cable Rooms and Battery Rooms) from the remainder of the plant or, o Containment from non-containment areas,

SHALL BE OPERABLE.

Note: (1) Fire Barriers are identified on drawing series CN-1105 ...

(2) A list of committed fire doors is located in Table 16.9-4.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

With one or more of the above required fire barrier penetrations and/or sealing devises inoperable, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months either establish a continuous fire watch on at least one side of the affected penetration, or verify the OPERABILITY of fire detectors on at least one side of the inoperable penetration and establish an hourly fire watch patrol.

If a continuous fire watch is required for the fire door (s) at the upper or lower containment entrances (AX352D, AX353D, AX701 or AX715A), this fire watch may be established using a closed circuit camera and monitor. The camera shall be adjusted so that the fire door can be fully viewed on the monitor. The monitor shall be placed in a constantly attended location. If the ability to view the open fire door with the camera and monitor is lost, the continuous fire watch shall be moved from the location of the monitor and re-established at the opened fire door.

O Chapter 16.9-5 Page 1 of 6 01/16/99

i TESTING REQUIRf MM:

a. At least once per 18 months the above required fire barrier penetrations and ,

sealing devices shall be verified OPERABLE by perfomiing a visual inspection of:

1. The exposed surfaces of each Fire Barrier; ;

ii. At least 10% of all fire dampers. Any abnormal changes or abnormal ,

degradation noted during the inspection shall be identified, investigated, and resolved through the Problem investigation Process.

Based on the results of the investigation process, additional dampers may be selected for inspection. Samples shall be selected such that ;

~

each fire damper will be inspectdd every 15 years; and iii. The committed fire barrier penetrations as identified in the predefined fire barrier penetration inspection schedule provided in the work control program. Any abnormal changes or abnormal degradation noted during the inspection shall be investigated through the Problem ;

investigation Process. Based on the results of the investigation i process, additional fire barrier penetrations may be selected for inspection. Samples shall be selected such that each fire barrier i penetration will be inspected every 15 years.

b. Each of the above required fire doors shall be verified OPERABLE by inspecting the closing mechanism and latches at least once per 6 months, and by verifying:

1. The position of each interior closed fire door at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and ii. That each locked closed fire door is closed at least once per 7 days.

REFERENCES:

1) Catawba FSAR, Section 9.5.1

2) Catawba SER, Section 9.5.1

3) Catawba SER, Supplement 3, Section 9.5.1

4) Catawba Fire Protection Review, as revised

5) Catawba Fire Protection Commitment index O !

l Chapter 16.9-5 Page 2 of 6 01/16/99

BASES:

The functional integrity of the fire barrier and associated penetration seals ensures that fires will be confined or adequately retarded so that the following criteria is achieved:

1 Fire will not spread from non safety related areas to safety related areas, e

Fire will not damage redundant analyzed post Fire Safe Shutdown equipment, Fire will not spread from the balance of plant to the control complex, e Fire will not spread from non-containment areas to containment areas.

1 The fire barriers and associated penetration seals are passive elements in the facility fire protection program and are subject to periodic inspections.

I 1

Fire barriers penetration seals, including cable / pipe penetration seals, fire doors, and fire dampers, are considered functional when the visually observed condition i indicates no abnormal change in appearance or abnormal degradation. An evaluation is performed to determine the cause of any identified fire barrier penetration seal abnormal change in appearance or abnormal degradation and the :

affect of this change on the ability of the fire barrier penetration seal to perform its !

function. Based on this evaluation additionalinspections may be performed. '

During periods of time when a barrier is not functional, either:

f (1) a continuous fire watch is required to be maintained in the vicinity of i the affected barrier, or (2) the fire detectors on at least one side of the affected barrier must be !

verified OPERABLE and an hourly fire watch patrol established, until l

the barrier is restored to functional status.

J!

This Selected Licensee Commitment is part of the Catawba Fire Protection Program ;

and therefore subject to the provisions of Section 2.C. of the Catawba Facility ,

Operating Lcenses. ,

h h

I O i Chapter 16.9-5 Page 3 of 6 01/16/99 i 1

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

TABLE 16.9-4 COMMITTED FIRE DOORS Elevation 543+0 t

.AX217D 52-53, BB AX217F 51, AA-BB ,

AX217G 52-53, BB AX228A 56-57 EE AX228B 57-58, EE AX248 57-58, QQ AX260B 61-62, BB-CC AX260F 62, AA-BB t

AX260G 61-62, BB-CC :

AX260H 61-62, BB-CC T527#1 52-43, BB-CC r

Elevation 554+0 AX354A 55, DD-EE l AX3548 59, DD-EE AX418 57, BB AX419 57, DD-EE !

AX420A 59, DD-EE ,

AX421A 55, DD-EE S102A 53-54, AA Elevation 560+0 AX352B- 53, CC-DD AX352C 53, CC-DD AX352D 46-47, BB-CC AX353 45-46, BB AX353B 45, AA-BB AX353C 45, AA-BB .

AX3938 61, CC-DD '

AX393C 61, CC-DD '

AX393D 67-68, BB-CC AX394 69, BB AX394B 69, AA-BB '

. AX3940 69, AA BB AX395 61, AA-BB i AX396 53, AA-BB AX415 45-46, CC-DD AX416 68-69, CC-DD AX417 57, QQ l i

1

)

Chapter 16.9-5 Page 4 of 6 01/16/99 l

/ TABLE 16.9-4 COMurrTED FIRE DOORS Elevation 574+0 AX515 54, BB AX516 56-57, . DD AX516A 57-58, DD AX516K 57, AA-BB AX517A 53-54, DD-EE AX517B 60-61, DD-EE AX5170 57, DD-EE AX517D 57, DD-EE AX517E 56-57, DD-EE AX518 60, BB S303A- 54, AA S304A 60, AA Elevation 577+0 AX513B 53, CC-DD AX514 45-46, BB AX5148 45-46, AA-BB AX517 57, EE

^*

AX525B

~'

56, 00 CD AX526D 58, OO A314#3 61, CC-DD AX533C 61. CC-DD AX534 69, BB AX534B 68-69, AA-BB AX535 61, AA-BB AX536 53, AA-BB AX656 53, CC-DD O

Chapter 16.9-5 Page 5 of 6 01/16/99

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

1 1

l l

)

TABLE 16.9 4 COMMITTED FIRE DOORS Elevation 594+0 t AX602 52, UU-VV AX627 62, UU-W !

AX630 58, OO 4 AX632' 57, OO l AX635 60-61, OO ;

AX635E 53-54, OO '

AX635F 53-54, OO i'

AX655 62-63, DD AX656C 61, CC-DD AX657 60-61, CC AX657B 52-53, CC-DD AX657F 60, DD-EE '

AX657G 57-58, DD-EE AX657H 54, DD-EE ;

AX657J 53, BB-CC i AX658B 51-52, DD S400 55-56, AA S406 58-59, AA Elevation 605+0 AX700B 50-51, JJ-KK ,

AX701 50-51, JJ-KK AX714B 63-64, JJ-KK AX720 50-51, HH-JJ AX721 63-64, HH-JJ AX715A 63-64, JJ-KK '

Nuclear Service Water Pump Structure i AX662A ,

)

i 1

l l

O Chapter 16.9-5 Page 6 of 6 01/16/99

.13& AUXILIARY SYSTEMS - FIRE PROTECTION SYSTEMS

! .1ga g FIRE DETECTION INSTRUMENTATION i-

! COMMITMENT:

i

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

l

i

. APPUCABILITY:

F )

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

i REMEDIAL ACTION: 1

a. With any, but not more than one-half the total in any fire zone, Function A fire l

. detection instruments shown in Table 16.9-3 inoperable, restore the l 1

inoperable instrument (s) to OPERABLE status within 14 days or within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months {

establish a fire watch patrol to inspect the zone (s) with the inoperable :

instrument (s) at least once per hour, unless the instrument (s) is located

'O inside the containment, then inspect that containment zone at least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or monitor the containment air temperature at least once per hour at the locations listed in Technical Specification Surveillance 3.6.5.1 and i

[

i 3.6.5.2. '

1 i b. With more than one-half of the Function A fire detection instruments in any !

^

fire zone shown in Table 16.9-3 inoperable, or with any Function B fire ]

detection instruments shown in Table 16.9-3 inoperable, or with any two or l more adjacent fire detection instruments shown in Table 16.9-3 inoperable, I l within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months establish a fire watch patrol to inspect the zone (s) with the !

inoperable instrument (s) at least once per hour, unless the instrument (s) is J

i- located inside the containment, then inspect that containment zone at least i l '- once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or monitor the containment air temperature at least once j per hour at the locations listed in Technical Specification Surveillance 3.6.5.1 !

and 3.6.5.2. i

)

TESTING REQUIREMENTS

a. . Each of the above required flame detection instruments shall be l demonstrated operable at least once per 6 months by the perfomiance of a !

VISUAL INSPECTION and at lest once per year by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST.

O 1 Chapter 16.9-6 Page 1 of 12 01/16/99

)

l l

/O V TESTING REQUIREMENTS (CON'T)

Each of the above required smoke detection instruments which are

accessible during plant operation shall be demonstrated OPERABLE at least once per 6 months by the performance of a VISUAL INSPECTION and at least once per year by performance of a TRIP ACTUATING DEVICE l OPERATIONAL TEST. Detectors which are not accessible during plant operation shall be demonstrated operable by the performance of a TRIP i ACTUATING DEVICE OPERATIONAL TEST. during each refueling outage.

!4ll spot type heat detectors which are accessible during plant operation shall be VISUALLY INSPECTED at least once per 6 months.

4

- l

i. For non-restorable spot-type detectors, at least two detectors ;

out of every hundred, or fraction thereof, shall be removed every 5 years and functionally tested. For each failure that occurs on the detectors removed, two additional detectors shall be removed and tested; and, ii. Restorable spot-type heat detectors which are accessible during plant operation, at least one detector on each signal initiating circuit shall be demonstratad OPERABLE at least once per 12 months by performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST. Diiferent detectors shall be selected for each test. Fire detectois which are not accessible during plant

operation shall be demonstrated OPERABLE by the performance of a TRIP ACTUATING DEVICE OPERATIONAL TEST during each refueling outage.

b. The NFPA Standard 72D supervised circuits supervision associated with the detector alarms of each of the above required fire detection instruments shall

, be demonstrated OPERABLE at least once per 6 months.

REFERENCES

1) Catawba FSAR, Section 9.5.1

2) Catawba SER, Section 9.5.1

3) Catawba SER, Supplement 2, Section 9.5.1

4) Catawba SER, Supplement 3, Section 9.5.1

5) Catawba Fire Protection Review, as Revised

6) Catawba Fire Protection Commitment Index BASES:

OPERABILITY of the detection instrumentation ensures that both adequate waming

, capability is available for prompt detection of fires and that Fire Suppression a

Chapter 16.9-6 Page 2 of 12 01/16/99

l BASES: (cont'd)

Systems, that are actuated by fire detectors, will discharge extinguishing agents in a ;

timely manner. Prompt detection and suppression of fires will reduce the potential l for damage to safety-related equipment and is an integral element in the overall facility Fire Protection Program.

Fire detectors that are used to actuate Fire Suppression Systems represent a more

)

critically important component of a plant's Fire Protection Program than detectors I that are installed solely te early fire waming and notification. Consequently, the l minimum number of OPERABLE fire detectors must be greater, j l

The loss of detection capability for Fire Suppression Systems, actuated by fire l

detectors, represents a significant degradation of fire protection for any area. As a result, the establishment of a fire watch patrol must be initiated at an earlier stage than would be warranted for the loss of detectors that provide only early fire waming. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to .

OPERABILITY. I EFA Zone 65 consists of 18 smoke detectors. Table 16.9-3 lists the Minimum Instruments Operable for EFA Zone 65 as 15. The basis for this difference is as l (D follows:

V l

. Two smoke detectors are located in the Auxiliary Service Building. These smoke detectors are not included in the Table 16.9-3 EFA Zone 65 Minimum Instruments Operable because they are located in a nonsafety-related area. ,

l

. The smoke detector located in the Spent Resin Batching Tank Room is not l included in the Table 16.9-3 EFA Zone 65 Minimum Instruments Operable because the Spent Resin Batching Tank Room is a High Rad Area with minimal combustible loading. Entering this room twice a year to perform smoke detector testing is not justified. The Spent Resin Batching Tank Room is enclosed by four foot thick reinforced concrete walls. The only access to this room is through a reinforced concrete hatch in the 594+0 Elevation floor slab. In the unlikely event that a significant fire should occur in the Spent Resin Batching Tank Room, the smoke detectors adjacent to (Elevation 577 - EFA Zone 65) or above (Elevation 594 - EFA Zone 84) the Spent Resin Batching Tank Room would provide an l alarm to the Control Room. The areas adjacent to the Spent Resin Batching I Room rely on the same assured train of Post Fire Safe Shutdown; therefore, a i fire in this room would have no adverse impact on Post Fire Safe Shutdown even in the unlikely event of fire extension into the adjacent areas. l EFA Zones 58 and 67: Each consist of 21 smoke detectors. Table 16.9-3 lists the p Minimum Instruments Operable for EFA Zones 58 and 67 as 19. The basis for this !

V difference is as follows:

I Chapter 16.9-6 Page 3 of 12 01/16/99 l

O V

e The smoke detectors located in the Unit 1 (Zone 58, rooms 476 and 477) and Unit 2 (Zone 67, rooms 467 and 468) Letdown Heat Exchanger and Moderating Heat Exchanger Rooms are not included in the Table 16.9-3 Minimum

!nstruments Operable because these rooms are High Rad Areas with minimal combustible loading. Entering these rooms twice a year to perform smoke detector testing is not justified. These rooms primarily consist of piping and a heat exchanger which are not significantly prone to fire damage. The Moderating Heat Exchanger Rooms are in a safety related area, but contain components which do not perform a safety related function nor a safe shutdown function. The Letdown Heat Exchanger Rooms contain safety related components, however, they do not perform a safe shutdown function. These rooms are enclosed by two to three foot thick concrete walls with a common labyrinth entrance. The outer, common entrance has a wire mesh door. In the unlikely event that a significant fire should occur in one of these heat exchanger rooms, the smoke detectors adjacent to these rooms would provide an alarm to the Control Room. The areas adjacent to these rooms rely on the same assured train of Post Fire Safe Shutdown; therefore, a fire in either of these rooms would have no adverse impact on Post Fire Safe Shutdown even in the unlikely event of fire extension into the adjacent areas.

EFA Zones 55 and 64: Each consist of 9 smoke detectors. Table 16.9-3 lists the Minimum Instrurnents Operable for EFA Zones 55 and 64 as 8. The basis for this

.O difference is as follows:

V y, y . The smoke detectors located in the Unit 1 (Zone 55, room 417) and Unit 2 (Zone 64, room 408) Fuel Pool Cooling (KF) Demineralizer Rooms are not included in the Table 16.9-3 Minimum Instruments Operable because the KF Demineralizer Rooms are High Rad Areas with minimal combustible loading. Entering these rooms twice a year to perform smoke detector testing is not justified. These rooms primarily consist of piping and a demineralizer which are not significantly prone to fire damage. These rooms are in a safety related area, but contain components which do not perform a safety related function nor a safe shutdown function. These rooms are enclosed by two and three foot thick concrete walls with a labyrinth entrance. The entrance has two doors. The outer door has a wire mesh above it and the inner door has an open grill. In the unlikely event that a significant fire should occur in one of the KF Demineralizer Rooms, the smoke detectors adjacent to these rooms would provide an alarm to the Control Room. The areas adjacent to the KF Demineralizer Rooms rely on the same assured train of Post Fire Safe Shutdown; therefore, a fire in either of these rooms would have no adverse impact on Post Fire Safe Shutdown even in the unlikely event of fire extension into the adjacent areas.

This Selected Licensee Commitment is part of the Catawba Fire Protection Program and therefore subject to the provisions of Section 2.C. of the Catawba Facility n Operating Licenses.

U Chapter 16.9-6 Page 4 of 12 01/16/99

O O O t

TABLE 16.9-3 FIRE DETECTION INSTRUMENTS FIRE DESCRIPTION LOCATION MINIMUM INSTRUMENTS OPERABLE

ZONE SMOKE FLAME HEAT FUNCTION **

1- ND Pump 1B GG-53 El.522 + 0 1 0 0 A 2 ND Pump 1 A FF-53 El.522 + 0 1 0 0 A 3 NS Pump 1B GG-54 El.522 + 0 3 0 0 A j 4 NS Pump 1A GG-55 El.522 + 0 2 0 0 A 5 ND Pump 2B GG-61 El.522 + 0 1 0 0 A !

6 ND Pump 2A FF-61 El.522 + 0 1 0 0 A 7 NS Pump 2B GG-60 El.522 + 0 3 0 0 A '

i 8 NS Pump 2A GG-59 El.522 + 0 2 0 0 A -

9 CA Pumps (Unit 1) BB-51 El.543 + 0 14 0 O(6) A(B) j 10 Mech Pen Room JJ-52 El.543 + 0 3 0 0 A !

11 Corridor / Caws NN-51 EI.543 + 0 6 0 0 A 12 NV Pump #1 (Unit 1) JJ-53 El.543 + 0 1 0 0 A 13 Ni Pump 1B HH-53 El.543 + 0 1 0 0 A 14 NI Pump 1A GG-53 El.543 + 0 1 0 0 A .

15 NV Pump 1B JJ-54 El.543 + 0 2 0 0 A 16 NV Pump 1 A JJ-55 El.543 + 0 2 0 0 A 17 Aisles / Cables KK-56 El.543 + 0 18 0 0 A 18 Aisier/ Cables EE-55 El.543 + 0 6 0 0 A ;

19 CA Pemps (Unit 2) BB-63 El.543 + 0 14 0 O(6) A(B) 20 Mech Pen Room JJ-62 El.543 + 0 3 0 0 A 21 Aisles / Cables NN-61 El.543 + 0 6 0 0 A '

22 NV Fump #2 (Unit 2) JJ-60 El.543 + 0 1 0 0 A 23 NI Pump 2B HH-60 El.543 + 0 1 0 0 A 24 NI Pump 2A GG-60 El.543 + 0 1 0 0 A 25 NV Pump 28 JJ-59 El.543 + 0 2 0 0 A

, i Chapter 16.9-6 Page 5 of 12 01/16/99 !

i

. _ _ _ _ _ _ _ _ _ _ - _ - _ _ - _ _ _ _ _ - - _ - _ _ _ _ _ _ - _ _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ . - - - _ - _ . - - . _ _ . .- - __ . ~_- -.. .-- - ?

~*

TABLE 16.9-3 FIRE DETECTION INSTRUMENTS FIRE DESCRIPTION LOCATION MINIMUM INSTRUMENTS OPERABLE

ZONE SMOKE FLAME HEAT FUNCTION **

26 NV Pump 2A JJ-58 El.543 + 0 2 0 0 A 27 Aisles / Cables KK-59 El.543 + 0 20 0 0 A 28 Aisles / Cables EE-58 El.543 + 0 6 0 0 A 29 Swgr Equip Room AA-50 El.560 + 0 7 0 0 A 30 Elect Pen Room CC-50 El.560+ 0 8 0 0 A 31 Corridor / Cables EE-53 El.560 + 0 5 0 0 A 32 Corridor / Cables KK-52 El.560 + 0 8 0 0 A

< 33 Corridor / Cables NN-54 El.560 + 0 to 0 0 A 34 Aisles / Cables JJ-56 El.560 + 0 14 0 0 A 35 Motor Control Centers GG-56 El.560 + 0 2 0 0 A 36 Cable Tray Access FF-56 El.568 + 0 2 0 0 A 37 Equip Batteries DD-55 El.554 + 0 5 0 0 A 38 Equip Batteries CC-55 El.544 + 0 5 0 0 A 39 Battery Room CC-56 El.554 + 0 17 0 0 A 41 Swgr Equip Room AA-64 El.560 + 0 7 0 0 A 42 Elect Pen Room CC-65 El.560 + 0 8 0 0 A 43 Corridor / Cables FF-61 El.560 + 0 5 0 0 A 44 Aisles / Cables KK-63 El.560 + 0 8 0 0 A 45 Aisles / Cables NN-60 El.560 + 0 13 0 0 A 46 Aisles / Cables HH-59 El.560 + 0 13 0 0 A 47 Motor Control Center GG-58 El.560 + 0 2 0 0 A 48 Cable Tray Access FF-58 El.560 + 0 2 0 0 A 49 Equip Batteries DD-60 El.560 + 0 5 0 0 A 50 Equip Batteries CC-60 El.560 + 0 5 0 0 A 51 Battery Room CC-E9 El.560 + 0 17 0 0 A Chapter 16.9-6 Page 6 of 12 01/16/99

f D h (V (b TABLE 16.9-3 FIRE DETECTION INSTRUMENTS FIRE DESCRIPTION LOCATION MINIMUM INSTRUMENTS OPERABLE

ZONE SMOKE. FLAME HEAT FUNCTION **

53 Swgr Equip Room AA-49 El.577 4 0 7 0 0 A 54 Aisles / Cables CC-50 El.577 + 0 10 0 0 A 55 Aisles / Cables NN-52 El.577 + 0 8 0 0 A(See Bases) 56 Aisles / Cables PP-55 El.577 + 0 13 0 0 A 57 Aisles / Cables LL-55 El.577 + 0 11 0 0 A

$8 Aisles / Cables HH-55 El.577 + 0 19 0 0 A (See Bases) 59 Motor Control Center EE-54 El.577 + 0 2 0 0 A 60 Cable Room CC-56 El.574 + 0 18 0 0 A 62 Swgr Equip Room AA-64 El.577 + 0 7 0 0 A 63 Elect Pen Room CC-64 El.577 + 0 10 0 0 A 64 Aisles / Cables PP-62 El.577 + 0 8 0 0 A(See Bases) 65 Aisles / Cables PP-59 E!.577 + 0 15 0 0 A(See Bases) 66 Aisles / Cables LL-59 El.577 + 0 11 0 0 A 67 Aisles / Cables HH-59 El.577 + 0 19 0 0 A (See Bases) 68 Motor Control Center FF-60 El.577 + 0 2 0 0 A 69 Cable Room CC-59 El.577 + 0 18 0 0 A 71 Elect Pen Room CC-51 El.594 + 0 10 0 0 A 72 Control Room CC-56 El.594 + 0 23 0 6 A 73 Vent Equip Room FF-50 El.594 + 0 9 0 0 A 74 Aisles / Cables LL-56 El.594 + 0 25 0 0 A 76 Aisles / Cables PP-54 El.594 + 0 15 0 0 A 79 Elect Pen Room BB-63 El.594 + 0 11 0 0 A 80 Control Room BB-59 El.594 + 0 22 0 6 A 81 Vent Equip Room FF-58 El.594 + 0 12 0 0 A 82 Aisles / Cables KK-58 El.594 + 0 27 0 0 A Chapter 16.9-6 Page 7 of 12 01/16/99 i

t O O O l

TABLE 16.9-3 FIRE DETECTION INSTRUMENTS FIRE DESCRIPTION LOCATION MINIMUM INSTRUMENTS OPERABLE *

! ZONE SMOKE FLAME HEAT FUNCTION "

