Text

Selected Licensee Commitment Manual
	ML103560105
Person / Time
Site:	Mcguire, McGuire
Issue date:	12/13/2010
From:	Repko R; Duke Energy Carolinas
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML103560105 (369)
	v • d • e

REGIS T. REPKO AkDuke Vice President

[WEnergy McGuire Nuclear Station Duke Energy MG01VP /12700 Hagers Ferry Rd.

Huntersville, NC 28078 980-875-4111 980-875-4809 fax regis.repko@duke-energy.com December 13, 2010 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555-0001

Subject:

Duke Energy Carolinas, LLC McGuire Nuclear Station, Units 1 and 2 Docket Nos. 50-369 and 50-370 Selected Licensee Commitment Manual Attached is a copy of the current McGuire Selected Licensee Commitment (SLC)

Manual. The SLC Manual is Chapter 16 of the McGuire Updated Final Safety Analysis Report.

Questions related to this submittal should be directed to Kay Crane, McGuire Regulatory Compliance at (980) 875-4306.

Regis T. Repko

Attachment:

McGuire Selected Licensee Commitment (SLC) Manual www. duke-energy.com

U.S. Nuclear.Regulatory Commission December 13, 2010 Page 2 xc w/attachment:

Mr. Jon H. Thompson, Project Manager (Addressee Only)

U.S. Nuclear Regulatory Commission 11555 Rockville Pike Rockville, MD 20852-2738 U.S. Nuclear Regulatory Commission Region II DRP/RPB1 Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 xc w/o attachment:

Mr. J.B. Brady Senior Resid6nt Inspector U.S. Nuclear Regulatory Commission McGuire Nuclear Station

U.S. Nuclear Regulatory Commission December 13, 2010 Page 3 Regis T. Repko affirms that he is the person who subscribed his name to the foregoing statement, and that all the matters and facts set forth herein are true and correct to the best of his knowledge.

Regis T. Repko, Site Vice President, McGuire Nuclear Station Subscribed and sworn to me: /0ý- 13-0ý0/0 Date

,"A &I alK.-I Notary Public I

My commission expires:

Date

ATTACHMENT McGuire Selected Licensee Commitment (SLC) Manual

TABLE OF CONTENTS SECTION TITLE PAGE

16.1 INTRODUCTION

16.1-1 16.2 APPLICABILITY 16.2-1 16.3 DEFINITIONS 16.3-1 16.4 COMMITMENTS RELATED TO REACTOR COMPONENTS 16.5 COMMITMENTS RELATED TO REACTOR COOLANT SYSTEM 16.5.1 Inventory Control - Reduced Inventory Operation 16.5.1-1 16.5.2 Power Systems and Decay Heat Removal - Reduced Inventory Operation 16.5.2-1 16.5.3 Reactivity Control - Reduced Inventory Operation 16.5.3-1 16.5.4 NCS Instrumentation - Reduced Inventory Operation 16.5.4-1 16.5.5 Containment.Closure - Reduced Inventory Operation 16.5.5-1 16.5.6 Safety Valves - Shutdown 16.5.6-1 16.5.7 Chemistry 16.5.7-1 16.5.8 Pressurizer 16.5.8-1 16.5.9 Structural Integrity 16.5.9-1 16.5.10 Reactor Vessel Head Vent System 16.5.10-1 16.6 COMMITMENTS RELATED TO ENGINEERED SAFETY FEATURES 16.6.1 Containment Sump 16.6.1-1 16.6.2 DELETED 6/11/03 16.6.3 Inlet Door Position Monitoring System 16.6.3-1 16.6.4 Safety Injection System Nozzles 16.6.4-1 16.7 COMMITMENTS RELATED TO INSTRUMENTATION 16.7.1 ATWS/AMSAC 16.7.1-1 16.7.2 Seismic Instrumentation 16.7.2-1 16.7.3 Meteorological Instrumentation 16.7.3-1 McGuire Units 1 and 2 Revision 36

TABLE OF CONTENTS SECTION TITLE PAGE 16.7.4 Loose-Part Detection Instrumentation 16.7.4-1 16.7.5 Turbine Overspeed Protection 16.7.5-1 16.7.6 Radiation Monitoring for Plant Operations 16.7.6-1 16.7.7 Movable Incore Detectors 16.7.7-1 16.7.8 Explosive Gas Monitoring Instrumentation 16.7.8-1 16.7.9 Rod Position Indication System - Shutdown 16.7.9-1 16.7.10 Rod Position Indication System - Test Exception 16.7.10- 1 16.7.11 Hydrogen Monitors 16.7.11- 1 16.8 COMMITMENTS RELATED TO ELECTRICAL POWER SYSTEMS 16.8.1 Containment Penetration Overcurrent Protective Devices 16.8.1-1 16.8.2 Switchyard Activities 16.8.2-1 16.8.3 DG Supplemental Testing Requirements 16.8.3-1 16.9 COMMITMENTS RELATED TO AUXILIARY SYSTEMS 16.9.1 Fire Suppression Water System 16.9.1-1 16.9.2 Spray and/or Sprinkler Systems 16.9.2-1 16.9.3 Halon Systems 16.9.3-1 16.9.4 Fire Hose Stations 16.9.4-1 16.9.5 Fire Rated Assemblies 16.9.5-1 16.9.6 Fire Detection Instrumentation 16.9.6-1 16.9.7 Standby Shutdown System 16.9.7-1 16.9.8 Groundwater Level Monitoring System 16.9.8-1 16.9.9 Boration Systems - Flow Path (Operating) 16.9.9-1 16.9.10 Boration Systems - Charging Pumps (Operating) 16.9.10-16.9.11 Borated Water Sources (Operating) 16.9.11-16.9.12 Boration Systems - Flow Path (Shutdown) 16.9.12-McGuire UInits 1 and 2 ii Revision 36

TABLE OF CONTENTS SECTION TITLE PAGE 16.9.13 Boration Systems - Charging Pumps (Shutdown) 16.9.13-1 16.9.14 Borated Water Sources (Shutdown) 16.9.14-1 16.9.15 Snubbers 16.9.15-1 16.9.16 Area Temperature Monitoring 16.9.16-1 16.9.17 Refueling Operations - Decay Time 16.9.17-1 16.9.18 Refueling Operations - Communications 16.9.18-1 16.9.19 Refueling Operations - Manipulator Crane 16.9.19-1 16.9.20 Crane Travel - Spent Fuel Pool Storage Building 16.9.20-1 16.9.21 Water Level - Spent Fuel Storage Pool 16.9.21-1 16.9.22 Switchgear Room Ventilation System (SGRVS) 16.9.22-1 16.9.23 Control Room Area Ventilation System (CRAVS) 16.9.23-1 16.9.24 Not Used 16.9.25 Refueling Operations - Containment Equip Hatch 16.9.25-1 16.10 COMMITMENTS RELATED TO STEAM AND POWER CONVERSION SYSTEMS 16.10.1 Steam Generator Pressure/Temperature Limitation 16.10.1-1 16.11 COMMITMENTS RELATED TO RADIOACTIVE WASTE MANAGEMENT AND RADIOLOGICAL EFFLUENT CONTROL 16.11.1 Liquid Effluents - Concentration 16.11.1-1 16.11.2 Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2-1 16.11.3 Dose - Liquid Effluents. 16.11.3-1 16.11.4 Liquid Radwaste Treatment System 16.11.4-1 16.11.5 Chemical Treatment Ponds 16.11.5-1 16.11.6 Dose Rate - Gaseous Effluents 16.11.6-1 16.11.7 Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7-1 McGuire Units 1 and 2 iii Revision 36

TABLE OF CONTENTS SECTION TITLE PAGE 16.11.8 Noble Gases 16.11.8-1 16.11.9 Dose - Iodine-131 and 133, Tritium and Radioactive Materials in Particulate Form 16.11.9-1 16.11.10 Gaseous Radwaste Treatment System 16.11.10-1 16.11.11 Solid Radioactive Waste 16.11.11-1 16.11.12 Total Dose 16.11.12-1 16.11.13 Radiological Environmental Monitoring Program 16.11.13-1 16.11.14 Land Use Census 16.11.14-1 16.11.15 Interlaboratory Comparison Program 16.11.15-1 16.11.16 Annual Radiological Environmental Operating Report 16.11.16-1 16.11.17 Radioactive Effluent Release Reports 16.11.17-1 16.11.18 Liquid Holdup Tanks 16.11.18-1 16.11.19 Explosive Gas Mixture 16.11.19-1 16.11.20 Gas Storage Tanks 16.11.20-1 16.12 COMMITMENTS RELATED TO RADIATION PROTECTION 16.12.1 In-Plant Iodine Monitoring 16.12.1-1 16.12.2 Sealed Source Contamination 16.12.2-1 16.13 COMMITMENTS RELATED TO CONDUCT OF OPERATIONS 16.13.1 Fire Brigade 16.13.1-1 16.13.2 Not Used 16.13.3 Not Used 1.6.13.4 Minimum Station Staffing Requirements 16.13.4-1 16.14 COMMITMENTS RELATED TO TESTING 16.14.1 Startup Reports 16.14.1-1 16.14.2 Reactor Trip Breaker and Solid State Protection 16. 14.2-1 System (SSPS) Logic Train Out of Service Commitments McGuire Units 1 and 2 iv Revision 36

SELECTED LICENSEE COMMITMENTS (SLC)

LOES SLCs ARE REVISED PER SECTION SECTION REVISION NUMBER DATE 16.1 REVISION 32 12/2/02 16.2 REVISION 90 3/13/07 16.3 REVISION 48 7/31/03 16.4 Not Issued 16.5.1 REVISION 115 0308/10 16.5.2 REVISION 0 12/14/99 16.5.3 REVISION 0 12/14/99 16.5.4 REVISION 55 3/23/04 16.5.5 REVISION 0 12/14/99 16.5.6 REVISION 0 12/14/99 16.5.7 REVISION 53 1/13/04 16.5.8 REVISION 0 12/14/99 16.5.9 REVISION 108 06/10/09 16.5.10 REVISION 38 4/9/03 16.6.1 REVISION 0 12/14/99 16.6.2 DELETED 6/11/03 16.6.3 REVISION 61 04/07/05 16.6.4 REVISION 27 06/12/02 16.7.1 REVISION 0 12/14/99 16.7.2 REVISION 80 10/1/05 16.7.3 REVISION 0 12/14/99 16.7.4 REVISION 1 4/11/00 16.7.5 REVISION 0 12/14/99 16.7.6 REVISION 103 12/03/08 16.7.7 REVISION 0 12/14/99 16.7.8 REVISION 77 10/12/05 16.7.9 REVISION 97 10/06/08 16.7.10 REVISION 0 12/14/99 16.7.11 REVISION 71 05/25/05 16.8.1 REVISION 98 11/1/07 16.8.2 REVISION 0 12/14/99 16.8.3 REVISION 65 12/21/04 16.9.1 REVISION 53 1/13/04 16.9.2 REVISION 81 6/15/06 16.9.3 REVISION 106 05/26/09 16.9.4 REVISION 83 7/12/06 16.9.5 REVISION 81 6/15/06 16.9.6 REVISION 107 5/18/09 16.9.7 REVISION 117 8/3/10 16.9.8 REVISION 96 9/10/07 16.9.9 REVISION 101 4/02/08 16.9.10 REVISION 13 2/26/01 16.9.11 REVISION 22 2/25/02 16.9.12 REVISION 101 4/02/08 McGuire Units 1 and 2 I Revision 101

SELECTED LICENSEE COMMITMENTS (SLC)

LOES SLCs ARE REVISED PER SECTION SECTION REVISION NUMBER DATE 16.9.13 Not Used - Rev 13 2/26/01 16.9.14 REVISION 22 2/25/02 16.9.15 REVISION 116 4/7/10 16.9.16 REVISION 111 09/09/09 16.9.17 REVISION 86 1/17/07 16.9.18 REVISION 0 12/14/99 16.9.19 REVISION 102 9/3/08 16.9.20 REVISION 8 11/30/00 16.9.21 REVISION 0 12/14/99 16.9.22 REVISION 109 8/13/09 16.9.23 REVISION 88 1/17/07 16.9.24 Not Used - Rev 74 6/20/05 16.9.25 REVISION 87 1/17/07 16.10.1 REVISION 56 4/6/04 16.11.1 REVISION 112 2/4/10 16.11.2 REVISION 84 7/19/06 16.11.3 REVISION 0 12/14/99 16.11.4 REVISION 0 12/14/99 16.11.5 REVISION 0 12/14/99 16.11.6 REVISION 112 2/4/10 16.11.7 REVISION 84 7/19/06 16.11.8 REVISION 0 12/14/99 16.11.9 REVISION 0 12/14/99 16.11.10 REVISION 0 12/14/99 16.11.11 REVISION 41 8/21/03 16.11.12 REVISION 67 2/28/05 16.11.13 REVISION 91 3/22/07 16.11.14 REVISION 21 1/17/02 16.11.15 REVISION 21 1/17/02 16.11.16 REVISION 1 4/11/00 16.11.17 REVISION 118 10/19/10 16.11.18 REVISION 0 12/14/99 16.11.19 REVISION 0 12/14/99 16.11.20 REVISION 0 12/14/99 16.12.1 REVISION 0 12/14/99 16.12.2 REVISION 0 12/14/99 16.13.1 REVISION 51 10/1/03 16.13.2 Not Used - Rev 75 8/05 16.13.3 Not Used - Rev 75 8/05 16.13.4 REVISION 58 5/11/04 16.14.1 REVISION 0 12/14/99 16.14.2 REVISION 104 3/18/09 McGuire Units 1 and 2 2 Revision 101

INTRODUCTION 16.1 16.0 SELECTED LICENSEE COMMITMENTS

16.1 INTRODUCTION

This chapter provides a single location in the UFSAR where certain selected licensee commitments are presented. The content of this chapter is based on the results of application of a set of criteria to determine the content of technical specifications. For purposes of administrative ease, this chapter is maintained in a separate manual, The McGuire Nuclear Station Selected Licensee Commitments Manual. Those previous technical specification requirements which did not meet the criteria are relocated in this chapter. McGuire Technical Specification 5.4 (Procedures and Programs) requires written procedures to be established, implemented, and maintained on these selected licensee commitments.

The control of the McGuire Nuclear Station selected licensee commitment program and manual shall be in accordance with an approved Nuclear System Directive. The manual is officially designated as Chapter 16 of the McGuire UFSAR. The original issue and subsequent revisions of the manual are approved by the station manager. Administrative requirements of the manual are the responsibility of the Regulatory Compliance Section.

Changes to these Selected Licensee Commitments shall be considered a change in an NRC commitment and shall be made only in accordance with the approved Nuclear System Directive for the Control of Selected Licensee Commitments and by use of the 10 CFR 50.59 Process.

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 UFSAR, but are not of such a level of importance that they need to be under the rigorous control provided by technical specifications.

This chapter includes 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 commitment is further described.

McGuire Units 1 and 2 16.1-1 Revision 32

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

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 REMEDIAL ACTION(S) shall be met, except as provided in SLC 16.2.11. 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 otherwise stated.

16.2.3 When a COMMITMENT is not met, except as provided in the associated REMEDIAL ACTIONS, the Station Manager and/or the Responsible Group Superintendent will determine any further actions.

16.2.4 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 conditions in the Applicability that are required to comply with REMEDIAL ACTIONS. Exceptions to this COMMITMENT are stated in the individual COMMITMENTS.

16.2.5 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)..."

McGuire Units 1 and 2 16.2-1 Revision 90

APPLICABILITY 16.2 16.2 APPLICABILITY (continued) 16.2.6 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, whether such failure is experienced during the performarice 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 16.2.8. Exceptions to these requirements are stated in the individual commitments or may be approved by the Station Manager and/or the Responsible Group Superintendent. Testing Requirements do not have to be performed on inoperable equipment or variables outside specified limits.

16.2.7 The specified Frequency for each Testing Requirement is met if the Test is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply.

If a Completion Time requires periodic performance on a "once per..." basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this SLC are stated in the individual SLCs.

16.2.8 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

.hours or up to the limit of the specified Frequency, whichever is greater. This delay period is permitted to allow performance of the Testing Requirement. A risk evaluation shall be performed for any Testing Requirement delayed greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , and the risk impact shall be managed.

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.

The clarifications provided by the McGuire Nuclear Station Technical Specification Bases for SR 3.0.3 are similarly applicable to SLC 16.2.8 and Testing Requirements.

McGuire Units 1 and 2 16.2-2 Revision 90

APPLICABILITY 16.2 16.2 APPLICABILITY (continued) 16.2.9 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 the specified frequency or as approved by the Station Manager and/or Responsible Group Superintendent. This provision shall not prevent entry into OPERATIONAL MODES or other specified conditions in the Applicability that are required to comply with REMEDIAL ACTIONS.

16.2.10 Testing Requirements shall apply to each unit individually unless otherwise indicated as stated in Specification 16.2.5 for individual commitments or whenever certain portions of a specification contain testing parameters different for each unit, which will be identified in parentheses or footnotes.

16.2.11 Equipment removed from service or declared inoperable to comply with REMEDIAL ACTIONS may be returned to service under administrative control solely to perform testing required to demonstrate its OPERABILITY or the OPERABILITY of other equipment. This is an exception to SLC 16.2.2 for the system returned to service under administrative control to perform the required testing to demonstrate OPERABILITY.

16.2-3 Revision 90 McGuire Units 1 McGuire Units and 2 1 and 2 16.2-3 Revision 90

DEFINITIONS 16.3 16.3 DEFINITIONS The definitions in the McGuire Technical Specifications apply to defined terms used herein.

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

AMSAC ATWS Mitigation System Activation Circuitry, the Westinghouse system for mitigating ATWS events.

ATWS An ANTICIPATED TRANSIENT WITHOUT SCRAM (ATWS) is an expected operational transient (such as loss of feedwater, loss of condenser vacuum, or loss of off site 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 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.

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

PROCESS CONTROL The PCP shall contain the correct formulas, sampling, PROGRAM (PCP) analyses, test, and determinations to be made to ensure that processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10CFR Parts 20, 61, and 71, state regulations, burial ground requirements, and other requirements governing the disposal of solid radioactive waste.

McGuire Units 1 and 2 16.3-1 Revision 48

DEFINITIONS 16.3 16.3 DEFINITIONS (continued)

PURGE or PURGING PURGE or PURGING 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 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.

SECURED Related to valve position indicates that:

1. For manual valves, the subject valve is locked in the desired position, or

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

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

SOURCE CHECK SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity or a simulated source of radioactivity such as a light emitting diode.

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

McGuire Units 1 and 2 16.3-2 Revision 48

DEFINITIONS 16.3 16.3 DEFINITIONS (continued)

VENTILATION EXHAUST A VENTILATION EXHAUST TREATMENT SYSTEM shall be TREATMENT SYSTEM 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 charcoal adsorbers and/or HEPA filters for the purpose of removing iodines or particulates from the gaseous exhaust stream prior to the release to the environment. Such a system is not considered to have any effect on noble gas effluents.

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

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 A WASTE GAS HOLDUP SYSTEM shall be any system 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 of reducing the total radioactivity prior to release to the environment.

McGuire Units 1 and 2 16.3-3 Revision 48

Inventory Control-Reduced Inventory Operation 16.5.1 16.5 REACTOR COOLANT SYSTEM (RCS) 16.5.1 Inventory Control-Reduced Inventory Operation COMMITMENT A detailed review of each outage schedule that involves operation at reduced inventory shall be conducted looking at evolutions which could perturb the RCS.

AND The RCS shall be properly vented when steam generator nozzle dams are in use or the RCS cold leg is open > 1 in2 .

AND The reactor shall be subcritical for at least 7 days, or as specified in Design Study CNDS-0242, or MGDS-0228/CNDS-0218.

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

Activities that could perturb the RCS during reduced inventory operation shall require prior notification of the operations shift manager.

APPLICABILITY RCS level < 60 inches (wide range) with irradiated fuel in the core.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RCS not properly A.1 Initiate action to provide the Immediately vented, required hot leg vent path.

AND A.2 Suspend all activities that could perturb RCS level or Immediately which may reduce the reliability of the operating ND loop.

McGuire Units 1 and 2 16.5.1-1 Revision 115

Inventory Control-Reduced Inventory Operation 16.5.1 TESTING REQUIREMENTS None BASES Generic Letter 88-17 and NUREG 1410 involve concerns associated with a loss of Residual Heat Removal (RHR) during RCS reduced inventory. Numerous events have occurred in the industry that resulted in a loss of RHR during reduced inventory operation. This is of great concern due to the potential for substantial core damage occurring in a relatively short time period. This SLC depicts those commitments that are extremely important to nuclear safety, however, are not presently covered by Technical Specifications.

Under the stated APPLICABILITY, when EITHER steam generator nozzle dams are in use OR the RCS cold leg side is opened with total opening of one square inch or greater, a hot leg vent path is required. The vent path may be satisfied by:

1. Hot leg nozzle dam not installed on the vented loop AND Removal of either of the following on the vented loop:

a. Hot leg diaphragm and manway OR

b. Cold leg diaphragm and manway (with associated cold leg nozzle dam installed*)

OR

2. Reactor vessel head is removed.

Installation of the associated cold leg nozzle dam avoids diversion of makeup flow (injected via a cold leg nozzle) intended for the core.

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 Nuclear Station.

3. Integrated Scheduling Management Procedure 3.1, Outage Planning and Execution Responsibilities McGuire Units 1 and 2 16.5.1-2 Revision 115

Inventory Control-Reduced Inventory Operation 16.

5.1 REFERENCES

(continued)

4. McGuire Nuclear Station responses to GL 88-17, dated January 3, 1989, February 2, 1989, March 10, 1989 and February 24, 1993.

5. Nuclear Station Directive NSD-403, Shutdown Risk Management (Modes 4, 5, 6, and No-Mode) per 10 CFR 50. 65(a)(4).

6. Design Study CNDS-0242, Catawba and McGuire Nuclear Stations, Shutdown By Decay Heat Level Before Mid-Loop Operation, Safety Analysis, Nuclear Engineering, Nuclear Services.

7. Design Study MGDS-0228/CNDS-0218, McGuire/Catawba Nuclear Stations, Loss of Decay Heat Removal With Steam Generator Mitigation, Safety Analysis, Engineering Support Section, Design Engineering Department.

8. PIP M-08-05725.

9. PIP M-09-04863.

McGuire Units 1 and 2 16.5.1-3 Revision 115

Power Systems and Decay Heat Removal - Reduced Inventory Operation 16.5.2 16.5 REACTOR COOLANT SYSTEM 16.5.2 Power Systems and Decay Heat Removal - Reduced Inventory Operation COMMITMENT Three power sources and two decay heat removal loops shall be available consisting of:

a. Two ND pumps available with one in operation,

b. Two trains of KC and RN pumps available with flow capacity sufficient to maintain stable core exit temperature, and either

c. Two independent buslines capable of supplying the 4160 V buses via normal or standby 7 KV/4160 V transformers and one DIG capable of supplying a 4160V bus, or

d. One busline capable of supplying one 4160V bus via normal or standby 7KV/4160V transformers and two D/G's and associated 4160V buses.

APPLICABILITY: RCS level < 60 inches (wide range) with irradiated fuel in the core.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Initiate action to restore the Immediately necessary power supplies to service.

AND A.2 Suspend all activities that Immediately could perturb RCS level or which may reduce the reliability of the operating ND loop.

AND A.3 Enter applicable Immediately Conditions and Required Actions of LCO 3.4.8 and LCO 3.9.6 for inoperable ND loops.

McGuire Units 1 and 2 16.5.2-1 Revision 0

Power Systems and Decay Heat Removal - Reduced Inventory Operation 16.5.2 TESTING REQUIREMENTS None BASES Generic Letter 88-17 and NUREG 1410 involve concerns 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 reduced inventory operation.

This is of great concern due to the potential for substantial core damage occurring in a relatively short time period. This SLC depicts those commitments that are extremely important to nuclear safety, however, are not presently covered by Technical Specifications.

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 Nuclear Station.

3. Integrated Scheduling Management Procedure 3.1, Outage Planning and Execution Responsibilities

4. McGuire Nuclear Station responses to GL 88-17, dated January 3, 1989, February 2, 1989, March 10, 1989 and February 24, 1993.

5. McGuire Station Directive 3.1.3 (MSD403) - Shutdown Risk Management Guidelines.

McGuire Units 1 and 2 16.5.2-2 Revision 0

Reactivity Control - Reduced Inventory Operation 16.5.3 16.5 REACTOR COOLANT SYSTEM 16.5.3 Reactivity Control - Reduced Inventory Operation COMMITMENT The following independent sources and makeup paths of borated water must be available:

a. One high head source from NV pump train A or train B taking suction on the FWST and capable of discharging to the RCS, and

b. One low head (gravity) source supplied from the FWST to the RCS.

APPLICABILITY: RCS level < 60 inches (wide range) with irradiated fuel in the core.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Initiate action to restore the Immediately required makeup sources.

AND A.2 Suspend all activities that Immediately could perturb RCS level or which may reduce the reliability of the operating ND loop.

TESTING REQUIREMENTS None McGuire Units 1 and 2 16.5.3-1 Revision 0

Reactivity Control - Reduced Inventory Operation 16.5.3 BASES Generic Letter 88-17 and NUREG 1410 involve concerns 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 reduced inventory operation.

This is of great concern due to the potential for substantial core damage occurring in a relatively short time period. This SLC depicts those commitments that are extremely important to nuclear safety, however, are not presently covered by Technical Specifications.

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 Nuclear Station.

3. Integrated Scheduling Management Procedure 3.1, Outage Planning and Execution Responsibilities

4. McGuire Nuclear Station responses to GL 88-17, dated January 3, 1989, February 2, 1989, March 10, 1989 and February 24, 1993.

5. McGuire Station Directive 3.1.3 (MSD403) - Shutdown Risk Management Guidelines.

McGuire Units 1 and 2 16.5.3-2 Revision 0

RCS Instrumentation - Reduced Inventory Operation 16.5.4 16.5 REACTOR COOLANT SYSTEM 16.5.4 RCS Instrumentation - Reduced Inventory Operation COMMITMENT Two independent RCS level indications consisting of any valid combination of Wide Range, Narrow Range, Sightglass, Upper RVLIS (Train A/B), Lower RVLIS (Train A/B) or RCS ultrasonic (Loop A/C) level instrumentation shall be provided. The Wide Range and Narrow Range RCS level instrumentation shall have level alarm setpoints for low and high level with trend capability AND The following additional instrumentation shall be provided:

a. Two core exit thermocouples operating while the reactor vessel head is in place, or

b. Two additional independent RCS level indications consisting of any combination of Wide Range, Narrow Range, Sightglass, Lower RVLIS (Train A/B), or RCS ultrasonic (Loop A/C) level instrumentation, when the core exit thermocouples are unavailable.

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

1. If RCS sightglass is being used as one of the alternate RCS level indications, then continuously monitor RCS sightglass level indication and record level at an interval no greater than 15 minutes during normal conditions. Water level monitoring should be capable of being performed either (a) by an operator in the Control Room OR (b) from a location other than the Control Room with provision for providing immediate water level values to an operator in the Control Room if significant changes occur.

2. The OAC computer points for the required thermocouples should be used for trending and alarm.

3. Removal of the last two thermocouples shall occur no sooner than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to reactor vessel head removal. Replacement of at least two thermocouples shall occur within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after reinstalling the reactor vessel head.

APPLICABILITY: RCS level < 60 inches (wide range) with irradiated fuel in the core.

McGuire Units 1 and 2 16.5.4-1 Revision 55

RCS Instrumentation - Reduced Inventory Operation 16.5.4 REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Initiate action to restore the Immediately required instrumentation.

AND A.2 Suspend all activities that Immediately could perturb RCS level or change RCS pressure.

TESTING REQUIREMENTS None BASES Generic Letter 88-17 and NUREG-1410 involve concerns associated with a loss of Residual heat Removal during NC system reduced inventory. Numerous events have occurred in the industry that resulted in loss of residual heat removal during reduced inventory operation.

This is of great concern due to the potential for substantial core damage occurring in a relatively short time period. This SLC depicts those commitments that are extremely important to nuclear safety, however, are not presently covered by Technical Specifications.

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 Nuclear Station.

3. Integrated Scheduling Management Procedure 3.1, Outage Planning and Execution Responsibilities

4. McGuire Nuclear Station responses to GL 88-17, dated January 3, 1989, February 2, 1989, March 10, 1989 and February 24, 1993.

5. McGuire Station Directive 3.1.3 (MSD403) - Shutdown Risk Management Guidelines.

McGuire Units 1 and 2 16.5.4-2 Revision 55

Containment Closure - Reduced Inventory Operation 16.5.5 16.5 REACTOR COOLANT SYSTEM 16.5.5 Containment Closure - Reduced Inventory Operation COMMITMENT The capability to close containment following a loss of RHR shall be assured. Containment closure completion shall be achievable prior to the onset of core boiling in the event ND is lost.

APPLICABILITY: RCS level < 60 inches (wide range) with irradiated fuel in the core.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Initiate action to ensure Immediately commitment can be met.

AND A.2 Suspend all activities that Immediately could perturb RCS level or which may reduce the reliability of the operating ND loop.

TESTING REQUIREMENTS None BASES Generic Letter 88-17 and NUREG 1410 involve concerns 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 reduced inventory operation.

This is of great concern due to the potential for substantial core damage occurring in a relatively short time period. This SLC depicts those commitments that are extremely important to nuclear safety, however, are not presently covered by Technical Specifications.

McGuire Units 1 and 2 16.5.5-1 Revision 0

Containment Closure - Reduced Inventory Operation 16.

5.5 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 Nuclear Station.

3. Integrated Scheduling Management Procedure 3.1, Outage Planning and Execution Responsibilities

4. McGuire Nuclear Station responses to GL 88-17, dated January 3, 1989, February 2, 1989, March 10, 1989 and February 24, 1993.

5. McGuire Station Directive 3.1.3 (MSD403) -Shutdown Risk Management Guidelines.

McGuire Units 1 and 2 16.5.5-2 Revision 0

Safety Valves - Shutdown 16.5.6 16.5 REACTOR COOLANT SYSTEM 16.5.6 Safety Valves - Shutdown COMMITMENT One pressurizer code safety valve shall be OPERABLE with lift settings > 2435 psig and < 2559 psig.

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

The lift setting pressure shall correspond to ambient conditions of the valve at normal operating temperature and pressure.

APPLICABILITY MODE 4 with any RCS cold leg temperature < 300'F, and MODE 5.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. No pressurizer Code A.1 Suspend all operations Immediately safety valve involving positive reactivity OPERABLE. changes.

AND A.2 Place an OPERABLE RHR Immediately loop into operation in the shutdown cooling mode.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.5.6.1 Verify the required pressurizer safety valve is In accordance with OPERABLE in accordance with the Inservice Testing the Inservice Program. Following testing, lift setting shall be > 2460 Testing Program psig and < 2510 psig.

McGuire Units 1 and 2 16.5.6-1 Revision 0

Safety Valves - Shutdown 16.5.6 BASES The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig. Each safety valve is designed 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 RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization. In addition, the Overpressure Protection System provides a diverse means of protection against RCS overpressurization at low temperatures.

Demonstration of the safety valve lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code. This SLC and Technical Specification 3.4.10 allow a +3% and -2% setpoint tolerance is allowed for OPERABILITY. The valves are reset to +1% during surveillance testing to allow for drift.

REFERENCES

1. ASME Boiler and Pressure Code,Section XI McGuire Units 1 and 2 16.5.6-2 Revision 0

Chemistry 16.5.7 16.5 REACTOR COOLANT SYSTEM 16.5.7 Chemistry COMMITMENT The Reactor Coolant System chemistry shall be maintained within the limits specified in Table 16.5.7-1.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more chemistry A. 1 Restore the parameter to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months parameters in excess of within steady state limit.

its Steady State Limit but within its Transient Limit in MODE 1, 2, 3, or 4.

B. One or more chemistry B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months parameters in excess of its Transient Limit in AND MODE 1, 2, 3, or 4.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months OR Required Action and associated Completion Time of Condition A not met.

(continued)

McGuire Units 1 and 2 16.5.7-1 Revision 53

Chemistry 16.5.7 REMEDIAL ACTIONS (continued)

C. --------- NOTE ------------ C.1 Initiate action to reduce the Immediately All Required Actions pressurizer pressure to <

must be completed 500 psig.

whenever this Condition is entered.

AND RCS chloride or fluoride C.2 Perform an engineering Prior to increasing concentration not within evaluation to determine the the pressurizer the Steady State Limits effects of the out-of-limit pressure > 500 psig for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in condition on the structural any condition other than integrity of the RCS. OR MODES 1, 2, 3 and 4.

Prior to entry to OR MODE 4 AND RCS chloride or fluoride concentration not within C.3 Determine that the RCS Prior to increasing Transient Limits in any remains acceptable for the pressurizer condition other than continued operation. pressure > 500 psig MODES 1, 2, 3 and 4.

OR Prior to entry to MODE 4 TESTING REQUIREMENTS TEST FREQUENCY TR 16.5.7.1 Verify RCS chemistry is within limits. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months McGuire Units 1 and 2 16.5.7-2 Revision 53

Chemistry 16.5.7 TABLE 16.5.7-1 RCS CHEMISTRY LIMITS STEADY-STATE TRANSIENT STEADY-STATE TRANSIENT PARAMETER LIMIT LIMIT Dissolved Oxygen (1) < 0.10 ppm < 1.00 ppm Chloride <0.15 ppm < 1.50 ppm Fluoride

0.15 ppm < 1.50 ppm Notes:

1. Oxygen limit and associated Testing Requirement not applicable with Tavg < 250 OF.

McGuire Units 1 and 2 16.5.7-3 Revision 53

Chemistry 16.5.7 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 structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride, and fluoride limits are time and temperature dependent. Corrosion studies show that 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 excess of the limits will be detected in sufficient time to take remedial action.

REFERENCES

1. McGuire Nuclear Station UFSAR, Section 18.2.4, Chemistry Control Program.

2. McGuire License Renewal Commitments MCS-1274.00-00-0016, Section 4.6, Chemistry Control Program.

McGuire Units 1 and 2 16.5.7-4 Revision 53

Pressurizer 16.5.8 16.5 REACTOR COOLANT SYSTEM 16.5.8 Pressurizer COMMITMENT The pressurizer temperature shall be limited to:

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

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

c. A maximum spray water temperature differential of 3201F.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. --------- NOTE ----------- A.1 Restore pressurizer 30 minutes All Required Actions temperature to within limits.

must be completed whenever this Condition AND is entered.

A.2 Perform engineering 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Pressurizer temperature evaluation to determine not within limits, effects of the out-of-limit condition on the structural integrity of the pressurizer.

AND A.3 Determine that the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months pressurizer remains acceptable for continued operation.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Reduce pressurizer 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months pressure to < 500 psig.

McGuire Units 1 and 2 16.5.8-1 Revision 0

Pressurizer 16.5.8 TESTING REQUIREMENTS TEST FREQUENCY TR 16.5.8.1 -NOTE ------------------

Only required to be performed during system heatup or cooldown operations.

Verify pressurizer temperatures are within limits. 30 minutes TR 16.5.8.2 ------------------- NOTE ------------------

Only required to be performed during auxiliary spray operations.

Verify spray water temperature differential within limit. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , BASES The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:

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

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

McGuire Units 1 and 2 16.5.8-2 Revision 0

Structural Integrity 16.5.9 16.5 REACTOR COOLANT SYSTEM 16.5.9 Structural Integrity COMMITMENT The structural integrity of ASME Code Class 1, 2 and 3 components shall be maintained.

APPLICABILITY: All MODES.

NOTE Separate Condition entry is allowed for each component.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Structural integrity of any A. 1 Restore structural integrity Prior to increasing ASME Code Class 1 of affected component(s) RCS temperature >

component(s) not within to within limit. 50°F above minimum limits. temperature required by NDT considerations OR A.2 Isolate the affected Prior to increasing component(s) from RCS temperature >

service. 50°F above minimum temperature required by NDT considerations McGuire Units 1 and 2 16.5.9-1 Revision 108

Structural Integrity 16.5.9 B. Structural integrity of any B.1 Enter the Operability Immediately ASME Code Class 2 or 3 Determination process to component(s) not within promptly confirm that limits. Structural Integrity is still maintained in the degraded or non-conforming condition.

OR B.2 Declare the affected Immediately component(s) inoperable.

__________________________________________________________ [ ___________________________________________________________ _____________________________________ 1~

TESTING REQUIREMENTS TEST FREQUENCY TR 16.5.9.1 Verify the structural integrity of ASME Code Class 1, 2, and In accordance with 3 components is in accordance with the Inservice the Inservice Inspection Program. Inspection Program BASES The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant.

These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a(g)(6)(i).

UFSAR Section 3.2, "Classification of Structures, Systems, and Components," defines and correlates code classes, safety classes, and Duke piping classes. In general, Class 1 applies to the reactor coolant pressure boundary. Class 2 applies to safety systems such as emergency core cooling, containment heat removal and cleanup, and containment isolation valves. Class 3 applies to the auxiliary feedwater system, spent fuel cooling, and air cleanup systems like control room ventilation. A complete listing of the Class 1, 2, and 3 systems and components are contained in the UFSAR Chapter 3 Tables.

This SLC applies to one or more ASME Code Class 1, 2, or 3 components in a degraded or nonconforming condition. A degraded or nonconforming condition of a system, structure, or component (SSC) is a condition in which there has been any loss of required quality or functional capability, failure to meet requirements of the regulations, or failure to conform to McGuire Units 1 and 2 16.5.9-2 Revision 108

Structural Integrity 16.5.9 all aspects of the licensing basis. Degraded or nonconforming conditions of SSCs discovered during the conduct of inservice inspections, maintenance or refueling activities, or during plant operation shall be evaluated to determine the affect on structural integrity.

References 4, 5, and 6 contain guidance and evaluation methods to be used in determining structural integrity and operability for ASME Code Class 1, 2, or 3 components. Structural integrity and operability may be restored by repair, replacement, or modification in accordance with ASME Section XI., In some cases, an operability determination may be an acceptable method of confirmation of operability. If structural integrity cannot be established promptly or the results are indeterminate, the component shall be declared inoperable and the appropriate Technical Specification action statement entered.

The NOTE above the REMEDIAL ACTIONS clarifies the application of this SLC. The CONDITIONS of this SLC may be entered independently for each component in a degraded or nonconforming condition.

BASES (continued)

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

REFERENCES

1. ASME Boiler and Pressure Vessel Code,Section XI, 1971 Edition and Addenda through Winter 1972.

2. 10 CFR 50.55a (g), Inservice Inspection Requirements.

3. UFSAR Section 3.2, Classification of Structures, Systems, and Components.

4. NRC RIS 2005-20, Revision 1, Revision to NRC Inspection Manual 9900 Technical Guidance, "Operability Determinations and Functionality Assessments for Resolution of Degraded and Nonconforming Conditions."

5. NRC Generic Letter 90-05, Guidance for Performing Temporary Non-code Repairs of ASME Code Class 1, 2, and 3 Piping.

6. NRC Regulatory Guide 1.147, lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1.

7. NSD 203, "Operability / Functionality".

McGuire Units 1 and 2 16.5.9-3 Revision 108

Reactor Vessel Head Vent System 16.5.10 16.5 REACTOR COOLANT SYSTEM 16.5.10 Reactor Vessel Head Vent System COMMITMENT Two reactor vessel head vent paths, each consisting of two valves in series powered from emergency buses, shall be OPERABLE and closed.

-NOTE Reactor head vent system may be aligned to support events where normal and excess letdown are unavailable.

APPLICABILITY MODES 1, 2, 3, and 4.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One reactor vessel head A.1 Initiate action to close and Immediately vent path inoperable, remove power from all valves in the inoperable flow path.

AND A.2 Restore the inoperable vent 30 days path to OPERABLE status.

B. Two reactor vessel head B.1 Initiate action to close and Immediately vent paths inoperable, remove power from all valves in the inoperable flow paths.

AND B.2 Restore at least one 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable vent path to OPERABLE status.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months McGuire Units 1 and 2 16.5.10-1 Revision 38

Reactor Vessel Head Vent System 16.5.10 TESTING REQUIREMENTS TEST FREQUENCY TR 16.5.10.1 Cycle each valve in the vent through one complete cycle 18 months of full travel from the control room during MODE 5 or 6.

TR 16.5.10.2 Verify flow through vent paths by venting during MODE 5 18 months or 6.

BASES Reactor Vessel Head Vents are provided to exhaust non-condensable gases from the primary system that could inhibit natural circulation core cooling. The Reactor Vessel Head Vent System further functions to provide inventory control for standby shutdown facility events ('A train only) and for events for which normal and excess letdown are unavailable.

The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function.

(Operability of the pressurizer steam space vent path is provided by ITS 3.4.11 and 3.4.12).

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

The Testing Requirement to verify Reactor Vessel Head Vent flowpath is qualitative as no specific size or flow rate is required to exhaust non-condensable gases. The function, capabilities, and testing requirements of the reactor coolant system vent systems are consistent with the requirements of Item Il.B.1 of NUREG-0737, "Clarification of TMI Action Plan Requirements", November 1980.

REFERENCES

1. NUREG-0737, "Clarification of TMI Action Plan Requirements", Item lI.B.1, November 1980.

2. PIP M97-3795.

3. PIP M00-0201.

McGuire Units 1 and 2 16.5.10-2 Revision 38

Containment Sump 16.6.1 16.6 ENGINEERED SAFETY FEATURES 16.6.1 Containment Sump COMMITMENT The containment sump shall be maintained free of loose debris.

APPLICABILITY MODES 1, 2, 3, and 4.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Debris found in A. 1 Remove debris from Prior to final exit from containment or in containment, containment containment sump.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.6.1.1 Perform a visual inspection for all accessible areas of the Prior to containment and verify no loose debris is present that establishing could be transported to the sump.. containment integrity' TR 16.6.1.2 Perform a visual inspection for affected areas of the At completion of containment and verify no loose debris is present that each containment could be transported to the sump. entry after containment integrity is established McGuire Units 1 and 2 16.6.1-1 Revision 0

Containment Sump 16.6.1 BASES Removal of identified debris from containment or from the containment sump is critical to the function of ECCS systems during the cold leg recirculation phase following a safety injection. Any loose debris (rags, trash, clothing, etc.) left in containment could be transported to the strainers in the containment sump during LOCA conditions causing a loss of suction or restriction of the ECCS pumps.

REFERENCES None.

McGuire Units 1 and 2 16.6.1-2 Revision 0

Inlet Door Position Monitoring System 16.6.3 16.6 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 ACTIONS CONDITION REQUIRED ACTION' COMPLETION TIME A. Inlet Door Position A.1.1 Verify the Ice Bed Immediately Monitoring System Temperature Monitoring inoperable. System is OPERABLE.

AND A.1.2 Verify ice bed temperature Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months

< 27 0 F.

AND A.1.3 Restore the Inlet Door 14 days Position Monitoring System to OPERABLE status.

OR A.2 Restore the Inlet Door 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months Position Monitoring System to OPERABLE status.

B. Required Action and B. 1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months McGuire Units 1 and 2 16.6.3-1 Revision 61

Inlet Door Position Monitoring System 16.6.3 TESTING REQUIREMENTS TEST FREQUENCY TR 16.6.3.1 Perform a CHANNEL CHECK. 7 days AND Once within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after receiving a door open alarm TR 16.6.3.2 Perform a TADOT. 18 months TR 16.6.3.3 Verify the Inlet Door Position Monitoring System correctly 18 months, when indicates'the status of each inlet door. each door is opened and reclosed during testing per TS 3.6.13 BASES The OPERABILITY of the Inlet Door Position Monitoring System ensures that the capability is available for monitoring the individual inlet door position. In the event the system is inoperable, the ACTION requirements provide assurance that the ice bed heat removal capacity will be retained within the specified limits.

TS 3.6.12 requires that the Ice Bed temperature be maintained less than or equal to 27 0 F.:

If the Ice Bed temperature chart recorder is out of service, or otherwise incapable of performing its design function, then alternate means, either independently or in combination, can be used to satisfy the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance of SR 3.6.12.1. The alternate means have proactive components (if any channel is failed or reads greater than 50 F of its associated channel) and includes the following:

1. Obtain manual temperature readings from multiplexer.

2. Use calibrated instrumentation to measure Ice Bed area/basket temperature.

3. Use calibrated thermography gun to measure ice bed area/basket temperature.

REFERENCES None.

McGuire Units 1 and 2 16.6.3-2 Revision 61

Safety Injection System (NI) 16.6.4 16.6. ENGINEERED SAFETY FEATURES 16.6.4 Safety Injection System Nozzles COMMITMENT: Reporting of ECCS Injections and Nozzle Usage Factors APPLICABILITY: MODES 1, 2, and 3.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME The ECCS is actuated and A Special Report shall be 90 days injects water into the Reactor prepared and submitted to Coolant System and the the NRC describing the current value of the usage circumstances of the factor for an affected safety actuation and the total injection nozzle exceeds accumulated actuation 0.70. cycles to date. The current value of the usage factor for each affected Safety Injection nozzle shall be provided.

TESTING REQUIREMENTS None BASES Duke letter, "Justification for Continued Operation with Seven Thermal Sleeves Removed,"

dated December 14, 1983, provided the results of evaluations performed by Westinghouse and Duke as a basis for continued operation of McGuire Units 1 and 2 without thermal sleeves installed in the reactor coolant system nozzles. The NRC's letter and safety evaluation (SER) dated December 30, 1986, concluded that continued operation was acceptable and that the requirements of License Condition (LC) 2.C.(4) had been completed. The NRC staff's acceptance recognized that McGuire's Technical Specification 3/4.5.2 required reporting of the usage factor of each nozzle if the value exceeded 0.70 and the ECCS actuated and injected water into the reactor coolant system. During implementation of Improved Technical Specifications, this reporting requirement was inadvertently removed.

McGuire Units 1 and 2 16.6.4-1 Revision 27

Safety Injection System (NI)

.16.6.4 By letter dated June 13, 2000, as supplemented on August 20, 2001 and September 10, 2001, McGuire submitted a proposed License Amendment Request (LAR). This purpose of this LAR was to delete LCs that had previously been completed. In the September 10, 2001 letter, McGuire committed to ensure proper notification is accomplished in the event the usage factor of affected safety injection nozzles exceeds the 0.70 value following an ECCS actuation and injection of water into the reactor coolant system. The NRC issued approved License Amendment 200/181 by letter dated December 5, 2001.

REFERENCES 1, McGuire License Amendment Request dated June 13, 2000, as supplemented by letters August 30, 2001 and September 10, 2001.

2.' NRC letter dated December 5, 2001, McGuire Nuclear Station, Units 1 and 2 Issuance of Amendments RE: License Conditions (TAC NOS. MA9297 and MA9298)/

McGuire Units 1 and 2 16.6.4-2 Revision 27

ATWS/AMSAC 16.7.1 16.7 INSTRUMENTATION 16.7.1 ATWS/AMSAC COMMITMENT The ATWS/AMSAC system shall be OPERABLE.

APPLICABILITY: MODE 1 above 40% RTP.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. ATWS/AMSAC system A.1 Restore ATWS/AMSAC 7 days inoperable, system to OPERABLE status.

OR A.2 Prepare and submit a 37 days Special Report outlining the cause of the malfunction and plans for restoring the system to OPERABLE status.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.7.1.1 Perform a CHANNEL OPERATIONAL TEST on the 18 months.

ATWS/AMSAC system.

McGuire Units 1 and 2 16.7.1-1 Revision 0

ATWS/AMSAC 16.7.1 BASES None REFERENCES

1. Final Design Description, ATWS Mitigation System Activation Circuitry, "AMSAC" Original Issue January 23, 1987, as revised.