84 Aisles / Cables NN-58 EL594 + 0 17 0 0 A 89 Fuel Pool Area (Unit 1) PP-50 El.605 + 10 19 7 0 A 90 Fuel Pool Area (Unit 2) PP-64 El.605 + 10 19 7 0 A 129 Fuel Pool Purge Room (Unit 1) NN-50 El.631 + 6 6 0 0 A- ,

131 Reactor Bldg 0*-45' Bel. El.565 + 3 4 0 0 A 132 Reactor Bldg 45'-90* Bel. El.565 + 3 3 0 0 A 133 Reactor Bldg 90'-135' Bel. El.565 + 3 4 0 0 A 134 Reactor Bldg 135'-180* Bel. El.565 + 3 5 0 0 A 135 Reactor Bldg 180'-225 Bel. El.565 + 3 4 0 0 A 136 Reactor Bldg 270*-315' Bel. El.565 + 3 3 0 0 A 137 Reactor Bldg 315'-O' Bel. El.565 + 3 8 0 0 A 138 Reactor Bldg 0*-45' Bel. El.586 + 3 6 0 0 A 139 Reactor Bldg 45'-90 Bel. El.586 + 3 4 0 0 A 8

140 Reactor Bidg 90 -135 Bel. El.565 + 3 3 0 0 A 141 Reactor Bldg 135'-180' Bel. El.586 + 3 8 0 0 A 142 Reactor Bldg 180 -225' Bel. El.586 + 3 5 0 0 A 143 Reactor Bldg 315'-08 Bel. El.586 + 3 5 0 0 A 144 Reactor Bidg O'-4 5' Bel. El.593 + 21/2 14 0 0 A 145 Reactor Bldg 45'-90' Bel. El.593 + 21/2 16 0 0 A 146 Reactor Bidg 90*-135' Bel. El.593 + 21/2 11 0 0 A 147 Reactor Bldg 135'-180' Bel. El.593 + 21/2 10 0 0 A 148 Reactor Bidg 180 -225' Bel. El.593 + 21/2 2 0 0 A Chapter 16.9-6 Page 8 of 12 01/16/99

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

O O O TABLE 16.9-3 FIRE DETECTION INSTRUMENTS FIRE DESCRIPTION LOCATION MINIMUM INSTRUMENTS OPERABLE

ZONE SMOKE FLAME HEAT FUNCTION "

149 Reactor Bidg 315"-0" Bel. El.593 + 21/2 7 0 0 A 150 Reactor Bldg (Unit 2) 0"-45 Bel. El.565 + 3 4 0 0 A 151 Reactor Bldg (Unit 2) 45'-90' Bel. El.565 + 3 3 0 0 A '

152 Reactor Bldg (Unit 2) 90'-135' Bel. El.5E3 + 3 4 0 0 A 153 Reactor Bldg (Unit 2) 135'-180' Bel. El.565 + 3 5 0 0 A i 154 Reactor Bidg (Unit 2) 180*-225* Bef. El.565 + 3 3 0 0 A 155 Reactor Bldg (Unit 2) 270'-315' Bel. El.565 + 3 4 0 0 A 156 Reactor Bldg (Unit 2) '

315'-O' Bel. El.565 + 3 6 0 0 A 157 Reactor Bldg (Unit 2) 0*-45" Bel. El.586 + 6 6 0 0 A 158 Reactor Bldg (Unit 2) 45'-90' Bel. El.586 + 6 4 0 0 A 159 Reactor Bldg (Unit 2) 90'-135 Bel. El.586 + 6 3 0 0 A 100 Reactor Bldg (Unit 2) 135'-180' Bel. El.586 + 6 8 0 0 A 161 Reactor Bldg (Unit 2) 180 -225' Bel. El.586 + 6 5 0 0 A i 162 Reactor Bldg (Unit 2) 315 -0 Bel. El.586 + 6 5 0 0 A 163 Reactor Bldg (Unit 2) 0'-45" Bel. El.593 + 21/2 13 0 0 A 164 Reactor Bldg (Unit 2) 45'-90' Bel. El.593 + 21/2 17 0 0 A 165 Reactor Bldg (Unit 2) 90'-135 Bel. El.593 + 21/2 13 0 0 A 166 Reactor Bldg (Unit 2) 135 -180 Bel. El.593 + 21/2 10 0 0 A 167 Reactor Bldg (Unit 2) 180 -225 Bel. El.593 + 21/2 2 0 0 A 168 Reactor Bldg (Unit 2) 315'-O' Bel. El.593 + 21/2 7 0 0 A 169 NC Pump 1 A Reactor Bldg El.593 + 21/2 0 0 1 A Chapter 16.9-6 Page 9 of 12 01/16/99

- n.

\

TABLE 16.9 3 FIRE DETECTION INSTRUMENTS FIRE DESCRIPTION LOCATION MINIMUM INSTRUMENTS OPERABLE

ZONE SMOKE FLAME HEAT FUNCTION "

170 NC Pump 1B Reactor Bldg El.593 + 21/2 0 0 1 A 171 NC Pump 1C Reactor Bldg El.593 + 21/2 0 0 1 A 172 NC Pump 1D Reactor Bldg El.593 + 21/2 0 0 1 A 173 NC Pump 2A 45' Bel. El.593 + 21/2 0 0 1 A 174 NC Pump 2B 135' Bel. El.593 + 21/2 0 0 1 A 175 NC Pump 2C 225' Bel. EiS93 + 21/2 0 0 1 A 176 NC Pump 2D 315' Bel. El.593 + 21/2 0 0 1 A 177 Filter Bed Unit 1B Reactor Bldg Bel. El.565 + 3 2 0 2 A 178 Filter Bed Unit 1 A Reactor Bldg Bel. El.565 + 3 2 0 2 A 179 Filter Bed Unit 2A Reactor Bldg El.565 + 3 2 0 2 A 180 Filter Bed Unit 2B Reactor Bldg El.565 + 3 2 0 2 A j 181a Annulus El.561 + 0 0 0 1 A 181b Annulus El.583 + 0 0 0 1 A 181c Annulus El.604 + 0 S 0 1 A 181d Annulus El.629 + 5 C 0 1 A 181e Annulus El.649 + 5 0 0 1 A 181f Annulus El.664 + 0 0 0 1 A 182a Annulus El.561 + 0 0 0 1 A 182b Annulus El.583 + 0 0 0 1 A 182c Annulus El.604 + 0 0 0 1 A 182d Annulus El.629 + 5 0 0 1 A 182e Annulus El.649 + 5 0 0 1 A 182f Annulus El.664 + 0 0 0 1 A 183 Fuel Pool Purge Room (Unit 2) NN-64 El.631 + 6 6 0 0 A Chapter 16.9-6 Page 10 of 12 01/16/99

1 k.

TABLE 16.9-3 FIRE DETECTION INSTRUMENTS FIRE DESCRIPTION LOCATION MINIMUM INSTRUMENTS OPERABLE

ZONE SMOKE FLAME HEAT FUNCTION **

~E12 Aisles / Cables GG-57 El.522 + 0 2 0 0 A 213 Aux Battery Room AA-55 El.544 + 0 4 0 0 A 214 Aux Control Power Batteries AA-59 El.560 + 0 4 0 0 A 215 D/G Corridor BB-45 El.556 + 0 3 0 0 A 216 D/G Corridor AA-45 El.556 + 0 2 0 0 A 217 D/G Corridor CC-71 El.560 + 0 3 0 0 A 218 D/G Corridor BB-71 Elli60 + 0 2 0 0 A 219 Mech Pen Room HH-52 El.577 + 0 6 0 0 A 220 Mech Pen Room JJ-62 El.577 + 0 6 0 0 A 222 Airlock Access (Unit 1) JJ-51 El.605 + 10 1 0 0 A 224 Airlock Access (Unit 2) JJ-63 El.605 + 10 1 0 0 A 225 RN Pump Structure West Section El.600 + 0 8 0 0 A 226 RN Pump Structure East Section EL600 + 0 8 0 0 A 231 Reactor Bldg (Unit 1) 260"-303' Bel. El.668 + 10 10 0 0 A 232 Reactor Bldg (Unit 2) 260'-303' Bel. El.668 + 10 10 0 0 A 184 HVAC duct for Rooms 331 & FF-53 El.543 + 0 1(Duct) 0 0 A 332 185 HVAC Duct for Rooms 203, MM-60 EL543 + 0 1(Duct) 0 0 A -

205,205A,206A,206B,207 &

209A 186 HVAC Duct for Rooms 301, NN-60 El.560 + 0 1 (Duct) 0 0 A 302,305, & 307 RFIA Diesel Generator 1 A EE-41 El.556 + 0 0 0 0(10) A(B)

RF1B Diesel Generator 1B AA-41 El.556 + 0 0 0 0(10) A(B)

RF2A Diesel Generator 2A EE-72 EI.556 + 0 0 0 0(10) A(B)

RF2B Diesel Generator 28 AA-72 El.ESS + 0 0 0 0(10) A(B)

Chapter 16.9-6 Page 11 of 12 01/16/99

O O O TABLE 16.9-3 FIRE DETECTION INSTRUMENTS The fire dectection instruments located within tiw containment are not required to be OPERABLE during the performance of Type A Containment Leakage Rate tests.

Function A: Early waming fire detection and notification only. Function B: Actuation of fire suppression system and early waming and notification.

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16.9 AUXsLIARY SYSTEMS 16.9-7 BORAT10N SYSTEMS FLOW PATHS - SHUTDOWN COMMITMENT:

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

a. A flow path from the boric acid tank via a boric acid transfer pump and a charging pump to the Reactor Coolant System if the boric acid storage tank in SLC 16.9.11a. is OPERABLE, or

b. The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if the refueling water storage tank in SLC 16.9.11b. is OPERABLE.

APPLICABILITY:

MODE 4 with any RCS cold leg temperature s 285 F, g MODES 5 and 6.

REMEDIAL ACTION:

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

l TESTING REQUIREMENTS:

At least one of the above required flow paths shall be demonstrated OPERABLE:

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

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

Chapter 16.9-7 Page 1 of 2 01/16/99 l

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REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The Boration System Flow Paths ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include a flow path and boric acid transfer pump.

In MODE 4 with any RCS cold leg temperature s 285 F, and in MODES 5 and 6, one Boron injection flow path is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional '

restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron injection flow path becomes inoperable. The boration capability of one path, in association with a charging pump and borated water source, is sufficient to provide a SHUTDOWN MARGIN of 1.3% Ak/k after xenon decay and cooldown to 200'F and of 1% Ak/k after xenon decay and cooldown from 200 F to 140 F.

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l 16.9 AUXILIARY SYSTEMS 16.9-8 BORATION SYSTEMS FLOW PATHS - OPERATING ,

COMMITMENT:

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

a. The flow path from the boric acid tanks via a boric acid transfer pump and a charging pump to the Reactor Coolant System, and j b. Two flow patns from the refueling water storage tank via ( harging pumps to the Reactor Coolant System.

APPLICABILITY:

MODES 1,2,3, MODE 4 with all RCS cold leg temperatures > 285'F.

REMEDIAL ACTION:

1 With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two boron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT l STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% Ak/k at !

200 F within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; restore at least two flow paths to OPERABLE status !

within the next 7 days or be in COLD SHUTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . l l

TESTING REQUIREMENTS:

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

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

b. At least once per 31 days by verifying that each valve (manual, power-operated, or automatic) in the flow path that is not locked, sealed, or !

otherwise secured in position, is in its correct position;

c. At least once per 18 months during shutdown by verifying that each 1 automatic valve in the flow path actuates to its correct position on a Safety q Injection test signal; and '

O I Chapter 16.9-8 Page 1 of 2 01/16/99

O G TESTING REQUIREMENTS (con't)

. d. At least once per 18 months by verifying that the flow path required by SLC 16.9-8.a delivers at least 30 gpm to the Reactor Coolant System.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

1 BASES:

The Boration System Flow Paths ensures that negative reactivity control is available l during each mode of facility operation. The components required to perform this function include separate flow paths and boric acid transfer pumps.

In MODES 1,2, and 3, and MODE 4 with RCS average temperature above 285'F, a minimum of two boron injection flow paths are required to ensure single functional capabili.y in the event an assumed failure renders one of the flow paths inoperable.

The boration capability of either flow path, in association with a charging pump and boratec'. water source, is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% Ak/k after xenon decay and cooldown to i 200'F.

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O Chapte: "6.9-8 Page 2 of 2 01/16/99

( 16.9 AUXILIARY SYSTEMS 16.9-9 BORATION SYSTEMS CHARGING PUMP - SHUTDOWN COMMITMENT:

One charging pump in the boron injection flow path required by SLC 16.9.7 shall be OPERABLE and capable of being powered from an OPERABLE emergency power source.

APPLICABILITY:

1 MODE 4 with any RCS cold leg temperature s 285 F, MODES 5 and 6.

I REMEDIAL ACTION: l l

With no charging pump OPERABLE or capable of being powered from an OPERABLE emergency power source, suspend all operations involving CORE ALTERATIONS or positive reactivity changes.

( TESTING REQUIREMENTS:

i The above required charging pump shall be demonstrated OPERABLE by verifying that a differential pressure across the pump of greater than or equal to 2380 psid is ,

developed when tested pursuant to Technical Specification 5.5.8. j REFERENCES-

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The Boration System Charging Pumps ensures that negative reactivity control is available during each mode of facility operation.

In MODE 4 with any RCS cold leg temperature s 285 F, and in MODES 5 and 6, one charging pump is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single charging pump becomes inoperable. The boration capability of one charging pump, in association with a flow path and borated water source, is sufficient to provide a

. Chapter 16.9-9 Page 1 of 2 01/16/99

BASES (con't)

SHUTDOWN MARGIN of 1.3% Ak/k after xenon decay and cooldown to 200 F and of 1% Ak/k after xenon decay and cooldown from 200 F to 140 F. -

When the temperature of one or more RCS cold legs drops below 285 F in Mode 4, the potential for low temperature overpressurization of the reactor vessel makes it necessary to render all but one charging pump or safety injection pump inoperable.

The Technical Specification 3.4.12 limitation for a maximum of one centrifugal charging or safety injection pump to be OPERABLE and the associated Surveillance Requirement to verify all charging pumps except the required OPERABLE pump to be inoperable below 285 F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV.

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() 16.9 AUXlLIARY SYSTEMS V

16.9-10 BORATION SYSTEMS CHARGING PUMPS - OPERATING COMMITMENT:

At least two charging pumps shall be OPERABLE.

APPLICABILITY:

MODES 1,2,3, MODE 4 with all RCS cold leg temperatures > 285 F.

REMEDIAL ACTIC J:

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

TESTING REQUIREMENTS: l At least two charging pumps chall be demonstrated OPERABLE by verifying that a l differential pressure across each pump of greater than or equal to 2380 psid is i developed when tested pursuant to Technical Specification 5.5.8. 1

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES: ,

The Boration System Charging Pumps ensures that negative reactivity control is available during each mode of facility operation.

In MODES 1,2, and 3, and MODE 4 with RCS average temperature above 285 F, two charging pumps are required to ensure single functional capability in the event an assumed failure rentrs one of the charging pumps inoperable. The boration !

capability of either charging pump, in association with a flow path and borated water source, is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% Ak/k after xenon decay and cooldown to 200 F.

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Chapter 16.9-10 Page 1 of 1 01/16/99 ;

4O 16.9 AUXILIARY SYSTEMS V

16.9-11 BORATION SYSTEMS BORATED WATER SOURCE - SHUTDOWN COMMITMENT:

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

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume as presented in the CORE OPERATING LIMITS REPORT,

2) A minimum boron concentration as presented in the CORE OPERATING LIMITS REPORT, and

3) A minimum solution temperature of 65 F.

b. The refueling water storage tank with:

1) A minimum contained borated water volume as presented in the CORE OPERATING LIMITS REPORT,

2) A minimum boron concentration as presented in the CORE OPERATING LIMITS REPORT, and

3) A minimum solution temperature of 70 F.

APPLICABILITY:

MODE 4 with any RCS cold leg temperature s 285 F, MODES 5 and 6.

REMEDIAL ACTION:

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

TESTING REQUIREMENTS:

The above required borated water source shall be demonstrated OPERABLE:

a. At least once per 7 days by.

Chapter 16.9-11 Page 1 of 3 01/16/99

(m; ) TEST NG REQUIREMENTS (con't)

1) Verifying the boron concentration of the water,

2) Verifying the contained borated water volume, and

3) Verifying the boric acid storage tank solution temperature when it is the source of borated water,

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the refueling water storage tank temperature when it is the source of borated water and the outside air temperature is less than 70 F.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The Boration System Borated Water Sources ensures that negative reactivity control is available during each mode of facility operation.

In MODE 4 with any RCS cold leg temperature s 285 F, and in MODES 5 and 6, one Borated Water Source is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Borated Water Source becomes inoperable. The boration capability of one borated water source, in association with a flow path and charging pump, is sufficient to provide a SHUTDOWN MARGIN (SDM) of 1% Ak/k after xenon decay and cooldown from 200 F to 140'F. To maintain SDM for this condition a minimum water volume at a minimum boron concentration, as presented in the CORE OPERATING LIMITS REPORT, is required from the boric acid storage tanks or the refueling water storage tank.

A minimum contained water volume and boron concentration, as presented in the CORE OPERATING LIMITS REPORT (COLR), is required to be available from the borated water sources in MODES 5 and 6. This volume is based on the required volume for maintaining SDM, unusable volume (to allow for a full suction pipe),

instrument error, and additional margin for conservatism as follows:

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Boric Acid Tank 1 Required Volume for Maintaining SDM presented in the COLR Unusable Volume, Vortexing, inst. Error 10,846 gallons i

Additional margin 569 gallons ,

Refuelina Water Storace Tank i

Required Volume for Maintaining SDM presented in the COLR Water Below the Nozzle-13,442 gallons Instrument inaccuracy 11,307 gallons Vortexing 13,247 gallons Additional Margin 3,504 gallons The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

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16.9 AUXILIARY SYSTEMS 16.9-12 BORATION SYSTEMS BORATED WATER SOURCES -

OPERATING 1 l

i COMMITMENT:

l As a minimum, the following borated water source (s) shall be OPERABLE as required by SLC 16.9.8:

a. A Boric Acid Storage System with:

1) A minimum contained borated water volume as presented in the CORE OPERATING LIMITS REPORT, l

2) A minimum boron concentration as presented in the CORE OPERATING LIMITS REPORT, and 1

3) A minimum solution temperature of 65'F. I I

b. The refueling water storage tank with:

g('~/ l

1) A minimum contained borated water volume as presented in the '

CORE OPERATING LIMITS REPORT or Technical Specification l Surveillance Requirement 3.5.4.2 whichever is larger,

2) A minimum boron concentration as presented in the CORE OPERATING LIMITS REPORT,

3) A minimum solution temperature of 70 F, and

4) A maximum solution temperature 100 F.

APPLICABILITY:

MODES 1,2, and 3, MODE 4 with all RCS cold leg temperatures > 285 F.

REMEDIAL ACTION:

a. With the Boric Acid Storage System inoperable and being used as one of the above required borated water sources, restore the system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in at least HOT STANDBY within the next 6

'O hours and borated to a SHUTDOWN MARGIN equivalant to at least 1% Ak/k l Chapter 16.9-12 Page 1 of 3 01/16/99

( REMEDIAL ACTION (con't) at 200 F; restore the Boric Acid Storage System to OPERABLE status within the next 7 days or be in COLD SHJTDOWN within the next 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ,

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

TESTING REQUIREMENTS:

Each borated water source shall be demonstrated OPERABLE:

1

a. At least once per 7 days by:

1) Verifying the boron concentration in the water,

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

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

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying the refueling water storage tank temperature when the outside air temperature is either less than 70 F or ,

greater than 100 F.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical 3 Specifications Amendments for Catawba, September 30,1998.

)

BASES:

The Boration System Borated Water Sources ensures that negative reactivity controlis available during each mode of facility operation.

In MODES 1,2, and 3, and MODE 4 with RCS average temperature above 285 F.,

a minimum of two borated water sources are required to ensure single functional capability in the event an assumed failure renders one of the sources inoperable. l The boration capability of either borated water source, in association with a flow '

path and charging pump, is sufficient to provide a SHUTDOWN MARGIN (SDM) from expected operating conditions of 1.3% Ak/k after xenon decay and cooldown to 200 F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions. To maintain SDM for this condition a O

Chapter 16.9-12 Page 2 of 3 01/16/99 l

l

BASES (con't) minimum water volume at a minimum boron concentration, as presented in the CORE OPERATING LIMITS REPORT, is required from the boric acid storage tanks or the refueling water storage tank.

A minimum contained water volume and boron concentration, as presented in the CORE OPERATING LIMITS REPORT (COLR), is required to be available from the borated water sources in MODES 1,2,3, and 4. This volume is based on the required volume for maintaining SDM, unusable volume (to allow for a full suction pipe), instrument error, and additional margin for conservatism as follows:

Boric Acid Tank '

Required Volume for Maintaining SDM presented in the COLR Additional Margin 1,303 gallons Unusable Volume (to maintain full suction pipe) 7,230 gallons 14" of water equivalent Vortexing (4" of water above top of suction pipe) 2,066 gallons Instrumentation Error (Based on Total Loop Acc. 1,550 gallons for 1 & 2 NV5740 loops)- 2" of water equivalent Refuelina Water Storaae Tank Required Volume for Maintaining SDM presented in the COLR Unusable Volume (below nozzle) 13,442 gallons Instrument inaccuracy 11,307 gallons 1

Vortexing 13,247 gallons Additional Margin 3,504 gallons The contained water volume limits include allowance for water not available because of discharge line location and other physical characteristics.

O Chapter 16.9-12 Page 3 of 3 01/16/99

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16.9 AUXILIARY SYSTEMS 16.9-13 SNUBBERS COMMITMENT:

All snubbers shall be OPERABLE. The only snubbers excluded from the requirements are those installed on non safety-related systems and then only if their i failure or failure of the system on which they are installed would have no adverse effect on any safety-related system.

APPLICABILITY:

MODES 1,2,3, and 4. MODES 5 and 6 for snubbers located on systems required '

OPERABLE in those MODES.