McGuire Units 1 and 2 16.7.1-2 Revision 0

Seismic Instrumentation 16.7.2 16.7 INSTRUMENTATION 16.7.2 Seismic Instrumentation COMMITMENT The seismic monitoring instrumentation shown in Table 16.7.2-1 shall be OPERABLE.

APPLICABILITY: At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more seismic A. 1 Restore inoperable 30 days monitoring instruments instrument to OPERABLE inoperable, status.

OR A.2 Prepare and submit a 40 days Special Report outlining the cause of the malfunction and plans for restoring the instrument(s) to OPERABLE status.

B. Seismic monitoring B. 1 Restore instrument to Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months instruments actuated OPERABLE status. following the during a seismic event > seismic event 0.01g. AND B.2 Retrieve data from Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months accessible actuated following the instruments and analyze to seismic event determine magnitude of vibratory ground motion.

AND (continued)

McGuire Units 1 and 2 16.7.2-1 Revision 80

Seismic Instrumentation 16.7.2 REMEDIAL ACTIONS (continued)

B. (continued) B.3 Prepare and submit a 10 days Special Report describing the magnitude, frequency spectrum, and resultant effect upon facility features important to safety.

TESTING REQUIREMENTS

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

Refer to Table 16.7.2-1 to determine which TRs apply for each Seismic Instrumentation.

TEST FREQUENCY TR 16.7.2.1 ---------------- NOTE ----------------------------------------

CHANNEL CHECK does not include 1IEEVD 1030 or 11EEVD 1040.

Perform CHANNEL CHECK. 31 days TR 16.7.2.2 Perform CHANNEL OPERATIONAL TEST. 6 months TR 16.7.2.3 Perform a CHANNEL CALIBRATION. 18 months McGuire Units 1 and 2 16.7.2-2 Revision 80

Seismic Instrumentation 16.7.2 TABLE 16.7.2-1 SEISMIC MONITORING INSTRUMENTATION INSTRUMENTS AND SENSOR MEASUREMENT REQUIRED TESTING LOCATIONS RANGE CHANNELS REQUIREMENTS

1. Triaxial Accelerographs 1.a 1IEEVD 1020 (Remote 0-2g 1 TR 16.7.2.1 Sensor A) Unit 1 TR 16.7.2.2 Containment Base Slab TR 16.7.2.3 1.b 11EEVD 1010 (Remote 0-2 g 1 TR 16.7.2.1 Sensor B) Unit 1 CA TR 16.7.2.2 Pump Room, TR 16.7.2.3 Elev. 716' - 0" 1.c llEEVD 1000 Control 0-2 g 1* TR 16.7.2.1 Room TR 16.7.2.2 TR 16.7.2.3 1.d 11EEVD 1030 Unit 1 0-2g 1 TR 16.7.2.2 Containment Bldg. Elev. TR 16.7.2.3 825' - 4", 00 1.e 1IEEVD 1040 Unit 1 0-2g 1 TR 16.7.2.2 Containment Bldg. Elev. TR 16.7.2.3 784'- 10". 00

With control room indication.

BASES 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. The capability is required to permit comparison of the measured response to that used in the design basis for the 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 for Earthquakes," Revision 2.

The seismic system records seismic data acquired by three MEMS triaxial accelerograph sensors connected to a network control center located in the control room. Two additional MEMS triaxial accelerograph sensor/recorders are located in Unit 1 containment. These sensor/recorders are stand-alone units and gather additional information during a seismic event. All of the MEMS triaxial sensor/recorders store seismic data in SRAM (Static random access memory) and can download it to a computer for analysis. Subsequent seismic events are likewise automatically captured and made available for data analysis.

The network control center (NCC2002, 11EECS 1000) in the control room provides online monitoring and data retrieval for the three MEMS triaxial accelerometers (sensors) connected to it (11IEEVD 1000, 1010, 1020). The NCC also continuously self-checks all McGuire Units 1 and 2 16.7.2-3 Revision 80

Seismic Instrumentation 16.7.2 BASES (continued) significant functions and power supply status on-line. Any failure will result in an alarm condition on the NCC and a remote alarm sent to the Unit 1 OAC.

The two stand-alone MEMS triaxial accelerograph sensor/recorders located in Unit 1 containment (1IEEVD 1030, 1040) will require the captured seismic data be downloaded to a computer manually, when they become accessible, after a seismic event. They are not required for the channel check every 31 days per TR 16.7.2.1 because they are not needed to detect a seismic event.

REFERENCES

1. Regulatory Guide 1.12, Instrumentation for Earthquakes, Revision 2.

2. 10 CFR Part 100, Appendix A.

McGuire Units 1 and 2 16.7.2-4 Revision 80

Meteorological Instrumentation 16.7.3 16.7 INSTRUMENTATION 16.7.3 Meteorological Instrumentation COMMITMENT The meteorological monitoring instrumentation channels shown in Table 16.7.3-1 shall be OPERABLE.

APPLICABILITY: At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A. 1 Restore inoperable 7 days Meteorological channel(s) to OPERABLE monitoring channels status.

inoperable.

OR A.2 Prepare and submit a 17 days Special Report outlining the cause of the malfunction and plans for restoring the channel(s) to OPERABLE status.

TESTING REQUIREMENTS

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

Refer to Table 16.7.3-1 to determine which TRs apply for each meteorological monitoring instrumentation.

TEST FREQUENCY TR 16.7.3.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TR 16.7.3.2 Perform a CHANNEL CALIBRATION. 6 months.

McGuire Units 1 and 2 16.7.3-1 Revision 0

Meteorological Instrumentation 16.7.3 TABLE 16.7.3-1 METEOROLOGICAL MONITORING INSTRUMENTATION INSTRUMENT AND LOCATION REQUIRED TESTING CHANNELS REQUIREMENTS

1. Wind Speed 1.a Meteorological Tower Nominal 1 TR 16.7.3.1 Elev. 801.88' TR 16.7.3.2 1.b Meteorological Tower Nominal 1 TR 16.7.3.1 Elev. 964.88' TR 16.7.3.2

2. Wind Direction 2.a Meteorological Tower Nominal 1 TR 16.7.3.1 Elev. 801.88' TR 16.7.3.2 2.b Meteorological Tower Nominal 1 TR 16.7.3.1 Elev. 964.88' TR 16.7.3.2

3. Air Temperature - Delta T 3.a Meteorological Tower Nominal 1 TR 16.7.3.1 Elev. 798.46' - 962.22' TR 16.7.3.2 McGuire Units 1 and 2 16.7.3-2 Revision 0

Meteorological Instrumentation 16.7.3 BASES 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 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.

REFERENCES

1. Regulatory Guide 1.23, "Onsite Meteorological Programs," February 1972.

McGuire Units 1 and 2 16.7.3-3 Revision 0

Loose-Part Detection System 16.7.4 16.7 INSTRUMENTATION 16.7.4 Loose-Part Detection System COMMITMENT The Loose-Part Detection System shall be OPERABLE.

APPLICABILITY: MODES 1 and 2 REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore inoperable 30 days Loose-Part Detection channel(s) to OPERABLE System monitoring status.

channels. inoperable.

OR A.2 Prepare and submit a 40 days Special Report outlining the cause of the malfunction and plans for restoring the channel(s) to OPERABLE status.

TESTING REQUIREMENTS

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

The following Testing Requirements are only required for one of the three channels in each monitored area, i.e., reactor lower vessel, reactor upper vessel, and each steam generator.

TEST FREQUENCY TR 16.7.4.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TR 16.7.4.2 Perform CHANNEL OPERATIONAL TEST, excluding 31 days setpoint verification.

TR 16.7.4.3 Perform a Ci-IANNEL CALIBRATION. 18 months.

McGuire Units 1 and 2 16.7.4-1 Revision 1

Loose-Part Detection System 16.7.4 BASES The OPERABILITY of the loose-part detection instrumentation ensures that 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 Testing 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.

The Testing Requirements on the Loose-Part Detection System are only required on one of the three channels associated with the reactor lower vessel area (channel 1, 2, or 3), one of the three channels associated with the reactor upper vessel area (channel 4, 5, or 6), and one channel associated with each steam generator (channel 8, 9, or 10 for SG-A, channel 12, 13, or 14 for SG-B, channel 16, 17, or 18 for SG-C, and channel 20, 21, or 22 for SG-D) during each required performance.

REFERENCES

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

McGuire Units 1 and 2 16.7.4-2 Revision 1

Turbine Overspeed Protection 16.7.5 16.7 INSTRUMENTATION 16.7.5 Turbine Overspeed Protection COMMITMENT At least one Turbine Overspeed Protection System shall be OPERABLE.

APPLICABILITY MODE 1.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One stop valve or one A.1 Restore inoperable valve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months governor valve per high OPERABLE status.

pressure turbine steam lead Inoperable. OR A.2 Close at least one valve in 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months the affected s!eam lead(s).

OR A.3 Isolate the turbine from the 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months steam supply.

B. One reheat stop valve or B.1 Restore inoperable valve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months one reheat intercept OPERABLE status.

valve per LP turbine steam lead inoperable. OR B.2 Close at least one valve in 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months the affected steam lead(s).

OR B.3 Isolate the turbine from the 78 hours9.027778e-4 days 0.0217 hours 1.289683e-4 weeks 2.9679e-5 months steam supply.

(continued)

McGuire Units 1 and 2 16.7.5-1 Revision 0

Turbine Overspeed Protection J 16.7.5 REMEDIAL ACTIONS (continued)

C. Turbine Overspeed C. 1 Isolate the turbine from the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Protection System steam supply.

otherwise inoperable.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.7.5.1 Perform inservice inspection in accordance with the In accordance with Turbine Overspeed Reliability Program. the Turbine Overspeed Reliability Program McGuire Units 1 and 2 16.7.5-2 Revision 0

Turbine Overspeed Protection 16.7.5 BASES 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 turbine overspeed of the turbine could generate potentially damaging missiles. All Category I structures except the new fuel vault at McGuire, are designed to withstand effects of turbine missiles without any adverse impact on the safety related equipment housed inside (UFSAR 3.5.2.7 and 10.2.3). To assure protection against turbine overspeed a "Turbine Overspeed Reliability Program" is implemented. Tests and inspections associated with this program will be performed in accordance with station procedures, maintenance work requests and/or outage work schedules as appropriate. All deviations from the program or deficiencies identified through the specified maintenance, calibration, or testing activities are evaluated by Duke Power Company to determine if operability of the system has been affected and appropriate action taken such as correcting the deviation or deficiency, performing compensatory action, or removing the turbine from service.

REFERENCES

1. McGuire Nuclear Station UFSAR Section 3.5.2.7 and 10.2.3 McGuire Units 1 and 2 16.7.5-3 Revision 0

Radiation Monitoring for Plant Operations 16.7.6 16.7 INSTRUMENTATION 16.7.6 Radiation Monitoring for Plant Operations COMMITMENT The radiation monitoring instrumentation channels shown in Table 16.7.6-1 shall be FUNCTIONAL.

APPLICABILITY As shown in Table 16.7.6-1.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more radiation A.1 Adjust setpoint to within the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months monitoring channels limit.

Alarm/Trip setpoint exceeding value shown OR in Table 16.7.6-1.

A.2 Declare the channel non- 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months functional.

B. One Containment B.1 Verify containment purge Immediately Atmosphere Gaseous system (VP) valves are Radioactivity monitoring maintained closed.

channel non-functional.

C. One or more Control C.1 Monitor at alternate Immediately Room Air Intake representative location Radioactivity monitoring (Unit Vent).

channel non-functional.

OR C.2 If Unit Vent monitor non- In accordance with functional, obtain grab SLC 16.11.7.

samples from Unit Vent in accordance with SLC 16.11.7.

(continued)

McGuire Units 1 and 2 16.7.6-1 Revision 103

Radiation Monitoring for Plant Operations 16.7.6 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more required D.1 Suspend all fuel movement Immediately channels for Spent Fuel operations in the fuel Handling Area, Reactor handling area being Building Fuel Handling monitored until Required Area or New Fuel Vault Acton D.2 is completed.

Fuel Handling Area Radiation Monitors non- AND functional.

D.2.1 Provide a portable Immediately continuous monitor with same Alarm Setpoint.

OR D.2.2 Provide RP continuous Immediately dose rate monitoring.

AND D.3 Restore non-functional 30 days monitors to FUNCTIONAL status.

E. One Spent Fuel Pool E.1 Verify the Fuel Handling Immediately Radioactivity monitoring Ventilation System (VF) channel non-functional. requirements in Technical Specification 3.7.12 are met.

F. Condenser Evacuation F.1 Ensure that all N-16 Immediately System Noble Gas Leakage Monitor (EMF-71, Activity Monitor (EMF- 72, 73, & 74) channels are

33) non-functional. FUNCTIONAL.

G. One or more N-16 G.1 Ensure that the Condenser Immediately Leakage Monitor (EMF- Evacuation System Noble 71, 72, 73, & 74) Gas Activity Monitor (EMF-channels non- 33) is FUNCTIONAL.

functional.

(continued)

McGuire Units 1 and 2 16.7.6-2 Revision 103

Radiation Monitoring for Plant Operations 16.7.6 CONDITION REQUIRED ACTION COMPLETION TIME H. Condenser Evacuation H.1 Initiate action to restore Immediately System Noble Gas online radiation monitor to Activity Monitor (EMF- FUNCTIONAL.

33) non-functional.

AND AND H.2 Perform TS-SR 3.4.13.2. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months One or more N-16 Leakage Monitor (EMF-71, 72, 73, & 74) channels non-functional.

TESTING REQUIREMENTS

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

Refer to Table 16.7.6-1 to determine which TRs apply for each Radiation Monitoring channel.

TEST FREQUENCY TR 16.7.6.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months TR 16.7.6.2 Perform CHANNEL OPERATIONAL TEST. 92 days TR 16.7.6.3 Perform CHANNEL OPERATIONAL TEST. 184 days TR 16.7.6.4 Perform a CHANNEL CALIBRATION. 18 months McGuire Units 1 and 2 16.7.6-3 Revision 103

Radiation Monitoring for Plant Operations 16.7.6 TABLE 16.7.6-1 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATION APPLICABLE REQUIRED ALARM/TRIP TESTING MONITOR MODES CHANNELS SETPOINT REQUIREMENTS

1. Containment 1,2,3,4,5,6 1 Must meet SLC TR 16.7.6.1 Atmosphere Gaseous 16.11-6 TR 16.7.6.2 Radioactivity-High (Low limits TR 16.7.6.4 Range EMF-39)

2. Spent Fuel Pool With irradiated 1 < 1.7 x 10-4 TR 16.7.6.1 Radioactivity-High (EMF- fuel in fuel pCi/ml TR 16.7.6.2

42) storage TR 16.7.6.4 areas or fuel building

3. Spent Fuel Handling With fuel in fuel 1 < 15 mR/hr TR 16.7.6.1 Area Radiation Monitor storage See Note (b) TR 16.7.6.3 (1EMF-17, 2EMF-4) areas or fuel TR 16.7.6.4 building

4. Reactor Building Fuel 6 1 < 15 mR/hr TR 16.7.6.1 Handling Area Radiation See Note (b) TR 16.7.6:3 Monitor (1EMF-16, TR 16.7.6.4 2EMF-3)

5. New Fuel Vault Fuel With fuel in New 1 < 15 mR/hr TR 16.7.6.1 Handling Area Radiation Fuel Vault See Note (b) TR 16.7.6.3 Monitors TR 16.7.6.4 (1EMF-20, 1EMF-21, 2EMF-7, 2EMF-8)

6. Control Room Air Intake 1,2,3,4,5,6 2 per station. < 3.4 x 10-4 TR 16.7.6.1 Radioactivity-High (EMF- pCi/ml TR 16.7.6.2 43a and 43b) TR 16.7.6.4

7. Condenser Evacuation 1 1 See Note (a) TR 16.7.6.1 System Noble Gas TR 16.7.6.3 Activity Monitor (EMF-33) TR 16.7.6.4

8. N-16 Leakage Monitor 1 (40-100% 4 (1/steamline) See Note (a) TR 16.7.6.1 (EMF-71, 72, 73 & 74) reactor TR 15.7.6.3 power) TR 16.7.6.4 (a) The setpoint is as required by the primary to secondary leak rate monitoring program.

(b) Setpoint can be elevated above 15 mR/hr based upon direction from approved station procedures.

McGuire Units 1 and 2 16.7.6-4 Revision 103

Radiation Monitoring for Plant Operations 16.7.6 BASES The FUNCTIONALITY of the radiation monitoring instrumentation for plant operations ensures that: (1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its setpoint, (2) the specified coincidence logic is maintained, and (3) sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. The radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions.

Once the required logic combination is completed, the system sends actuation signals to initiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems.

Remedial Action C.

Control room air intake radioactivity monitoring channels (EMF-43a and 43b) are noble gas beta detection channels. The alternate representative monitoring locations are both unit vent noble gas monitors 1EMF-36L and 2EMF-36L with continuous indication below TRIP 2 ALARM setpoint. During periods of non-functionality of either 1EMF-36L or 2EMF-36L, the alternate monitoring is satisfied by grab sampling of the unit vent in accordance with SLC 16.11.7.

Remedial Action D.

Fuel assemblies are stored and handled in areas of the plant discussed below. Radiation monitoring is provided for these areas to detect excessive radiation levels and will provide an alarm to alert personnel if a potential radiation hazard is present.

1. Unit 1 and 2 Spent Fuel Pool; includes the cask pool area, the new fuel elevator, the fuel transfer tube area and the spent fuel storage are/racks.

2. Unit 1 and 2 Reactor Building; includes the fuel transfer tube area, the reactor core and the refueling canal.

3. Unit 1 and 2 Fuel Building; includes the new fuel vault area.

Performance of Required Acton D.1 shall not preclude completion of movement of a component to a safe position. When a fuel handling area radiation monitor channel becomes non-functional, an alternate means is required for determining dose rate and alerting individuals to excessive radiation levels. This can be accomplished by either a portable monitor with same alarm setpoint located within the area monitored by the inoperable channel or using Radiation Protection personnel performing continuous monitoring of area dose rate using a hand-held dose rate meter. This hand-held meter will not provide an alarm, but relies upon RP personnel to alert individuals of excessive radiation levels.

Certain evolutions may result in a higher gamma dose rate field, resulting in the need to adjust the alarm setpoint above the nominal alarm/trip setpoint (15 mR/hr). An approved station procedure controls adjustment of this setpoint to a higher value that still ensures individuals are alerted to the presence of excessive radiation levels.

McGuire Units 1 and 2 16.7.6-5 Revision 103

Radiation Monitoring for Plant Operations 16.7.6 Remedial Action F, G and H.

The condenser evacuation system noble gas activity monitor (EMF-33) and main steam line N-16 monitors (EMF- 71, 72, 73, & 74) are used for online monitoring of primary-to-secondary leak rate. These radiation monitors provide the preferred means to accomplish Technical Specification Surveillance SR 3.4.13.2 while in Mode 1. For the condenser evacuation system noble gas activity monitor (EMF-33) or main steam line N-16 monitor to be considered functional for primary to secondary leakage monitoring the monitor must be sensitive to at least 30 gallons per day (GPD) leakage rate.

REFERENCES

1. Technical Specification 3.4.13 - RCS Operational Leakage.

2. NSD-513 - Primary to Secondary Leak Monitoring Program, Revision 5.

3. 10CFR50.68 - Criticality Accident Requirements

4. Duke letter dates July 29, 2004 - RAI Response, TS 3.7.15 and TS 4.3 Changes.

5. NRC Safety Evaluation Report dated March 17, 2005 - Amendments Nos. 225/207

6. MCS-1 578-VC-00-0001, DBD for VC/YC System

7. MCTC-1578-VC-R001-001, TAC for Radiation Monitors EMF-43a and EMF-43b McGuire Units 1 and 2 16.7.6-6 Revision 103

Movable Incore Detectors 16.7.7 16.7 INSTRUMENTATION 16.7.7 Movable Incore Detectors COMMITMENT The Movable Incore Detection System 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 When the Movable Incore Detection System is used for:

a. Recalibration of Excore Neutron Flux Detection System,

b. Monitoring the QUADRANT POWER TILT RATIO, or

c. Measurement of FQ(Z) and F NAH.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Movable Incore A.1 Suspend use of the Immediately Detection System Movable Incore Detection inoperable. System for the applicable monitoring or calibration functions.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.7.7.1 Normalize each movable incore detector output. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months McGuire Units 1 and 2 16.7.7-1 Revision 0

Movable Incore Detectors 16.7.7 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.

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

The Testing Requirements require that each detector be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing detector output when the system is required for the specified activities.

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

REFERENCES

1. WCAP-8648, June 1976.

McGuire Units 1 and 2 16.7.7-2 Revision 0

Explosive Gas Monitoring Instrumentation 16.7.8 16.7 INSTRUMENTATION 16.7.8 Explosive Gas Monitoring, Instrumentation COMMITMENT One hydrogen monitor and two oxygen monitors for the in-service hydrogen recombiner train shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of SLC 16.11.19 are not exceeded.

APPLICABILITY During WASTE GAS HOLDUP SYSTEM operation.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required hydrogen A.1 Suspend oxygen supply to Immediately monitor inoperable, the recombiner.

AND A.2 Restore the required 14 days hydrogen monitor to OPERABLE status.

B. One required oxygen B.1 Restore the required 14 days monitor inoperable, oxygen monitor to OPERABLE status.

(continued)

McGuire Units 1 and 2 16.7.8-1 Revision 77

Explosive Gas Monitoring Instrumentation 16.7.8 CONDITION REQUIRED ACTION COMPLETION TIME C. Two required oxygen C.1.1 Suspend oxygen supply to Immediately monitors inoperable. the recombiner.

AND C.1.2 Isolate all hydrogen inputs Immediately to the system.

AND C.1.3 Restore the required 14 days oxygen monitor(s) to OPERABLE status.

OR C.2.1 Suspend oxygen supply to Immediately the recombiner.

AND C.2.2 Suspend Reactor Coolant Immediately system degas.

AND C.2.3 Obtain and analyze grab Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months samples of the in service WG Decay Tank for oxygen concentration.

AND C.2.4 Restore the required 14 days oxygen monitor(s) to OPERABLE status.

(continued)

McGuire Units 1 and 2 16.7.8-2 Revision 77

Explosive Gas Monitoring Instrumentation 16.7.8 CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D. 1 Prepare and submit a 30 days associated Completion Special Report to the NRC Time not met. explaining why the inoperability was not corrected in the time specified.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.7.8.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months TR 16.7.8.2 Perform CHANNEL OPERATIONAL TEST. 31 days TR 16.7.8.3 ------------------ NOTES ------------------

1. The CHANNEL CALIBRATION of the hydrogen monitor shall include the use of standard gas samples corresponding to alarm setpoints in accordance with the manufacturer's recommendations.

2. The CHANNEL CALIBRATION of the oxygen monitor shall include the use of standard gas samples in accordance with the manufacturer's recommendations.

3. A standard gas sample of nominal 4 volume percent oxygen, balance nitrogen, shall be used in the calibration to check linearity of the oxygen monitor.

Perform a CHANNEL CALIBRATION. 92 days McGuire Units 1 and 2 16.7.8-3 Revision 77

Explosive Gas Monitoring Instrumentation 16.7.8 BASES The gas instrumentation is provided for monitoring and controlling the concentrations of potentially explosive gas mixtures in the WASTE GAS HOLDUP SYSTEM.

SLC 16.7.8 requires that one hydrogen and two oxygen monitors per train be OPERABLE in the Waste Gas (WG) System to ensure that explosive gas mixtures are within the limits of SLC 16.11.19 thereby preventing explosive gas concentrations.

Only one recombiner train is in service at a time. Therefore, the requirement for one hydrogen and two oxygen monitors shall apply only to the train in service.

For applicability purposes, WASTE GAS HOLDUP SYSTEM operation is defined as when the system is receiving any hydrogen inputs. Reactor Coolant system (RCS) degas is the recirculation of gases through the Volume Control Tank (VCT) using the Waste Gas Holdup system to remove hydrogen and oxygen from the RCS.

The requirement for oxygen monitors may be satisfied for Train "A" by using two of the following three monitors:

OWGMT5790 0WGMT6210 OWGMT6211 The requirement for hydrogen monitors may be satisfied for Train "A" by using 0WGMT5590.

The requirement for oxygen monitors may be satisfied for Train "B" by using two of the following three monitors:

OWGMT5780 0WGMT621 0 OWGMT6211 The requirement for hydrogen monitors may be satisfied for Train "B" by using 0WGMT5580.

Loops OWGMT5580 and 0WGMT5590 have both hydrogen and oxygen monitoring capability. The oxygen monitoring portion of these two loops shall not be used to satisfy the oxygen monitor requirements of either train because these oxygen monitors measure the oxygen concentration at the recombiner inlet after the addition of bulk oxygen. This is not representative of the Waste Gas System as defined in SLC 16.11.19. These oxygen monitors will be used for the operation of the hydrogen recombiners but will not be used to satisfy the requirements of SLC 16.7.8.

REFERENCES

1. UFSAR, Section 11.3

2. UFSAR, Section 15.7

3. Catalytic Hydrogen Recombiner Operational Manual, MCM-1201.04-0174.

McGuire Units 1 and 2 16.7.8-4 Revision 77

Rod Position Indication System - Shutdown 16.7.9 16.7 INSTRUMENTATION 16.7.9 Rod Position Indication System - Shutdown COMMITMENT One rod position indicator (excluding demand position indication) shall be OPERABLE and capable of determining the control rod position within +

12 steps for each shutdown or control rod not fully inserted.

MODES 3, 4 and 5 with the reactor trip breakers in the closed position APPLICABILITY with rods not fully inserted and capable of withdrawal.

-NOTE For testing or trouble shooting, alternate methods may be used to ensure there is no possibility of rod motion. These methods are pulling fuses, sliding links in the rod control cabinets or removal of CRDM head cables.

After one of these alternate methods is used, the reactor trip breakers may remain in the closed position.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required Open the reactor trip breakers. Immediately rod position indicators inoperable.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.7.9.1 -------------- NOTE ---------------------------------

Reactor trip breakers shall be closed to perform this test.

For each rod, verify upon rod withdrawal that one train Once prior to initial of the rod position indicating system responds criticality after each appropriately to rod motion. removal of reactor vessel head.

McGuire Units 1 and 2 16.7.9-1 Revision 97

Rod Position Indication System - Shutdown 16.7.9 BASES OPERABILITY of the Rod Position Indicating System is defined as its ability to determine rod position within 12 steps when rods are stationary and thereby ensure compliance with the control rod alignment and insertion limits.

TR 16.7.9.1:

Following removal of the reactor vessel head, one train of the Rod Position Indicating System shall be verified to be functioning. prior to initial criticality. Verification of one Rod Position Indication train responds appropriately to rod withdrawal provides reasonable assurance that the Rod Position Indication System is accurately,indicating rod positions. Rods shall be withdrawn that amount necessary to demonstrate that the position indication system is capable of determining rod position within +/- 12 steps. A note is provided to allow~the reactor trip breakers to be closed when performing this test.

REFERENCES PIP M06-1987 McGuire Units 1 and 2 16.7.9-2 Revision 97

Rod Position Indication System - Test Exception 16.7.10 16.7 INSTRUMENTATION 16.7.10 Position Indication System - Test Exception COMMITMENT The limitations of SLC 16.7-9 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.

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

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

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Position Indication A.1 Open reactor trip breakers. Immediately System inoperable.

OR More than one bank of rods withdrawn.

McGuire Units 1 and 2 16.7.10-1 Revision 0

Rod Position Indication System -Test Exception 16.7.10 TESTING REQUIREMENTS TEST FREQUENCY and the Rod Verify Demand Position Indication System

~1~-

TR 16.7.10.1 Verify Demand Position Indication System and the Rod Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Position Indication Systems agree prior to the start of rod drop time

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

b. Within 24 steps during rod motion. AND Once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during rod drop time measurements BASES This special test exception permits the Position Indication Systems to be inoperable during rod drop time measurements. The exception is required since the data necessary to determine the rod drop time are derived from the induced voltage in the position indicator coils as the rod is dropped. This induced voltage is small compared to the normal voltage and, therefore, cannot be observed if the Position Indication Systems remain OPERABLE.

REFERENCES None.

McGuire Units 1 and 2 16.7.10-2 Revision 0

Hydrogen Monitors 16.7.11 16.7 INSTRUMENTATION 16.7.11 Hydrogen Monitors COMMITMENT The Hydrogen Monitors shall be OPERABLE.

APPLICABILITY: MODES 1, 2 and 3.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One Hydrogen Monitor A.1 Restore channel to 30 days channel inoperable. OPERABLE status.

B. Required Action and B.1 Prepare and submit a 14 days associated Completion Special Report to the Time of Condition A not Commission outlining the met. preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the channel to OPERABLE status.

C. Two Hydrogen Monitor C.1 Restore one Hydrogen 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months channels inoperable. Monitor channel to OPERABLE status.

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

D.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months McGuire Units 1 and 2 16.7.11-1 Revision 71

Hydrogen Monitors 16.7.11 TESTING REQUIREMENTS TEST FREQUENCY TR 16.7.11.1 Perform CHANNEL CALIBRATION. 92 days BASES The Hydrogen Monitors are provided to detect high hydrogen concentration conditions that represent a potential for containment breach from a hydrogen explosion during accident conditions. With the elimination of the design basis LOCA hydrogen release (Ref. 5), the Hydrogen Monitors are no longer required to mitigate design basis accidents. The Hydrogen Monitors are now classified as Regulatory Guide 1.97, Category 3 instrumentation. The Hydrogen Monitors are used to assess the degree of core damage during a severe accident and confirm that random or deliberate ignition has taken place.

The OPERABILITY of the Hydrogen Monitors ensures that there is sufficient information available on unit parameters to monitor and assess unit status and behavior following an accident. The availability of the Hydrogen Monitors is important so that responses to corrective actions can be observed and the need for, and the magnitude of, further actions can be determined. Two OPERABLE channels ensure no single failure prevents operators from getting the information necessary for them to determine the safety status of the unit.

These SLC requirements were relocated from the Technical Specifications as a result of License Amendments 227 and 290 for Units 1 and 2, respectively..

REFERENCES 1. Letter from NRC to G.R. PetersonG.R. Peterson, Duke, License Amendments 227and 209 for Units 1 and 2, respectively, dated April 4April 4, 2005.

2. McGuireMcGuire Updated Final Safety Analysis Report Section 1.8.

3. Regulatory Guide 1.97, Rev. 2.

4. NUREG-0737, Supplement 1, "TMI Action Items."

5. 10 CFR 50.44, "Combustible gas control for nuclear power reactors."

McGuire Units 1 and 2 16.7.11-2 Revision 71

Containment Penetration Conductor Overcurrent Protective Devices 16.8.1 16.8 ELECTRICAL POWER SYSTEMS 16.8.1 Containment Penetration Conductor Overcurrent Protective Devices COMMITMENT All containment penetration conductor overcurrent protective devices shown in Table 16.8.1-1 and Table 16.8.1-2 shall be OPERABLE.

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

REMEDIAL ACTIONS

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

1. Separate Condition entry is allowed for each penetration circuit.

2. Enter applicable Conditions and Required Actions for systems made inoperable by containment penetration conductor overcurrent devices.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1.1 De-energize the circuit(s) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months containment penetration by tripping the associated overcurrent protection redundant circuit breaker or device(s) inoperable. removing the redundant fuse(s).

AND A.1.2 Verify the associated Once per 31 days redundant protective device(s) to be tripped or removed.

OR A.2.1 De-energize the circuit(s) 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by racking out the inoperable circuit breaker or removing the inoperable protective device(s).

AND A.2.2 Verify the inoperable Once per 31 days device(s) are racked out or removed.

(continued)

McGuire Units 1 and 2 16.8.1-1 Revision 98

Containment Penetration Conductor Overcurrent Protective Devices 16.8.1 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months TESTING REQUIREMENTS

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

1. All containment penetration conductor overcurrent protective devices in Table 16.8.1-1 and Table 16.8.1-2 shall be demonstrated OPERABLE by performance of the following Testing Requirements.

2. TR 16.8.1.1, 16.8.1.2, and 16.8.1.3 are only required to be performed for 10% of the circuit breakers within each voltage level on a rotating basis during each surveillance interval.

TEST FREQUENCY TR 16.8.1.1 Perform a CHANNEL CALIBRATION of associated 18 months protective relays for medium voltage circuits (4 - 15 kV).

TR 16.8.1.2 ----------------

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

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

Perform an integrated system functional test on each 18 months medium voltage (4 -15 kV) circuit breaker which includes simulated automatic actuation of the system and verifying that each relay and associated circuit breakers and control circuits function as designed.

(continued)

McGuire Units 1 and 2 16.8.1-2 Revision 98

Containment Penetration Conductor Overcurrent Protective Devices 16.8.1 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.8.1.3 -NOTES-

1. Circuit breakers selected for functional testing shall be selected on a rotating basis.

2. For each circuit breaker found inoperable during functional tests, an additional representative sample of 10% of all the defective type shall be functionally tested until no more failures are found or all of that type have been functionally tested.

3. Lower voltage circuit breakers found inoperable during functional testing shall be restored to OPERABLE status prior to resuming operation.

Perform a functional test of lower voltage circuit breakers 18 months using the nominal trip setpoint and response time values in Table16.8.1-1 and Table 16.8.1-2.

TR 16.8.1.4 Perform fuse inspection and maintenance program. 18 months TR 16.8.1.5 Perform inspection and preventive maintenance on each 60 months circuit breaker in accordance with manufacturer's recommendations.

BASES The tables listed in this commitment were relocated from the McGuire Technical Specifications with the approval of the U.S. Nuclear Regulatory Commission. Any additions, deletions, or revisions to the table are considered a change in a commitment, can only be changed using the 10 CFR 50.59 process, and shall be performed pursuant to applicable procedure.

Containment electrical penetrations and penetration conductors are protected by either de-energizing circuits not required during reactor operation or by demonstrating the OPERABILITY of Primary and Backup overcurrent protection devices during periodic surveillance. Primary and Backup overcurrent protection devices are redundant to each other.

McGuire Units 1 and 2 16.8.1-3 Revision 98

Containment Penetration Conductor Overcurrent Protective Devices 16.8.1 BASES (continued)

Electrical penetrations serve a mechanical integrity function in forming part of the containment pressure boundary. Redundant protective devices provide a means of maintaining this mechanical integrity, which ensures proper protection assuming a single random failure of one of the protective devices. In the event a Containment Penetration Conductor Overcurrent Protective device becomes inoperable, the affected electrical penetration must be de-energized. The method of de-energization must include the use of at least one protective device that cannot be adversely affected by a single active failure.

Acceptable methods of de-energization the circuit(s) are tripping the associated redundant circuit breaker, removing the associated redundant fuses, racking out the inoperable circuit breaker, or removing the inoperable circuit breaker or fuse. Opening the inoperable circuit breaker and verifying all phases are open is not an acceptable means of de-energizing the circuit based on concerns With internal breaker integrity after interrupting a rated fault current.

The 31 day Completion Time to reverify that devices are removed or tripped in inoperable circuits is acceptable considering the fact'that the devices are operated under administrative control and the probability of misalignment is low.

The Surveillance Requirements applicable to lower voltage circuit breakers provide assurance of breaker reliability by testing at least one representative sample of each manufacturer's brand of circuit breaker. Testing of these circuit breakers consists of injecting a current in excess of the breaker's nominal setpoint and measuring the response time. The measured response time is compared to the manufacturer's data to ensure that it is less than or equal to a value specified by the manufacturer. Each manufacturer's molded case and metal clad circuit breakers are grouped into representative samples which are then tested on a rotating basis to ensure that all breakers are tested. If a wide variety exists within any manufacturer's brand of circuit breakers, it is necessary to divide that manufacturer's breakers into groups and treat each group as a separate type of breaker for surveillance purposes.

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

REFERENCES

1. IEEE Standard 242-1975 McGuire Units 1 and 2 16.8.1-4 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED

1. 6900 VAC-Swgr Primary Bkr-RCP1A 5.0 15.4 @ 25A Reactor Coolant Pump 1A Backup Bkr-1TA-5 5.0 16.5 @ 20A Primary Bkr RCP1B 5.0 15.4 @ 25A Reactor Coolant Pump 1B Backup Bkr-1TB-5 5.0 16.5 @ 20A Primary Bkr RCP1C 5.0 15.4 @ 25A Reactor Coolant Pump 1C Backup Bkr-1TC-5 5.0 16.5 @ 20A Primary Bkr RCP1D 5.0 15.4 @ 25A Reactor Coolant Pump 1D Backup Bkr-1TD-5 5.0 16.5 @ 20A

2. 600 VAC-MCC 1EMXA-2 1D Primary Bkr 20 45 or 70 @ 60A

NC Pump 1C Thermal Barrier Backup Fuse 20 NA Outlet Auto Isol VIv 1KC345A 1EMXA-2 1E Primary Bkr 20 45 or 70 @ 60A

NC Pump 1A Thermal Barrier Backup Fuse 20 NA Outlet Auto Isol Vlv 1 KC394A 1EMXA-2 2A Primary Bkr 20 45 or 70 @ 60A

Cont Air Return Fan 1A Backup Fuse 20 NA Damper 1RAF-D-2 1EMXA-2 2B Primary Bkr 20 45 or 70 @ 60A

N2 to Prt Cont Isol Inside VIv Backup Fuse 20 NA 1NC54A 1EMXA-2 2C Primary Bkr 20 45 or 70 @ 60A

RCP Mtg Brg Oil Fill Isol VIv Backup Fuse 20 NA 1NC196A 1EMXA-2 3A Primary Bkr 30 45 or 70 @ 90A

Accumulator 1A Disch Isol Vlv Backup Fuse 30 NA 1N154A 1EMXA-2 3B Primary Bkr 30 45 or 70 @ 90A

Accumulator 1C Disch Isol Backup Fuse 30 NA VIv 1N176A

_____ I

___ ~.___ I

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-5 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1EMXA-2 3C Primary Bkr 20 45 or 70 @ 60A

Test Hdr Inside Cont Isol VIv Backup Fuse 20 NA 1N195A 1EMXA-2 4B Primary Bkr 20 45 or 70 @ 60A

PALS Pnl Smple Ret to Cont.

Backup Fuse 20 NA Isol VIv 1WL-1302A 1EMXA-2 4C Primary Bkr 20 45 or 70 @ 60A

Accum 1A Vent to 1NC34 for Backup Fuse 20 NA Blkout VIv 1 N1430A 1EMXA-2 5A Primary Bkr 20 45 or 70 @ 60A

RN Containment Isolation Vlv Backup Fuse 20 NA 1 RN253A 1EMXA-2 5B Primary Bkr 20 45 or 70 @ 60A

RN Containment Isolation VIv Backup Fuse 20 NA 1 RN276A 1EMXA-2 7A Primary Bkr 20 45 or 70 @ 60A

S/G 1A Upper Shell Sample Backup Fuse 20 NA Cont Isol VIv 1NM187A 1EMXA-2 7B Primary Bkr 20 45 or 70 @ 60A

S/G 1A Blowdown Line Backup Fuse 20 NA Sample Cont Isol VIv 1NM190A 1EMXA-2 7C Primary Bkr 20 45 or 70 @ 60A

SG 1C Upper Shell Sample Backup Fuse 20 NA Cont Isol VIv 1NM207A 1EMXA-2 8A Primary Bkr 20 45 or 70 @ 60A

SG 1C Blowdown Line Line Sample Cont Isol VIv Backup Fuse 20 NA 1NM210A 1EMXA-3 2C Primary Bkr 20 45 or 70 @ 60A

RV Containment Isolation VIv Backup Fuse 20 NA 1RV76A 1EMXA-3 3A Primary Bkr 20 45 or 70 @ 60A

H2 Purge Exhaust Cont Backup Fuse 20 NA Vessel Isol VIv 1VE5A

_ _ _ _ _ _ _ I _ _ _ _ b _ _ _ _ I _ _ _ _ _ _

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-6 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1EMXA-3 4A Primary Bkr 20 45 or 70 @ 60A H2 Skimmer Fan 1A Suction Backup Fuse 20 NA Isol VIv 1VX1A 1EMXA-3 5B Primary Bkr 20 45 or 70 @ 60A

RCDT Pump Disch Cont Isol Backup Fuse 20 NA VIv 1WL2A 1EMXA-3 5C Primary Bkr 20 45 or 70 @ 60A

RCDT Vent Cont Isol VIv Backup Fuse 20 NA 1WL39A 1EMXA-3 6A Primary Bkr 20 45 or 70 @ 60A

RB Sump Pump Disch Cont Backup Fuse 20 NA Isol VIv 1WL64A 1EMXA-3 6B Primary Bkr 20 45 or 70 @ 60A

Cont Vent Unit Condensate Backup Fuse 20 NA Cont Isol VIv 1WL321A 1EMXA-4 1B Primary Bkr 20 45 or 70 @ 60A

NC Pump Seal Return Cont Backup Fuse 20 NA Vlv 1NV94AC 1EMXA-4 3C Primary Bkr 30 45 or 70 @ 90A

NC Loop 1C Discharge to ND Backup Fuse 30 NA System Cont Isol VIv 1ND2A,C 1EMXA-5 1B Primary Bkr 20 45 or 70 @ 60A

Pzr Liquid Sample Line Inside Backup Fuse 20 NA Cont Isol VIv 1NM3A,C 1EMXA-5 2C Primary Bkr 20 45 or 70 @ 60A

Pzr Steam Sample Line Backup Fuse 20 NA Inside Cont Isol VIv 1 NM6A,C 1EMXA-5 2D Primary Bkr 20 45 or 70 @ 60A

NC Hotleg 1 D Sample Line Backup Fuse 20 NA Cont Isol VIv 1 NM25A,C 1EMXA-5 3B Primary Bkr 20 45 or 70 @ 60A NC Hotleg 1A Sample Line Backup Fuse 20 NA Cont Isol VIv 1NM22A,C

______________ [

________ I_________ I

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-7 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1EMXB-4 1B Primary Bkr 20 45 or 70 @ 60A

NC Pump 1B Thermal Barrier Backup Fuse 20 NA Outlet Auto Isol VIv 1 KC364B 1EMXB-4 1C Primary Bkr 20 45 or 70 @ 60A

NC Pump 1D Thermal Barrier Backup Fuse 20 NA Auto Isol VIv 1KC413B 1EMXB-4 2A Primary Bkr 20 45 or 70 @ 60A

NC Pumps Return Hdr Pend Backup Fuse 20 NA Inside Isol Vlv 1KC424B 1EMXB-4 2B Primary Bkr 20 45 or 70 @ 60A

Reactor Bldg Drn Hdr Inside Backup Fuse 20 NA Cont Isol VIv 1KC429B 1EMXB-4 2C Primary Bkr 30 45 or 70 @ 90A

Accumulator 1B Disch Isol Vlv Backup Fuse 30 NA 1N165B 1EMXB-4 3D Primary Bkr 30 45 or 70 @ 90A

Accumulator 1D Disch Isol Backup Fuse 30 NA Vlv 1N188B 1EMXB-4 3E Primary Bkr 20 45 or 70 @ 60A

Hotleg Inj Check 1 N1124, Backup Fuse 20 NA 1N1128 Test Isol VIv 1 N1122B 1EMXB-4 4A Primary Bkr 20 45 or 70 @ 60A

Cont Air Return Fan 1B Backup Fuse 20 NA Damper 1RAF-D-4 1EMXB-4 4C Primary Bkr 20 1 45 or 70 @ 60A

NI Accum 1A Sample Line Backup Fuse 20 NA Inside Cont Isol VIv 1NM72B 1EMXB-4 5A Primary Bkr 20 45 or 70 @ 60A

NI Accum 1B Sample Line Backup Fuse 20 NA Inside Cont Isol Vlv 1NM75B 1EMXB-4 5B Primary Bkr 20 45 or 70 @ 60A

NI Accum 1C Sample Line Backup Fuse 20 NA Inside Cont Isol Vlv 1NM78B

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-8 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1EMXB-4 5C Primary Bkr 20 45 or 70 @ 60A

Accum 1B Vent to 1NC32 for Backup Fuse 20 NA Blkout VIv 1N1431 B 1EMXB-4 6A Primary Bkr 20 45 or 70 @ 60A

NI Accum 1D Sample Line Backup Fuse 20 NA Inside Cont Isol Vlv 1NM81 B 1EMXB-4 6B Primary Bkr 20 45 or 70 @ 60A

SG 1B Upper Shell Sample Backup Fuse 20 NA Cont Isol Vlv 1NM197B 1EMXB-4 6C Primary Bkr 20 45 or 70 @ 60A

SG 1B Bowdown Line Sample Backup Fuse 20 NA Cont Isol VIv 1NM200B 1EMXB-4 7B Primary Bkr 20 45 or 70 @ 60A

SG 1D Upper Shell Sample Backup Fuse 20 NA Cont Isol Vlv 1NM217B 1EMXB-4 7C Primary Bkr 20 45 or 70 @ 60A

SG 1D Blowdown Line Smple Backup Fuse 20 NA Cont Isol VIv 1 NM220B 1EMXB-5 1B Primary Bkr 20 45 or 70 @ 60A

RV Containment Isolation Vlv Backup Fuse 20 NA 1RV33B 1EMXB-5 1C Primary Bkr 20 45 or 70 @ 60A

H2 Skimmer Fan 1B Suction Backup Fuse 20 NA Isol Vlv 1VX2B 1EMXC-1A Primary Bkr 200 250 @ 600A Lower Containment Cooling Backup Fuse 200 NA Unit No. 1A (Normal Source) 1EMXC-2A Primary Bkr 200 250 @ 600A Lower Containment Cooling Backup Fuse 200 NA Unit No. 1C (Normal Source) 1EMXC-3B Primary Bkr 30 45 or 70 @ 90A

Pzr Cavity Booster Fan 1A Backup Fuse 30 NA (Normal Source)

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-9 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1EMXC-3C Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No. 1A (Normal Source)

Backup Fuse 100 NA 1EMXC-3D Primary Bkr 100 11'0 or 150 @ 300A* Control Rod Drive Vent Fan Backup Fuse 100 NA No. 1C (Normal Source) 1EMXC-4C Primary Bkr 90 110 or 125 @ 270A* Containment Air Return Fan Backup Fuse 90 NA No. 1A (CARF-1A) 1EMXC-4D Primary Bkr 90 110 or 125 @ 270A* Hydrogen Recombiner No. 1A Backup Fuse 90 NA 1EMXC-6A Primary Bkr 40 45 or 70 @ 120A* Containment Pipe Tunnel Backup Fuse 40 NA Booster Fan CPT-BF-1A 1EMXC-6B Primary Bkr 30 45 or 70 @ 90A* Upper Containment Air Backup Fuse 30 NA Handling Unit 1A 1EMXC-6C Primary Bkr 30 45 or 70 @ 90A* Upper Containment Air Hdlg Backup Fuse 30 NA Unit IC 1EMXC-6D Primary Bkr 90 110 or 125@270A* Hydrogen Skimmer Fan No.