REMEDIAL ACTION; With one or more snubbers inoperable, within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months replace or restore the inoperable snubber (s) to OPERABLE status and perform an engineering evaluation perTESTING REQUIREMENT 16.9-13.g on the attached component or declare the ,

attached system inoperable and follow the appropriate Technical Specification ACTION statement for that system.

TESTING REQUIREMENTSJ Each snubber shall be demonstrated OPERABLE by performance of the following augmented inservice inspection program.

a. Inspection Tvoes I As used in this specification, " type of snubber" shall mean snubbers of the same design and manufacturer, irrespective of capacity.

b. Visual inspections Snubbers are categorized as inaccessible or accessible during reactor l operation. Each of these categories (inaccessible and accessible) may be inspected independently according to the schedule determined by Table .

16.9-13A. The visual inspection interval for each category of snubber shall )

be determined based upon the criteria provided in Table 16.9-13A and the i first inspection interval determined using this criteria shall be based upon the l previous inspection interval as established by the requirements in effect i before Technical Specification Amendment 88 (Unit 1) and 82 (Unit 2). l Chapter 16.9-13 Page 1 of 10 01/16/99

I TESTING REQUIREMENTS (con't) l

c. VisualInspection Acceptance Criteria '

h Visual inspections shall verify that: (1) the snubber has no visible indications of damage or impaired OPERABILITY, (2) attachments to the foundation or '

supporting structure are functional, and (3) fastenets for the attachment of the snubber to the component and to the snubber anchorage are functional.

Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that: (i) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type that may be !

generically susceptible; and (ii) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per Testing Requirement 1 16.9-13f. All snubbers found connected to an inoperable common hydraulic fluid reservoir shall be counted as unacceptable and may be reclassified as acceptable for determining the next inspection interval provided that criterion (i) and (ii) above are met. A review and evaluation shall be performed and '

documented to justify continued operation with an unacceptable snubber. If l continued operation cannot be justified, the snubber shall be declared '

inoperable and the REMEDIAL ACTION requirements shall be met. ,

O d. Refuelina Outaoe inspections At each refueling, the systems which have the potential for a severe dynamic event, specifically, the Main Steam System (upstream of the main steam 1 isolation valves) the main steam safety and power-operated relief valves and piping, Auxiliary Feedwater System, main steam supply to the auxiliary j feedwater pump turbine, and the letdown and charging portion of the CVCS System shall be inspected to determine if there has been a severe dynamic event. In the case of a severe dynamic event, mechanical snubbers in that system which experienced the event shall be inspected during the refueling outage to assure that the mechanical snubbers have freedom of movement and are not frozen up. The inspection shall consist of verifying freedom-of-motion using one of the following: (1) manually induced snubber movement; or (2) evaluation of in-place snubber piston setting; or (3) strcking the mechanical snubber through its full range of travel. If one or more mechanical snubbers are found to be frozen up during this inspection, those snubbers shall be replaced or repaired before retuming to power. The requirements of Testing Requirement 16.7-13b. are independent of the requirements of this Testing Requirement.

O Chapter 16.9-13 Page 2 of 10 01/16/99

1 I

TESTING REQUIREMENTS (con't)

e. Functional Tests During the first refueling shutdown and at least once per 18 months thereafter during shutdown, a representative sample of snubbers of each type shall be tested using one of the following sample plans. The large-bore steam generator hydraulic snubbers shall be treated as a separate type (population) for functional test purposes. A 10% random sample shall be tested at least once per 18 months during refueling with continued testing based on a failure evaluation. The sample plan shall be selected prior to the test period and 1 cannot be changed during the test period. The NRC shall be notified in writing of the sample p;an selected for each snubber type prior to the test l period or the sample plan used in the prior test period shall be implemented:

1) At least 10% of all snubbers shall be functionally tested either in-place or in a bench test. For each snubber of a type that does not meet the functional test acceptance criteria of Testing Requirement 16.9-13f, an additional 10% of all snubbers shall be functionally tested until no more failures are found or until all snubbers have been functionally tested; or O 2) A representative sample of all snubbers shall be functionally tested in accordance with Figure 16.9-13A. "C"is the total number of snubbers of a type found not meeting the acceptance requirements of Testing Requirement 16.9-13f. The cumulative number of snubbers tested is denoted by "N". At the end of each day's testing, the new values of "N" and "C" (previous day's total plus current day's increments) shall be plotted on Figure 16.9-13A. If at any time the point plotted falls in the " Accept" region, testing of snubbers of that type may be terminated. When the point plotted lies in the " Continue Testing" region, additional snubbers of that type shall be tested until the point falls in the " Accept" region or all the snubbers of that type have been tested; or

3) An initial representative sample of 55 snubbers shall be functionally tested. For each snubber type which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until the total number tested is equal to the initial sample size multiplied by the factor,1 + C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. The results from this sample plan shall be plotted ,

using an " Accept"line which follows the equation N = 55(1 + C/2). !

Each snubber point should be plotted as soon as the snubber is l

tested. If the point plotted falls on or below the " Accept" O j i

Chapter 16.9-13 Page 3 of 10 01/16/99

O O TESTING REQUIREMENTS (con't) line, testing may be terminated. If the point plotted falls above the

" Accept" line, testing must continue until the point falls it: the " Accept" region or all the snubbers of that type have been tested.

Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later time provided all snubbers tested with the failed equipment during the day of equipment failure are retested. The representative sample selected for the functional test sample plans shall be randomly selected from all snubbers and reviewed before beginning the testing.

The review shall ensure, as far as practicable, that they are representative of the various configurations, operating environments, range of size, and capacity of snubbers. Snubbers placed in the same location as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing, additional sampling is required due to failure of only one type of snubber, the functional test results shall be reviewed at that time to determine if additional samples should be limited to the type of snubber which has failed the functional testing.

p V f. FunctionalTest Acceptance Criteria The snubber functional test shall verify that:

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

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

3) For mechanical snubbers, the force required to initiate or maintain motion of the snubber is within the specified range in both directions of travel; and

4) For snubbers specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement.

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

vs Chapter 16.9-13 Page 4 of 10 01/16/99

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O U/ TESTING REQUIREMENTS (con't) l

g. Functional Test Failure Analysis i

An engineering evaluation shall be made of each failure to meet the 1 functional test acceptance criteria to determine the cause of the failure. The

', )

results of this evaluation shall be used, if applicable, in selecting snubbers to i be tested in an effort to determine the OPERABILITY of other snubbers irrespective of type which may be subject to the same failure mode. l For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if i the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that I the component remains capabl'a of meeting the designed service.

If any snubber selected for functional testing either fails to lock up or fails to 1 move, i.e., frozen-in-place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be functionally tested. This Testing Requirement shall be independent of the requirements stated in Testing Requirement 16.9-13e (3 for snubbers not meeting the functional test acceptance criteria.

b)

h. Functional Testino of Repaired and Replaced Snubbers Snubbers which fail the visual inspection or the functional test acceptance i criteria shall be repaired or replaced. Replacement snubbers and snubbers i which have repairs which might affect the functional test results shall be tested to meet the functional test criteria before installation in the unit.

Mechanical snubbers shall have met the acceptance criteria subsequent to their most recent service, and the freedom-of-motion test must have been performed within 12 months before being installed in the unit. I 1

i. Snubber Service Life Proaram j l

The scivice performance of all snubbers shall be monitored. If a service j lifetime limit is associated (established) with any snubber (or critical part) l based on manuf acturer's information, qualification tests, or historical service i resulis, then the service life shall be monitored to ensure that the service life is not exceeded between surveillance inspections. Established snubber service life shall be extended or shortened based on monitored test results and failure history. The replacements (snubbers or critical parts) shall be documented and the documentation shall be retained. Records of the p service lives of all hydraulic and mechanical snubbers, including the date at G'

Chapter 16.9-13 Page 5 of 10 01/16/99

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

i l

TESTING REQUIREMENTS (con't) l which the service life commences, and associated installation and maintenance records shall be retained for the duration of the Unit Operating ;

License, i

REFERENCES:

,1.

Letter from NRC to Gary R. Peterson, Duke, issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998. ;

o )

BASES: ;

All snubbers are required OPERABLE to ensure that the structuralintegrity of the ,

Reactor Coolant System and all other safety-related systems is maintained during l and following a seismic or other event initiating dynamic loads.

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

For example, mechanical snubbers utilizing the same design features of the 2-kip, !

10-kip, and 100-kip capacity manufactured by Company "A" are of the same type.

The same design mechanical snubbers manufactured by Company "B" for the !

purposes of this Commitment would be of a different type, as would hydraulic l snubbers from either manufacturer.

A list of individual snubbers with detailed information of snubber location and size f' and of system affected shall be available at the plant in accordance with Section 50.71(c) of 10 CFR Part 50. The addition or deletions of any hydraulic or .

mechanical snubber shall be made in accordance with Section 50.59 of 10 CFR Part 50. I i

The visual inspection frequency is based upon maintaining a constant level of i snubber protection during an earthquake or severe transient. Therefore, the ;

required inspection interval varies inversely with the observed snubber failures and i is determined by the number of inoperable snubbers found during an inspection. In ;

order to establish the inspection frequency for each type of snubber, it was assumed ;

that the frequency of snubber failures and initiating events are constant with time ;

and that the failure of any snubber on that system could cause the system to be i unprotected and to result in failure during an assumed initiating event. Inspections !

performed before that interval has elapsed may be used as a new reference point to l determine the next inspection. However, the results of such early inspections ;

performed before the original required time interval has elapsed (nominal time less l

25%) rnay liot be used to lengthen the required inspection interval. Any inspection :

whose results require a shorter inspection interval will override the previous schedule. The acceptance criteria are to be used in the visual inspection to r determine OPERABILITY of the snubbers.

O J I

Chapter 16.9-13 Page 6 of 10 01/16/99 r

.I

BASES (cont'd)

- To provide assurance of snubber functional reliability, one of three functional testing methods are used with the stated acceptance criteria: ;

1. Functionally test 10% of a type of snubber with an additional 10% tested for each functional testing failure, or

2. Functionally test a sample size and determine sample acceptance using Figure 16.9-13A, or i

3. Functionally test a representative sample size and determine sample ;

acceptance or rejection using the stated equation.

Figure 16.9-13A was developed using "Wald's Sequential Probability Ratio Plan" as ;

described in " Quality Control and Industrial Statistics" by Acheson J. Duncan.

l Permanent or other exemptions from the inspection program for individual snubbers may be granted by the Commission if a justifiable basis for exemption is presented i and, if applicable, snubber life testing was performed to qualify the snubber for the applicable design conditions. Snubbers so exempted shall be listed in the list of l individual snubbers indicating the extent of the exemptions. r The service life of a snubber is established via manufacturer input and information j through consideration of the snubber service conditions and associated installation '

and maintenance records (newly installed snubbers, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.).

l The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and l' operating conditions. These records will provide statistical bases for future 4

consideration of snubber service life. i h

i i

4 O ;

Chapter 16.9-13 Page 7 of 10 01/16/99

I TABLE 16.9-13A SNUBBER VISUAL INSDECTION INTERVAL NUMBER OF UNACCEPTABLE SNUBBERS Population Column A Column B Column C or Category Extend Interval Repeat interval Reduce Interval (Notes 1 and 2) (Notes 3 and 6) (Notes 4 and 6) (Notes 5 and 6) 1 0 0 1 80 0 0 2 100 0 1 4 150 0 3 8 200 2 5 13 300 5 12 25 400 8 18 36 500 12 24 48 O 750 1000 or greater 20 29 40 56 78 109 Note 1: The next visual inspection interval for a snubber population or category size shall be determined based upon the previous inspection interval and the number of unacceptable snubbers found during that interval.

Snubbers may be categorized, based upon their accessibility during power operation, as accessible or inaccessible. These categories may be examined separately or jointly. However, the licensee must make and document that decision before any inspection and shall use that decision as the basis upon which to determine the next inspection interval for that category.

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

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

Chapter 16.9-13 Page 8 of 10 01/16/99

TABLE 16.9-13A (Continued)

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

Note 5: If the number of unacceptable snubbers is equal to or greater than the number in Column C, the next inspection interval shall be two-thirds of the previous interval. However, if the number of unacceptable snubbers is less than the number in Column C but greater than the number in Column B, the next interval shall be reduced proportionally by interpolation, that is, the previous interval shall be reduced by a factor that is one-third of the ratio of the difference between the number of unacceptable snubbers found during the previous interval and the number in Column B to the difference in the numbers in Columns B and C.

Note 6: The provisions of SLC 16.2.6 are applicable for all inspection intervals up to and including 48 months.

O O

Chapter 16.9-13 Page 9 of 10 01/16/99

O i 10 9

L 3

8 i

7 6

i C 5 4

CONTINUE i

3 TESTING ,/

1 2 ,

' ACCEPT

1 g

}

, 0 10 20 30 40 50 60 70 80 90 100 )

l N

'l FIGURE 16.9-13A SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST O

Chapter 16.9-13 Page 10 of 10 01/16/99

. Igj AUXILIARY SYSTEMS 16.9-14 LAKE WYLIE WATER TEMPERATURE COMMITMENT:

The water temperature of Lake Wylie shall be s 92 F when aligned to the Nuclear Service Water System.

APPLICABILITY:

MODES 1,2,3, and 4.

REMEDIAL ACTION:

With the requirements of the above COMMITMENT not satisfied, at least one loop of nuclear service water shall be aligned to the Standby Nuclear Service Water Pond, immediately.

TESTING REQUIREMENTS:

h" At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during the months of July, August and September while the Nuclear Service Water System is aligned to Lake Wylie record the water temperature of Lake Wylie, as measured in the discharge path of an operating Nuclear Service Water pump.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

2. UFSAR 9.2.5, Ultimate Heat Sink.

BASES:

Two sources of water are available to the Nuclear Service Water System (NSWS);

the Standby Nuclear Service Water Pond (SNSWP) (assured source) and Lake Wylie (normal source). To ensure that the NSWS initial temperature assumptions in the peak containment pressure analysis are met, temperature of the water sources are monitored. During periods of time while Lake Wylie temperature is greater than 92 F, the emergency procedure for transfer of ECCS flow paths to cold leg recirculation directs the operator to align at least one train of containment spray to be cooled by a loop of Nuclear Service Water which is aligned to the SNSWP. The specified REMEDIAL ACTIONS are consistent with the actions required in the Chapter 16.9-14 Page 1 of 2 01/16/99

. i 4

d BASES fcon't) emergency procedures. Should the SNSWP exceed 91.5*F during unit operation, the unit is required to shutdown in accordance with Technical Specification 3.7.9. I

!- Swapover to the SNSWP is required at 92 F rather than 91.5'F because Lake Wylie ;

j is not subject to subsequent heatup due to recirculation, as is the SNSWP; hence

the 100 F design basis maximum temperature is not approached.

l

. 1 l l 4

l 1

l l

O Chapter 16.9-14 Page 2 of 2 01/16/99

16.9 AUXILIARY SYSTEMS 16.9-15 AUXlLIARY BUILDING FILTERED EXHAUST SYSTEM FILTER COOLING BYPASS VALVES COMMITMENT:

The Auxiliary Building Filtered Exhaust System filter cooling bypass valves shall be OPERABLE.

APPLICABILITY:

MODES 1,2,3, and 4.

REMEDIAL ACTION:

With the above COMMITMENT not met, take the Required Actions of Technical Specification 3.7.12.

TESTING REQUIREMENTS:

At least once per 18 months, verify that the filter cooling bypass valves can be manually opened.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

D_ASES:

Filter cooling bypass lines are provided in the filter system to provide a small cooling air flow through the non-operating filter train to limit the possibility of iodine desorption and auto-ignition from radioactivity induced heat in the adsorber section as described in Regulatory Guide 1.52, Revision 2. This COMMITMENT provides assurance that this path will be available if required by manually opening the valves once per 18 months.

O Chapter 16.9-15 Page 1 of 1 01/16/99 '

ita AUXILIARY SYSTEMS i

l 16.9 FUEL HANDLING VENTILATION EXHAUST SYSTEM FILTER COOLING BYPASS VALVES !

( i

COMMITMENT: .

. The Fuel Handling Ventilation Exhaust System filter cooling bypass valves shall be ;

t OPERABLE.

i* l APPLICABILITY: .

Whenever irradiated fuel is in the storage pool. :

I: i REMEDIAL ACTION: :

l I With the above COMMITMENT not met, take the Required Actions of Technical ,

i Specification 3.7.13. .

! TESTING REQUIREMENTS:

At least once per 18 months for each train, verify that the filter cooling bypass valves ,

can be manually opened. t

REFERENCES:

i i

j 1. Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical -

i Specifications Amendments for Catawba, September 30,1998.

i :

BASES: 1

1 Filter cooling bypass lines are provided in the filter system to provide a small cooling I

! air flow through the non operating filter train to limit the possibility of iodine desorption and auto-ignition from radioactivity induced heat in the adsorber section as described in Regulatory Guide 1.52, Revision 2. This COMMITMENT provides

{ assurance that this path will be availab!s if required by manually opening the valves

, once per 18 months.

O 4

. Chapter 16.9-10 Page.1 of 1 01/16/99

)

iga AUXILIARY SYSTEMS 16.9-17 REFUELING OPERATIONS - DECAY TIME l t

COMMITMENT:

The reactor shall be suberitical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , j APPUCABILITY: !

During movement of irradiated fuel in the reactor vessel. ,

REMEDIAL ACTION:

With the reactor subcritical for less than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , suspend all operations involving movement of irradiated fuel in the reactor vessel. 1 TESTING REQUIREMENTS:

The reactor shall be deterrnined to have been suberitical for at least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by I verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor vessel. )

REFERENCES:

1. . Letter from NRC to Gary R. Peterson, Duke, Issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

l BASES:

The minimum requirement for reactor suberitice!.ity prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to !

aliaw the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the safety analyses.

4 i

O Chapter 16.9-17 Page 1 of 1 01/16/99

1 1gj AUX 1UARY SYSTEMS l 16,9-18 REFUEUNG OPERATIONS - COMMUNICATIONS 1

COMMITMENT:

Direct communications shall be maintained between the control room and personnel !

at the refueling station. i APPUCABluTY: !

During CORE ALTERATIONS. '

REMEDIAL ACTION: :

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

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

Direct communications between the control room and personnel at the refueling stations shall be demonstrated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to the start of and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months during CORE ALTERATIONS. '

REFERENCES:

9

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

{

The requirement for communications capability ensures that refueling station !

personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.

i

=

Chapter 16.9-18 Page 1 of 1 01/16/99 !

.w, - . . ., ,- - e. , - - - ---- - - , _ _ . . . _ .- ..-m.1-v- .-- ,

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

16.9 AUXILIARY SYSTEMS 16.9-19 REFUELING OPERATIONS - MANIPULATOR CRANE ,

l COMMITMENT: ,

The reactor building manipulator crane and an auxiliary hoist shall be used for movement of fuel assemblies or control rods and shall be OPEP.AR.E with: '

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

1) A minimum capacity of 3250 pounds, and

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

b. Auxiliary hoists used for latching, unlatching and drag load testing of control 1 rods having:

1) A minimum capacity of 1000 pounds, and

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

APPLICABILITY:

During movement of fuel assemblies and control rods within the reactor vessel.

REMEDIAL ACTION:

With the requirements for crane and/or hoist OPERABILITY not satisfied, suspend use of any inoperable manipulator crane and/or auxiliary hoist from operations involving the movement of fuel assemblies and control rods within the reactor vessel.

TESTING REQUIREMENTS: !

1. Each manipu!ator crane used for movement of fuel assemblies within the reactor vessel shall be demonstrated OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of such operations by performing a load test of at least 3250 pounds and demonstrating an automatic load cutoff when the crane load exceeds 2850 pounds.

I O

I Chapter 16.9-19 Page 1 of 2 01/16/99 l

f TESTING REQUIREMENTS (con't) 1

2. Each auxiliary hoist and associated load indicator used for movement of control rods or control rod drag load testing within the reactor vessel shall be ;

demonstrated OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months prior to the start of such '

operations by performing a load test of at least 1000 pounds. !

REFERENCES:

l

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

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

(3 V

Chapter 16.9-19 Page 2 of 2 01/16/99

16.9 AUXILIARY SYSTEMS 16.9-20 REFUELING OF$ATIONS - CRANE TRAVEL - SPENT FUEL STORAGE POOL DUILDING COMMITMENT:

Loads in excess of 3000 pounds'shall be prohibited fram travel over fuel assemblies in the storage pool. The requirements cf Technical Specification Limiting Condition for Operation 3.8.2 shall be met whenever loads are moved over the storage pool.

APPLICABILITY:

With fuel assemblies in the storage pool.

REMEDIAL ACTION:

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

TESTING REQUIREMENTS:

The weight of each load, other than a fuel assembly and control rod, shall be verified to be less than or equal to 3000 pounds prior to moving it over fuel assemblies.* -

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

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

t

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

Chapter 16.9-20 Page 1 of 1 01/16/99 '

l 16.9 AUXILIARY SYSTEMS 16.9-21 REFUELING OPERATIONS - STORAGE POOL WATER LEVEL COMMITMENT:

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

APPLICABILITY:

Whenever irradiated fuel assemblies are in the storage pool.

REMEDIAL ACTION:

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

TESTING REQUIREMENTSj (D

\- 'j The water level in the storage pool shall be determined to be at least its minimum

) required depth at least once per 7 days when irradiated fuel assemblies are in the j fuel storage pool.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

_ BASES:

The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis.

Chapter 16.9-21 Page 1 of 1 01/16/99

, ,~., -

I

,1ga AUXILIARY SYSTEMS 16.9-22 CONTROL ROOM AREA VENTILATION SYSTEM -INTAKE, 3 ALARMS- '

COMMITMENT:

l The Control Room Area Ventilation Systems intake Alarms shall be OPERABLE.

APPLICABILITY:

2 All MODES. l REMEDIAL ACTION: (Units 1 and 2) :

MODES 1,2,3 and 4:

With the Control Room Area Ventilation System Intake Alarms inoperable re? lore the inoperable system to OPERABLE status within 7 days or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within the folk aing 30 i hours.

f- ;

( MODES 5 and 6: 1 With the Control Room Area Ventilation Systems intake Alarms inoperable, suspend all operations involving CORE ALTERATIONS or positive reactivity changes. I NOTE: The provisions of SLC 16.2.3 are not applicable.

TESTING REQUIREMENTS:

The Control Room Area Ventilation System Intake Alarms shall be demonstrated OPERABLE at least once per 18 months by verifying that on a High Radiation-Air intake, or Smoke Density-High test signal, an alarm is received in the control room.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, issuance of improved Technical Specification-s Amendments for Catawba, September 30,1998.

O Chapter 16.9-22 Page 1 of 2 01/16/99

BASES:

The Control Room Area Ventilation System Intake Alarms provide operator information relative to smoke and radiation concentrations at each control room intake. Operators use this information to align the Control Room Area Ventilation System to ensure that the control room will remain habitable for operations personnel during and following accident conditions.

O t

Chapter 16.9-22 Page 2 of 2 01/16/99

l 1

l (J

16.9 AUXILIARY SYSTEMS - FIRE PROTECTION SYSTEMS I 16.9-23 FIRE HYDRANTS

_ \

COMMITMENT:

Fire Hydrants located at the RN Pump Structure (FH61 and FH62) shall be I OPERABLE:

APPLICABILITY:

Whenever equipment in the RN Pump Structure is required to be OPERABLE.

REMEDIAL ACTION:

With one of the subject fire hydrants inoperable, the other fire hydrant shall be verified as OPERABLE. The inoperable fire hydrant shall be tagged as such and restored to OPERABLE status in a timely manner. With both of the subject fire hydrants inoperable, within one shift establish a once per shift fire watch patrol for the RN Pump Structure.

TESTING REQUIREMENTS i

i Each of the above subject fire hydrants shall be demonstrated OPERABLE:

a. By verifying that each valve in the flow path is in the correct position. The l frequency of the verification shall be determined by the performance based criteria stated in the Bases Section. l

b. At least once per 12 months by cycling each testable valve in the flow path through at least one complete cycle cf full travel. I

c. At least once per 6 months by a visual inspection of the hydrants to assure the hydrants show no signs of physical damage.

d. At least once per 6 months by performing a system flush of the RN Pump Structure fire protection hydrant piping to verify no flow blockage by fully opening the hydraulically most remote RN Pump Structure Hydrant.

REFERENCES:

1. Catawba FSAR, Section 9.5.1

/ 2. Catawba SER, Section 9.5.1

3. Catawba Fire Protection Review, as revised j Chapter 16.9-23 Page 1 of 3 05/04/98-

p REFERENCES (cont'd}

4. Catawba Fire Protection Commitment index

5. NFPA 25 and 291 BASES:

The OPERABILITY of the Fire Suppression Systems 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 system, spray and/or sprinklers, CO2, 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.

The proper positioning of RF/RY valves is critical to delivering fire suppression water at the source of the fire as quickly as possible. The option of increasing or decreasing the frequency of valve position verification allows the ability to optimize plant operational resources. Should an adverse trend develop with RF/RY valve positions the frequency of verification shall be increased. Similarly, if the RF/RY valve position trends are positive, the frequency could be decreased. Through O

d programmed trending of RF/RY as found valve positions, the RF/RY System will be maintained at predetermined reliability standards. The RF/RY System Engineer is responsible for trending and determining verification frequencies based on the following:

Initially the frequency will be month'y.

Annually review the results of the completed valve position verification procedures.

l If the results demonstrate that the valves are found in the correct position at least 99% of the time, the frequency of conducting the valve position verification may be decreased from - monthly to quarterly or - quarterly to semiannually or - l semiannually to annually - as applicable. The frequency shall not be extended beyond annually (plus grace period).

If the results demonstrate that the valves are not found in the correct position at least 99% of the time, the frequency of conducting the valve position verification shall be increased from - annually to semiannually or - semiannually to quarterly or - quarterly to monthly - as applicable. The valve position verification need not be conducted more often than monthly. I In the event that portions of the Fire Suppression Systems are inoperable, attemate

('} backup fire-fighting equipment is required to be made available in the affected areas V

Chapter 16.9-23 Page 2 of 3 05/04/98 l

l l

R&SEli (cont'd) until the inoperable equipment is restores to service. When the inoperable fire-fighting equipment is intended for use as a backup means of fire suppression, a longer period of time is allowed to provide an attemate means for fire suppression.

\

This Selected Licensee Commitment is part of the Catawba Fire Protection Program ;

and therefore subject to the provisions of Section 2.C. of the Catawba Facility i Operating Licenses. '

I O

t i

i i

O Chapter 16.9-23 Page 3 of 3 05/04/98

l 16.10 STEAM AND POWER CONVERSION SYSTEM 16.10-1 STEAM VENTTO ATMOSPHERE COMMITMENT:

Four steam generator PORV safety-related gas supply systems shall be OPERABLE with both nitrogen bottles per S/G PORV, pressurized to greater than or {

equal to 2100 psig.

APPLICABILITY:

l l

Modes 1, 2, 3, 4*

i REMEDIAL ACTION:

a. With one nitrogen bottle on one or more S/Gs less than 2100 psig, immediately start corrective action to retum the nitrogen supply to '

OPERABLE. Work to retum the nitrogen supply to OPERABLE status should continue without interruption. '

b. With two nitrogen bottles on one or more S/Gs less than 2100 psig, consider the PORV(s) inoperable and refer to Technical Specification 3.7.4 for the Required Action.

TESTING REQUIREMENTS:

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by verifying that both nitrogen bottles per S/G PORV has a pressure greater than or equal to 2100 psig.

REFERENCES:

1. Design Basis Specification for the Catawba Main Steam System, Main Steam Vent to Atmosphere and Main Steam Bypass to Condenser System, Section 20.3.4

2. PIR 0-C90-0304

3. Branch Technical Position RSB5-1

4. CNC-1223.43-01-0011, rev 1 When Steam Generators are being used for decay heat removal.

Chapter 16.10-1 Page 1 of 2 01/16/99

BASES:

Design Engineering calculation CNC-1223.43-01-0011, rev 1, demonstrates that with one nitrogen bottle charged to at least 2100 psig, sufficient nitrogen exists to meet the Tech Spec Design basis of the S/G PORVs.

A revision to calculation CNC-1223.43-01-0011 also demonstrates that with two nitrogen bottles charged to as least 2100 psig, sufficient nitrogen exists to meet the Branch Technical Position RSB5-1 of supporting a controlled cooldown to the point where residual heat removal system can be put in service with or without offsite power following an earthquake.

The COMMITMENT for having both nitrogen bottles pressurized to greater than or equal to 2100 psig and the REMEDIAL ACTION, is adequate to ensure the intent of our FSAR commitment to Branch technical Position RSB5-1 is met.

O O

Chapter 16.10-1 Page 2 of 2 01/16/99 l

d d

16.10 STEAM AND POWER CONVERSION SYSTEM 16.10-2 CONDENSER CIRCULATING WATER SYSTEM COMMITMENT:

With the Condenser Cooling Water System (RC) partially or completely filled, the j

4 system boundaries within the Turbine Building and RC pump pit shall be in their normal alignment.

APPLICABILITY:

All Plant conditions which require the availability of the 6900/4160V Essential Transformers: (SATA, SATB,1 ATC,1 ATD,2ATC,2ATD) for EITHER Train and for EITHER Unit.

REMEDIAL ACTION:

Restore the RC System boundaries to the normal cornmitment alignment in accordance with the Risk Assessment Matrix priorities.

TESTING REQUIREMENTS: j None

REFERENCES:

1. CNS FSAR, Section 10.4.5.3

2. WPM 607, Maintenance Rule Assessment of Equipment out of Service

3. 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.

4. SAAG File: 160 Severe Accident Analysis Report, CNS Probability Risk l Assessment (PRA) Risk Significant SSCs for the Maintenance Rule Chapter 16.10-2 Page 1 of 2 01/15/97

BASES:

The effective implementation of the Maintenance Rule,10 CFR 50.65, requires the continuous assessment of systems determined Risk Significant in the protection against Core Damage or Radiation Release. It has been determined through PRA numerical methods that this system function provides a significant contribution to the defense in the prevention of a Loss of Offsite Power Event. This SLC serves two purposes.

(1) It defines the Risk Significant concems of the Condenser Circulating Water, RC System integrity with respect to flooding EITHER Units 6900/4160V Essential Transformers. A failure to control the RC system inventory while partially or completely full has the potential consequence of degrading the power function of the 6900/4160 V Essential Transformers for either or both units. Damage to these transformers may result in either the Loss of Offsite Power (LOOP) or a significant decrease in the defense of Accident Mitigating Equipment. The concem from this event includes either RC System of Unit 1 or Unit 2 leading to the affect on either Unit / Train transformers.

(2) This SLC also provides a method of tracking this function for intersystem configuration control of the Essential Transformers and their susceptibility to flooding through support of WPM 607 and 10 CFR 50.65.

O 1

l i

O Chapter 16.10-2 Page 2 of 2 01/15/97

I

] 16.11 RADIOLOGICAL EFFLUENTS CONTROLS l

16.11-1 LIQUID EFFLUENTS i l

COMMITMENT:

1 The concentration of radioactive material released in liquid effluents to 3

UNRESTRICTED AREAS (see Figure 16.11-1) shall be limited to ten times the l

' l 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 2 x 104 microcurie /mi l i

total activity. '

l 4

APPLICABILITY: l At all times.

REMEDIAL ACTION:

l With the concentration of radioactive material released in liquid effluents to j UNRESTRICTED AREAS exceeding the above limits, immediately restore the 1 j concentration to within the above limits.

TESTING REQUIREMENTS:

Radioactive liquid wastes shall be sampled and analyzed according to the sampling and analysis program of Table 16.11-1.

The results of the radioactivity analyses shall be used in accordance with the methodology and parameters in the ODCM to assure that the concentrations at the point of release are maintained within the limits of SLC 16.11-1.

REFERENCES:

1. Catawba Offsite Dose Calculation Manual

2. 10 CFR Part 20, Appendix B J

BASES:

i The basic requirements for Selected Licensee Commitments conceming effluents from nuclear power reactors are stated in 10 CFR 50.36a. These requirements indicate that compliance with effluent Selected Licensee Commitments will keep average annual releases of radioactive material in effluents to small percentages of Chapter 16.11-1 Page 1 of 7 01/16/99

- - - . . . - - . _ - .. _ _ - . - - _ . _ = - ___ .-.

[ BASES (con't) the limits specified in the old 10 CFR 20.106 (new 10 CFR 20.1302). These requirements further indicate that operational flexibility is allowed, compatible with considerations of health and safety, 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 11 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 flexibility, best efforts shall be exerted to keep levels of radioactive materials in effluents as low as is reasonably achievable (ALARA) as set forth in 10 CFR 50, Appendix 1.

As stated in the Introduction to Appendix B of the new 10 CFR 20, the liquid effluent concentration (EC) limits given in Appendix B, Table 2, Column 2, are based on an annual dose of 50 mrem. Since a release concentration corresponding to a limiting dose rate of 500 mrem / year has been acceptable as a SLC limit for liquid effluents, which applies at all times as an assurance that the limits of 10 CFR 50, Appendix I .

are not likely to be exceeded, it should not be necessary to reduce this limit by a factor of 10.

Operational history at Catawba has demonstrated that the use of the concentration values associated with the old 10 CFR 20.106 as SLC limits has resulted in C calculated maximum individual doses to a MEMBER OF THE PUBLIC that are small percentages of the limits of 10 CFR 50, Appendix 1. Therefore, the use of concentration values which correspond to an annual dose of 500 mrem (ten times the concentration values stated in the new 10 CFR 20, Appendix B, Table 2, Column 2) should not have a negative impact on the ability to continue to operate within the limits of 10 CFR 50, Appendix I and 40 CFR 190.

Having sufficient operational flexibility is especially important in establishing a basis for effluent monitor setpoint calculations. As discussed above, the concentrations stated in the new 10 CFR 20, Appendix B, Table 2, Column 2, relate to a dose of 50 mrem in a year. When applied on an instantaneous basis, this corresponds to a dose rate of 50 mrem / year. This low value is impractical upon which to base effluent monitor setpoint calculations for many liquid effluent release situations when monitor background, monitor sensitivity, and monitor performance must be taken into account.

Therefore, to accommodate operational flexibility needed for effluent releases, the limits associated with SLC 16.11-1 are based on ten times the concentrations stated in the new 10 CFR 20, Appendix B, Table 2, Column 2, to apply at all times. The i multiplier of ten is proposed because the annual dose of 500 mrem, upon which the ;

concentrations in the old 10 CFR 20, Appendix B, Table 11, Column 2, are based, is ;

a factor of 10 higher than annual dose of 50 mrem, upon which the concentrations in the new 10 CFR 20, Appendix B, Table 2, Column 2, are based. Compliance with Chapter 16.11-1 Page 2 of 7 01/16/99

BASES (con't) the limits of the new 10 CFR 20.1301 will be demonstrated by operating within the limits of 10 CFR 50, Appendix I and 40 CFR 190. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-135 is the controlling radioisotope and its MPC in air (submersion) was converted to an equivalent concentration in water using the methods described in Intemational Commission on Radiological Protection (ICRP) Publication 2.

This commitment applies to the release of radioactive materials in liquid effluents from all units at the site.

The required detection capabilities for radioactive materials in liquid waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry," Annal. Chem. 40. 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

O O

Chapter 16.11-1 Page 3 of 7 01/16/99

TABLE 16.11-1 (Page 1 of 3) g RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LOWER LIMIT OF MINIMUM TYPE OF DETECTION LIQUID SAMPLING ANALYSIS ACTIVITY (LLD)N i RELEASE TYPE FREQUENCY FREQUENCY ANALYSIS ( Ci/ml)

1. Batch Waste P P Release Each Batch Each Batch Principle Gamma 5x10-7 Tankst2) Emitters ( )

1-131 1x10* l Any tank P M Dissolved and 1x10*

which One Batch /M Entrained Gases discharges (Gamma emitters) liquid wastes by either liquid effluent l monitor, !

EMF-49 or EMF-57

(' P M H-3 1x10* ;

Each Batch Composite") i Gross Alpha 1 x10 I h

P Q Sr-89, Sr-90 5x10*

Each Batch Composite")

2. Continuous W Principal Gamma Releases

Continuous

Composite (') Emitters (') 5x10

l

a. Conventional Waste Water Treatment Line

b. Turbine Building Sump Demineralizer Skid, EMF-31*

l-131 1x10* l I

Chapter 16.11-1 Page 4 of 7 01/16/99

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

1 i

)

l M M Dissolved and 1x10* l Grab Sample Entrained Gases (Gamma Emitters) 1 l

M H-3 1x10*

Continuous (*) Composite (6)

Gross Alpha 1x10-'

O Sr-89, Sr-90 5x10*

Continuous (*) Composite (')

l

During use of demineralizer (use EMF-31 in off-normal mode).

l 1

l O i l

l l

l i

l O

Chapter 16.11-1 Page 5 of 7 01/1G/99

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

k

( TABLE 16.11-1 (Page 2 of 3) 1 TABLE NOTATIONS (1) The LLD is defined, for purposes of these commitments, as the smallest 4

concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation: 1 g 4.66s, E. V 2.22 x 106 Y exp (-Aar)

Where: '

LLD = the "a priori" lower limit of detection (microcurie per unit mass or volume),

so= the standard deviation of the background counting rate or of the i counting rate of a blank sample as appropriate (counts per minute),

l E = the counting efficiency (counts per disintegration),

V = the sample size (units of mass or volume),

6 2.22 x 10 = the number of disintegrations per minute per microcurie, Y = the fractional radiochemical yield, when applicable, A = the radioactive decay constant for the particular radionuclide (sec"), and At = the elapsed time between midpoint of sample collection and time of counting (sec).

Typical values of E, V, Y and At shall be used in the calculation.

It should be recognized that the LLD is defined as an g oriori (before the fact) limit representing the capability of a measurement system and not as an g posteriori (after the fact) limit for a particular measurement.

O Chapter 16.11-1 Page 6 of 7 01/16/99

l TABLE 16.11-1 (Page 3 of 3) {

i TABLE NOTATIONS (Continued) j (2) A batch release is the discharge of liquid wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated, and then thoroughly !

mixed to assure representative sampling. !

i (3) The principal gamma emitters for which the LLD specification applies include l the following radionuclides: Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, Cs- ,

134, Cs-137, and Ce-141. The LLD for Ce-144 is 5x104 Ci/ml. This list i

does not mean that only these nuclides are to be considered. Other gamma l peaks that are identifiable, together with those of the above nuclides, shall j also be analyzed and reported in the Radioactive Effluent Release Report '

pursuant to Technical Specification 5.6.3 in the format outlined in Regulatory l Guide 1.21, Appendix B, Revision 1, June 1974.

l (4) A composite sample is one in which the quantity of liquid sampled is l proportional to the quantity of liquid waste discharged and in which the i method of sampling employed results in a specimen that is representative of the liquids released.

O (5) A continuous release is the discharge of liquid wastes of a nondiscrete ;

volume, e.g., from a volume of a system that has an input flow during the l

continuous release.

/i (6) 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 flow of the effluent stream. Prior to analyses, all !

samples taken for the composite shall be thoroughly mixed in order for the !

composite sample to be representative of the effluent release. !

i 3

i l

l l

1 l

O l Chapter 16.11-1 Page 7 of 7 01/16/99 l

_j

O 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-2 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION COMMITMENT:

The radioactive liquid effluent monitoring instrumentation channels shown in Table 16.11-2 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of SLC 16.11-1 are not exceeded. The Alarm / Trip Setpoints of these channels shall be detemlined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY:

At all times.

REMEDIAL ACTION:

a. With a radioactive liquid effluent monitoring instrumentation channel g Alarm / Trip Setpoint less conservative than required by the above g specification, immediately suspend the release of radioactive liquid effluents monitored by the affected channel, or declare the channel inoperable.

b. With less than the minimum number of radioactive liquid effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 16.11-2. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Radioactive Effluent J Release Report pursuant to Technical Specification 5.6.3 why this {

inoperability was not corrected within the time specified. j TESTING REQUIREMENTS:

)

Each radioactive liquid effluent monitoring instrumentation channel shall be i demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION and CHANNEL OPERATIONAL TEST l operations at the frequencies shown in Table 16.11-3.

REFERENCES:

i

1. Catawba Offsite Dose Calculation Manual

2. 10 CFR Part 20 b

d 3. 10 CFR Part 50, Appendix A Chapter 16.11-2 Page 1 of 6 01/16/99

O BASES:

The radioactive liquid effluent instrumentation is provided to monitor and control, as applicable, the releases of radioactive materials in liquid effluents during actual or i potential releases of liquid effluents. The Alarm / Trip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the Alarm / Trip will occur prior to exceeding 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.

O l

i l

1 l

l l

O Chapter 16.11-2 Page 2 of 6 01/16/99 1

I

O O O TABLE 16.11-2 (Page 1 of 2)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM l CHANNELS l INSTRUMENT OPERABLE ACTION l 1. Radioactivity Monitors Providing Alarm And Automatic Termination of Release

!- a. Waste Liquid Discharge Monitor (Low Range - EMF-49) 1 per station C l l '

j b. Turbine Building Sump Monitor (Low Range - EMF-31) 1 E l l c. Deleted

d. Monitor Tank Building Uguid Discharge Monitor (EMF-57) 1 per station C l l

2. Continuous Composite Samplers And Sampler Flow Monitor

a. Conventional Waste Water Treatment Line 1 per station E l

b. Turbine Building Sump 1 per station E* l

3. Flow Rate Measurement Devices

a. Waste Liquid Effluent Line 1 per station D l

b. Conventional Waste Water Treatmant Line 1 per station D l

c. Low Pressure Serv *ce Water Minimum Flow Interlock 1 per station D l

d. Monitor Tank Building Waste Liquid Effluent Line 1 per station D l

e. Turbine Building Sump Demineralizer Skid Totalizer 1 per station D* l Chapter 16.11-2 Page 3 of 6 01/16/99 '

TABLE 16.11-2 (Page 2 of 2)

TABLE NOTATIONS i

During use of demineralizer (EMF-31 in off-normal mode)

REMEDIAL ACTION STATEMENTS

)

ACTION C - With the number of channels OPERABLE less than required by the l f Minimum Channels OPERABLE requirement, effluent releases via this l pathway may continue for up to 14 days provided that prior to initiating !

a release:

i

a. At least two independent samples are analyzed in accordance with !

SLC 16.11-1; and !

i

b. At least two technically qualified members of the facility staff i independently verify- '

1) The discharge line valving; and, !

i 2)- The manual portion of the computer input for the release rate !

O calculations performed on the computer, or the entire release rate calculations if such calculations are performed manually.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION D - With the number of channels OPERABLE less than required by the l !

Minimum Channels OPERABLE requirement, effluent releases via this l

pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during actual releases. Pump perfomiance curves generated in place may be used to estimate flow. !

l ACTION E - With the number of channels OPERABLE less than required by the l l Minimum Channels OPERABLE requirement, effluent releases via this !

pathway may continue for up to 30 days provided grab samples are i

analyzed for radioactivity at a lower limit of detection of no more than l 10 microcurie /mi- l

a. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the l

secondary coolant is greater than 0.01 microcurie / gram DOSE !

EQUIVALENT l-131, or

'l

b. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the l secondary coolant is less than or equal to 0.01 microcurie / gram O DOSE EQUIVALENT l-131.

Chapter 16.11-2 Page 4 of 6 01/16/99

O O O

~

TABLE 16.11-3 (Page 1 of 2)

RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL CHANNEL SOURCE CHANNEL OPERATIONAL INSTRUMENT CHECK CHECK CALIBRATION TEST

1. Radioactivity Monitors Providing Alarm and Automatic Termination of Release

a. Waste Liquid Discharge Monitor (Low Range - EMF-49) D P R(2) O(1)

b. Turbine Building Sump Monitor (Low Range - EMF-31) D M R(2) O(1)

c. Deleted

d. Monitor Tank Building Liquid Discharge Monitor (EMF-57) D P R(2) O(1)

2. Continuous Composite Samplers and Sampler Flow Monitor

a. Conventional Waste Water Treatment Line D(3) N.A. R~ N.A.

b. Turbine Building Sump D(3) N.A. R N.A.

3. Flow Rate Measurement Devices

a. Waste Liquid Effluent Line D(3) N.A. R N.A.

b. Conventional Waste Water Treatment Line D(3) N.A. R N.A.

c. Low Pressure Service Water Minimum Flow D(3) N.A. R O Interlock

d. Monitor Tank Building Waste Liquid Effluent Line D(3) N.A. R O

e. Turbine Building Sump Demineralizer Skid Totalizer D(3) N.A. R N.A.

Chapter 16.11-2 Page 5 of 6 01/16/99

o TABLE 16.11-3 (PEge 2 of 2)

TABLE NOTATIONS (1) The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic

l isolation of this patiiway and control room alarm antmciation occur if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint; or,

b. Circuit failure (alarm only); or,

c. Instrument indicates a downscale failure (alarm only).

(2) The initial CHANNEL CAllBRATION shall be performed using one or more of the reference standards certified by the National Bureau of St?ndmos (NBS) or using standards that have been obtained from suppliers that participata in measurement assurance activities with NBS. These standards shall permit calibrating the l system overits intended range of energy and measurement range. For subsequent CHANNEL CAllBRAITON, sources that have been related to the

( initial calibration shall be used.