Backup Fuse 90 NA 1A 1EMXC-7C Primary Bkr 30 45 or 70 @ 90A* Upper Cont Return Air Fan Backup Fuse 30 NA No. 1C 1EMXC-7D Primary Bkr 20 45 or 70 @ 60A* Pzr Pwr Oper Relief Isol VIv Backup Fuse 20 NA 1NC33A 1EMXC-8C Primary Bkr 20 45 or 70 @ 60A* Incore Instrumentation Rm Air Backup Fuse 20 NA HdIg Unit 1A (Normal Source)

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-10 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1EMXC-8D Primary Bkr 20 45 or70 @ 60A* Upper Containment Return Backup Fuse 20 NA Air Fan No. 1A 1EMXD-1A Primary Bkr 200 250 @ 600A Lower Containment Cooling Backup Fuse 200 NA Unit No. 1 B (Normal Source) 1EMXD-2A Primary Bkr 200 250 @ 600A Lower Containment Cooling Backup Fuse 200 NA Unit No. 1D (Normal Source) 1EMXD-3B Primary Bkr 40 45 or 70 @ 120A* Containment Pipe Tunnel Backup Fuse 40 NA Booster Fan CPT-BF-1 B 1EMXD-3C Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan Backup Fuse 100 NA No. 1B (Normal Source) 1EMXD-3D Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan Backup Fuse 100 NA No. 1D (Normal Source) 1EMXD-4C Primary Bkr 90 110 or 125 @ 270A* Containment Air Return Fan Backup Fuse 90 NA No. 1 B (CARF-1 B) 1EMXD-4D Primary Bkr 90 110 or 125 @ 270A* Hydrogen Recombiner No. 1B Backup Fuse 90 NA 1EMXD-6C Primary Bkr 30 45 or 70 @ 90A* Upper Containment Air Hdlg Backup Fuse 30 NA Unit No. 1B 1EMXD-6D Primary Bkr 30 45 or 70 @ 90A* Upper Containment Air Hdlg Backup Fuse 30 NA Unit No. 1D 1EMXD-6E Primary Bkr 90 110 or 125 @ 270* Hydrogen Skimmer Fan No.

Backup Fuse 90 NA 1B

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-11 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1EMXD-7B Primary Bkr 30 45 or 70 @ 90A* Upper Cont Return Air Fan Backup Fuse 30 NA No. 1D 1EMXD-7C Primary Bkr 20 45 or 70 @ 60A* Pzr No. 1 Pwr Oper Safety Backup Fuse 20 NA Relief Isol Vlv 1 NC31 B 1EMXD-7D Primary Bkr 20 45 or 70 @ 60A* Pzr No. 1 Pwr Oper Safety Backup Fuse 20 NA Relief Isol Vlv 1NC35B 1EMXD-8A Primary Bkr 30 45 or 70 @ 90A* PZR Cavity Booster Fan 1B Backup Fuse 30 NA (Normal Source) 1EMXD-8B Primary Bkr 20 45 or 70 @ 60A* Incore Instrumentation Rm Air Backup Fuse 20 NA HdIg Unit 1B (Normal Source) 1EMXD-8C Primary Bkr 20 45 or 70 @ 60A* Upper Containment Return Backup Fuse 20 NA Air Fan 1B 1EMXD-8D Primary Bkr 30 45 or 70 @ 90A* NC Loop 1C Disch to ND Backup Fuse 30 NA System Cont Isol Vlv 1ND1 B 1MXM-FIA Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 1LR14 Backup Fuse 40 NA 1MXM-F1B Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 1LR15 Backup Fuse 40 NA 1MXM-F1C Primary Bkr 30 45 or 70 @ 90A* S/G Compt. Fan 1C Backup Fuse 30 NA 1MXM-F1D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU WA1 Blower A Backup Fuse 20 NA

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-12 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1MXM-F1E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A2 Blower A Backup Fuse 20 NA 1MXM-F2A Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 1LR16 Backup Fuse 40 NA 1MXM-F2B Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 1LR17 Backup Fuse 40 NA 1MXM-F2C Primary Bkr 25 45 or 70 @ 75A* Reactor Bldg Equip Hdlg 5 Backup Fuse 25 NA Ton Jib Crane 1MXM-F2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A3 Blower A Backup Fuse 20 NA 1MXM-F2E Primary Bkr 20 45 or 70 @ 60A* Ice Cont AHU 1A4 Blower A Backup Fuse 20 NA 1MXM-F3A Primary Bkr 20 45 or 70 @ 60A* Ice Cont AHU 1A5 Blower A Backup Fuse 20 NA 1MXM-F3B Primary Bkr 20 45 or 70 @ 60A* Ice Cont AHU 1A6 Blower A Backup Fuse 20 NA 1MXM-F3C Primary Bkr 20 45 or 70 @ 60A* Incore Inst Room Sump Backup Fuse 20 NA Pump 1 1MXM-F3D Primary Bkr 100 110 or 150 @ 300A* Upper Cont Welding Recpt Backup Fuse 100 NA 1MXM-F4A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A7 Blower A Backup Fuse 20 NA

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-13 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP

,SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1 MXM-F4B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A8 Blower A Backup Fuse 20 NA 1MXM-F4D Primary Bkr 100 110 or 150 @ 300A* Welding Feeder Backup Fuse 100 NA 1MXM-F5C Primary Bkr 50 110 or 150 @ 150A* Ice Cond Floor Cooling Backup Fuse 50 NA Defrost Heater 1A 1MXM-F6C Primary Bkr 60 110 or 150 @ 180A* Reactor Coolant Drain Tank Backup Fuse 60 NA Pump lA 1MXM-F7A Primary Bkr 20 45 or 70 @ 60A* Ice Cond-AHU 1A9 Blower A Backup Fuse 20 NA 1MXM-F7B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1Al0 Blower A Backup Fuse 20 NA 1MXM-F7C Primary Bkr 20 45 or 70 @ 60A* Lower Cont Aux Charcoal Backup Fuse 20 NA Filter Fan 1A 1MXM-F8A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1Al1 Blower A Backup Fuse 20 NA 1MXM-F8B Primary Bkr 20 45 or 70 @ 60A* - Ice Cond AHU 1A12 Blower A Backup Fuse 20 NA 1MXM-F8C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A13 Blower A Backup Fuse 20 NA 1MXM-R1A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B1 Blower A Backup Fuse 20 NA

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-14 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1MXM-R1B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B2 Blower A Backup Fuse 20 NA 1MXM-R1C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B3 Blower A Backup Fuse 20 NA 1MXM-RID Primary Bkr 30 45 or 70 @ 90A* RCP 1A Oil Lift Pump No. 1 Backup Fuse 30 NA 1MXM-R2A Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 1LR12 Backup Fuse 40 NA 1MXM-R2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B4 Blower A Backup Fuse 20 NA 1MXM-R2E Primary Bkr 30 45 or 70 @ 90A* RCP 1B Oil Lift Pump No. 1 Backup Fuse 30 NA 1MXM-R3D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B5 Blower A Backup Fuse 20 NA 1MXM-R3E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B6 Blower A Backup Fuse 20 NA 1MXM-R3F Primary Bkr 30 45 or 70 @ 90A* RCP 1C Oil Lift Pump No. 1 Backup Fuse 30 NA 1MXM-R4D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B7 Blower A Backup Fuse 20 NA 1MXM-R4E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B8 Blower A Backup Fuse 20 NA

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-15 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1MXM-R4F Primary Bkr 30 45 or 70 @ 90A* RCP 1 D Oil Lift Pump No. 1 Backup Fuse 30 NA 1MXM-R5B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B9 Blower A Backup Fuse 20 NA 1MXM-R5C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B1 0 Blower A Backup Fuse 20 NA 1MXM-R5D Primary Bkr 175 200 @ 525A Ice Cond Equip Pwr Pnlbd 1A Backup Fuse 175 NA 1MXM-R6A Primary Bkr 20 45 or 70 @ 60A* Rod Cntrl Cluster Change Backup Fuse 20 NA Fixture Hoist Drive 1MXM-R6B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B311 Blower A Backup Fuse 20 NA 1MXM-R6E Primary Bkr 150 110 or 230 @ 450A* 175 Ton Polar Crane Backup Fuse 150 NA 1MXM-R7A Primary Bkr 20 45 or 70 @ 60A* Stud Tensioner Hoist Backup Fuse 20 NA 1MXM-R7B Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive 1A Backup Fuse 20 NA 1MXM-R7D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B1 2 Blower A Backup Fuse 20 NA 1MXM-R7E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1Bi13 Blower A Backup Fuse 20 NA

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-16 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1MXM-R8A Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive 1B Backup Fuse 20 NA 1MXM-R8B Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive IC Backup Fuse 20 NA 1MXM-R8D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B14 Blower A Backup Fuse 20 NA 1MXM-R8E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B1 5 Blower A Backup Fuse 20 NA 1MXMA-1B Primary Bkr 30 45 or 70 @ 90A* Pzr Cavity Booster Fan 1B Backup Fuse 30 NA (Alt Source) 1MXMA-1D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A14 Blower A Backup Fuse 20 NA 1MXMA-1E Primary Bkr 20 45 or 70 @ 60A* Cont Floor & Equip Sump 1A Backup Fuse 20 NA Pump WA1 1MXMA-2A Primary Bkr 25 45 or 70 @ 75A* RCPM Maintenance Crane Backup Fuse 25 NA Recpt lA, 1B, 1C, & 1D 1MXMA-2B Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 1LR6 Backup Fuse 25 NA 1MXMA-2C Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 1LR18 Backup Fuse 40 NA 1MXMA-2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1Al 5 Blower A Backup Fuse 20 NA

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-17 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES), (SECONDS) SYSTEM POWERED 1MXMA-3A Primary Bkr 25 45 or 70 @ 75A* Lighting ,Pnlbd 1LR9 Backup Fuse 25 NA 1MXMA-3B Primary Bkr 20 45 or 70 @ 60A* Ice Cond Equip Access Door Backup Fuse 20 NA 1A 1MXMA-3C Primary Bkr 50 110 or 150 @ 150A* Ice Cond Floor Cooling Pump Backup Fuse 50 NA 1A 1MXMA-3D Primary Bkr 20 45 or 70 @ 60A* Cont Floor & Equip Sump 1B Backup Fuse 20 NA Pump 1B1 1MXN-F1A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU WA1 Blower B Backup Fuse 20 NA 1MXN-F1B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A2 Blower B Backup Fuse 20 NA 1MXN-FIC Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A3 Blower B Backup Fuse 20 NA 1MXN-F1D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A4 Blower B Backup Fuse 20 NA 1MXN-F2A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A5 Blower B Backup Fuse 20 NA 1MXN-F2B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A6 Blower B Backup Fuse 20 NA 1MXN-F2C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A7 Blower B Backup Fuse 20 NA 1MXN-F2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A8 Blower B Backup Fuse 20 NA

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-18 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1MXN-F3A Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 1LR1 Backup Fuse 25 NA 1MXN-F3B Primary Bkr 30 45 or 70 @ 90A* S/G Compt. Fan 1B Backup Fuse 30 NA 1MXN-F3C Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 1LR2 Backup Fuse 25 NA 1MXN-F3D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A9 Blower B Backup Fuse 20 NA 1MXN-F3E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU WA1O Blower B Backup Fuse 20 NA 1MXN-F4A Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive No 1D Backup Fuse 20 NA 1MXN-F4B Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive No. 1E Backup Fuse 20 NA 1MXN-F4C Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive No. 1F Backup Fuse 20 NA 1MXN-F4D Primary Bkr 20 45 or 70 @ 60A* Cont Floor & Equip Sump 1A Backup Fuse 20 NA Pump 1A2 1MXN-F5C Primary Bkr 60 110 or 150 @ 180A* Reactor Coolant Drain Tank Backup Fuse 60 NA Pump 1B 1MXN-F6B Primary Bkr 20 45 or 70 @ 60A* Cont Floor & Equip Sump 1B Backup Fuse 20 NA Pump 1B2 I___ I ___ [

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-19 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1MXN-F6C Primary Bkr 50 110 or 150 @ 150A* Ice Cond Floor Cooling Backup Fuse 50 NA Defrost Htr 1B 1MXN-F7A Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 1LR4 Backup Fuse 25 NA 1MXN-F7B Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 1LR5 Backup Fuse 25 NA 1MXN-F7C Primary Bkr 20 45 or 70 @ 60A* Fuel Transfer Sys Reactor Backup Fuse 20 NA Side Fdr 1MXN-F7D Primary Bkr 20' 45 or 70 @ 60A* Ice Cond AHU 1Al1 Blower B Backup Fuse 20 NA 1MXN-F8B Primary Bkr 30 45 or 70 @ 90A* S/G Compt. Fan 1A Backup Fuse 30 NA 1MXN-F8D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A12 Blower B Backup Fuse 20 NA 1MXN-F8E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A13 Blower B Backup Fuse 20 NA 1MXN-R1D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B1 Blower B Backup Fuse 20 NA 1MXN-R1E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B2 Blower B Backup Fuse 20 NA 1MXN-R1F Primary Bkr 30 45 or 70 @ 90A* RCP 1A Oil Lift Pump No. 2 Backup Fuse 30 NA

_____ I

___ I___ I

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-20 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1MXN-R2C Primary Bkr 30 45 or 70 @ 90A* Reactor Cavity Manipulator Backup Fuse 30 NA Crane 1MXN-R2F Primary Bkr 30 45 or 70 @ 90A* RCP 1B Oil Lift Pump No. 2 Backup Fuse 30 NA 1MXN-R3A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B3 Blower B Backup Fuse 20 NA 1MXN-R3B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B4 Blower B Backup Fuse 20 NA 1MXN-R3C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B5 Blower B Backup Fuse 20 NA 1MXN-R3D Primary Bkr 30 45 or 70 @ 90A* RCP 1C Oil Lift Pump No. 2 Backup Fuse 30 NA 1MXN-R4A Primary Bkr 50 110 or 150 @ 150A* Ice Cond Bridge Crane Backup Fuse 50 NA 1MXN-R4B Primary Bkr 30 45 or 70 @ 90A* RB Equip Hatch Hoist No. 1 Backup Fuse 30 NA 1MXN-R4C Primary Bkr 30 45 or 70 @ 90A* S/G Compt. Fan 1D Backup Fuse 30 NA 1MXN-R4D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B6 Blower B Backup Fuse 20 NA 1MXN-R4E Primary Bkr 30 45 or 70 @ 90A* RCP 1D Oil Lift Pump No.2 Backup Fuse 30 NA 1MXN-R5D Primary Bkr 175 200 @ 525A Ice Cond Equip Pwr Pnlbd 1B

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-21 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED Backup Fuse 175 NA 1MXN-R6A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B7 Blower B Backup Fuse 20 NA 1MXN-R6B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B8 Blower B Backup Fuse 20 NA 1MXN-R6C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B9 Blower B Backup Fuse 20 NA 1MXN-R6D Primary Bkr 100 110or150@300A* Welding Fdr Backup Fuse 100 NA 1MXN-R7A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B10 Blower B Backup Fuse 20 NA 1MXN-R7B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1 B1 1 Blower B Backup Fuse 20 NA 1MXN-R7C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B12 Blower B Backup Fuse 20 NA I MXN-R7D Primary Bkr 50 110 or 150 @ 150A* Ice Cond Floor Cooling Pump Backup Fuse 50 NA 1B 1MXN-R8D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B13 Blower B Backup Fuse 20 NA 1MXN-R8E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1B14 Blower B Backup Fuse 20 NA 1MXN-R8F Primary Bkr 20 45 or 70.@ 60A* Ice Cond AHU 11B15 Blower B Backup Fuse 20 NA

___________ I _______ I.

_______ I___________

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-22 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1MXNA-2A Primary Bkr 30 45 or 70 @ 90A* Pzr Cavity Booster Fan 1A Backup Fuse 30 NA (Alt Source) 1MXNA-2B Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 1LR7 Backup Fuse 25 NA 1MXNA-2C Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 1LR8 Backup Fuse 25 NA 1MXNA-2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A14 Blower B Backup Fuse 20 NA 1MXNA-2E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 1A15 Blower B Backup Fuse 20 NA 1MXNA-3A Primary Bkr 20 45 or 70 @ 60A* 2 Ton CRDM HdIg Jib Crane Backup Fuse 20 NA 1MXNA-3C Primary Bkr 20 45 or 70 @ 60A* NC Pump Motor Drain Tank Backup Fuse 20 NA Pump No. 1 1MXNA-3D Primary Bkr 20 45 or 70 @ 60A* Ice Cond Equip Access Door Backup Fuse 20 NA B SMXA-F4A Primary Bkr 15 45 or 70 @ 45A* Unit 1 Emergency Personnel Backup Fuse 15 NA Lock SMXC-7D Primary Bkr 15 45 or 70 @ 45A* Unit 1 Personnel Lock Backup Fuse 15 NA SMXG-F3G Primary Bkr 20 45 or 70 @ 60A* Standby Makeup Pump to Backup Fuse 20 NA Cont Sump Isol VIv 1NV1O12C

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-23 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED SMXG-F4G Primary Bkr 20 45 or 70 @ 60A* Standby Makeup Pump to NC Backup Fuse 20 NA Pump Seals Isol Vlv 1NV1013C SMXG-F5A Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 28, 55, &

Backup Fuse 90 NA 56

3. 600 VAC-Press Htr Pwr PnIs Backup Press Htr Pwr Pnl 1A-1A Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 1, 2, & 22 Backup Fuse 90 NA Backup Press Htr Pwr Pnl 1A-1B Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 5, 6, & 27 Backup Fuse 90 NA Backup Press Htr Pwr PnI 1A-1C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 9, 10, &

Backup Fuse 90 NA 32 Backup Press Htr Pwr Pnl 1A-2C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 11, 12, &

Backup Fuse 90 NA 35 Backup Press Htr Pwr PnI 1A-2D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 13, 14, &

Backup Fuse 90 NA 37 Backup Press Htr Pwr Pni 1A-2E Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 17, 18, &

Backup Fuse 90 NA 42 Backup Press Htr Pwr PnI 1B-1A Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 21, 47, &

Backup Fuse 90 NA 48 Backup Press Htr Pwr Pnl 1B-1B

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-24 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 26, 53, &

Backup Fuse 90 NA 54 Backup Press Htr Pwr PnI 1B-1C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 31, 59, &

Backup Fuse 90 NA 60 Backup Press Htr Pwr PnI 1B-2C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 36, 65, &

Backup Fuse 90 NA 66 Backup Press Htr Pwr Pnl 1B-2D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 41, 71, &

Backup Fuse 90 NA 72 Backup Press Htr Pwr PnI 1B-2E Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 46, 77, &

Backup Fuse 90 NA 78 Backup Press Htr Pwr Pnl 1C-1A Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 7, 8, &30 Backup Fuse 90 NA Backup Press Htr Pwr PnI 1C-1B Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 19, 20, &

Backup Fuse 90 NA 45 Backup Press Htr Pwr Pnl 1C-ic Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 24, 51, &

Backup Fuse 90 NA 52 Backup Press Htr Pwr Pni 1C-1D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 29, 57, &

Backup Fuse 90 NA 58 Backup Press Htr Pwr PnI 1C-2C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 34, 63, &

Backup Fuse 90 NA 64

__ _ _ _ __ __ I _ _ _ _ _ _ J _ _ _ _ _ _ I _ _ _ _

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-25 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED Backup Press Htr Pwr Pni 1C-2D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 39, 69, &

Backup Fuse 90 NA 70 Backup Press Htr Pwr Pni 1C-2E Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 44, 75, &

Backup Fuse 90 NA 76 Backup Press Htr Pwr Pnl 1D-1A Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 3, 4, & 25 Backup Fuse 90 NA Backup Press Htr Pwr Pnl 1D-1B Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 15, 16, &

Backup Fuse 90 NA 40 Backup Press Htr Pwr Pnl 1D-ic Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 23, 49, &

Backup Fuse 90 NA 50 Backup Press Htr Pwr PnI 1D-2C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 33, 61, &

Backup Fuse 90 NA 62 Backup Press Htr Pwr Pn1 1D-2D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 38, 67, &

Backup Fuse 90 NA - 68 Backup Press Htr Pwr Pnl 1D-2E Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 43, 73, &

Backup Fuse 90 NA 74

4. 120 VAC-Panelboards 1KM-1 Primary Bkr 30 45 or 70 @ 90* RCP 1A Space Htr Backup Fuse 30 NA 1KM-2 Primary Bkr 30 45 or 70 @ 90* RCP 1 C Space Htr HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-26 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED Backup Fuse 30 NA 1KN-1 Primary Bkr 30 45 or 70 @ 90* RCP 1B Space Htr Backup Fuse 30 NA 1KN-2 Primary Bkr 30 45 or 70 @ 90* RCP 1D Space Htr Backup Fuse 30 NA 1KN-27 Primary Bkr 20 36 or 70 @ 60* Fuel Handling Control Backup Fuse 20 NA Console 1KN-31 Primary Bkr 20 36 or 70 @ 60* Incore Inst. 120 VAC Outlet Backup Fuse 20 NA Receptacles

5. 250 VDC-Lighting RB Deadlight Pnlbd 1DLD # 1 Primary Bkr 20 40 @ 60 Ltg Pnl Nos. 1LR1 & 1LR2 Backup Fuse 20 NA RB Deadlight Pnlbd 1DLD # 3 Primary Bkr 20 40 @ 60 Ltg Pnl Nos. 1LR4, 1LR5, &

Backup Fuse 20 NA 1LR6 RB Deadlight Pnlbd 1DLD # 4 Primary Bkr 20 40 @ 60 Ltg Pnl Nos. 1LR7, 1LR8, &

Backup Fuse 20 NA 1LR9 RB Deadlight Pnlbd 1DLD # 6 Primary Bkr 20 40 @ 60 Ltg Pnl Nos.,1LR12 Backup Fuse 20 NA RB Deadlight Pnlbd 1DLD # 7 Primary Bkr 20 40 @ 60 Ltg Pnl Nos. 1LR16 Backup Fuse 20 NA RB Deadlight Pnlbd 1DLD # 9 Primary Bkr 20 40 @ 60 Ltg PnI Nos. 1LR18 & 1LR17 Backup Fuse 20 NA

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-27 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED

6. VAC - LC Outage Power Fuse Box 1A Primary Fuse 150 NA Upper Containment Safety Backup Fuse 150 Switch 1A Primary Fuse 150 NA Upper Containment Safety Backup Fuse 150 NA Switch 1B Primary Fuse 200 NA Lower Containment Outage Backup Fuse 200 NA Power Distr. Pnl 1A Primary Fuse 225 NA Lower Containment Outage Backup Fuse 225 NA Power Distr. Pnl 1B

7. 600 VAC - Containment HVAC Alternrate Feeders 1VTB-1A Primary Bkr 20 45 or 70 @ 60A* Incore Instrumentation Rm Air HdIg Unit 1A (Alt Source) 1VTF-1A Primary Fuse 20 NA Incore Instrumentation Rm Air Hdlg Unit 1A (Alt Source) lVTB-1B Primary Bkr 20 45 or 70 @ 60A* Incore Instrumentation Rm Air Hdlg Unit 1B (Alt Source) 1VTF-1 B Primary Fuse 20 NA Incore Instrumentation Rm Air Hdlg Unit 1B (Alt Source) 1VRB-1A Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No. 1A (Alt Source) 1VRF-1A Primary Fuse 100 NA Control Rod Drive Vent Fan No. 1A (AIt Source) lVRB-1 B Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No. 1 B (Alt Source)

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-28 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED 1VRF-1B Primary Fuse 100 NA Control Rod Drive Vent Fan No. 1 B (Alt Source) 1VRB-1C Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No. 1C (Alt Source) 1VRF-1C Primary Fuse 100 NA Control Rod Drive Vent Fan No. 1C (AIt Source) 1VRB-1D Primary Bkr 100 110 or 150@ 300A* Control Rod Drive Vent Fan No. 1 D (AIt Source) 1VRF-1D Primary Fuse 100 NA Control Rod Drive Vent Fan No. 1D (Alt Source) 1VLB-1A Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit No. 1A (Alt Source) 1VLF-1A Primary Fuse 200 NA Lower Containment Cooling Unit No. 1A (Alt Source) 1VLB-1B Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit No. 1B (Alt Source) 1VLF-1B Primary Fuse 200 NA Lower Containment Cooling Unit No. 1 B (Alt Source) 1VLB-1C Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit No. 1C (Alt Source) 1VLF-1C Primary Fuse 200 NA Lower Containment Cooling Unit No. 1C (Alt Source) 1VLB-1D Primary Bkr 200 250 @ 600A Lower Containment Cooling

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-29 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-1 UNIT 1 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING RESPONSE TIME DEVICE NO. & LOCATION (AMPERES) (SECONDS) SYSTEM POWERED Unit No. 1D (Alt Source) 1VLF-1D Primary Fuse 200 NA Lower Containment Cooling Unit No. 1D (Alt Source)

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-30 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION

1. 6900 VAC-Swgr Primary Bkr- 5.0 15.4 @ 25A Reactor Coolant Pump 2A RCP2A Backup Bkr-2TA-5 5.0 16.5 @ 20A Primary Bkr 5.0 15.4 @ 25A Reactor Coolant Pump 2B RCP2B Backup Bkr-2TB-5 5.0 16.5 @ 20A Primary Bkr 5.0 15.4 @ 25A Reactor Coolant Pump 2C RCP2C Backup Bkr-2TC-5 5.0 16.5 @ 20A Primary Bkr 5.0 15.4 @ 25A Reactor Coolant Pump 2D RCP2D Backup Bkr-2TD-5 5.0 16.5 @ 20A

2. 600 VAC-MCC 2EMXA-2 1D Primary Bkr 20 45 or 70 @ 60A* NC Pump 2C Thermal Barrier Backup 20 NA Outlet Auto Isol VIv 2KC345A Fuse 2EMXA-2 1E Primary Bkr 20 45 or 70 @ 60A* NC Pump 2A Thermal Barrier Backup 20 NA Outlet Auto Isol VIv 2KC394A Fuse 2EMXA-2 2A Primary Bkr 20 45 or 70 @ 60A* Cont Air Return Fan 2A Damper Backup 20 NA 2RAF-D-2 Fuse 2EMXA-2 2B Primary Bkr 20 45 or 70 @ 60A* N2 to Prt Cont Isol Inside Vlv Backup 20 NA 2NC54A Fuse 2EMXA-2 2C Primary Bkr 20 45 or 70 @ 60A* RCP Mtg Brg Oil Fill Isol VIv Backup 20 NA 2NC196A Fuse 2EMXA-2 3A Primary Bkr 30 45 or 70 @ 90A* Accumulator 2A Disch Isol VIv Backup 30 NA 2N154A Fuse

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-31 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION 2EMXA-2 3B Primary Bkr 30 45 or 70@ 90A* Accumulator 2C Disch Isol VIv Backup 30 NA 2NI76A Fuse 2EMXA-2 3C Primary Bkr 20 45 or 70 @ 60A* Test Hdr Inside Cont Isol Vlv Backup 20 NA 2NI95A Fuse 2EMXA-2 4B Primary Bkr 20 45 or 70 @ 60A* PALS Pnl Smple Ret to Cont.