(3) CHANNEL CHECK shall consist of verifying indication of flow during periods of release. CHANNEL CHECK shall be made at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months on days on which continuo s, periodic, or batch releases are made.

For EMF-57, the alarm annunciation is in the Monitor Tank Building Control Room and on the MTB O Control Panel Remote Annurr;iator panet.

Chapter 16.11-2 Page 6 of 6 01/16/99

16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-3 QQEE COMMITMENT:

The dose or dose commitment to a MEMBER OF THE PUBLIC from radioactive materials in liquid effluents released, from each unit, to UNRESTRICTED AREAS (see Figure 16.11-1) shall be limited:

a. During any calendar quar 1er to less than or equal to 1.5 mrem to the whole body and to less than or equaMo 5 mrem to any organ, and

b. During any calendar year to less than or equal to 3 mrem to the whole body and to less than or equal to 10 mrem to any organ.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

O. With the calculated dose from the release of radioactive materials in liquid effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report that identifies the cause(s) for exceeding the limit (s) and l defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits. This Special Report shall also include: (1) the results of radiological analyses of the drinking water source, and (2) the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR Part 141, Safe Drinking Water Act.

TESTING REQUIREMENTS:

Cumulative dose contributions from liquid effluents for the current calendar quarter and the current calendar year shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

\ The requirements of REMEDIAL ACTRON (1) and (2) are applicable only if drinking water supply is taken from the receiving water body within 3 miles downstream of the plant discharge.

Chapter 16.11-3 Page 1 of 2 01/16/99

4 1

3- !

REFERENCES:

1. Catawba Offsite Dose Calculation Manual

2. 40 CFR Part 141 f

3. 10 CFR Part 50, Appendix l I l

l E t d

This commitment is provided to implement the requirements of Sections ll.A, Ill.A !

i and IV.A of Appendix 1,10 CFR Part 50. The Limiting Condition for Operation i implements the guides set forth in Section ll.A of Appendix 1. The REMEDIAL l

[ ACTION statements provide the required operating flexibility and at the same time 4

implement the guides set forth in Section IV.A of Appendix l to assure that the ;

releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will j j be kept "as low as is reasonably achievable". Also, for fresh water sites with '

e drinking water supplies that can be potentially affected by plant operations, there is

~ !

reasonable assurance that the operation of the facility will not result in radionuclide j

, concentrations in the finished drinking water that are in excess of the requirements ;

. of 40 CFR Part 141. The dose calculation methodology and parameters in the '

- ODCM implement the requirements in Section Ill.A of Appendix i that conformance l

! with the guides of Appendix I be shown by calculational procedures based on >

models and data, such that the actual exposure of a MEMBER OF THE PUBLIC

through appropriate pathways is unlikely to be substantially underestimated. The - -

g equations specified in the ODCM for calculating the doses due to the actual releese rates of radioactive materials in liquid effluents are consistent with the methodology j

provided in Regulatory Guide 1.109, " Calculation of Annual Doses to Man from j Routine Releases 6f Reactor Effluents for the Purpose of Evaluating Compliance 4 with 10 CFR Part 50, Appendix I, " Revision 1, October 1977 and Regulatory Guide

1.113," Estimating Aquatic Dispersion of Effluents from Accidental and Routine 7 . Reactor Releases for the Purpose of implementing Appendix I", April 1977.

/

This commitment applies to the release of radioactive materials in liquid effluents j from each unit at the site. When shared Radwaste Treatment Systems are used by

. more than one unit on a site, the wastes from all units are mixed for shared j

' treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to commitments, these allocations from shared Radwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

O Chapter 16.11-3 Page 2 of 2 01/16/99

16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-4 LIQUID RADWASTE TREATMENT SYSTEM

~ j l

COMMITMENT: '

The Liquid Radwaste Treatment System shall be OPERABLE and appropriate portions of the system shall be used to reduce releases of radioactivity when the l projected doses due to the liquid effluent, from each unit, to UNRESTRICTED l AREAS (see Figure 16.11-1) would exceed 0.06 mrem to the whole body or 0.2 mrem to any organ in a 31-day period.

APPLICABILITY:

At all times.

REMEDIAL ACTION: !

With radioactive liquid waste being discharged without treatment and in excess of the above limits and any portion of the Liquid Radwaste Treatment System not in operation, prepare and submit to the Commission within 30 days a Special Report l that includes the following infomiation:

1. Explanation of why liquid radwaste was being discharged without treatment, identification of any inoperable equipment or subsystems, and the reason for the inoperability,

2. Action (s) taken to restore the inoperable equipment to OPERABLE status, and

3. Summary description of action (s) taken to prevent a recurrence.

TESTING REQUIREMENTS:

Doses due to liquid releases from each unit to UNRESTRICTED AREAS shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Liquid Radwaste Treatment Systems are not being fully utilized.

The installed Liquid Radwaste Treatment System shall be considered OPERABLE by meeting SLC 16.11-1 and 16.11-3.

Chapter 16.11-4 Page 1 of 2 01/16/99

REFERENCES:

1. Catawba Otisite Dose Calculation Manual 1

2. 10 CFR Part 50, Appendix A

3. 10 CFR Part 50, Appendix !

BASES:

The OPERABILITY of the Liquid Radwaste Treatment System ensures that this system will be available for use whenever liquid effluents require treatment prior to release to the environment. The requirement 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 commitment implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objective given in Section ll.D of Appendix I to 10 CFR Part 50. The specified limits goveming the use of appropriate portions of the Liquid Radwaste Treatment System were specified as ,

a suitable fraction of the dose design objectives set forth in Section ll.A of Appendix 1,10 CFR Part 50, for liquid effluents.

O V

This commitment applies to the release of radioactive materials in liquid effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to SLCs, these allocations from shared Radwaste l Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

A U

Chapter 16.11-4 Page 2 of 2 01/16/99

1 O

Q 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-5 CHEMICAL TREATMENT PONDS l

l COMMITMENT: :

The quantity of radioactive material contained in each chemical treatment pond shall be limited by the following expression:

264 E A j

< 1.0 V j (C, x10)

I excluding tritium and dissolved or entrained noble gases, Where:

Aj = pond inventory limit for single radionuclide "j", in Curies C3 = 10 CFR 20, Appendix B, Table 2, Column 2, concentration for q single radionuclide "j", microCuries/ml; b V = design volume of liquid and slurry in the pond, in gallons; and 264 = conversion unit, microCuries/ Curie per milliliter / gallon.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

With the quantity of radioactive materialin any of the above listed ponds exceeding the above limit, immediately suspend all additions of radioactive material to the pond and initiate corrective action to reduce the pond contents to within the limit.

TESTING REQUIREMENTS:

The quantity of radioactive material contained in each batch of resin / water sluny to be transferred to the chemical treatment ponds shall be determined to be within the above limit by analyzing a representative sample of the batch to be O

O Chapter 16.11-5 Page 1 of 3 08/01/94

I t

TESTING Phau:REMENTS (con't) transferred to the chemical treatment ponds and shall be limit:d by the expression:

E c j -

< 0.006 t j (C j x10) 1 Where:

cj = radioactive resin / water slurry concentration for radionuclide "j" entering the UNRESTRICTED AREA chemical treatment ponds, in microCuries/ milliliter; and Cj = 10 CFR 20, Appendix B, Table 2, Column 2, concentration for single radionuclide "j", in microCuries/ milliliter.

REFERENCES:

1. Catawba Offsite Dose Calculation Manual

2. 10 CFR Part 20, Appendix B

3. 10 CFR Part 50, Appendix !

BASES:

The inventory limits of the chemical treatment ponds (CTP) are based on limiting the consequences of an uncontrolled release of the pond inventory. The expression in this commitment assumes the pond inventory is uniformly mixed, that the pond is located in an uncontrolled area as defined in 10 CFR Part 20, and that the concentration limit in Note 1 to Appendix B of 10 CFR Part 20 applies.

The batch limits of resin / water sluny transferred to the CTP assure that radioactive material transferred to the CTP are "as low is reasonably achievable"

in accordance with 10 CFR 50.36a. The expression in SLC 16.11-6 assures no batch will be transferred to the CTP unless the sum of the ratios of the activity of the radionuclides to their respective concentration limitation is less than the ratio ;

of the 10 CFR Part 50, Appendix 1, Section ll.A, total body dose level to the instantaneous whole body dose rate limitation, or that:

I ci 3 mrem /yr

= 0.006 d y I (C jx 10) 500 mrem /yr ,

Chapter 16.11-5 Page 2 of 3 08/01/94

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

l l

?

I BASES (cp.p'll Where: l i

cj s-radioactive resin / water slurry concentration for radionuclide "j" l entering the UNRESTRICTED AREA CTP, in !

riicroCuries/ milliliter; and, i Ci = E CFR Part 20, Appendix B, Table 2, Column 2, concentration t

for single radionuclide "j", in microCuries/ milliliter. i The filter /demineralizers using powdered resin and the blowdown demineralizer !

are backwashed or sluiced to a holding tank. The tank will be agitated to obtain l a representative sample of the resin inventory in the tank. A known weight of the j wet, drained resin (moisture content approximately 55 to 60%, bulk density of j about 58 pounds per cubic foot) will then be counted. The concentration of the j resin slurry to be pumped to the chemical treatment ponds will then be ;

determined by the formula: j cj= Oi Wn V7 i i

Where: l Q, = concentration of radioactive materials in wet, drained resin for l radionuclide "j", excluding tritium, dissolved or entrained noble gases, and radionuclides with less than an 8-day half-life. The analysis shall include at least Ce-144, Cs-134, Cs-137, Co-58 and Co-60,in microCuries/ gram. Estimates of the Sr-89 and ,

Sr-90 batch concentration shall be included based on the most i recent monthly composite analysis (within 3 months);

Wn = total weight of resin in the storage tank in grams (determined from chemistry logs procedures); and, '

Vr = total volume of resin water mixture in storage tank to be transferred to the chemical treatment ponds in milliliters.

The batch limits provide assurance that activity input to the CTP will be minimized, and a means of identifying radioactive material in the inventory limitation of this commitment.

O Chapter 16.11-5 Page 3 of 3 08/01/94

l

( 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-6 GASEOUS EFFLUENTS l DOSE RATE COMMITMENT: ;

i The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figure 16.11-1) shall be limited to the following:

a. For noble gases: Less than or equal to 500 mrem /yr to the whole body ar.d less than or equal to 3000 mrem /yr to the skin; and,

b. For lodine-131, for lodine-133, for tritium, and for all radionuclides in particulate form with half-lives greater than 8 days: Less than or equal to 1500 mrem /yr to any organ.

APPLICABillTY:

At all times.

REMEDIAL ACTION:

With the dose rate (s) exceeding the above limits, immediately restore the release rate to within the above limit (s).

TESTING REQUIREMENTS:

The dose rate due to noble gases in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM.

Tlie dose rate due to lodine-131, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents shall be determined to be within the above limits in accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with the sampling and analysis program specified in Table 16.11-4.

Chapter 16.11-6 Page 1 of 7 01/16/99

l O air a ac==:

i

1. Catawba Offsite Dose Calculation Manual '

L 2. 10 CFR Part 20, Appendix B [

3. 10 CFR Part 20 -

BASES:

The basic requirements for Selected Licensee Commitments concoming effluents ,

from nuclear power reactors are stated in 10 CFR 50.36a. These requirements ;

indicate that compliance with effluent Selected Licensee Commitments will keep >

average annual releases of radioactive material in effluents to small percentages of ,

the limits specified in the old 10 CFR 20.106 (new 10 CFR 20.1301). These ,

requirements further indicate that operational flexibility is allowed, compatible with [

considerations of health and safety, which may temporarily result in releases higher i

than such small percentages, but still within the limits specified in the old 10 CFR 20.106 which references Appendix B, Table ll concentrations (MPCs).- These !'

referenced concentrations are specific values which relate to an annual dose of 500 mrems, it is further indicated in 10 CFR 50.36a that when using operational flexibility, best efforts shall be exerted to keep levels of radioactive materials in l O effluents as low as is reasonably achievable (ALARA) as set forth in 10 CFR 50, Appendix 1.

As stated in the_ Introduction to Appendix B of the new 10 CFR 20, the gaseous !

effluent concentration (EC) limits given in Appendix B, Table 2, Column 1, are based i on an annual dose of 50 mrems for isotopes for which inhalation or ingestion is '

limiting or 100 mrems for isotopes for which submersion (noble gases) is limiting.

Since release concentrations corresponding to limiting dose rates less than or equal ;

to 500 mrems/ year to the whole body,3000 mrems/ year to the skin from noble i 7 gases, and 1500 mrems/ year to any organ from lodine-131, lodine-133, tritium and E all radionuclides in particulate form with half-lives greater than eight days at the site boundary has been acceptable as a SLC limit for gaseous eff!uents to assure that l the limits of 10 CFR 50, Appendix l and 40 CFR 190 are not likely to be exceeded, it i should not be necessary to restrict the operational flexibility by incorporating the dose rate associated with the EC value for isotopes based on inhalation / ingestion (50 mrems/ year) or the dose rate associated with the EC value for isotopes based on submersion (100 mroms/ year).

Having sufficient operational flexibility is especially ! nportant in establishing a basis for effluent monitor setpoint calculations. As discussed above, the concentrations stated in the new 10 CFR 20, Appendix B, Table 2, Column 1, relate to a dose of 50 or 100 mrems in a year. When applied on an instantaneous basis, this corresponds to a dose rate of 50 or 100 mrems/ year.

Chapter 16.114 Page 2 of 7 01/16/99

1 I

BASES (con't) 1 These low values are impractical upon which to base effluent monitor setpoint '

calculations for many gaseous effluent release situations when monitor background, monitor sensitivity, and monitor performance must be taken into account.

Therefore, to accommodate operational flexibility needed for effluent releases. the limits associated with gaseous release rate SLCs will be maintained at the current l instantaneous dose rate limit for noble gases of 500 mrems/ year to the whole body l and 3000 mrems/ year to the skin; and for lodine-131, for lodine-133, for tritium, and i for all radionuclides in particulate form with half-lives greater than 8 dLys, an instantaneous dose rate limit of 1500 mrems/ year to any organ.

1 Compliance with the limits of the new 10 CFR 20.1301 will be demonstrated by operating within the limits of 10 CFR 50, Appendix 1 and 40 CFR 190. Operational !

history at Catawba has demonstrated that the use of the dose rate values listed ,

above (i.e., 500 mrems/ year,3000 mrems/ year, and 1500 mrems/ year) as SLC l limits has resulted in calculated maximum individual doses to MEMBERS OF THE 1 PUBLIC that are small percentages of the limits of 10 CFR 50, Appendix I and 40 l CFR 190. For MEMBERS OF THE PUBLIC who may at times be within the SITE BOUNDARY, the occupancy of that MEMBER OF THE PUBLIC will usually be sufficiently low to compensate for any increase in the atmospheric diffusion factor O above that for the SITE BOUNDARY. Examples of calculations for such MEMBERS OF THE PUBLIC, with the appropriate occupancy factors, shall be given in the ODCM. The specified release rate limits restrict, at all times, the corresponding gamma and beta dose rates above background to a MEMBER OF THE PUBLIC at or beyond the SITE BOUNDARY to less than or equal to 500 mrem / year to the whole body or to less than or equal to 3000 mrem / year to the skin. These release I rate limits also restrict, at all times, the corresponding thyroid dose rate above background to a child via the inhalation pathway to less than or equal to 1500 ;

mrem / year.

)

This commitment applies to the release of radioactive materials in gaseous effluents from all units at the site. i The required detection capabilities for radioactive material in gaseous waste samples are tabulated in terms of the lower limits of detection (LLDs). Detailed discussion of the LLD, and other detection limits can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry", Anal. Chem. 40, 586-93 (1968), and Hartwell, J. K., " Detection Limits for Radioanalytical Counting Techniques", Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

O Chapter 16.11-6 Page 3 of 7 01/16/99

O O O TABLE 16.11-4 (Pace 1 of 4)

RADIOACTIVE GASEOUS WASTE SAMPUNG AND ANALYSIS PROGRAM Gaseous Release Type Sampling Minimum Type of Activity Analysis Lower Limit of Frequency Analysis Detection (LLD)"I Frequency ( Ci/ml)

1. Waste Gas Storage Tank P P Principal Gamma Emitters'"' 1x1C*

Each Tank Each Tank Grab Sample

2. Containment Purge P P Each PURGEW Each PURGEW Principal Gamma Emitters" 1x10" Grab Sample M H-3 (oxide) 1x10*

3. Unit Vent W'"'"' W W Principal Gamma Emitters'#' 1x10*

Grab Sample H-3 (oxide) 1x10* ,

4. Containment Air Release and D** D** Principal Gamma Emitters

1x10*

Addition System Grab Sample M H-3 (oxide) 1x10*

5. All Release Types as listed in 3 Continuous

D l-131 1x10'"

above. Charcoal Sample I-133 1x10*

Continuous" DW Principal Gamma Emitters

1x10 Particulate Sample Continuous" M Gross Alpha" 1 x10-"

Composite Particulate Sample Continuous" Q Sr-89, Sr-90 1 x10-"

Composite Particulate Sample Chapter 16.11-6 Page 4 of 7 01/16/99

l TABLE 16.11-4 (Pace 2 of 4)

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Gaseous Release Type Sampling Minimum Type of Activity Analysis Lower Limit of j Frequency Analysis Detection (LLD)m Frequency ( Ci/ml)

6. Waste Monitor Tank Building W W Principal Gamma Emitters W 1x10+

Ventilation Exhaust Grab Sample i

H-3 (oxide) 1x10*

Continuous *' W l-131 1 x10'"

Charcoal Sample l-133 1x10'"

Continuous *' W Principal Gamma Emitters

1x10'*

Particulate Sample ,

Continuous *' M Gross Alpha 1 x10'"

Composite Particulate Sample Continuous *' O Sr-89, Sr-90 1 x10'"

Composite Particulate Sample Chapter 16.11-6 Page 5 of 7 01/16/99

TABLE 16.11-4 (Pace 3 of 4)

TABLE NOTATIONS (1) THE LLD is defined, for purposes of these commitments, as the smallest concentration of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system, which may include radiochemical separation:

LLD =

E V 2.22 x10' Y exp(-Abt)

Where:

LLD =

the "a priori" lower limit of detection (microcurie per unit mass or volume);

so =

the standard deviation of the background counting rate or of the O counting rate of a blank sample as appropriate (counts per minute);

E =

the counting efficiency (counts per disintegration);

V =

the sample size (units of mass or volume);

6 2.22 x 10 = the number of disintegrations per mir ute per microcurie; Y =

the fractional radiochemical yield, when applicable; A =

the radioactive decay constant for the particular radionuclide (sec"); and At =

the elapsed time between midpoint of sample collection and time of counting (sec).

Typical values of E, V, Y and At shall be used in the calculation.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an a posteriori (after the fact) limit for a particular measurement.

Chapter 16.114 Page 6 of 7 01/16/99

r TABLE 16.11-4 (Paae 4 of 4) !

i TABLE NOTATIONS (Continued) l (2) '

The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-133, Xe-133m, Xe-135, and Xe- '

138 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, I-131, Cs-134, Cs-137, and Ce-141 in lodine and particulate releases. The LLD for Ce-144 is 5x10* pCi/ml. This list does not mean that only these ,

nuclides are to be considered. Other gamma peaks that are identifiable, ;

together with those of the above nuclides, shall also be analyzed and i reported in the Radioactive Effluent Release Report, pursuant to Technical Specification 5.6.3, in the format outlined in Regulatory Guide 1.21, Appendix B, Revision 1, June 1974. ;

(3)' Sampling and analysis shall also be performed following shutdown, staitup, or a THERMAL POWER stabilization (power level constant at desired power J level) after a THERMAL POWER change exceedin'g 15% of RATED i THERMAL POWER within a 1-hour period, for at least one of the three gaseous release types with this notation.

[

(4) Tritium grab samples shall be taken at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the O- refueling canalis flooded.

1 (5) Required sampling and analysis frequency during effluent release via this pathway.

(6) The ratio of the sample flow volume to the sampled stream flow volume shall !

be known for the time period covered by each dose or dose rate calculation -

made in accordance with SLCs 16.11-6,16.11-8, and 16.11-9. l (7)' Samples shall be changed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and analyses shall be completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after changing, or after removal from sampler.

(8) The composite filter (s) will be analyzed for alpha activity by analyzing one filter per week to ensure that at least four filters are analyzed per collection period.

O Chapter 16.11-6 Page 7 of 7 01/16/99

= -. - ..

16.11 RADIOLOGICAL EFFLUENTS CONTROLS '

l 16.11-7 RADIOACTIVE GASEOUS EFFLUENT MONITORING

! INSTRUMENTADON i COMMITMENT The radioactive gaseous effluent monitoring instrumentation channels shown in Table 16.11-5 shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of SLC 16.11-6 are not exceeded. The Alarm / Trip Setpoints of these i channels meeting SLC 16.11-6 shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

APPLICABILITY:

As shown in Table 16.11-5.

REMEDIAL ACTION:

a. With a radioactive gaseous effluent monitoring instrumentation channel g Alarm / Trip Setpoint less conservative than required by the above specification, immediately suspend the release of radioactive gaseous effluents monitored by the affected channel, or declare the channel inoperable,

b. With less than the minimum number of radioactive gaseous effluent monitoring instrumentation channels OPERABLE, take the ACTION shown in Table 16.11-5. Restore the inoperable instrumentation to OPERABLE status within the time specified in the ACTION, or explain in the next Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3 why this inoperability was not corrected within the time specified.

TESTING REQUIREMENTS:

Each radioactive gaseous effluent monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK, SOURCE CHECK, CHANNEL CAllBRATION, and CHANNEL OPERATIONAL TEST l operations at the frequencies shown in Table 16.11-6.

REFERENCES:

1. Catawba Offsite Dose Calculation Manual i

2. 10 CFR Part 20 f

Chapter 16.11-7 Page 1 of 9 01/16/99

l l

BASES:

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 of gaseous effluents. The AlamVTrip Setpoints for these instruments shall be calculated and adjusted in accordance with the methodology and parameters in the ODCM to ensure that the alarm / trip will occur prior to exceeding the limits of 10 CFR Part 20. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, l 63, and 64 of Appendix A to 10 CFR Part 50. The sensitivity of any noble gas 1 activity monitor used to show compliance with the gaseous effluent release requirements of SLC 16.11-8 shall be such that concentrations as low as 1 x 10-6 i pCi/cc are measurable.