Backup 20 NA Isol Vlv 2WL-1 302A Fuse 2EMXA-2 4C Primary Bkr 20 45 or 70 @ 60A* Accum-2A Vent to 2NC34 for Backup 20 NA Blkout Vlv 2NI430A Fuse 2EMXA-2 5A Primary Bkr 20 45 or 70 @ 60A* RN Containment Isolation Vlv Backup 20 NA 2RN253A Fuse 2EMXA-2 5B Primary Bkr 20 45 or 70 @ 60A* RN Containment Isolation VIv Backup 20 NA 2RN276A Fuse 2EMXA-2 7A Primary Bkr 20 45 or 70 @ 60A* S/G 2A Upper Shell Sample Backup 20 NA Cont Isol Vlv 2NM187A Fuse 2EMXA-2 7B Primary Bkr 20 45 or 70 @ 60A* S/G 2A Blowdown Line Sample Backup 20 NA Cont Isol Vlv 2NM190A Fuse 2EMXA-2 7C Primary Bkr 20 45 or 70 @ 60A* SG 2C Upper Shell Sample Cont Backup 20 NA Isol Vlv 2NM207A Fuse 2EMXA-2 8A Primary Bkr 20 45 or 70 @ 60A* SG 2C Blowdown Line Line

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-32 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup 20 NA Sample Cont Isol Vlv 2NM210A Fuse 2EMXA-3 1A Primary Bkr 20 45 or 70 @ 60A* Lower Cont Vent Unit discharge Backup 20 NA cont isol vlv 2RV76A Fuse 2EMXA-3 3A Primary Bkr 20 45 or 70 @ 60A* H2 Purge Exhaust Cont Vessel Backup 20 NA Isol Vlv 2VE5A Fuse 2EMXA-3 4A Primary Bkr 20 45 or 70 @ 60A* H2 Skimmer Fan 2A Suction Isol.

Backup 20 NA Vlv 2VX'1A Fuse 2EMXA-3 5B Primary Bkr 20 45 or 70 @ 60A* RCDT Vent Cont Isol Vlv Backup 20 NA 2WL2A Fuse 2EMXA-3 5C Primary Bkr 20 45 or 70 @ 60A* RCDT Vent Cont Isol Vlv Backup 20 NA 2WL39A Fuse 2EMXA-3 6A Primary Bkr 20 45 or 70 @ 60A* RB Sump Pump Disch Cont Isol Backup 20 NA VIv 2WL64A Fuse 2EMXA-3 6B Primary Bkr 20 45 or 70 @ 60A* Cont Vent Unit Condensate Cont Backup 20 NA Isol Vlv 2WL321A Fuse 2EMXA-4 1B Primary Bkr 20 45 or 70 @ 60A* NC Pump Seal Return Cont Vlv Backup 20 NA 2NV94AC Fuse 2EMXA-4 3C Primary Bkr 30 45 or 70 @ 90A* NC Loop 2C Discharge to ND System Cont Isol VIv 2ND2A,C Backup 30 r NA Fuse HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-33 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION 2EMXA-5 1B Primary Bkr 20 45 or 70 @ 60A* Pzr Liquid Sample Line Inside Backup 20 NA Cont Isol VIv 2NM3A,C Fuse 2EMXA-5-2C Primary Bkr 20 45 or 70 @ 60A* Pzr Steam Sample Line Inside Backup 20 NA Cont Isol VIv 2NM6A,C Fuse 2EMXA-5 2D Primary Bkr 20 45 or 70 @ 60A* NC Hotleg 2D Sample Line Cont Backup 20 NA Isol VIv 2NM25A,C Fuse 2EMXA-5 3B Primary Bkr 20 45 or 70 @ 60A* NC Hotleg 2A Sample Line Cont Backup 20 NA Isol VIv 2NM22A,C Fuse 2EMXB-4 1B Primary Bkr 20 45 or 70 @ 60A* NC Pump 2B Thermal Barrier Backup 20 NA Outlet Auto Isol VIv 2KC364B Fuse 2EMXB-4 1C Primary Bkr 20 45 or 70 @ 60A* NC Pump 2D Thermal Barrier Backup 20 NA Auto Isol VIv 2KC413B Fuse 2EMXB-4 2A Primary Bkr 20 45 or 70 @ 60A* NC Pumps Return Hdr Pend Backup 20 NA Inside Isol VIv 2KC424B Fuse 2EMXB-4 2B Primary Bkr 20 45 or 70 @ 60A* Reactor Bldg Drn Hdr Inside Backup 20 NA Cont Isol VIv 2KC429B Fuse 2EMXB-4 2C Primary Bkr 30 45 or 70 @ 90A* Accumulator 2B Disch Isol VIv Backup 30 NA 2NI65B Fuse 2EMXB-4 3D Primary Bkr 30 45 or 70 @ 90A* Accumulator 2D Disch Isol VIv

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-34 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup 30 NA 2NI88B Fuse 2EMXB-4 3E Primary Bkr 20 45 or 70 @ 60A* Hotleg Inj Check 2NI124, Backup 20 NA 2NI128 Test Isol Vlv 2NI122B Fuse 2EMXB-4 4A Primary Bkr 20 45 or 70 @ 60A* Cont Air Return Fan 2B Damper Backup 20 NA 2RAF-D-4 Fuse 2EMXB-4 4C Primary Bkr 20 45 or 70 @ 60A* NI Accum 2A Sample Line Backup 20 NA Inside Cont Isol VIv 2NM72B Fuse 2EMXB-4 5A Primary Bkr 20 45 or 70 @ 60A* NI Accum 2B Sample Line Backup 20 NA Inside Cont Isol Vlv 2NM75B Fuse 2EMXB-4 5B Primary Bkr 20 45 or 70 @ 60A* NI Accum 2C Sample Line Backup 20 NA Inside Cont Isol Vlv 2NM78B Fuse 2EMXB-4 5C Primary Bkr 20 45 or 70 @ 60A* Accum 2B Vent to 2NC32 for Backup 20 NA BIkout Vlv 2NI431 B Fuse 2EMXB-4 6A Primary Bkr 20 45 or 70 @ 60A* NI Accum 2D Sample Line Backup 20 NA Inside Cont Isol VIv 2NM81 B Fuse 2EMXB-4 6B Primary Bkr 20 45 or 70 @ 60A* SG 2B Upper Shell Sample Cont Backup 20 NA Isol VIv 2NM197B Fuse 2EMXB-4 6C Primary Bkr 20 45 or 70 @ 60A* SG 2B Bowdown Line Sample Backup 20 NA Cont Isol Vlv 2NM200B Fuse 2EMXB-4 7B Primary Bkr 20 45 or 70 @ 60A* SG 2D Upper Shell Sample Cont

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-35 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup 20 NA Isol VIv 2NM217B Fuse 2EMXB-4 7C Primary Bkr 20 45 or 70 @ 60A* SG 2D Blowdown Line Sample Cont Isol VIv 2NM220B Backup 20 NA Fuse 2EMXB-5 1B Primary Bkr 20 45 or 70 @ 60A* Lower cont vent unit supply cont Backup 20 NA isol vlv 2RV33B Fuse 2EMXB-5 1C Primary Bkr 20 45 or 70 @ 60A* H2 Skimmer Fan 2B Suction Isol Backup 20 NA VIv 2VX2B Fuse 2EMXC-1A Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit Backup 200 NA No. 2A (Normal Source)

Fuse 2EMXC-2A Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit Backup 200 NA No. 2C (Normal Source)

Fuse 2EMXC-3C Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No.

Backup 100 NA 2A (Normal Source)

Fuse 2EMXC-3D Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No.

Backup 100 NA 2C (Normal Source)

Fuse 2EMXC-4C Primary Bkr 90 110 or 125 @ 270A* Containment Air Return Fan No.

Backup 90 NA 2A (CARF-2A)

Fuse 2EMXC-4D Primary Bkr 90 110 or 125 @ 270A* Hydrogen Recombiner No. 2A Backup 90 NA Fuse

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-36 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION 2EMXC-6A Primary Bkr 40 45 or 70 @ 120A* Containment Pipe Tunnel Backup 40 NA Booster Fan CPT-BF-2A Fuse 2EMXC-6B Primary Bkr 30 45 or 70 @ 90A* Upper Containment Air Handling Backup 30 NA Unit 2A Fuse 2EMXC-6C Primary Bkr 30 45 or 70 @ 90A* Upper Containment Air Hdlg Unit Backup 30 NA 2C Fuse 2EMXC-6D Primary Bkr 90 110 or 125 @ 270A* Hydrogen Skimmer Fan No. 2A Backup 90 NA Fuse 2EMXC-7A Primary Bkr 30 45 or 70 @ 90A* PZR Cavity Booster Fan 2A 30 NA (Normal Source)

Backup Fuse 2EMXC-7B Primary Bkr 20 45 or 70 @ 60A* Upper Containment Return Air Backup 20 NA Fan No. 2A Fuse 2EMXC-7C Primary Bkr 20 45 or 70 @ 60A* Upper Cont Return Air Fan No.

Backup 20 NA 2C Fuse 2EMXC-7D Primary Bkr 20 45 or 70 @ 60A* Pzr Pwr Oper Relief Isol Vlv Backup 20 NA 2NC33A Fuse 2EMXC-8C Primary Bkr 20 45 or 70 @ 60A* Incore Instrumentation Rm Air Backup 20 NA Hdlg Unit 2A (Normal Source)

Fuse 2EMXD-1A Primary Bkr 200 250 @ 600A Lower Containment CoolinQ Unit

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-37 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup 200 NA No. 2B (Normal Source)

Fuse 2EMXD-2A Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit Backup 200 NA No. 2D (Normal Source)

Fuse 2EMXD-3B Primary Bkr 40 45 or 70 @ 120A* Containment Pipe Tunnel Backup 40 NA Booster Fan CPT-BF-2B Fuse 2EMXD-3C Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No.

Backup 100 NA 2B (Normal Source)

Fuse 2EMXD-3D Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No.

Backup 100 NA 2D (Normal Source)

Fuse 2EMXD-4C Primary Bkr 90 110 or 125 @ 270A* Containment Air Return Fan No.

2B (CAR F-2B) 90 NA Backup Fuse 2EMXD-4D Primary Bkr 90 110 or 125 @ 270A* Hydrogen Recombiner No. 2B Backup 90 NA Fuse 2EMXD-5C Primary Bkr 20 45 or 70 @ 60A* Upper Containment Return Air Backup 20 NA Fan 2B Fuse 2EMXD-6C Primary Bkr 30 45 or 70 @ 90A* Upper Containment Air Hdlg Unit Backup 30 NA No. 2B Fuse 2EMXD-6D Primary Bkr 30 45 or 70 @ 90A* Upper Containment Air Hdlg Unit Backup 30 NA No. 2D Fuse 2EMXD-6E

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-38 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Primary Bkr 100 110 or 150 @ 300A* Hydrogen Skimmer Fan No. 2B Backup 100 NA Fuse 2EMXD-7B Primary Bkr 20 45 or 70 @ 60A* Upper Cont Return Air Fan No.

Backup 20 NA 2D Fuse 2EMXD-7C Primary Bkr 20 45 or 70 @ 60A* Pzr Pwr Oper Safety Relief Isol Backup 20 NA Vlv 2NC31 B Fuse 2EMXD-7D Primary Bkr 20 45 or 70 @ 60A* Pzr Pwr Oper Safety Relief Isol

,Backup 20 NA Vlv 2NC35B Fuse 2EMXD-8A Primary Bkr 40 45 or 70 @ 120A* PZR Cavity Booster Fan 2B Backup 40 NA (Normal Source)

Fuse 2EMXD-8B Primary Bkr 20 45 or 70 @ 60A* Incore Instrumentation Rm Air Backup 20 NA Hdlg Unit 2B (Normal Source)

Fuse 2EMXD-8D Primary Bkr 30 45 or 70 @ 90A* NC Loop 2C Disch to ND Backup 30 NA System Cont Isol Vlv 2ND1B Fuse 2MXM-FlC Primary Bkr 50 110 or 150 @ 150A* Ice Cond Floor Cooling Defrost Backup 50 NA Heater 2A Fuse 2MXM-F2A Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 2LR14 Backup 40 NA Fuse 2MXM-F2B Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 2LR15 Backup 40 NA Fuse HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-39 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION 2MXM-F2C Primary Bkr 30 45 or 70 @ 90A* Pzr Cavity Booster Fan 2B (Alt Backup 30 NA Source)

Fuse 2MXM-F2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A1 Blower A Backup 20 NA Fuse 2MXM-F2E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A2 Blower A Backup 20 NA Fuse 2MXM-F3A Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 2LR16 Backup 40 NA Fuse 2MXM-F3B Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 2LR17 Backup 40 NA Fuse 2MXM-F3C Primary Bkr 25 45 or 70 @ 75A* Reactor Bldg Equip Hdlg 5 Ton Backup 25 'NA Jib Crane Fuse 2MXM-F3D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A3 Blower A Backup 20 NA Fuse 2MXM-F3E Primary Bkr 20 45 or 70 @ 60A* Ice Cont AHU 2A4 Blower A Backup 20 NA Fuse 2MXM-F4A Primary Bkr 20 45 or 70 @ 60A* Ice Cont AHU 2A5 Blower A Backup 20 NA Fuse 2MXM-F4B Primary Bkr 20 45 or 70 @ 60A* Ice Cont AHU 2A6 Blower A Backup 20 NA Fuse

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-40 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION 2MXM-F4C Primary Bkr 20 45 or 70 @ 60A* Incore Inst Room Sump Pump Backup 20 NA Fuse 2MXM-F4D Primary Bkr 100 110 or 150 @ 300A* Upper Cont Welding Recpt Backup 100 NA Fuse 2MXM-F5A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A7 Blower A Backup 20 NA Fuse 2MXM-F5B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A8 Blower A Backup 20 NA Fuse 2MXM-F5D Primary Bkr 100 110 or 150 @ 300A* Welding Feeder Backup 100 NA Fuse 2MXM-F6A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A9 Blower A Backup 20 NA Fuse 2MXM-F6B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A1 0 Blower A Backup 20 NA Fuse 2MXM-F6C Primary Bkr 30 45 or 70 @ 90A* Lower Cont Aux Charcoal Filter Backup 30 NA Fan 2A Fuse 2MXM-F7A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A1 1 Blower A Backup 20 NA Fuse 2MXM-F7B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A12 Blower A

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-41 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup 20 NA Fuse 2MXM-F7C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A13Blower A Backup 20 NA Fuse 2MXM-F8C Primary Bkr 60 110 or 150 @ 180A* Reactor Coolant Drain Tank Backup 60 NA Pump 2A Fuse 2MXM-R1A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B1 Blower A Backup 20 NA Fuse 2MXM-R1B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B2 Blower A Backup 20 NA Fuse 2MXM-RIC Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B3 Blower A Backup 20 NA Fuse 2MXM-R 1D Primary Bkr 30 45 or 70 @ 90A* RCP 2A Oil Lift Pump No. 1 Backup 30 NA Fuse 2MXM-R2A Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 2LR12 Backup 40 NA Fuse 2MXM-R2C Primary Bkr 20 45 or 70 @ 160A* RCPM Maintenance Crane Backup 20 NA Recpt 2A, 2B, 2C, & 2D Fuse 2MXM-R2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B4 Blower A Backup 20 NA Fuse 2MXM-R2E Primary Bkr 30 45 or 70 @ 90A* RCP 2B Oil Lift Pump No. 1

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-42 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup 30 NA Fuse 2MXM-R3D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B5 Blower A Backup 20 NA Fuse 2MXM-R3E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B6 Blower A Backup 20 NA Fuse 2MXM-R3F Primary Bkr 30 45 or 70 @ 90A* RCP 2C Oil Lift Pump No. 1 Backup 30 NA Fuse 2MXM-R4D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B7 Blower A Backup 20 NA Fuse 2MXM-R4E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B8 Blower A Backup 20 NA Fuse 2MXM-R4F Primary Bkr 30 45 or 70 @ 90A* RCP 2D Oil Lift Pump No. 1 Backup 30 NA Fuse 2MXM-R5B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B9 Blower A Backup 20 NA Fuse 2MXM-R5C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B130 Blower A Backup 20 NA Fuse 2MXM-R5D Primary Bkr 175 200 @ 525A Ice Cond Equip Pwr Pnlbd 2A Backup 175 NA Fuse 2MXM-R6A HFB or HFD Circuit Breakerý'Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-43 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Primary Bkr 20 45 or 70 @ 60A* Rod Cntrl Cluster Change Backup 20 NA Fixture Hoist Drive Fuse 2MXM-R6B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B131 Blower A Backup 20 NA Fuse 2MXM-R6D Primary Bkr 150 110 or 230 @ 450A* 175 Ton Polar Crane Backup 150 NA Fuse 2MXM-R7A Primary Bkr 20 45 or 70 @ 60A* Stud Tensioner Hoist Backup 20 NA Fuse 2MXM-R7B Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive 2A Backup 20 NA Fuse 2MXM-R7C Primary Bkr 30 45 or 70 @ 90A* S/G Comp 2D Fan Backup 30 NA Fuse 2MXM-R7D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B12 Blower A Backup 20 NA Fuse 2MXM-R7E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B13 Blower A Backup 20 NA Fuse 2MXM-R8A Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive 2B Backup 20 NA Fuse 2MXM-R8B Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive 2C Backup 20 NA Fuse

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-44 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION 2MXM-R8C Primary Bkr 30 45 or 70 @ 90A* S/G Comp 2A Fan Backup 30 NA Fuse 2MXM-R8D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B14 Blower A Backup 20 NA Fuse 2MXM-R8E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B15 Blower A Backup 20 NA Fuse 2MXMA-1D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A14 Blower A Backup 20 NA Fuse 2MXMA-1E Primary Bkr 20 45 or 70 @ 60A* Cont Floor & Equip Sump 2A Backup 20 NA Pump 2A1 Fuse 2MXMA-2B Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 2LR6 Backup 40 NA Fuse 2MXMA-2C Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 2R18 Backup 40 NA Fuse 2MXMA-2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A15 Blower A Backup 20 NA Fuse 2MXMA-3A Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 2LR9 Backup 25 NA Fuse 2MXMA-3B Primary Bkr 20 45 or 70 @ 60A* Ice Cond Equip Access Door 2A

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-45' Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup 20 NA Fuse 2MXMA-3C Primary Bkr 50 110 or 150 @ 150* Ice Cond Floor Cooling Pump Backup 50 NA 2A Fuse 2MXMA-3D Primary Bkr 20 45 or 70 @ 60A* Cont Floor & Equip Sump 2B Backup 20 NA Pump 2B1 Fuse 2MXN-F1C Primary Bkr 60 110 or 150 @ 180A* Reactor Coolant Drain Tank Backup 60 NA Pump 2B Fuse 2MXN-F2A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A1 Blower B Backup 20 NA Fuse 2MXN-F2B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A2 Blower B Backup 20 NA Fuse 2MXN-F2C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A3 Blower B Backup 20 NA Fuse 2MXN-F2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A4 Blower B Backup 20 NA Fuse 2MXN-F3A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A5 Blower B Backup 20 NA Fuse 2MXN-F3B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A6 Blower B Backup 20 NA Fuse 2MXN-F3C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A7 Blower B HFB or HFD Circuit Breaker Test Response Time, Respectively.'

McGuire Units 1 and 2 16.8.1-46 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. &- (SECONDS) SYSTEM POWERED LOCATION Backup 20 NA Fuse 2MXN-F3D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A8 Blower B Backup 20 NA Fuse 2MXN-F4A Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 2LR1 Backup 25 NA Fuse 2MXN-F4B Primary Bkr 30 45 or 70 @ 90A* S/G Comp 2C Fan Backup 30 NA Fuse 2MXN-F4C Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 2LR2 Backup 25 NA Fuse 2MXN-F4D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A9 Blower B Backup 20 NA Fuse 2MXN-F4E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A10 Blower B Backup 20 NA Fuse 2MXN-F5A Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive No 2D Backup 20 NA Fuse 2MXN-F5B Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive No. 2E Backup 20 NA Fuse 2MXN-F5C Primary Bkr 20 45 or 70 @ 60A* Incore Inst Drive No. 2F Backup 20 NA Fuse 2MXN-F5D

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-47 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Primary Bkr 20 45 or 70 @ 60A* Cont Floor & Equip Sump 2A Backup 20 NA Pump 2A2 Fuse 2MXN-F6A Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 2LR4 Backup 25 NA Fuse 2MXN-F6B Primary Bkr 40 45 or 70 @ 120A* Lighting Pnlbd 2LR5 Backup 40 NA Fuse 2MXN-F6C Primary Bkr 20 45 or 70 @ 60A* Fuel Transfer Sys Reactor Side Backup 20 NA Fdr Fuse 2MXN-F6D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A1 1 Blower B Backup 20 NA Fuse 2MXN-F7B Primary Bkr 30 45 or 70 @ 90A* Pzr Cavity Booster Fan 2A (Alt Backup 30 NA Source)

Fuse 2MXN-F7D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A12 Blower B Backup 20 NA Fuse 2MXN-F7E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A13 Blower B Backup 20 NA Fuse 2MXN-F8B Primary Bkr 20 45 or 70 @ 60A* Cont Floor & Equip Sump 2B Backup 20 NA Pump 2B2 Fuse 2MXN-F8C Primary Bkr 50 110 or 150 @ 150A* Ice Cond Floor Cooling Defrost Backup 50 NA Htr 2B Fuse

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-48 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION 2MXN-R1 D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B1 Blower B Backup 20 NA Fuse 2MXN-R1E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B2 Blower B Backup 20 NA Fuse 2MXN-R1F Primary Bkr 30 45 or 70 @ 90A* RCP 2A Oil Lift Pump No. 2 Backup 30 NA Fuse 2MXN-R2C Primary Bkr 30 45 or 70 @ 90A* Reactor Cavity Manipulator Backup 30 NA Crane Fuse V 2MXN-R2F Primary Bkr 30 45 or 70 @ 90A* RCP 2B Oil Lift Pump No. 2 Backup 30 NA Fuse 2MXN-R3A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B3 Blower B Backup 20 NA Fuse 2MXN-R3B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B4 Blower B Backup 20 NA Fuse 2MXN-R3C Primary Bkr 20. 45 or 70 @ 60A* Ice Cond AHU 2B5 Blower B Backup 20 NA Fuse 2MXN-R3D Primary Bkr 30 45 or 70 @ 90A* RCP 2C Oil Lift Pump No. 2 Backup 30 NA Fuse

_________ I.

__________ I_________ I _______________

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-49 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION 2MXN-R4A Primary Bkr 50 110 or 150 @ 150A* Ice Cond Bridge Crane Backup 50 NA Fuse 2MXN-R4B Primary Bkr 30 45 or 70 @ 90A* RB Equip Hatch Hoist Backup 30 NA Fuse 2MXN-R4C Primary Bkr 25 45 or 70 @ 75A* S/G Comp 2B Fan Backup 25 NA Fuse 2MXN-R4D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B6 Blower B Backup 20 NA Fuse 2MXN-R4E Primary Bkr 30 45 or 70 @ 90A* RCP 2D Oil Lift Pump No.2 Backup 30 NA Fuse 2MXN-R5D Primary Bkr 175 200 @ 525A Ice Cond Equip Pwr Pnlbd 2B Backup 175 NA Fuse 2MXN-R6A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B7 Blower B Backup 20 NA Fuse 2MXN-R6B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B8 Blower B Backup 20 NA Fuse 2MXN-R6C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B9 Blower B Backup 20 NA Fuse 2MXN-R6D Primary Bkr 100 110 or 150 @ 300A* Welding Fdr Backup 100 NA Fuse

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-50 Revision 98

Containment Penetration OVercurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION 2MXN-R7A Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B130 Blower B Backup 20 NA Fuse 2MXN-R7B Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B131 Blower B Backup 20 NA Fuse 2MXN-R7C Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B12 Blower B Backup 20 NA Fuse 2MXN-R7D Primary Bkr 50 110 or 150 @ 150A* Ice Cond. Floor Cooling Pump Backup 50 NA 2B Fuse 2MXN-R8D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B13 Blower B Backup 20 NA Fuse 2MXN-R8E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B14 Blower B Backup 20 NA Fuse 2MXN-R8F Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2B15 Blower B Backup 20 NA Fuse Backup 30 NA Fuse 2MXNA-2B Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 2LR7 Backup 25 NA Fuse 2MXNA-2C Primary Bkr 25 45 or 70 @ 75A* Lighting Pnlbd 2LR8 Backup 25 NA Fuse 2MXNA-2D Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A14 Blower B

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-51 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup 20 NA Fuse 2MXNA-2E Primary Bkr 20 45 or 70 @ 60A* Ice Cond AHU 2A15 Blower B Backup 20 NA Fuse 2MXNA-3A Primary Bkr 20 45 or 70 @ 60A* 2 Ton CRDM Hdlg Jib Crane Backup 20 NA Fuse 2MXNA-3C Primary Bkr 20 45 or 70 @ 60A* NC Pump Motor Drain Tank Backup 20 NA Pump No. 2 Fuse 2MXNA-3D Primary Bkr 20 45 or 70 @ 60A* Ice Cond Equip Access Door 2B Backup 20 NA Fuse SMXD-3E Primary Bkr 15 45 or 70 @ 45* Unit 2 Personnel Lock Backup 15 NA Fuse SMXG-R3G Primary Bkr 20 45 or 70 @ 60A* Standby Makeup Pump to Cont Backup 20 NA Sump Isol Vlv 2NV1012C Fuse SMXG-R4F Primary Bkr 20 45 or 70 @ 60A* Standby Makeup Pump to NC Backup 20 NA Pump Seals Isol Vlv 2NV1013C Fuse SMXV-2F Primary Bkr 15 45 or 70 @ 45A* Unit 2 Emergency Personnel Backup 15 NA Lock Fuse

3. 600 VAC-Press Htr Pwr PnIs Backup Press Htr Pwr Pnl 2A-1A

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-52 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 1, 2, & 22 Backup 90 NA Fuse Backup Press Htr Pwr PnI 2A-1B Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 5, 6, & 27 Backup 90 NA Fuse Backup Press Htr Pwr Pni 2A-1C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 9, 10, & 32 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2A-2C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 11, 12, & 35 Backup 90 NA Fuse Backup Press Htr Pwr PnI 2A-2D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 13, 14, & 37 Backup 90 NA Fuse Backup Press Htr Pwr Pni 2A-2E Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 17, 18, & 42 Backup 90 NA Fuse Backup Press Htr Pwr PnI 2B-1A Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 21, 47, & 48 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2B-1B Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 26, 53, & 54 Backup 90 NA Fuse Backup Press Htr Pwr Pnf 2B-1C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 31, 59, &60 Backup 90 NA Fuse

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-53 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup Press Htr Pwr Pnl 2B-2C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 36, 65, & 66 Backup 90 NA Fuse Backup Press Htr Pwr Pni 2B-2D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 41, 71, & 72 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2B-2E Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 46, 77, & 78 Backup 90 NA Fuse Backup Press Htr Pwr PnI 2C-1A Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 7, 8, & 30 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2C-1B Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 19, 20, & 45 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2C-1C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 24, 51, & 52 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2C-1 D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 29, 57,.& 58 Backup 90 NA Fuse Backup Press Htr Pwr PnI 2C-2C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 34, 63, & 64 Backup 90 NA Fuse

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-54 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup Press Htr Pwr Pni 2C-2D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 39, 69, & 70 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2C-2E Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 44, 75, & 76 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2D-1A Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 3, 4, & 25 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2D-1 B Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 15, 16, & 40 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2D-1C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 23, 49, & 50 Backup 90 NA Fuse SMXG-R5A Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 28, 55, & 56 Backup 90 NA Fuse Backup Press Htr Pwr Pnl 2D-2C Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 33, 61, & 62 Backup 90 NA Fuse Backup Press Htr Pwr PnI 2D-2D Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 38, 67, & 68 Backup 90 NA Fuse

_______ [________ J_______ I ___________

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-55 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Backup Press Htr Pwr Pnl 2D-2E Primary Bkr 90 110 or 125 @ 270A* Pressurizer Heaters 43, 73, & 74 Backup 90 NA Fuse

4. 120 VAC-Panelboards 2KM-19 Primary Bkr 20 45 or 70 @ 60A* RCP 2A Space Htr Backup 20 NA Fuse 2KM-20 Primary Bkr 20 45 or 70 @ 60A* RCP 2C Space Htr Backup 20 NA Fuse 2KN-19 Primary Bkr 20 45 or 70 @ 60A* RCP 2B Space Htr Backup 20 NA Fuse 2KN-20 Primary Bkr 20 45 or 70 @ 60A* RCP 2D Space Htr Backup 20 NA Fuse 2KN-27 Primary Bkr 20 36 or 70 @ 60A* Fuel Handling Control Console, Backup 20 NA Fuse

5. 250 VDC-Lighting RB Deadlight Pnlbd 2DLD # 1 Primary Bkr 20 40 @ 60A Ltg Pnl Nos. 2LR1 & 2LR2 Backup 20 NA" Fuse RB Deadlight Pnlbd 2DLD # 3 Primary Bkr 20 40 @ 60A Ltg PnI Nos. 2LR4, 2LR5, &

Backup 20 NA 2LR6 Fuse HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-56 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION RB Deadlight Pnlbd 2DLD # 4 Primary Bkr 20 40 @ 60A Ltg Pnl Nos. 2LR7, 2LR8, &

Backup 20 NA 2LR9 Fuse RB Deadlight Pnlbd 2DLD # 6 Primary Bkr 20 40 @ 60A Ltg PnI Nos. 2LR12 Backup 20 NA Fuse RB Deadlight Pnlbd 2DLD #77 Primary Bkr 20 40 @ 60A Ltg Pnl Nos. 2LR16 Backup Fuse 20 NA RB Deadlight Pnlbd 2DLD # 9 Primary Bkr 20 40 @ 60A Ltg Pnl Nos. 2LR18 & 2LR17 Backup Fuse 20 NA

6. VAC - LC Outage Power Fuse Box 2A Primary Fuse 150 NA Upper Containment Safety Backup Fuse 150 NA Switch 2A Ltg PnI Nos. 2LR1

& 2LR2 Primary Fuse 150 NA Upper Containment Safety Backup Fuse 150 NA Switch 2B Primary Fuse 200 NA Lower Containment Outage Backup Fuse 200 NA Power Distr PnI 2A Primary Fuse 225 NA Lower Containment Outage Backup Fuse 225 NA Power Distr PnI 2B

7. 600 VAC -

Containment HVAC Alternate Feeders 2VTB-2A

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-57 Revision 98

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Primary Bkr 20 45 or 70 @ 60A* Incore Instrumentation Rm Air Hdlg Unit 2A (Alt Source) 2VTF-2A Primary 20 NA Incore Instrumentation Rm Air Fuse Hdlg Unit 2A (Alt Source) 2VTB-2B Primary Bkr 20 45 or 70 @ 60A* Incore Instrumentation Rm Air Hdlg Unit 2B (Alt Source) 2V-F-2B Primary 20 NA Incore Instrumentation Rm Air Fuse Hdlg Unit 2B (Alt Source) 2VRB-2A Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No.

2A (Alt Source) 2VRF-2A Primary 100 NA Control Rod Drive Vent Fan No.

Fuse 2A (Alt Source) 2VRB-2B Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No.

2B (Alt Source) 2VRF-2B Primary 100 NA Control Rod Drive Vent Fan No.

Fuse 2B (Alt Source) 2VRB-2C Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No.

2C (Alt Source) 2VRF-2C Primary 100 NA Control Rod Drive Vent Fan No.

Fuse 2C (Alt Source) 2VRB-2D Primary Bkr 100 110 or 150 @ 300A* Control Rod Drive Vent Fan No.

2D (Alt Source) 2VRF-2D

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-58 Revision 98 r--

Containment Penetration Overcurrent Protective Devices 16.8.1 TABLE 16.8.1-2 UNIT 2 Containment Penetration Conductor Overcurrent Protective Devices TRIP SETPOINT OR CONT. RATING (AMPERES) RESPONSE TIME DEVICE NO. & (SECONDS) SYSTEM POWERED LOCATION Primary. 100 NA Control Rod Drive Vent Fan No.

Fuse 2D (Alt Source) 2VLB-2A Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit

__ No. 2A (Alt Source) 2VLF-2A Primary 200 NA Lower Containment Cooling Unit Fuse No. 2A (Alt Source) 2VLB-2B Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit No. 2B (Alt Source) 2VLF-2B Primary 200 NA Lower Containment Cooling Unit Fuse No. 2B (Alt Source) 2VLB-2C Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit No. 2C (Alt Source) 2VLF-2C Primary 200 NA Lower Containment Cooling Unit Fuse No. 2C (Alt Source) 2VLB-2D Primary Bkr 200 250 @ 600A Lower Containment Cooling Unit No. 2D (Alt Source) 2VLF-2D Primary 200 NA Lower Containment Cooling Unit Fuse No. 2D (Alt Source)

HFB or HFD Circuit Breaker Test Response Time, Respectively.

McGuire Units 1 and 2 16.8.1-59 Revision 98

Switchyard Activities 16.8.2 16.8 ELECTRICAL POWER SYSTEMS 16.8.2 Switchyard Activities COMMITMENT Switchyard activities that may affect the availability and reliability of offsite power shall be identified as important to safe plant operation.

APPLICABILITY: Modes 1 through 6 REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Abnormal equipment A.1 Restore equipment to As soon as possible operating conditions normal conditions and/or and/or alignment, alignments.

TESTING REQUIREMENTS None BASES From the probabilistic risk assessment of McGuire Units 1 and 2, it may be concluded that it is important to minimize the risk of a loss of offsite power (LOOP) event, and it is important to be able to restore offsite power following a LOOP event. The identified risk significant activities are a result of an engineering review to determine those systems or actions that are significant to help maximize the availability and r~liability of offsite power. The activities are combinations or alignment and design considerations and good practices, as well as lessons learned from past industry events that have been initiators of or contributors to LOOP events.

This SLC was created to provide a method of tracking the switchyard systems for the purposes of supporting WPM 607 (Maintenance Rule Assessment of Equipment Out of Service) and 10 CFR 50.65 (Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants).

McGuire Units 1 and 2 16.8.2-1 Revision 0

Switchyard Activities 16.8.2 BASES (continued)

Switchyard activities that may affect the availability and reliability of off site power include the following:

230KV Switchyard

1. Work on equipment within the Unit 1 busline boundary. The busline boundary includes the structures, supporting structures, bus, equipment, and hardware from the high side windings of the Unit 1 main step-up transformers up to and including the busline breakers 8, 9, 11 and 12 and their associated disconnects.

2. Work on the protective relaying and/or controls and their cables for equipment within the Unit 1 busline boundary.

525KV Switchyard

1. Work on equipment within the Unit 2 busline boundary. The busline boundary includes the structures, supporting structures, bus, equipment, and hardware from the.

high side windings of the Unit 2 main step-up transformers up to and including the busline breakers 58, 59, 61 and 62 and their associated disconnects.

2. Work on the protective relaying and/or controls and their cables for equipment within the Unit 2 busline boundary.

Shared Systems between 230KV and 525KV Switchyards

1. Work on the following Switchyard AC auxiliary equipment: (Reference One Line Diagram MC 801-02)

a. AC load centers and their associated transformers:

MCOESILXSTA MCOESILXSTB MCOESILXSTC MCOESILXSTD

b. AC load center feeder circuits to Unit 1 and/or Unit 2 busline boundary equipment:

Panelboard SPA breakers 6 and 7 Panelboard SPB breakers 8 and 9 Panelboard SPC breakers 7, 8 and 9 Panelboard SPD breaker 1.1

c. AC load center feeder circuits to battery chargers:

Panelboard SPA breaker 23 Panelboard SPB breaker 23 Panelboard SPC breaker 18 BASES (continued)

McGuire Units 1 and 2 16.8.2-2 Revision 0

Switchyard Activities 16.8.2

2. Work on the following Switchyard 125V DC system equiprnment that includes the batteries, chargers, distribution bus, and panelboards. This equipment excludes the panelboard feeders which are addressed in #3 below: (Reference MC 802-01).

a. Batteries:

MCOESHBASY1 MCOESHBASY2

b. Battery Chargers:

MCOESHBCSY1 MCOESHBCSY2 MCOESHBCSY3

c. 125 VDC Switchyard Distribution Centers:

SY-DC1 SY-DC2

d. 125 VDC Switchyard Panelboards:

DYA DYB DYC DYD DYE DYF DYG DYH DYI DYJ DYK DYL DYM DYN DYO DYP

3. Work on the Switchyard 125V DC panelboard feeders that serve the Unit 1 and/or Unit 2 busline boundary equipment: (Reference MC 802-01)

a. 125 VDC Switchyard Panelboard Feeder Breakers:

DYA - 11, 12 DYB - 5, 6 DYC -1,2, 3, 4, 16, 17, 18, 19 DYD -1,3,4,5,6 DYE-9, 10, 12, 13 DYF -4,5, 6, 7, 14, 15, 16, 17 DYH - 1,3, 4, 5, 6 DYI - 11, 12 DYJ - 5, 6, 13, 14, 15, 18, 19 DYK -20 DYL - 5, 6, 7, 8,13 DYM - 9, 10, 12, 13, 18, 19, 20 DYN - 14, 15, 16 DYO -20 DYP - 6, 7, 9,10 General.

Cranes or other heavy equipment that have the potential to touch or affect any or all of the four buslines as they are moved in or out of the switchyard, or anywhere within the switchyard where they could touch or affect the buslines.

McGuire Units 1 and 2 16.8.2-3 Revision 0

Switchyard Activities 16.

8.2 REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 8

2. McGuire Nuclear Station, Technical Specifications and Bases Section 3.8

3. Nuclear System Directive 409, Nuclear Generation Department/Power Delivery Department Switchyard Interface Agreement

4. Nuclear System Directive 502, Corporate Conduct of Operations in the Switchyard

5. McGuire Units 1 and 2 PRA Risk Significant SSC's for the Maintenance Rule, MCC 1535.00-00-0006, SAAG File 208

6. NSAC-203 (EPRI), Losses of Off-Site Power at U. S. Nuclear Power Plants through 1993.

7. MC-801-02 One Line Diagram 230/525KV Switchyard 480/277 AC Load Centers (Rev 21)

8. MC-802-01 One Line Diagram 230/525KV Switchyard 125V DC System (Rev 27)

9. WPM 607, Maintenance Rule Assessment of Equipment Out of Service

10. 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.

McGuire Units 1 and 2 16.8.2-4 Revision 0

DG Supplemental Testing Requirements 16.8.3 16.8 ELECTRICAL POWER SYSTEMS 16.8.3 Diesel Generator (DG) Supplemental Testing Requirements COMMITMENT The DG supplemental testing requirements specified below shall be met.

APPLICABILITY: MODES 1, 2, 3, 4, 5, and 6 NOTE --------------------

The testing requirements for the DG batteries are not required in MODES 5 and 6.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Declare DG inoperable. Immediately TESTING REQUIREMENTS TEST FREQUENCY TR 16.8.3.1 Verify the electrolyte level of each DG battery is above 7 days the plates.

TR 16.8.3.2 Verify overall DG battery voltage is > 125 volts under a 7 days float charge.

TR 16.8.3.3 Verify DG batteries and battery racks show no visual 18 months indication of physical damage or abnormal deterioration.

(continued)

McGuire Units 1 and 2 16.8.3-1 Revision 65

DG Supplemental Testing Requirements 16.8.3 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.8.3.4 Verify DG battery'to-battery and terminal connections are 18 months clean, tight, free of corrosion and coated with anti-corrosion material.

TR 16.8.3.5 Perform DG battery service test 18 months TR 16.8.3.6 Remove accumulated water from DG day tank. After each run of

>1 hour TR 16.8.3.7 Perform DG inspection, during shutdown, in accordance 18 months with manufacturer's recommendations for this class of standby service.

TR 16.8.3.8 Verify that the fuel oil transfer pump transfers fuel from 18 months each fuel storage tank to the day tank of each DG via the installed cross-connection lines.

TR 16.8.3.9 -------------------- NOTE----------------

This Testing Requirement may be performed in conjunction with periodic pre-planned preventative maintenance activity that causes the DG to be inoperable provided that performance of the Testing Requirement does not increase the time the DG would be inoperable for the maintenance activity alone.

Verify, during shutdown, that the turning gear engaged or 18 months emergency stop features prevent DG starting only when required.

TR 16.8.3.10 Perform a pressure test of those portions of the diesel 10 years fuel oil system designed to ASME Section III, subsection ND in accordance with applicable NRC-approved ASME code requirements.

TR 16.8.3.11 For each fuel oil storage tank: 10 years

a. Drain the fuel oil;

b. Remove the sediment; and

c. Clean the tank.

TR 16.8.3.12 Verify DG battery temperature is >45 0 F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months McGuire Units 1 and 2 16.8.3-2 Revision 65

DG Supplemental Testing Requirements 16.8.3 BASES The Testing Requirements for demonstrating the OPERABILITY of the diesel generators are in accordance with the recommendations of Regulatory Guides and Generic Letters referenced below.

TR 16.8.3.9 is modified with a note. This TR specifies that it is to be performed during shutdown. This note allows the TR to be performed during preplanned Preventative Maintenance (PM) activities that would result in the diesel generator being inoperable. This TR can be performed at that time as long as it does not increase the time the diesel generator is inoperable for the PM activity that is being performed. The note is only applicable at that time. The provision of the note shall not be utilized for operational convenience.

Since the McGuire emergency diesel generator manufacturer (Nordberg) is no longer in business, McGuire engineering is the designer of record. Therefore, in the absence of manufacturer recommendations, McGuire engineering will determine the appropriate actions required for nuclear class diesel service taking into account McGuire diesel generator maintenance and operating history and industry experience where applicable.

Draining of the DG fuel oil stored in the supply tanks, removal of accumulated sediment, and tank cleaning are required at 10 year intervals by Regulatory Guide 1.137 (Ref. 7),

paragraph 2.f. TR 16.8.3.11 also requires the performance of the ASME Code,Section XI (Ref. 8), examinations of the tanks. To preclude the introduction of surfactants in the fuel oil system, the cleaning should be accomplished using sodium hypochlorite solutions, or their equivalent, rather than soap or detergents. This TR is for preventive maintenance. The presence of sediment does not necessarily represent a failure of this TR, provided that accumulated sediment is removed during performance of the Test.

The DG fuel oil storage tanks are currently deferred from the requirements of the Federal EPA regulations for underground storage tanks (USTs) on the basis that they are controlled through other programs and requirements. McGuire Environmental, Health and Safety group must be consulted regarding any changes to the testing requirements of these tanks to ensure Federal UST regulations are continued to be met.

Verification that the D/G Battery temperature is greater than or equal to 45 0 F will ensure sufficient battery capacity to perform its design function. This is based on the Diesel Generator Battery and Charger Sizing Calculation. Since a Battery Area Temperature indication is representative of the actual Diesel Generator Battery Temperature, use of this parameter is acceptable. However, if the temperature indication from this instrument is low

(< 450 F) or the instrument is out of service, the actual D/G Battery Temperature must be determined. This may be accomplished by measuring the battery skin temperature, the battery electrolyte temperature, or the internal battery compartment temperature.

REFERENCES

1. Regulatory Guide 1.9, Selection of Diesel Generator Set Capacity for Standby Power

, Supplies, March 10, 1971.

McGuire Units 1 and 2 16.8.3-3 Revision 65

DG Supplemental Testing Requirements 16.8.3

2. Regulatory Guide&1.108, Periodic Testing of Diesel Generator Units Used as Onsite Electric Power Systems at Nuclear Power Plants, Revision 1, August 1977.

3. Regulatory Guide 1.137, Fuel-Oil Systems for Standby Diesel Generators, Revision 1, October 1979.

4. Generic Letter 84-15, which modified the testing frequencies specified in Regulatory Guide 1.108.

5. Generic Letter 93-05, which reduced the surveillance requirements for testing of Diesel Generators during power operation.

6. Generic Letter 94-01, which removed the accelerated testing and special reporting requirements for Emergency Diesel Generators.

7. Regulatory Guide 1.137.

8. ASME, Boiler and Pressure Vessel Code, Section Xl.

9. McGuire Nuclear Station UFSAR, Section 18.2.4, Chemistry Control Program.

10. McGuire License Renewal Commitments MCS-1274-00-00-0016, Section 4.6, Chemistry Control Program..

11. MCC-1 381.05-00-0195, The 125 VDC Diesel Generator Battery and Charger Sizing Calculation.

McGuire Units 1 and 2 16.8.3-4 Revision 65

Fire Suppression Water System 16.9.1 16.9 AUXILIARY SYSTEMS 16.9.1 Fire Suppression Water System COMMITMENT The Fire Suppression Water System shall be OPERABLE with:

a. Fire suppression pump C and one other fire suppression pump, with their discharge aligned to the fire suppression header, and

b. An OPERABLE flow path capable of taking suction from Lake Norman and transferring water through distribution piping with OPERABLE sectionalizing control or isolation valves to the yard hydrants, the last valve ahead of the water flow alarm device on each sprinkler or hose standpipe, and the last valve ahead of the deluge valve on each Deluge or Spray System required to be OPERABLE per SLC 16.9.2 and 16.9.4.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fire suppression pumps A. 1 Restore one pump (A or B) 7 days A and B inoperable, and its associated water supply to OPERABLE OR status.