O l

i 1

1 O

Chapter 16.11-7 Page 2 of 9 01/16/99

O O O TABLE 16.11-5 (Pace 1 of 4)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION

. Minimum Channels Instrument Operable Applicability Action t

1. Waste Gas Holdup System

a. Noble Gas Activity Monitor- Providing Alarm and Automatic Termination of i Release (Low Range - EMF-50) 1 per station C l

b. Effluent System Flow Rate Measuring Device 1 per station D l

2. Condenser Evacuation System Noble 1 1,2,3,4,# H l' ,

Gas Activity Monitor (Low Range - EMF-33) 'l i

3. Vent System

a. Noble Gas Activity Monitor 1 E l (Low Range - EMF-36)

b. lodine Sampler (EMF-37) 1 G l

c. Particulate Sampler (EMF-35) 1 G l

d. Flow Rate Monitor 1 D l

e. Sampler Flow Rate Monitor 1 D l I

Chapter 16.11-7 Page 3 of 9 01/16/99

O O O l

TABLE 16.11-5 (Pace 2 of 4) l RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION Minimum Channels Instrument Operable Applicability Action

4. Containment Purge System Noble Gas Activity Monitor- Providing Alarm and Automatic Termination of 1 F

Release (Low Range- EMF-39) l

5. Containment Air Release and Addition System Noble Gas Activity Monitor- Providing Alarm 1 C

(Low Range - EMF-39) l

6. Monitor Tank Building HVAC

a. Noble Gas Activity Monitor- 1 per station **

E l Providing Alarm (EMF-58)

b. Monitor Tank Building 1 per station "

D l Effluent Flow Rate Measuring Device Chapter 16.11-7 Page 4 of 9 01/16/99

-- -)

l TABLE 16.11-5 (Paae 3 of 4) 1 TABLE NOTATIONS i

At all times except when the isolation valve is closed and locked.

l At all times.

i

Apply Action Hb in Modes 5 and 6 l

l ACTION STATEMENTS ACTION C - With the number of channels OPERABLE less than required the l Minimum Channels OPERABLE requirement, the contents of the tank (s) may be released to the environment for up to 14 days provided that prior to initiating the release either:

a. Vent system noble gas activity monitor providing alarm and automatic termination of release (Low Range - EMF 36) has at ;

least one channel OPERABLE; or, i O !

O b. At least two independent samples of the tank's contents are l at:alyzed, and at least two technically qualified members of the facility staff independently verify:

1. The discharge valve lineup; and

2. The manual portion of the computer input for the release rate calculations performed on the computer, or the entire release rate calculations if such calculations are performed manually.

Otherwise, suspend release of radioactive effluents via this pathway.

ACTION D - With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided the flow rate is estimated at least once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

ACTION E - With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement, effluent releases via this pathway may continue for up to 30 days provided grab samples Chapter 16.11-7 Page 5 of 9 01/16/99

TABLE 16.11-5 (Pace 4 of 4)

TABLE NOTATIONS are taken at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

ACTION F- With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement, immediately suspend PURGING of radioactive effluents via this pathway.

ACTION G - With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement, effluent releases via the affected pathway may continue for up to 30 days provided samples are continuously collected with auxiliary sampling equipment as required in Table 16.11-4.

ACTION H - With the number of channels OPERABLE less than required by the l Minimum Channels OPERABLE requirement:

a. Effluent release via the CSAE System (ZJ) may continue for up l to 30 days provided grab samples are taken at least once per k 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and these samples are analyzed for radioactivity within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and

b. Gaseous effluent releases via the BB system atmospheric vent l valve (BB27) in the off normal mode may continue for up to 30 days provided grab samples of steam generator water are analyzed for radioactivity at a lower limit of detection of no more than 1E-7 microcurie /ml:

1. At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the specific activity of the l secondary coolant is greater than 0.01: microcurie / gram DOSE EQUIVALENT I-131, or

2. At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months when the specific activity of the l secondary coolant is less than or equal to 0.01 microcurie / gram DOSE EQUIVALENT l-131.

O Chapter 16.11-7 Page 6 of 9 01/16/59

O O O '

TABLE 16.11-6 (Pace 1 of 3)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS :

Modes For ;

Channel Which Channel Source Channel Operational Surveillance ,

instrument Check Check Calibration Test is Required

1. Waste Gas Holdup System

a. Noble Gas Activity Monitor- Providing Alarm and Automatic Termination of ;

Release (Low Range - EMF-50) P P(4) R(3) O(1)

b. Effluent System Flow Rate Measuring P N.A. R N.A.

Device

2. Condenser Evacuation System Noble Gas Activity Monitor '

(Low Range - EMF-33) D M(4) R(3) O(1) 1,2,3,4

3. Vent System

a. Noble Gas Activity Monitor (Low Range - EMF-36) D M(4) R(3) Q(2)

b. lodine Sampler (EMF-37) W N.A. N.A. N.A. *

c. Particulate Sampler (EMF-35) W N.A. N.A. N.A. *

d. Flow Rate Monitor D N.A. R N.A.

Chapter 16.11-7 Page 7 of 9 01/16/99

O O O TABLE 16.11-6 (Paae 2 of 3)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS Modes For Channel Wh!ch ,

Channel Source Channel Operational Survrillr nce Instrument Check Check Calibration Test is Rs,1uired

e. Sampler Flow Rate Monitor D N.A. R N.A. *

4. Containment Purge System Noble Gas activity Monitor- Providing Alarm and Automatic Termination of Re! ease (Low Range- EMF-39) D P(4) R(3) O(1)

5. Containment Air Release and Addition System Noble Gas Activity Monitor-Providing Alarm (Low Range - EMF-39) D P(4) R(3) O(1)

6. Monitor Tank Building HVAC

a. Noble Gas Activity Monitor-Providing Alarm (EMF-58) D M R(3) O(2)

b. Discharge Flowinstrumentation D N.A. R N.A. "

Chapter 16.11-7 Page 8 of 9 01/16/99

O D TABLE 16.116 (Pace 3 of 3)

TABLE NOTATIONS At all times except v, hen the isolation valve is closed and locked.

At all times.

1.

The CHANNEL OPERATIONAL TEST shall also demonstrate that automatic l isolation of this pathway and control room alarm annunciation occur if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm / Trip Setpoint; or,

b. Circuit failure (Alarm only); or,

c. Instrument indicates a downscale failure (Narm only).

2. The CHANNEL OPERATIONAL TEST shall also demonstrate that control l room alarm annunciation" occurs if any of the following conditions exists:

a. Instrument indicates measured levels above the Alarm Setpoint; or,

b. Circuit failure; or,

c. Instrument indicates a downscale failure.

3. The initial CHANNEL CAllBRATION shall be performed using one or more of the reference standards certified by the National Bureau of Standards (NBS) or using standards that have been obtained from suppliers that participate in measurement assurance activities with NBS. These standards shall permit calibrating the system over its intended range of energy and measurement range. For subsequent CHANNEL CALIBRATION, sources that have been related to the initial calibration shall be used.

4. A source check for these channels shall be the qualitative assessment of channel response when the channel sensor is exposed to a light emitting diode.

For EMF-58, the alarm annunciation is in the Monitor Tank Building Control Room and on the MTB Control Panel Remote Annunciator Panel.

Chapter 16.11-7 Page 9 of 9 01/16/99

[ 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11-8 DOSE - NOBLE GASES COMMITMENT:

The air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 16.11-1) shall be limited to the following:

a. During any calendar quar 1er: Less than or equal to 5 mrads for gamma radiation and less than or equal to 10 mrad for beta radiation; and,

b. During any calendar year: Less than or equal to 10 mrad for gamma radiation and less than or equal to 20 mrad for beta radiation.

APPLICABILITY:

At all times, i REMEDIAL ACTION:

With the calculated air dose from radioactive noble gases in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report that identifies the cause(s) for exceeding the limit (s) and l defines the corrective actions that have been taken to reduce the releases and the proposed corrective actions to be taken to assure that subsequent releases will be in compliance with the above limits.

TESTING REQUIREMENTS:

Cumulative dose contributions for the current calendar quarter and current calendar year for noble gases shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

REFERENCES:

1. Catawba Offsite Dose Calculation Manual

2. 10 CFR Part 50, Appendix l i

Chapter 16.11-8 Page 1 of 2 01/16/99

BASES:

This commitment is provided to implement the requirements of Sections 11.8, Ill.A and IV.A of Appendix 1,10 CFR Par 150. The Limiting Condition for Operation implements the guides set forth in Section ll.B of Appendix 1. The REMEDIAL ACTION statement provides the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix l to assure that the releases of radioactive materialin gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable". The Testing Requirements implement the requirements in Section Ill.A of Appendix 1 that conformance with the guides of Appendix I be shown by calculational procedures based on models and data such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate pathways is unlikely to be substantially underestimated. The dose calculation methodology and parameters established in the ODCM for calculating the doses due to the actual release rates of radioactive noble gases in gaseous effluents are consistent with the methodology provided in Regulatory Guide 1.109,

" Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for J the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1 1, October 1977 and Regulatory Guide 1.111, " Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water Cooled Reactors", Revision 1, July 1977. The ODCM equations provided for i

determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditMns.

, This commitment applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared 4

treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions frorr, each unit based I

, on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each of the i radioactives waste producing units sharing the Radwaste Treatment System. For determining conformance to commitments, these allocations from shared Radwaste

, Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

l l

Chapter 16.11-8 Page 2 of 2 01/16/99

16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11-9 DOSE -IODINE-131. IODINE-133. TRITlUM. AND RADIOACTIVE MATERIAL IN PARTICULATE FORM COMMITMENT:

The dose to a MEMBER OF THE PUBLIC from lodine-131, lodine-133, tritium, and all radionuclides in particulate form with half-lives greater than 8 days in gaseous effluents released, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 16.11-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 7.5 mrem to any organ; and,

b. During any calendar year: Less than or equal to 15 mrem to any organ.

APPLICARILITY:

At all times.

( REMEDIAL ACTION:

With the calculated dose from the release of lodine-131, lodine-133, tritium, and radionuclides in particulate fomt with half-lives greater than 8 days, in gaseous effluents exceeding any of the above limits, prepare and submit to the Commission within 30 days a Special Report that identifies the cause(s) for exceeding the limit l and defines the corrective actions that have been taken to reduce the releases and ,

the proposed corrective actions to be taken to assure that subsequent releases will ;

be in compliance with the above limits. i TESTING REQUIREMENTS:

Cumulative dose contributions for the current calendar quarter and current calendar year for lodine-131, lodine-133, tritium and radionuclides in particulate form with half-lives greater than 8 days shall be determined in accordance with the methodology and parameters in the ODCM at least once per 31 days.

REFERENCES:

1. Catawba Offsite Dose Calculation Manual

2. 10 CFR Part 50, Appendix l I

Chapter 16.11-9 Page 1 of 2 01/16/99 1 I

p)

( BASES:

This commitment is provided to implement the requirements of Sections 11.C, Ill.A and IV.A of Appendix 1,10 CFR Part 50, and are the guides set forth in Section ll.C of Appendix 1. The REMEDIAL ACTION statements provide the required operating flexibility and at the same time implement the guides set forth in Section IV.A of Appendix 1 to assure that the releases of radioactive materials in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable". The ODCM calculational methods specified in the Surveillance Requirements implement the reqeirements in Section Ill.A of Appendix 1 that conformance with the guides of l Appendix i be chown by calculational procedures based on models and data, such that the actual exposure of a MEMBER OF THE PUBLIC through appropriate :

pathways is unlikely to be substantially underestimated. The ODCM calculational methodology and parameters for calculating the doses due to the actual release rates of the subject materials are consistent with the methodology provided in Regulatory Guide 1.109," Calculation of Annual Doses to Man from Routine Releases of Reactor Effluents for the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I", Revision 1, October 1977 and Regulatory Guide 1.111, ;

" Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors", Revision 1, July 1977. These equations also provide for determining the actual doses based upon the historical average atmospheric conditions. The release rate commitments for

/7 lodine-131, lodine-133, tritium, and radionuclides in particulate form with half-lives O greater than 8 days are dependent upon the existing radionuclide pathways to man in the areas at and beyond the SITE BOUNDARY. The pathways that were examined in the development of the calculations were: (1)individualinhalation of airbome radionuclides, (2) deposition of radionuclides onto green leafy vegetation with subsequent consumption by man, (3) deposition onto grassy areas where milk animals and meat-producing animals graze with consumption of the milk and meat by man, and (4) deposition on the ground with subsequent exposure of man.

i This commitment applies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared i treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases may be allocated equally to each .-J the radioactive waste producing units sharing the Radwaste Treatment Sycam. For determining conformance to commitments, these allocations from shared Rs iwaste Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

/3 l V '

l Chapter 16.11-9 Page 2 of 2 01/16/99 i

i 16.11 }]!.;NOLOGICAL EFFLUENT CONTROLS 16.11-10 Gk..EOUS RADWASTE TREATMENT SYSTEM l COMMITMEN1, The VENTILATION EMMAUST TREATMENT SYSTEM and the WASTE GAS HOLDUP SYST EU, shall be OPERABLE arid appropriate portions of these systems shall be used to re ' n releases of radioactivity when the projected doses in 31 days due to gaMo t efiluent releases, from each unit, to areas at and beyond the SITE BOUNDAFi/ pe Figure 16.11-1) would exceed either:

a. 0.2 mrad to air from gan:ma radiation; or, !

b. 0.4 mrad to air from beta radiation; or,

c. 0.3 mrem to any orgar of a MEMBER OF THE PUBLIC. !

APPLICABILITY:

At all times.

O REMEDIAL ACTION:

With radioactive gaseous waL being discharged without treatment and in excess l of the above limits, prepare and submit to the Commission within 30 days a Special l .

Report that includes the following information: i

1. Identification of any inon arable equipme nt or subsystems, and the reason for the inoperability;

2. Action (s) taken to restore the inoperable equipment to OPERABLE status; i and, ;

3. Summary description of action (s) taken to prevent a recurrence.

TESTING REQUIREMENTS:

Doses due to gaseous releases from each unit to areas at and beyond the SITE BOUNDARY shall be projected at least once per 31 days in accordance with the methodology and parameters in the ODCM when Gaseous Radwaste Treatment l Systems are not being fully utilized. ;

O Chapter 16.11-10 Page 1 of 2 01/16/99

TESTING REQUIREMENTS (con't)

The installed VENTILATION EXHAUST TREATMENT SYSTEM and WASTE GAS HOLDUP SYSTEM shall be considered OPERABLE by meeting SLCs 16.11-6, 16.11-8, or 16.11-9.

REFERENCES:

1. Catawba Offsite Dose Calculation Mcnual

2. 10 CFR Part 50, Appendix I BASES:

The OPERABILITY of the WASTE GAS HOLDUP SYSTEM and the VENTILATION EXHAUST TREATMENT SYSTEM ensures that the systems will be available for use whenever gaseous effluents require treatment prior to release to the environment. The requirement that the appropriate portions of these systems be used, when specified, provides reasonable assurance that the releases of radioactiva materials in go.seous effluents will be kept "as low as is reasonably achievable". This comrnitment implements the requirements of 10 CFR 50.36a, ,

General Design Criterion 60 of Appendix A to 10 CFR Part 50, and the design i objectives given in Section ll.D of Appendix ! to 10 CFR Part 50. The specified -

limits goveming the use of appropriate portions of the systems were specified as a ,

suitable fraction of the dose design objectives set forth in Sections ll.B and ll.C of '

Appendix 1,10 CFR Part 50, for gaseous effluents.

This commitment appiies to the release of radioactive materials in gaseous effluents from each unit at the site. When shared Radwaste Treatment Systems are used by more than one unit on a site, the wastes from all units are mixed for shared treatment; by such mixing, the effluent releases cannot accurately be ascribed to a specific unit. An estimate should be made of the contributions from each unit based on input conditions, e.g., flow rates and radioactivity concentrations, or, if not practicable, the treated effluent releases mcy be allocated equally to each of the radioactive waste producing units sharing the Radwaste Treatment System. For determining conformance to commitments, these allocations from shared Radwaste .

Treatment Systems are to be added to the releases specifically attributed to each unit to obtain the total releases per unit.

l O ;

Chapter 16.11-10 Page 2 of 2 01/16/99

16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11-11 SOLID RADIOACTIVE WASTES l

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

APPLICABILITY:

i At all times.

1 REMEDIAL ACTION: ,

a. With SOLIDIFICATION or dewatering not meeting disposal site and shipping and transportation requirements, suspend shipment of the inadequately processed wastes and correct the PROCESS CONTROL PROGPAM, the procedures and/or the Solid Radwaste System as necessary b g ' vent O recurrence. ;

b. With SOLIDIFICATION or dewatering not performed in accordance o a the PROCESS CONTROL PROGRAM, test the improperly processed waste in each container to ensure that it meets burial ground and shipping requirements and take appropriate administrative action to prevent recurrence.

l TESTING REQUIREMENTS:

SOLIDIFICATION of at least one representative test specimen from at least every tenth batch of each type of wet radioactive wastes (e.g., filter sludges, spent resins, evaporator bottoms, boric acid solutions and sodium sulfate solutions) shall be verified in accordance with the PROCESS CONTROL PROGRAM:

a. If any test specimen fails to verify SOLIDIFICATION, the SOLIDIFICATION of the batch under test shall be suspended until such time as additional test specimens can be obtained, attemative SOLIDIFICATION parameters can be determined in accordance with the PROCESS CONTROL PROGRAM, and a subsequent test verifies SOLIDIFICATION. SOLIDIFICATION of the batch may then be resumed O

Chapter 16.11-11 Page 1 of 2 01/16/99

(

TESTING REQUIREMENTS (con't) using the attemative SOLIDIFICATION parameters determined by the PROCESS CONTROL PROGRAM;

b. If the initial test specimen from a batch of waste fails to verify SOLIDIFICATION, the PROCESS CONTROL PROGRAM shall provide for the collection and testing of representative test specimens from each consecutive batch of the same type of wet waste until at least three consecutive initial test specimens demonstrate SOLIDIFICATION. The PROCESS CONTROL PROGRAM shall be modified as required to assure l SOLIDIFICAh0N of subsequent batches of waste; and,

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

REFERENCES:

1. Catawba Offsite Dose Calculation Manual

2. 10 CFR Part 50, Appendix A

3. 10 CFR Part 50 BASES:

This commitment implements the requirements of 10 CFR 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 pH, waste / liquid / SOLIDIFICATION agent / catalyst ratios, waste oil content, waste principal chemical constituents, and mixing and curing times.

O Chapter 16.11-11 Page 2 of 2 01/16/99

16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11-12 TOTAL DOSE COMMITMENT:

The annual (calendar year) dose or dose commitment to any MEMBER OF THE PUBLIC due to releases of radioactivity and to radiation from uranium fuel cycle sources shall be limited to less than or equal to 25 mrem to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

With the calculated doses from the release of radioactive materials in liquid or gaseous effluents exceeding twice the limits of SLCs 16.11-Sa,16.11-3b,16.11-8a, 16.11-8b,16.11-9a, or 16.11-9b, calculations shall be made including direct radiation contributions from the units and from outside storage tanks to determine whether the above limits of this commitment have been exceeded. If such is the case, prepare and submit to the Commission within 30 days a Special Report that l defines the corrective action to be taken to reduce subsequent releases to prevent recurrence of exceeding the above limits and includes the schedule for achieving conformance with the above limits. This Special Report, as defined in 10 CFR t 20.405c, shall include an analysis that estimates the radiation exposure (dose) to a MEMBER OF THE PUBLIC from uranium fuel cycle sources, including all effluent pathways and direct radiation, for the calendar year that includes the release (s) covered by this report. It shall also describe levels of radiation and concentrations of radioactive materialinvolved, and the cause of the exposure levels or concentrations. If the estimated dose (s) exceeds the we limits, and if the release condition resulting in violation of 40 CFR Part 190 has ,

iiready been corrected, the Special Report shall include a request for a varianc o accordance with the provisions of 40 CFR Part 190. Submittal of the report is considered a timely 1 request, and a variance is granted until staff action on the request is complete.

TESTING REQUIREMENTS:

1 Cumulative dose contributions from liquid and gaseous effluents shall be !

determined in accordance with SLCs 16.11-3,16.11-8 and 16.11-9, and in l accordance with the methodology and parameters in the ODCM. I Ci V

l Chapter 16.11-12 Page 1 of 2 01/16/99 i

i I

l i- I t

TESTING REQUIREMENTS: (cont'd) l

.,' Cumulative ' dose contributions from direct radiation from the units and from 1 1 radwaste storage tanks shall be determined in accordance with the methodology i and parameters in the ODCM. This requirement is applicable only under conditions '

} set forth in the REMEDIAL ACTION of this commitment. l l

REFERENCES:

I r 1. Catawba Offsite Dose Calculation Manual ,

j 2. 10 CFR Part 20

)

3. 40 CFR Part 190 i

L BASES; This commitment is provided to meet the dose limitations of 40 CFR Part 190 that j have been incorporated into 10 CFR Part 20 by 46 FR 18525. .The commitment i . requires the preparation and submittal of a Special Report whenever the calculated I doses due to releases of radioactivity and to radiation from uranium fuel cycle i

sources exceed 25 mrem to the whole body or any organ, except the thyroid, which shall be limited to less than or equal to 75 mrem.

i For sites containing up to four reactors, it is highly unlikely that the resultant dose to

] a MEMBER OF THE PUBLIC will exceed the dose limits of 40 CFR Part 190 if the

[ individual reactors remain within twice the dose design objectives of Appendix 1, and l if direct radiation doses from the units and from outside storage tanks are kept small. The Special Report will describe a course of action that should result in the limitation of the annual dose to a MEMBER OF THE PUBLIC to within the 40 CFR j Part 190 limits. For the purposes of the Special Report, it may be assumed that the o dose commitment to the MEMBER OF THE PUBLIC from other uranium fuel cycle sources is negligible, with the exception that dose contributions from other nuclear

fuel cycle facilities at the same site or within a radius of 8 km must be considered. If l the dose to any MEMBER OF THE PUBLIC is estimated to exceed the requirements

- of 40 CFR Part 190, the Special Report with a request for a variance (provided the

release conditions resulting in violation of 40 CFR Part 190 have not already been !

l corrected), in accordance with the provisions of 40 CFR 190.11 and 10 CFR !

i

~

20.2203(a)(4), is considered to be a timely request and fulfills the requirements of 40 CFR part 190 until NRC staff action is completed. The variance only relates to ,

j' the limits of 40 CFR Part 190, and does not apply in any way to the other i requirements for' dose limitation of 10 CFR Part 20, as addressed in SLC 16.11-1 l 3 and 16.11-6. An individual is not considered a MEMBER OF THE PUBLIC during 1

any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

- Chapter 16.11-12 Page 2 of 2 01/16/99 i

16.11 RAIOLOGICAL EFFLUENT CONTROLS ,

16.11-13 MONITORING PROGRAM COMMITMENT:

The Radiological Environmental Monitoring Program shall be conducted as specified in Table 16.11-7.

l APPLICABILITY: l At all times. I REMEDIAL ACTION:

a. With the Radiological Environmental Monitoring Program not being conducted as specified in Table 16.11-7, prepare and submit to the Commission, in the Annual Radiological Environmental Operating Report required by Technical Specification 5.6.2, a description of the reasons for not conducting the program as required l and the plans for preventing a recurrence.

b. With the level of radioactivity as the result of plant effluents in an environmental sampling medium at a specified location exceeding the repe%g levels of Table 16.11-7 when averaged over any calendar quarter, prepare tsd submit to the Commission within 30 days a Special Report that identifies the cause(s) for l exceeding the limit (s) and defines the corrective actions to be taken to reduce radioactive effluents so that the potential annual dose to a MEMBER OF THE PUBLIC is less than the calendar year limits of SLC 16.11-3,16.11-8, and 16.11-

9. When more than one of the radionuclides in Table 16.11-7 are detected in the sampling medium, this report shall he cabmitted if:

concentration (1) c2ncertation (2) reporting level (1) reporting level (2) + .. 21.0 When radionuclides other than those in Table 16.11-7 are detected and are the result of plant effluents, this report shall be submitted if the potential annual dose

to a MEMBER OF THE PUBLIC from all radionuclides is equal to or greater than the calendar yearlimits of SLC 16.11-3,16.11-8 and 16.11-9. This report is not required if the measured level of radioactivity was not the I

b)

The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shan be indicated in this report.

1 Chapter 16.11-13 Page 1 of 14 01/16/99 l

REMEDIAL ACTION (con t) result of plant effluents; however, in such an event, ine condition shall be

reported and described in the Annual Radiological Environmenta! Operating

Report required by Technical Specification 5.6.2.

l

c. With milk or fresh leafy vegetation samples unavailable from one or more of the sample locations required by Table 16.11-7, identify specific locations for 4

obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. The specific focations from which samples were unavailable may then be deleted from the monitoring program. Pursuant to Technical Specification 5.5.1, submit in the next l l'

Radioactive Effluent Release Report documentation for a change in the ODCM including a revised figure (s) and table for the ODCM reflecting the new location (s) with supporting information identifying the cause of the unavailability of samples and justifying the selection of the new location (s) for obtaining samples.

JE, STING REQUIREMENTS:

'i ne radiological environmental monitoring samples shall be collected pursuant to Table 16.11-7 from the specific locations given in the table and figure (s) in the ODCM, and shall be analyzed pursuant to the requirements of Table 16.11-7 and the detection capabilities required by Table 16.11-8.

REFERENCES:

1. Catawba Offsite Dose Calculation Manual

2. 10 CFR Part 50 Appendix l

^

BASES:

The Radiological Environmental Monitoring Program required by this commitment provides representative measurements of radiation and of radioactive materials in those exposure pathways and for those radionuclides that lead to the highest potential radiation exposures of MEMBERS OF THE PUBLIC resulting from the plant operation.

This monitoring program implementsSection IV.B.2 of Appendix l to 10 CFR Part 50 and 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 the modeling of the environmental exposure pathways. Guidance for this monitoring program is provided by the Radiological Assessment Branch Technical Position on Environmental Monitoring.

The initially specified monitoring program will be effective for at least the first 3 years of commercial Chapter 16.11-13 Page 2 of 14 01/16/99

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

_B.A TES (con't) ooeration. Following this period, program changes may be initiated based on operational experience.

The required detection capabilities for environmental sample analyses are tabulated in terms of the lowerlimits of detection (LLDs). The LLDs required by Table 16.11-8 are considered optimum for routine environmental measurements in industrial laboratories.

It should be recognized that the LLD is defined as an a priori (before the fact) limit representing the capability of a measurement system and not as an g costeriori (after the fact) limit for a particular measurement.

Detailed discussion of the LLD, and other detection limits, can be found in HASL Procedures Manual, HASL-300 (revised annually), Currie, L. A., " Limits for Qualitative Detection and Quantitative Determination - Application to Radiochemistry", Anal. Chem.

4Q,586-93 (1968), and Hartwell, J. K.," Detection Limits for Radioanalytical Counting Techniques", Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

O O

Chapter 16.11-13 Page 3 of 14 01/16/99

O O O TABLE 16.11-7 (Pace 1 of 8)

RADIOLOG! CAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF REPRESENTATIVE SAMPLING AND ,

EXPOSURE PATHWAY SAMPLES AND SAMPLE COLLECTION TYPE AND FREQUENCY AND/OR SAMPLE LOCATIONSW FREQUENCY OF ANALYSIS

1. Direct Radiation" Forty routine monitoring stations Quarterly Gamma dose quartety '

either with two or more dosimeters or with one instrument for '

measuring and recording dose rate continuously, placed as follows:

An inner ring of stations, one in each meteorological sectorin the generalarea of the SITE BOUNDARY; An outer ring of stations, one in each meteorological sectorin the '

6- to 8-km range from the site; and. '

The balance of the stations to be placed in specialinterest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations.

Chapter 16.11-13 Page 4 of 14 01/16/99

O O O TABLE 16.11-7 (Pace 2 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM u

NUMBER OF REPRESENTATIVE ' SAMPLING AND EXPOSURE PATHWAY SAMPLES AND SAMPLE COLLECTION TYPE AND FREQUENCY AND/OR SAMPLE LOCATIONSW FREQUENCY OF ANALYSIS

2. Airbome Radioiodine and Samples from five locations. Continuous sarr.pler operation with Radiciodine Canister Particulates sample c ,iiection weekly, or more I-131 analysis weekly.

frequentlyif required by dust Three samples from close to the loading.

three SITE BOUNCARY locations, Particulate Sampler in different sectors, of the highest Gross beta radioactivity calculated annual average ground- analysis following filter level D/O; change;* and gamma isotopic analysisW of One sample from the vicinity of a compor2e (bylocation) community having the highest qum ,&.

calculated annualaverage ground-level D/Q; and One sample from a control ,

location, as for example 15 to 3u km distant and in the least prevalent wind direction.

Chapter 16.11-13 Page 5 of 14 01/16/99

O O O TABLE 16.11-7 (Pane 3 of 8)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NUMBER OF EXPOSURE PATHWAY REPRESENTATIVE SAMPLING AND AND/OR SAMPLE SAMPLES AND SAMPLE COLLECTION TYPE AND FREQUENCY LOCATIONSW FREQUENCY OF ANALYSIS

3. Waterborne

a. Surface" One sample upstream. Composite sample over 1-month Gamma isotopic analysis ") .

One sample downstream. period *. monthly. Composite for tritium analysis quarterly. I

b. Ground Samples from one or two sources Quarterly Gamma isotopic"3 and tritium onlyif likely to be affectedm analysis quarterly.

r

c. Drinking One sample of each of one to Composite sample over 2-week l-131 analysis on each composite three of the nearest water supplies period

when 1-131 analysis is when the dose calculated for the that could be affected byits performed; r.onthly composite consumption of the water is discharge. otherwise. greater than 1 mrem per year". i Composite for gross beta and One sample from a control gamma isotopic analyses "3 location. monthly. Composite for tritium t analysis quarterly.

d. Sediment from Shoreline One sample from downstream Semiannually. Gamma isotopic analysis"3 area with existing or potential semiannually.

recreational value.

Chapter 16.11-13 Page 6 of 14 01/16/99

O O O .

1 TABLE 16.11-7 (Pace 4 of 8) :

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM I NUMBER OF !

REPRESENTATIVE SAMPLING AND EXPOSURE PATHWAY SAMPLES AND SAMPLE COLLECTION TYPE AND FREQUENCY l AND/OR SAMPLE LOCATIONSW FREQUENCY OF ANALYSIS

4. InDestion

a. Milk Samples from milking animals in Semimonthly when animals are on Gamma isotopic") and I-131 ;

three locations within 5-km pasture; monthly at other times. analysis semi-monthly when i distance having the highest dose animals are on pasture; monthly at !

potential. If there are noN, then other times.

one sam $ from mileir/, %simals

[

in each of three areev oetwer.- 5 -

to 8 km distant wheee doses are ,

calculated to be preater than 1 ;

mrem per yeer". One sample '

from milking animals at a control !

location 15 to 30 km distant and in the feast prevalent wind direction. l l

l

b. Fish and Invertebrates One sample each of a predatory Sample in season, or Gamma isotopic analysis"I on ,

species, a bottom feeder and a *emiannuallyif theyare not edible portions.

i forage species in vicinity of plant seaconal. !

discharge area.

[

i One sample each of a predatory >

species, a bottom feeder and a I forage species in areas not influenced by plant discharge.

Chapter 16.11-13 Page 7 of 14 01/16/99 i i

l l

l O O O l .

i l TABLE 16.11-7 (Pace 5 of 8)

( RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM l

NUMBER OF REPRESENTATIVE SAMPLING AND EXPOSURE PATHWAY SAMPLES AND SAMPLE COLLECTION TYPE AND FREQUENCY AND/OR SAMPLE LOCATIONSW FREQUENCY OF ANALYSIS

4. Ingestion (Continued) i

c. Food Products One sample of each principal At time of harvestm. Gamma isotopic analyses M on class of food products from any edible portion.

area that is irrigated bywaterin which liquid plant wastes have been discharged.

Samples of three different kinds of Monthly, when available. Gamma isotopicmand 1-131 broad leaf vegetation grown analysis.

neares? each of two different offsite locations of highest predicted annual average ground level D/O if milk sampling is not ,

performeo.

One sample of each of the similar Monthly, when available. Gamma isotopicMand 1-131 broad leaf vegetation grown 15 to analysis.

30 km distant in the least prevalent wind direction if milk sampling is not performed.

Chapter 16.11-13 Page 8 of 14 01/16/99

1 1

) TABLE 16.11-7 (Pace 6 of 8) i TABLE NOTATIONS

! 1. Specific parameters of distance and direction sector from the centerline of the j station, and additional description where pertinent, shall be provided for each l

] and every sample location in Table 16.11-7 in a table and figure (s) in the ,

, ODCM. Refer to NUREG-0133," Preparation of Radiological Effluent I j Technical Specifications for Nuclear Power Plants", October 1978, and to

)

Radiological Assessment Branch Technical Position, Revision 1, November '

1979. Deviations are permitted from the required sampling schedule if specimens are unobtainable due to circumstances such as hazardous )

conditions, seasonal unavailability, and malfunction of automatic sampling j equipment. If specimens are unobtainable due to sampling equipment malfunction, effort shall be made to complete corrective action prior to the i end of the next sampling period. All deviations from the sampling schedule

)

i shall be documented in the Annual Radiological Environmental Operating i

Report pursuant to Technical Specification 5.6.2. It is recognized that, at l times, it may not be possible or practicable to continue to obtain samples of

the media of choice at the most desired location or time, in these instances

[ suitable attemative media and locations may be chosen for the particular 1 .

pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program. In lieu of any Licensee i Event Report required 10 CFR 50.73 and pursuant to Technical Specification i 5.6.3, identify the cause of the unavailability of samples for that pathway and identify the new location (s) for obtaining replacement samples in the next

Radioactive Effluent Release Report and also include in the report a revised figuro(s) and table for the ODCM reflecting the new location (s).

2. One or more instrumento, such as a pressurized ion chamber, for measuring and recording dose rate continuously may be used in place of, or in addition to, integrating dosimeters. For the purposes of this table, a thermoluminescent dosimeter (TLD) is considered to be one phosphor, two or )

, more phosphors in a packet are considered as two or more dosimeters. Film l

badges shall not be used as dosimeters for measuring direct radiation. (The j 40 stations is not an absolute number. The number of direct radiation !

monitoring stations may be reduced according to geographical limitations, ;

e.g., at an ocean site, some sectors will be over water so that the number of i

, dosimeters may be reduced accordingly. The frequency of analysis or j readout for TLD systems will depend upon the characteristics of the specific

system used and should be selected to obtain optimum dose information j within minimal fading.)

5 iO ;

Chapter 16.11-13 Page 9 of 14 01/16/99

TABLE 16.11-7 (Pace 6 of 8)

TABLE NOTATIONS (Cont'd)

3. Airbome particulate sample filters shall be analyzed for gross beta l

radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron :

daughter decay. If gross beta activity in air particulate samples is greater l than 10 times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

1

4. Gamma isotopic analysis means the identification and quantification of !

gamma-emitting radionuclides that may be attributable to the effluents from I the facility. I

5. The " upstream sample" sha!l be taken at a distance beyond significant ;

influence of the discharge. The " downstream" sample shall be taken in an )

area beyond but near the mixing zone. " Upstream" samples in an estuary must be taken far enough upstream to be beyond the plant influence. Salt '

water shall be sampled only when the receiving water is utilized for recreational activities.

'O V

6. A composite sample is one in which the rate at which the liquid sampled is uniform and in which the method of sampling employed results in a specimen that is representative of the time averaged concentration at the location being sampled. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g., hourly) relative to the compositing i period (e.g., monthly) in order to assure obtaining a representative sample.

7. Groundwater samples shall be taken when this source is tapped for drinking or irrigation purposes in areas where the hydraulic gradient or recharge properties are suitable for contamination.

8. The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the OPCM.

9. If harvest occurs more than once a year, sampling shall be performed during each discrete harvest. If harvest occurs continuously, sampling shall be monthly. Attention shall be paid to including samples of tuberous and root food products.

Chapter 16.11-13 Page 10 of 14 01/16/99

O O O TABLE 16.11-7 (Pace 8 of 8)

REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS ANALYSIS WATER AIRBOURNE FISH Milk FCOD (pCill) PARTICULATE (pCi/kg, wet) (pCill) PRODUCTS OR GASES (pCi/kg, wet)

(pCi/m*)

H-3 20,000(" .

Mn-54 1,000 30,000 Fe-59 400 10,000 Co-58 1,000 30,000 Co-60 300 10,000 Zn-65 300 20,000 Zr-Nb-95 400 +

l-131 2 0.9 3 100 !

Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 i

Ba-La-140 200 300 5

(1) For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 i pCi/l may be used. ;

i Chapter 16.11-13 Page 11 of 14 01/16/99 !

t

~

O O O LOWER LIMIT OF DETECTION (LLD/*I ANALYSIS WATER AIRBORNE FISH MILK FOOD SEDIMENT (pCi/l) PARTICULATE (pCl/kg, wet) (pCi/l) PRODUCTS (pCl/kg, dry) .

OR GASES (pCi/kg, wet)

(pCi/m*)

Gross Beta 4 0.01 H-3 2000*

Mn-54 15 130 Fe-59 30 260 Co-58, 60 15 130 Zn-65 30 260 Zr-Nb-95 15 1-131 1W 0.07 1 60 !

Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-La-140 15 15 ;

If no drinking water pathway exists, a value of 3000 pCi/l may be used.

)

i Chapter 16.11-13 Page 12 of 14 01/16/99

TABLE 16.11-8 (Pace 2 of 3)

TABLE NOTATIONS

1. This list does not mean that only these nuclides are to be considered. Other peaks that are identifiable, together with those of the above nuclides, shall also be analyzed and reported in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2. l

2. Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.

3. The LLD is defined, for purposes of these commitments, as the smallest concentrations of radioactive material in a sample that will yield a net count, above system background, that will be detected with 95% probability with only 5% probability of falsely concluding that a blank observation represents a i

"real" signal.

For a particular measurement system, which may include radiochemical separation:

O LLD = 4.66s

E V 2.22 Y exp(- A At)

Where:

LLD = the "a priori" lower limit of detection (picoCuries per unit mass or volume);

so = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (counts per minute);

E = the counting efficiency (counts per disintegration);

V = the sample size (units of mass or volume);

2.22 = the number of disintegrations per minute per picocurie; Y = the fractional radiochemical yield, when applicable; A = the radioactive decay constant the particular radionuclide (sec");

and, 4

Chapter 16.11-13 Page 13 of 14 01/16/99

TABLE 16.11-8 (Pace 3 of 3)

TABLE NOTATIONS (Cont'd)

)

At = the elapsed time between environmental collection, or end of the I sample collection period, and time of counting (sec). )

Typical values of E, V, Y and At should be used in the calculation.

i lt should be recognized that the LLD is defined as an a Driori (before the fact) )

limit representing the capability of a measurement system and not as an a Dosteriori (after the fact) limit for a particular measurement. Analyses shall be )

performed in such a manner that the stated LLDs will be achieved under ;

routine conditions. Occasionally background fluctuations, unavoidable small '

sample sizes, the presence of interfering nuclides, or other uncontrollable c!rcumstances may render these LLDs unachievable. In such cases, the :

contributing factors shall be identified and described in the Annual l

Radiological Environmental Operating Report pursuant to Technical '

Specification 5.6.2.

l

4. LLD for drinking water samples. If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.

l l

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O i Chapter 16.11 Page 14 of 14 01/16/99 j

l

t _16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11-14 LAND USE CENSUS COMMITMENT:

A Land Use Census shall be conducted and shall identify within a distance of 8 km (5 miles) the location in each of the 16 meteorological sectors of the nearest milk animal, 2

the nearest residence, and the nearest garden' of greater than 50 m2 (500 ft ) producing broad leaf vegetation.

APPLICABILITY:

At all times.

REMEDIAL ACTION:

a. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment greater than the values currently being calculated in SLC 16.11-9, identify the new location (s) in the next Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3.

O, b. With a Land Use Census identifying a location (s) that yields a calculated dose or dose commitment (via the same exposure pathway) 20% greater than at a location from which samples are currently being obtained in accordance with SLC 16.11-13, add the new location (s) within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. The sampling location (s), excluding the control station location, having the lowest calculated dose or dose commitment (s), via the same exposure pathway, may be deleted from this monitoring program after October 31 of the year in which this Land Use Census was conducted. Pursuant to Technical Specification 5.5.1, submit in the next Radioactive Effluent Release Report l documentation for a change in the ODCM including a revised figure (s) and table (s) for the ODCM reflecting the new location (s), with information supporting the change in the sampling locations.

Broad leaf vegetation sampling of at least three different kinds of vegetation may be performed at the

^ SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Os in lieu of 4

) the garden census. Commitments for broad leaf vegetation sampling in Table 16.11-7.4.c shall be followed, including analysis of control samples.

Chapter 16.11-14 Page 1 of 2 01/16/99

i t TESTING REQUIREMENTS:

, The Land Use Census shall be conducted during the growing season at least once

per 12 months using that information that will provide the best results, such as by a l

" door-to-door survey, aerial survey, or by consulting local agriculture authorities. The !

results of the Land Use Census shall be included in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2. l ;

REFERENCES:

I j 1. Catawba Offsite Dose Calculation Manual 2.~ 10 CFR Part 50, Appendix I BASES:

I !

This commitment is provided to ensure that changes in the use of areas at and I beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program given in the ODCM are made if ,

j required by the results of this census. The best information from the door-to-door l

~

survey, from aerial survey or from consulting with local agricultural authorities shall j

be used. This census satisfies the requirements of Section IV.B.3 of Appndix ! to

!f '

10 CFR Part 50. Restrict;ng the census to gardens of greater than 50 m provides

] assurance that significant exposure pathways via leafy vegetables will be identified

and monitored since a garden of this size is the minimum required to produce the j quantify (26 kg/ year) of leafy vegetables assumed in Regulatory Guide 1.109 for consumption by a child. To determine this minimum garden size, the following .

! assumptions were made: (1) 20% of the garden was used for growing broad leaf i

vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 j kg/m*.

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Chapter 16.11-14 Page 2 of 2 01/16/99 l

I 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11-15 INTERLABORATOR COMPARISON PROGRAM COMMITMENT:

Analyses shall be performed on all radioactive materials, supplied as part of an l Interlaboratory Comparison Program that has been approved by the Commission, i that correspond to samples required by Table 16.11-7. !

APPLICABILITY: l At all times.

REMEDIAL ACTION: I With analyses not being performed as required above, report the corrective actions taken to prevent a recurrence to the Commission in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2. l s TESTING REQUIREMENTS:

(

The Interlaboratory Comparison Program shall be described in the ODCM. A summary of the results obtained as part of the above required Interlaboratory Comparison Program shall be included in the Annual Radiological Environmental Operating Report pursuant to Technical Specification 5.6.2. l

REFERENCES:

1. Catawba Offsite Dose Calculation Manual

2. 10 CFR Part 50, Appendix i BASES:

The requirement for participation in an approved Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive materialin environmental sample matrices are performed as part of the quality assurance program for environmental monitoring in order to demonstrate that the results are valid for the purposes of Section IV.B.2 of Appendix i to 10 CFR Part 50.

O V

Chapter 16.11-15 Page 1 of 1 01/16/99

A d

16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11-16 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT AND RADIOACTIVE EFFLUENT RELEASE REPORT l 1

l COMMITMENT:

16.11-16.1 ANHUAL RADIOLOGICAL ENVIRONMENTAL OPERATING

REPORT' s

l Routine Annual Radiological Environmental Operating Reports covering the l L operation of the unit during the previous calendar year shall be submitted prior to May 15 of each year. l ;

i

! The Annual Radiological Environmental Operating Reports shall include summaries, interpretations, and an analysis of trends of the results of the radiological

environmental surveillance activities for the report period, including a comparison

. with preoperational studies, with operational controls as appropriate, and with i

! previous environmental surveillance reports, and an assessment of the observed !

impacts of the plant operation on the environment. The reports shall also include i the results of the land use census. i O The Annual Radiological Environmental Operating Reports shall include the results I of analysis of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the Table and Figures in the ODCM, as well as summarized and tabulated results of these analyses and measurements in the format of the table in the Radiological Assessment Branch Technical Position, Revision 1, November 1979. In the event ;

that some individual 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 possible in a supplementary report.

The reports shall also include the following: a summary description of the ,,

Radiological Environmental Monitoring Program; at least two legible maps covering all sampling locations keyed to a table giving distances and directions from the centerline of one reactor the results of licensee participation in the Interlaboratory Comparison Program, required by SLC 16.11-15; discussion of all

A single submittal may be made for the station.

O " One map shall cover stations near the SITE BOUNDARY, and a second map shall include the more distant stations.

Chapter 16.11-16 Page 1 of 4 01/16/99

1 fh (v) COMMITMENT (con't) deviations from the sampling schedule of Table 16.11-7; and discussion of all analyses in which the LLD required by Table 16.11-8 was not achievable.

16.11-16.2 RADIOACTIVE EFFLUENT RELEASE REPORT (See Note)

The Radioactive Effluent Release Report covering the operation of the unit during l the previous calendar year shall be submitted before May 1 of each year. The Radioactive Effluent Release Reports shallinclude a summary of the quantities of l radioactive liquid and gaseous effluents and solid waste released from the unit.

The Radioactive Effluent Release Report shallinclude an annual summary of hourly l meteorological data collected over the previous year. This annual summary may be either in the form of an hour-by-hour listing on magnetic tape of wind speed, wind direction, atmospheric stability, and precipitation (if measured), or in the form of joint frequency distributions of wind speed, wind direction, and atmospheric stability. [In lieu of submission with the Radioactive Effluent Release Report, the licensee has l the option of retaining this summary of required meteorological data on site in a file that shall be provided to the NRC upon request.) This same report shallinclude an assessment of the radiation doses due to the radioactive liquid and gaseous effluents released from the unit or station during the previous calendar year. This O same report shall also include an assessment of the radiation doses from V radioactive liquid and gaseous effluents to MEMBER OF THE PUBLIC due to their activities inside the SITE BOUNDARY during the report period. All assumptions used in making these assessments, i.e., specific activity, exposure time and location, shall be included in these reports. The meteorological conditions concurrent with the time of release of radioactive materials in gaseous effluents, as determined by sampling frequency and measurement, shall be used for determining the gaseous pathway doses. The assessment of radiation doses shall be performed in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

The Radioactive Effluent Release Report shall also include an assessment of l

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 direct radiation, for the previous calendar year to show conformance with 40 CFR Part 190,

Environmental Radiation Protection Standards for Nuclear Power Operation". Acceptable methods for calculating the dose contribution from liquid and gaseous effluents are given in Regulatory Guide 1.109, Rev.1, October 1977.

!o Chapter 16.11-16 Page 2 of 4 01/16/99

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

1 l

COMMITMENT (con't)

The Radioactive Effluent Release Reports shallinclude the following informatioa for l each type of solid waste shipped offsite during the report period:

a. Total Container volume, in cubic meters,

b. Total Curie quantity (determined by measurement or estimate),

c. Principal radionuclides (determined by measurement or estimate),

d. Type of waste (e.g., dewatered spent resin, compacted dry waste, evaporator bottoms),

e. Number of shipments, and

f. Solidification agent or absorbent (e.g., cement or other approved 4 agents (media)).

i i

The Radioactive Effluent Release Reports shall include a list and description of l unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during l the reporting period to the PROCESS CONTROL PROGRAM (PCP) and 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 pursuant to SLC 16.11-14. i Note:

i A single submittal may be made for the station. The submittal should combine !

those sections that are common to both units. '

I O

Chapter 16.11-16 Page 3 of 4 01/16/99

!O s

SITE ROUNDARY PERIMETER FENCE

- yy t

NUCLEAR SERVICE F_

WATER POND

. 4 N 2500 FT.' R. .

- I t

EXCLUSION ',

BOUNDARY

\ 1 lNTAKE STRUCTURE ,

1 g_ _ , ..... . .......... : :'e v

; m ; . .-', ; -: :

REACTOR STATION VENTS TURRINE BLDGS- i g

f i: BLDGS. { GASEOUS RELEASE ,

I l POIN'l3 EL. 718.75) 4 s 2:

7 l I " Tc" J cOOuNG

( it YARD l TOWER 1

- Jj ACCESS no s$.

O OOOI^"

U_

x y =s '

--. O

\ CENTER

. OtSCHARGE STRUCTURE l (Lt0UID RE POINT)

\

, g ~w_CO RO t m j _

1 1

Figure 16.11-1 UNRESTRICTED AREA AND SITE BOUNDARY FOR RADIOACTIVE EFFLUENTS Chapter 16.11-16 Page 4 of 4 01/16/99

.. _ m . _. ._ . - _ _ _ _ _ _ _ . _ _ . _ . . _ . _ _ _ . _ , _ _ _ _ . _ _ _ _ .-

F i

16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-17 LIQUID HOLDUP TANKS COMMITMENT:

i The quantity of radioactive material contained in each temporary unprotected ,

outdoor tank shall be limited to less than or equal to 10 Curies, excluding tritium and dissolved or entrained noble gases.

APPLICABILITY:

r At all times. i REMEDIAL ACTION:

With the quantity of radioactive materialin any of the above tanks exceeding the l above limit, immediately suspend all additions of radioactive material to the tank, j within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit, and describe the events leading to this condition in the next Radioactive Effluent Pelease Report, pursuant to Technical Specification 5.6.3.

O TESTING REQUIREMENTS:

The quantity of radioactive material contained in each of the above tanks shall be f

determined to be within the above limit by analyzing a representative sample of the tank's contents at least oace per 7 days when radioactive materials are being added :

to the tank.

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

2. Technical Specification 5.5.12, Explosive Gas and Storage Tank Radioactivity Monitoring Program.

BASES:

The tanks included in this COMMITMENT are all those outdoor radwaste tanks that are not surrounded by liners, dikes or walls capable of holding the tank contents and O

Chapter 16.11-17 Page 1 of 2 01/16/99

i 1

BASES (con't) j

that do not have tank overflows and surrounding e.rsa drains connected to the Liquid Radwaste Treatment System.

Restricting the quantity of radioactive materer s.ntained in the specif.ied tanks .

provides assurance that in the avent of an unco .u;hed release of the tank's

contents, the resulting concentrations would b. esa than the limits of 10 CFR Part i 20, Appendix B, Table ll, Column 2, at the noa. y' totable water supply and the

nearest surface water supply in an U.PESTRI~ dD AREA.

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O Chapter 16.11-17 Page 2 of 2 01/16/99

I

) 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-18 EXPLOSIVE GAS MIXTURE ,.

COMMITMENT:

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

$ APPLICABILITY:

At all times.

REMEDIAL ACTION:

4

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

b. With the concentration of oxygen in the WASTE GAS HOLDUP SYSTEM greate: ther 4% by volume and the hydrogen concentration greater than 4%

by wlume, immediately suspend all additions of waste gases to the system ,

4 ard reduce !he concentration of oxygen to less than or equal to 4% by l v,>lume; then take REMEDIAL ACTION a. above.

TE. STING REQUIREMENTS:

The concentrations of hydrogen and oxygen in the WASTE GAS HOLDUP SYSTEM shall be determined to be within the above limits by continuously monitoring ii'is waste gases in the WASTE GAS HOLDUP SYSTEM with the hydrogen and oxygen j monitors required OPERABLE by Table 16.11-20A of SLC 16.11-20.

REFERENCES:

4

! 1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Techr.ical Specifications Amendments for Catawba, September 30,1998.

2. Technical Specification 5.5.12, Explosive Gas and Storage Tank Radioactivity Monitoring Program.

O l

i Chapter 16.11-18 Page 1 of 2 01/16/99 l

' BASES: )

This COMMITMENT is provided to ensure that the concentration of potentially explosive gas mixtures contained in the WASTE GAS HOLDUP SYSTEM is maintained below the flammability limits of hydrogen and oxygen. (Automatic control features are included in the system to prevent the hydrogen and oxygen :

concentrations from reaching these flammability limits. These automatic control '

features include isolation of the source of hydrogen and/or oxygen, automatic diversion to recombiners, orinjection of dilutants to reduce the concentration below the flammability Wits.) Maintaining the concentration of hydrogen and oxygen below their f!ais mability limits provides assurance that the releases of radioactive ;

materials will Ns cortrolled in conformance with the requirements of General Design Criterion 60 af Appendix A to 10 CFR Part 50.

i 1

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o ll Chapter 16.11-18 Page 2 of 2 01/16/99 I

i

16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-19 GAS STORAGE TANKS COMMITMENT:

The quantity of radioactivity contained in each gas storage tank shall be limited to less than or equal to 97,000 Curies of noble gases (considered as Xe-133 equivalent).

APPLICABILITY:

At all times.

REMEDIAL ACTION:

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

leading to this condition in the next Radioactive Effluent Release Report pursuant to Technical Specification 5.6.3.

TESTING REQUIREMENTS:

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

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

2. Technical Specification 5.5.12, Explosive Gas and Storage Tank Radioactivity Monitoring Program.

BASES:

l The tanks included in this COMMITMENT are those tanks for which the quantity of l radioactivity contained is not limited directly or indirectly by another COMMITMENT.

J Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the l

Chapter 16.11-19 Page 1 of 2 01/16/99

BASES (con't) resulting whole body exposure to a MEMBER OF THU PUBLIC at the nearest SITE ,

BOUNDARY will not exceed 0.5 rem. This is consistent with Standard Review Plan 11.3, Branch Technical Position ETSB 11-5, " Postulated Radioactive Releases Due to a Waste Gas System Leak or Failure", in NUREG-0800, July 1981. l t

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Chapter 16.11-19 Page 2 of 2 01/16/99 l

l

16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-20 EXPLOSIVE GAS MONITORING INSTRUMENTATION

, COMMITMENT:

The explosive gas monitoring instrumentation channels shown in Table 16.11-20A shall be OPERABLE with their Alarm / Trip Setpoints set to ensure that the limits of SLC 16.11-18 are not exceeded.

APPLICABILITY:

1 During WASTE GAS HOLDUP SYSTEM operation.

REMEDIAL ACTION: 1

a. With an explosive gas monitoring instrumentation channel Alarm / Trip setpoint less conservative than required by the above COMMITMENT, declare the channel inoperable and take the REMEDIAL ACTION shown in Table I 16.11-20A.

b. With less than the minimum number of explosive gas monitoring

~.

instrumentation channels operable, take the REMEDIAL ACTION shown in Table 16.11-20A. Restore the inoperable instrumentation to OPERABLE status within 30 days, or if unsuccessful, prepare and submit a Special Repod to the Commission to explain why this inoperability was not corrected within the time specified.

TESTING REQUIREMENTS:

Each explosive gas monitoring instrumentation channel shall be demonstrated

OPERABLE by performance of the CHANNEL CHECK, CHANNEL CAllBRATION, and CHANNEL OPERATIONAL TEST operations at the frequencies shown in Table 16.11-20B.

4 4

REFERENCES:

l

1. Letter from NRC to Gary R. Peterson, Duke, issuance of Improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

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

Chapter 16.11-20 Page 1 of 3 01/16/99 i

TABLE 16.11-20A EXPLOSIVE GAS MONITORING INSTRUMENTATION MINIMUM CHANNELS REMEDIAL INSTRUMENT OPERABLE ACT ON WASTE GAS HOLDUP SYSTEM ;

Explosive Gas Monitoring System ^

, a. Hydrogen Monitors 1/ train per station C

b. Oxygen Monitors 2/ train per station D

) TABLE NOTATIONS 4

4 REMEDIAL ACTION STATEMENTS O

O ACTION C - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend oxygen supply to the recombiner.

ACTION D - With the number of channels OPERABLE one less than required by 1 the Minimum Channels OPERABLE requirement, operation of this system may continue provided grab samples are taken and analyzed at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . With both channels inoperable, operation may continue provided grab samples are taken and analyzed at least i once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during degassing operations ar.d at least once per 24 '

(

hours during other operations.

Chapter 16.11-20 Page 2 of 3 01/16/99

O O O TABLE 16.11-20B EXPLOSIVE GAS MONITORING INSTRUMENTATION TESTING REQUIREMENTS CHANNEL CHANNEL CHANNEL OPERATIONAL INSTRUMENT CHECK CALIBRATION TEST WASTE GAS HOLDUP SYSTEM Explosive Gas Monitoring System ,

a. Hydrogen Monitors 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 92 days (1) 31 days

b. Oxygen Monitors 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 92 days (2) 31 days ,

(1) The CHANNEL CALIBRATION shall include the use of standard gas samples in accordance with the manufacturer's recommendations. In addition, a standard gas sample of nominal four volume percent hydrogen, balance nitrogen, shall be used in the calibration to check linearity of the hydrogen analyzer.

(2) The CHANNEL CAllBRATION shallinclude the use of standard gas samples in accordance with the 6 manufacturer's recommendations. In addition, a standard gas sample of nominal four percent oxygen, balance nitrogen, shall be used in the calibration to check linearity of the oxygen analyzer.

i Chapter 16.11-20 Page 3 of 3 01/16/99 1:

I 16.11 RADIOLOGICAL EFFLUENTS CONTROLS 16.11-21 plAJOR CHANGES TO LIQUID. GASEOUS. AND SOLID RADWASTE TREATMENT SYSTEMS 1

COMMITMENT:

Licensee-initiated major changes to the Radwaste Treatment Systems (liquid, !

gaseous, and solid):

a. Shall be reported" to the Commission in the Radioactive Effluent Release l

Report for the period in which the evaluation was reviewed by the Station l Manager. The discussion of each change shall contain:

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

2) Su'l'icient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

3) A detailed description of the equipment, components, and processes

(

involved and the interfaces with other plant systems;

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

5) An evaluation of the change, which shows the expected maximum exposures to a MEMBER OF THE PUBLIC in the UNRESTRICTED AREA and to the general population that differ from those previously estimated in the License application and amendments thereto;

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

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

t.icensees may choose to submit the information called for in this Commitment as part of the periodic UFSAR update.

Chapter 16.11-21 Page 1 of 2 01/16/99

i COMMITMENT (con't)

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

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

REMEDIAL ACTIONS:

)

None 1 l

TESTING REQUIREMENTS:

None

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical ,

Specifications Amendments for Catawba, September 30,1998. !

.i BASES:

O None 4

O Chapter 16.11-21 Page 2 of 2 01/16/99

16.12 RADIATION PROTECTION 16.12-1 SEALED SOURCE CONTAMINATION

! COMMITMENT:

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

APPLICABILITY

At all times.

REMEDIAL ACTIONS:

! With a sealed source having removable contamination in excess of the above limits, immediately withdraw the sealed source from use and either:

I

1. Decontaminate and repair the sealed source, or

[

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

TESTING REQUIREMENTS:

a. Test Requirements - Each sealed source shall be tested for leakage and/or ,

contamination by: l

1. The licensee, or i

2. Other persons specifically authorized by the Commission or an Agreement State.

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

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

O Chapter 16.12-1 Page 1 of 2 01/16/99

TESTING REQUIREMENTS (con't)

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

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

and b) in any form other than gas. ;

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

3. Startup sources and fission detectors - Each sealed startup source ;

and fission detector shall be tested within 31 days prior to being i subjected to core flux or installed in the core and following repair or maintenance to the source.

l

c. Reports - A report shall be prepared and submitted to the Commission on an i f annual basis if sealed source or fission detector leakage tests reveal the presence of greater than or equal to 0.005 microcurie of removable contamination.

l

REFERENCES:

1. Letter from NRC to Gary R. Peterson, Duke, Issuance of improved Technical Specifications Amendments for Catawba, September 30,1998.

BASES:

The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(a)(3) limits for plutonium. This limitation will ensure that leakage from Byproduct, Source, and Special Nuclear Material sources will not exceed allowable intake values.

Sealed sources are classified into three groups according to their use, with Testing Requirements commensurate with the probability of damage to a source in that group. Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested l unless they are removed from the shielded mechanism. '

Chapter 16.12-1 Page 2 of 2 01/16/99

1 Igda CONDUCT OF OPERATIONS 4

l 16.13-1 FIRE BRIGADE ,

! COMMITMENT:

A site Fire Brigade of at least five members shall be maintained onsite at all times. :

The Fire Brigade shall not include three members of the minimum shift crew necessary for safety shutdown of the unit and any personnel required for other

essential functions during a fire emergency.

4 APPLICABILITY: '

At all times, i i i

REMEDIAL ACTION

! l

j. With the Fire Brigade composition less than the minimum requirements for a period !

of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence, take

] immediate action to fill the required positions.

l TESTING PROCEDURES:

j 1. Operations Management Procedure 2-2

REFERENCES:

1. Catawba FSAR, Chapter 13.2

2. Catawba SER, Section 9.5.1 and Appendix D 4

j- 3. Catawba Fire Protection Review, as revised

4. Catawba Fire Protection Commitment Index BASES:

The primary purpose of the Fire Brigade Training Program is to develop a group of I station employees skilled in fire prevention, fire fighting techniques, first aid procedures, and emergency response. They are trained and equipped to function as a team for the fighting of fires. The station fire brigade organization is intended to l be self sufficient with respect to fire fighting activities.

C)

Chapter 16.13-1 Page 1 of 2 12/90

4 BASES (con't)

The Fire Brigade Training program provides for initial training of all new fire brigade members, quarterly classroom training and drills, annual practical training, and j

. leadership training for fire brigade leaders.

This Selected Licensee Commitment is part of the Catawba Facility Operating j License Conditions #6 for NPF-52 and #8 for NPF-35. -

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____._______..________-._a

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Q 16.13 CONDUCT OF OPERATIONS 16.13-2 TECHNICAL REVIEW AND CONTROL COMMITMENT:

4 A Technical Review and Control Program covering the preparation, review, and

, approval of documents important to station operation shall be established and maintained for the site.

APPLICABILITY:

This commitment is applicable at all times and applies to the review and control activities described in items a through j as listed below. 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 2

. standard is as listed in Table 17-1 of the Duke Power Topical Report, Quality Assurance Program, Duke-l.A).

i

a. The preparation, review, and approval of station procedures shall be done in n accordance with the Duke Power Topical Repoli, Quality Accurance '

4 i ) Program. Individuals responsible for these reviews shall be members of the supervisory staff assigned to the site, be previously designated by the Site Vice President as a Qualified Reviewer, and successfully complete the site Qualified Reviewer training program. Review of environmental radiological

. analysis procedures, shall be performed by the General Manager, Environmental Services or a designee. Each such review shall include a ,

determination of whether or not additional, cross-disciplinary review shall be '

performed by the appropriately designated site review personnel. i

b. Proposed modifications shall be designed and the design reviewed in accordance with the Duke Power Topical Report, Quality Assurance Program. The proposed modification design, the design review, and design approval shall be in accordance with ANSI N45.2.11 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).

V Chapter 16.13-2 Page 1 of 3 01/16/99

l l

n

U APPLICABILITY: (cont'd)

c. Proposed changes to the station Technical Specifications and/or Facility

{

Operating License shall be prepared in accordance with the Duke Power Topical Report, Quality Assurance Program. Each proposed Technical Specifications (incNding affected Bases) and/or Facility Operating License change shall be reviewed by the Plant Operations Review Committee i (PORC) and the Nuclear Safety Review Board (NSRB) prior to submittal to the Nuclear Regulatory Commission. Proposed changes to the Technical i Specifications and/or Facility Operating License shall be approved by the l i

Station Manager, or for the Station Manager by a designated manager or l

company officer. License Amendment Request cover letters shall be signed by an officer of Duke Energy Corporation. l

d. Proposed tests and experiments which affect station nuclear safety and are not addressed in the FSAR 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 i prepared which evaluates the occurrence and which provides recommendations to prevent recurrence. Such reports shall be approved by the Manager, Safety Assurance and provided to the Site Vice President and V the Plant Operations Review Committee (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 the environs, including the preparation and forwarding of reports covering evaluation, recommendations, and disposition of the corrective action to prevent recurrence to the Site Vice President, and to the 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 (ODCM), and Radwaste Treatment Systems.

I. The Manager, Safety Assurance shall ensure the performance of a review by a i knowledgeable individual / organization of the Fire Protection program and implementing procedures and submittal of recommended change-s to the U]

Manager, Human Resources.

Chapter 16.13-2 Page 2 of 3 01/16/99

e. ;

= :

, I Y

O i

are'ica iuTv: <o et > !

! j. Reports documenting each of the activities performed under this commitment f

.shall be maintained. Copies shall be provided to the NSRB. !

REMEWAL ACDONS

Not Applicable I

E8ElEl i i i

a.- The requirements contained in this Selected Licensee Commitment were 6

l. relocated from the Catawba Technical Specifications with the approval of the l ;

j U. S. Nuclear Regulatory Commission. Changes to this SLC shall be considered a change in an NRC commitment and shall be made only in !

i accordance with the approved Nuclear System Directive for the Control of l !

!. Selected Licensee Commitments and by use of the 10 CFR 50.59 evalue. tion !

! process.

] l

b. This SLC implements the review requirements of ANSI N18.7-1976/ANS-3.2 l ll '

and ANSI N45.2.11-1974 as referenced in the Duke Power Company Topical i Report, Quality Assurance Program, Duke-1 -A.

e ,

l\ TESTING _Rj9U!REMENTS:

]

Not Applicable '

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 Operatienal Phase of Nuclear Power Plants.

3. ANSI N45.2.11-1974, Quality Assurance Requirements for the Design of ,

Nuclear Power Plants i

4. Nuclear System Directive 221, Facility Operating License and Technical i Specificatien Amendments / Selected Licensee Commitments / Technical 1 Specificat!ons Bases Changes l

5. Nuclear System Directive 209,10CFR 50.59 Evaluations

6. 10CFR 50.59 1

7. Nuclear System Directive 703, Administrative Instructions for Station Procedures Chapter 16.13-2 Page 3 of 3 01/16/99

I 16.13 CONDUCT OF OPERATIONS 16.13-3 PLANT OPERATIONS REVIEW COMMITTEE COMMITMENT:

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 most responsible for engineering support of station operation and maintenance, or designated altemates. The PORC Chairperson, members, and attemate members shall be qualified in accordance with ANSI N18.1-1971 and be appointed by the Site Vice President. The quorum necessary for conducting the PORC functions shall consist of the Chairperson, or his/her designated attemate, and at least three other PORC members including alternates.

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.

(J l APPLICABILITY:

a. The PORC shall be responsible for reviewing the following prior to final approval:

1. All proposed tests and experiments which affect station nuclear safety and are not addressed in the FSAR or Technical Specif! cations;

2. Operability evaluations resulting in a Justification for Continued Operation and a proposal for discretionary enf arcement;

3. Operability evaluations resulting in the decisic.n that affeated systems, structures or components are OPERABLE but degraded; and

4. All prcposed changes to the station Technical Specifications, Bases, or Fa;ility Operating License.

b. The PORC shall be responsible for reviewing the effectiveness of corrective actions for:

1. Lhenseo Event Reports and Special Reports made to the NRC; C}

\,

Chapter 16.13-3 Page 1 of 2 01/16/99

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

O APPLICABILITY (con't)

2. Violations of Technical Specifications;

3. Special reviews and investigations as requested by the Site Vice Pree' dent; and

4. Reports on unplanned onsite releases of radioactive material to the environs.

c. The PORC shall review additional programs, procedures and plant activities as directed by the Site Vice President.

1 REMEDIAL ACTIONS- '

1 Not Applicable I i

TESTING REQUIREMENTS:

Not Applicable BASES: '

f

a. The PORC shall be establishe,d to recommend to the Station Manager

' approval or disapproval of the items listed under APPLICABILITY prior to their final approval.

i b. The PORC shall report to the Site Vice President on the areas of 4

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 I

i Chapter 16.13-3 Page 2 of 2 01/16/99 i

ML20199H406

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16.1 INTRODUCTION

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