Water supply to pumps A and B inoperable.

B. Fire suppression pump B.1 Restore pump to 7 days C inoperable. OPERABLE status.

OR B.2 Verify fire suppression 7 days pumps A and B and their water supplies are OPERABLE and at least one can be aligned to the blackout diesel generator.

(continued)

McGuire Units 1 and 2 16.9.1-1 Revision 53

Fire Suppression Water System 16.9.1 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Primary automatic C.1 Verify secondary automatic Immediately starting function for one starting function for each or more required fire affected fire suppression suppression pump(s) pump is OPERABLE.

inoperable.

OR C.2 Place at least one fire Immediately suppression pump in continuous operation.

D. Secondary automatic D.1 Verify primary automatic Immediately starting function for one starting function for each or more required fire affected fire suppression suppression pump(s) pump is OPERABLE.

inoperable.

OR D.2 Place at least one fire Immediately suppression pump in continuous operation.

E. Both primary and E.1 Place at least one fire Immediately secondary automatic suppression pump in starting functions for one continuous operation.

or more required fire suppression pump(s) inoperable.

OR Jockey pumps unable to maintain system header pressure.

F. Fire Suppression Water F. 1 Establish a backup fire 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months System inoperable for suppression water system.

reasons other than Condition A, B, C, D, or E.

(continued)

McGuire Units 1 and 2 16.9.1-2 Revision 53

Fire Suppression Water System 16.9.1 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G. 1 Restore the system to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months associated Completion OPERABLE status.

Time not met.

OR G.2.1 Be in MODE 3. 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months AND G.2.2 Be in MODE 4. 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months AND G.2.3 Be in MODE 5. 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.1.1 Start each fire pump (A & B, or C) and operate for > 15 15 days on a minutes on recirculation flow.. STAGGERED TEST BASIS TR 16.9.1.2 Verify each manual, power operated, or automatic valve 31 days in flow path is in its correct position.

TR 16.9.1.3 Perform a system flush of the outside distribution loop 6 months and verify no flow blockage.

(continued)

McGuire Units 1 and 2 16.9.1-3 Revision 53

Fire Suppression Water System 16.9.1 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.9.1.4 Cycle each testable valve in flow path through one 12 months complete cycle of full travel.

TR 16.9.1.5 Verify each automatic valve in the flow path actuates to 18 months its correct position in response to a simulated automatic actuation signal.

TR 16.9.1.6 Verify each pump develops > 2500 gpm at a system 18 months pressure of > 125 psig.

TR 16.9.1.7 Cycle each valve in flow path that is not testable during 18 months plant operation through one complete cycle of full travel.

TR 16.9.1.8 Verify each fire suppression pump starts automatically in 18 months response to a simulated automatic actuation signal.

TR 16.9.1.9 Perform a system flow test in accordance with NFPA Fire 3 years Protection Handbook, 1 4 th ed., Section 11, Chapter 5.

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, Halon, and fire hose stations. The collective capability of the Fire Suppression Systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program.

In the event that portions of the Fire Suppression Systems are inoperable, alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. 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 alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression. For McGuire Nuclear Station, fire pumps A and B serve as a backup for each other. Pump C is located separately with an independent dedicated power supply.

McGuire Units 1 and 2 16.9.1-4 Revision 53

Fire Suppression Water System 16.9.1 BASES (continued)

The Testing Requirements (TR) provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met. Compliance with the testing requirements of SLC 16.9.1 ensures the main fire pumps meet all specified testing mandated by the 1978 National Fire Protection Association Code (Licensing Basis Code). Additional testing is conducted under the scope of TR 16.9.1.6 to gather pump operational data for the purpose of performance trending.

TR 16.9.1.7 requires cycling each valve in the flow path that is not testable during plant operation through one complete cycle of full travel. Although 1RF823 (Unit 1) and 1 RF834 (Unit 2) are Containment Isolation check valves in the flow path, these valves are excluded from this testing requirement for the following reasons:

1. Check valves do not perform a sectionalizing control or isolation function.

2. 1RF823 and 1 RF834 do not perform a dedicated fire protection system function.

3. NFPA 25 states that each control valve shall be operated through its full range and returned to its normal position. NFPA 25 recommends inspection of check valves internally to verify that all components operate properly, move freely, and are in good condition.

4. This exclusion is consistent with industry practices.

5. During Unit refueling outages, the Fire Suppression Water System including the check valves has been available for use.

6. Reactor Building fire hose stations are inspected every 36 months requiring opening hose valves, allowing flow through the check valves and verifying the fire protect system flow path.

7. The most common failure mode for these check valves will not affect the ability of the valve to open.

In the event the Fire Suppression Water System becomes inoperable, immediate corrective measures must be taken since this system provides the major fire suppression capability of the plant. These corrective measures include unit shutdown if a backup fire suppression water system is not established as required.

Regulatory codes and standards mandate that the fire suppression water system has automatic starting function to preclude the necessity of manual operator action. The fire suppression pumps have dual auto-start design functions to meet this requirement. The primary auto-start control circuit (ORFLP5000) will start fire pumps at higher pressure setpoints than those associated with the secondary auto-start control circuits (ORYPS5010 for pump A, ORYPS5020 for pump B and ORYPS5030 for pump C). Either primary or secondary auto-start control circuit is fully capable of providing the required automatic starting function.

Since the requirement for fire suppression pump automatic starting functions is intended to provide a high level of system standby readiness, it is equally acceptable to place at least one pump in continuous operation if all automatic starting functions are inoperable. Likewise, if the fire suppression water system jockey pumps are unable to maintain system header pressure, it is acceptable to maintain system OPERABILITY by placing at least one pump in continuous operation.

McGuire Units 1 and 2 16.9.1-5 Revision 53

Fire Suppression Water System 16.9.1 BASES (continued)

This selected licensee commitment is part of the McGuire Fire Protection Program and therefore subject to the provisions of McGuire Facility Operating License Conditions C.4 (Unit 1) and C.7 (Unit 2).

REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 9.5.1

2. McGuire Nuclear Station SER Supplement 2, Chapter 9.5.1 and Appendix D

3. McGuire Nuclear Station SER Supplement 5, Chapter 9.5.1 and Appendix B

4. McGuire Fire Protection Review, as revised

5. McGuire Nuclear Station SER Supplement 6, Chapter 9.5.1 and Appendix C

6. Fire Protection System OP/i/A/6400/02A

7. McGuire Nuclear Station Facility Operating Licenses, Unit 1 License Condition C.(4) and Unit 2 License Condition C.(7)

8. Fire Protection Handbook, 14th Edition, Published by the National Fire Protection Association, Chapter 5, Section 11

9. McGuire Nuclear Station UFSAR, Section 18.2.8, Fire Protection Program.

10. McGuire License Renewal Commitments MCS-1274.00-00-0016, Section 4.13, Fire Protection Program.

McGuire Units 1 and 2 16.9.1-6 Revision 53

Spray and/or Sprinkler Systems 16.9.2 16.9 AUXILIARY SYSTEMS 16.9.2 Spray and/or Sprinkler Systems COMMITMENT Spray and/or Sprinkler Systems in Table 16.9.2-1 shall be OPERABLE.

APPLICABILITY Whenever equipment protected by the Spray/Sprinkler System is required to be OPERABLE.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Establish a continuous fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Spray and/or Sprinkler watch with backup fire Systems inoperable in suppression equipment.

an area in which redundant systems or components could be damaged.

B. One or more required B. 1 Establish fire watch patrol. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Spray and/or Sprinkler Systems inoperable in AND areas other than Once per hour Condition A. thereafter TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.2.1 Verify each manual, power operated, or automatic valve 31 days in flow path which is accessible during plant operation is in its correct position.

(continued)

McGuire Units 1 and 2 16.9.2-1 Revision 81

Spray and/or Sprinkler Systems 16.9.2 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.9.2.2 Cycle each testable valve in flow path through one 12 months complete cycle of full travel.

TR 16.9.2.3 Verify each automatic valve in the flow path actuates to 18 months its correct position in response to a simulated automatic Fire Detection signal.

TR 16.9.2.4 Cycle each valve in flow path that is not testable during 18 months plant operation through one complete cycle of full travel.

TR 16.9.2.5 Perform a visual inspection of the dry pipe spray and 18 months sprinkler headers to verify their integrity.

TR 16.9.2.6 Perform a visual inspection of each nozzle's spray area to 18 months verify the spray pattern is not obstructed.

TR 16.9.2.7 Verify each manual, power operated, or automatic valve 18 months in flow path which is not accessible during plant operation is in its correct position.

TR 16.9.2.8 Perform an air flow test through each open head 3 years spray/sprinkler header and verify each open head nozzle is unobstructed.

McGuire Units 1 and 2 16.9.2-2 Revision 81

Spray and/or Sprinkler Systems 16.9.2 TABLE 16.9.2-1 Spray and Sprinkler Systems Elevation Building Room No. Equipment Elevation 695 ft Auxiliary 501 RHR Pump 1A 500 RHR Pump 1B 506 RHR Pump 2A 507 RHR Pump 2B 508 Corridor Elevation 716 ft Auxiliary 600 Aux. FW Pump Room - Unit 1 649 Nuclear Service Water Pumps 627 Centrifugal Charging Pump 1A 630 Centrifugal Charging Pump 1B 601 Aux. FW Pump Room - Unit 2 634 Centrifugal Charging Pump 2A 637 Centrifugal Charging Pump 2B 648 Cable Shaft Elevation 733 ft Auxiliary 723 Component Cooling Pumps 701 Battery Room Trench Area Elevation 750 ft Auxiliary 801 Cable Room - Unit 1 801C Cable Room.- Unit 2 806 Component Cooling Pumps 803A 1 ETA HVAC Equip Room 805A 2ETA HVAC Equip Room Elevation 725 ft Reactor Pipe Corridor Lower Containment Ventilation Filters Elevation 738 ft Reactor Annulus McGuire Units 1 and 2 16.9.2-3 Revision 81

Spray and/or Sprinkler Systems 16.9.2 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, Halon, and fire hose stations. The collective capability of the Fire Suppression Systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program.

In the event that portions of the Fire Suppression Systems are inoperable, alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. 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 alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression.

The Testing Requirements (TS) provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met.

TR 16.9.2.4 requires cycling each valve in the flow path that is not testable during plant operation through one complete cycle of full travel. Although 1RF823 (Unit 1) and 1 RF834 (Unit 2) are Containment Isolation check valves in the flow path, these valves are excluded from this testing requirement for the following reasons:

1. Check valves do not perform a sectionalizing control or isolation function.

2. 1RF823 and 1RF834 do not perform a dedicated fire protection system function.

3. NFPA 25 states that each control valve shall be operated through it's full range and returned to it's normal position. NFPA 25 recommends inspection of check valves internally to verify that all components operate properly, move freely, and are in good condition.

4. This exclusion is consistent with industry practices.

5. During Unit refueling outages, the Fire Suppression Water System including the check valves has been available for use.

6. Reactor Building fire hose stations are inspected every 36 months requiring opening hose valves, allowing flow through the check valves and verifying the fire protect system flow path.

7. The most common failure mode for these check valves will not affect the ability of the valve to open.

This selected licensee commitment is part of the McGuire Fire Protection Program and therefore subject to the provisions of McGuire Facility Operating License Conditions C.4 (Unit 1) and C.7 (Unit 2).

McGuire Units 1 and 2 16.9.2-4 Revision 81

Spray and/or Sprinkler Systems 16.

9.2 REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 9.5.1

2. McGuire Nuclear Station SER Supplement 2, Chapter 9.5.1 and Appendix D

3. McGuire Nuclear Station SER Supplement 5, Chapter 9.5.1 and Appendix B

4. McGuire Fire Protection Review, as revised

5. McGuire Nuclear Station SER Supplement 6, Chapter 9.5.1 and Appendix C

6. MCFD-1599 1.0 through 3.01

7. McGuire Nuclear Station Facility Operating Licenses, Unit 1 License Condition C.(4) and Unit 2 License Condition C.(7)

8. McGuire Nuclear Station UFSAR, Section 18.2.8, Fire Protection Program.

9. McGuire License Renewal Commitments MCS-1274.00-00-0016, Section 4.13, Fire Protection Program.

McGuire Units 1 and 2 16.9.2-5 Revision 81

Halon Systems 16.9.3 16.9 AUXILIARY SYSTEMS 16.9.3 Halon Systems COMMITMENT The following Halon Systems shall be OPERABLE:

a. Elevation 716 ft. - Auxiliary Building Room No. Equipment 600B Turbine Driven Aux. FW Pump - Unit 1 601B Turbine Driven Aux. FW Pump - Unit 2

b. Elevation 733 ft. - Auxiliary Building Room No. Equipment 703-704 Diesel Generators - Unit 1 714-715 Diesel Generators - Unit 2 APPLICABILITY Whenever equipment protected by the Halon System is required to be OPERABLE.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A. 1 Establish a continuous fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Halon Systems watch with backup fire inoperable in an area in suppression equipment.

which redundant systems or components could be damaged.

B. One or more required B. 1 Establish fire watch patrol. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Halon Systems inoperable in areas other AND than Condition A.

Once per hour thereafter McGuire Units 1 and 2 16.9.3-1 Revision 106

Halon Systems 16.9.3 TESTING REQUIREMENTS TEST FREQUENCY

'TR 16.9.3.1 Verify each manual, power operated, or automatic valve 31 days in flow path is in its correct position.

TR 16.9.3.2 Verify Halon storage tank weight > 95% of full charge 6 months weight and pressure > 90% of full charge pressure.

TR 16.9.3.3 Verify system actuates upon receipt of a simulated 18 months manual and automatic actuation signal and damper closure devices receive an actuation signal upon system operation.

TR 16.9.3.4 Perform a flow test through headers and nozzles to 18 months assure no blockage.

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, Halon, and fire hose stations. The collective capability of the Fire Suppression Systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program.

In the event that a Halon System becomes inoperable, compensatory actions are required to be taken in the affected areas until the inoperable equipment is restored to service.

For Rooms 704 (1B D/G), 714 (2A D/G) and 715 (2B D/G), Condition B is applied if the Halon system is declared inoperable since there is no impact to redundant systems and components.

For Room 703 (1A D/G), Condition A is applied if the Halon system is declared inoperable since there is impact on redundant equipment (1CA-42B power cable).

The Testing Requirements provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met. An allowance is made for ensuring a sufficient volume of Halon in the Halon storage tanks by verifying either the weight or the level of the tanks. Level measurements are made by either a UL or FM approved method.

The main bank (1 cylinder for the TDCA Pump Room, 8 cylinders for the D/G Room) or the reserve bank (1 cylinder for the TDCA Pump Room, 8 cylinders for the D/G Room) provides McGuire Units 1 and 2 16.9.3-2 Revision 106

Halon Systems 16.9.3 a sufficient quantity of halon to totally flood any of the TD CA Pump Rooms or Diesel Generator Rooms with the required design concentrations. Therefore, the Halon System is OPERABLE with the system aligned to either the main or the reserve bank of cylinders.

The system is aligned to the main or reserve bank of cylinders by means of a local manual toggle switch.

TR 16.9.3.1 requires that valves in the flow path for the required halon systems be verified to be in their correct position. Although the selector valves and the cylinder valves are in the flow path, these valves are excluded from this testing requirement for the following reasons:

1. There is no visible means of determining valve position,

2. The valves are spring loaded piston actuators which fail closed and require halon discharge header pressure to open (Selector Valves Only),

3. There is no credible means to mis-position these valves other than actual actuation of the halon system,

4. These valves are an integral component of the actuation circuitry for the halon system, which is tested per TR 16.9.3.3, and

5. This exclusion is consistent with fire protection industry practices.

This selected licensee commitment is part of the McGuire Fire Protection Program and therefore subject to the provisions of McGuire Facility Operating License Conditions C.4 (Unit 1) and C.7 (Unit 2).

REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 9.5.1

2. McGuire Nuclear Station SER Supplement 2, Chapter 9.5.1 and Appendix D

3. McGuire Nuclear Station SER Supplement 5, Chapter 9.5.1 and Appendix B

4. McGuire Fire Protection Review, as revised

5. McGuire Nuclear Station SER Supplement 6, Chapter 9.5.1 and Appendix C

6. MCM-1206.07-35

7. MC-1599 4.0, MC-2599-4.0

8. MCEE-120.08.07

9. MCEE-120.16.07

10. MCEE-133-00.17 McGuire Units 1 and 2 16.9.3-3 Revision 106

Halon Systems 16.9.3

11. McGuire Nuclear Station Facility Operating Licenses, Unit 1 License Condition C.(4) and Unit 2 License Condition C.(7)

McGuire Units 1 and 2 16.9.3-4 Revision 106

Fire Hose Stations 16.9.4 16.9 AUXILIARY SYSTEMS 16.9.4 Fire Hose Stations COMMITMENT The fire hose stations shown in Table 16.9.4-1 shall be OPERABLE.

APPLICABILITY Whenever equipment in areas protected by the fire hose stations is required to be OPERABLE.

REMEDIAL ACTIONS

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

1. One outlet of the wye shall be connected to the standard length of hose provided for the hose station. The second outlet of the wye shall be connected to a length of hose sufficient to provide coverage for the area left unprotected by the inoperable hose station.

2. Where it can be demonstrated that the physical routing of the fire hose would result in a recognizable hazard to operatirng technicians, plant equipment, or the hose itself, the fire hose shall be stored in -a roll at the outlet of the OPERABLE hose station.

3. Signs shall be mounted above the gate wye(s) to identify the proper hose to use.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more fire hose A.1 Provide gated wye(s) on 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months stations inoperable in an nearest OPERABLE hose area in which the hose is station(s).

the primary means of fire suppression.

B. One or more fire hose B.1 Provide gated wye(s) on 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months stations inoperable in an nearest OPERABLE hose area in which the hose is station(s).

not the primary means of fire suppression.

McGuire Units 1 and 2 16.9.4-1 Revision 83

Fire Hose Stations 16.9.4 TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.4.1 Perform visual inspection of the fire hose stations, Quarterly or in accessible during plant operations, to assure all required accordance with equipment is at the station and the fire hose shows no the performance physical damage. based criteria stated in the Bases TR 16.9.4.2 Perform a visual inspection of the fire hose stations not 18 months accessible during plant operations to assure all required equipment is at the station and the fire hose shows no physical damage.

TR 16.9.4.3 Remove each fire hose for inspection and reracking. 18 months TR 16.9.4.4 Inspect all fire hose gaskets and replace degraded 18 months gaskets in the couplings.

TR 16.9.4.5 Open each hose station valve partially to verify valve 3 years OPERABILITY and no flow blockage.

TR 16.9.4.6 Conduct a hose hydrostatic test at a pressure > 150 psig 3 years or > 50 psig above maximum fire main operating pressure, whichever is greater.

McGuire Units 1 and 2 16.9.4-2 Revision 83

Fire Hose Stations 16.9.4 TABLE 16.9.4-1 Page 1 of 2 FIRE HOSE STATIONS Number Location Elevation (feet) 157 55-FF 695 158 57-FF 695 175 51-LL/MM 716 176 55-MM 716 177 55-QQ 716 178 58/59-MM 716 179 61-LL 716 180 52-CC 716 181 54-C6 716 182 58-CG 716 183 59-CC /DD 716 167 51 -JJ/KK 733 168 52-MM/NN 733 169 55-NN 733 170 57-LL 733 171 54-HH 733 172 58HH 733 173 60-MM/NN 733 174 61 -JJ/KK 733 887 53-DD 733 889 51/52-DD 733 890 51-BB 733 891 40-CC 733 892 43/44-DD 733 893 40-AA/BB 733 894 44-AA/BB 733 895 46-BB 733 897 60-DD 733 898 61-BB 733 899 66-BB 733 900 68-AA/BB 733 901 72-BB 733 902 68/69-DD 733 903 72-DD 733 904 58-CC/DD 733 McGuire Units 1 and 2 16.9.4-3 Revision 83

Fire Hose Stations 16.9.4 TABLE 16.9.4-1 Page 2 of 2 FIRE HOSE STATIONS Number Location Elevation (feet) 913 45-AA/BB 733 914 66BB 733 1184 56-JJ 733 161 50/51-MM 750 162 54/55-LL 750 163 54-JJ 750 164 56-QQ 750 165 58-LL/MM 750 166 61-MM 750 302 60-KK 750 303 52-GG 750 961 45-BB 750 962 46-CC 750 963 51-BB 750 964 51-BB 750 965 51-CC 750 966 56-DD 750 967 67-BB 750 968 66-CC 750 969 61-CC 750 970 61-BB 750 971 58-BB 750 972 57-DD 750 1185 58-JJ 750 184 54-KK 767 185 54-MM 767 186 50/51-MM 767 191 56-CG 767 192 58-JJ 767 193 60-MM 767 194 61/62-MM 767 974 51-BB 767 975 61-BB 767 McGuire Units 1 and 2 16.9.4-4 Revision 83

Fire Hose Stations 16.9.4 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, Halon, and fire hose stations. The collective capability of the Fire Suppression Systems is adequate to minimize potential damage to safety-related equipment and is a major element in the facility fire protection program.

In the event that portions of the Fire Suppression Systems are inoperable, alternate backup fire-fighting equipment is required to be made available in the affected areas until the inoperable equipment is restored to service. 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 alternate means of fire fighting than if the inoperable equipment is the primary means of fire suppression.

The Testing Requirements provide assurance that the minimum OPERABILITY requirements of the Fire Suppression Systems are met.

The location of the required equipment at the fire hose station and the physical condition of fire hose is critical to fire brigade operations. The option of increasing or decreasing the frequency of the fire hose station inspections, based on hose performance, allows the ability to optimize plant resources. Should an adverse trend develop with fire hose station equipment or fire hose condition, the frequency of the inspection shall be increased.

Similarly if the fire hose station equipment or fire hose condition trends are positive, the frequency of verification could be decreased. Through programmed trending of fire hose station inspections, fire hose stations will be maintained at predetermined reliability standards. The option to modify the frequency of SLC TR 16.9.4.1 is the responsibility of the Site Fire Protection Engineer via trending analysis of previous inspection results based on the following:

Annual review of the results of the completed fire hose station inspection procedures.

- If the results demonstrate that the fire hose stations are found acceptable at least 99% of the time over the 3 year rolling period, 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 (including grace period).

- If the results demonstrate 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 quarterly or - quarterly to monthly - as applicable. The verification need not be conducted more often than monthly.

This commitment is part of the McGuire Fire Protection Program and therefore subject to the provisions of McGuire Facility Operating License Conditions 2.C.(4) (Unit 1) and 2.C.(4)

(Unit 2).

McGuire Units 1 and 2 16.9.4-5 Revision 83

Fire Hose Stations 16.

9.4 REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 9.5.1

2. McGuire Nuclear Station SER Supplement 2, Chapter 9.5.1 and Appendix D

3. McGuire Nuclear Station SER Supplement 5, Chapter 9.5.1 and Appendix B

4. McGuire Fire Protection Review, as revised

5. McGuire Nuclear Station SER Supplement 6, Chapter 9.5.1 and Appendix C

6. McGuire Nuclear Station Facility Operating Licenses, Unit 1 License Condition C.(4) and Unit 2 License Condition 2.C.(4).

7. McGuire Nuclear Station UFSAR, Section 18.2.8, Fire Protection Program.

8. McGuire License Renewal Commitments MCS-1274.00-00-0016, Section 4.13, Fire Protection Program.

McGuire Units 1 and 2 16.9.4-6 Revision 83

Fire Rated Assemblies 16.9.5 16.9 AUXILIARY SYSTEMS 16.9.5 Fire Rated Assemblies COMMITMENT All fire rated assemblies (walls, flbors/ceilings, cable tray enclosures and other fire barriers) separating:

a. Redundant analyzed Post Fire Safe Shutdown Equipment, or

b. Control Complex (i.e., Control Room, Cable Rooms and Battery Rooms) from the remainder of the plant, or

c. Safety related from non-safety related areas, or

d. Containment from non-containment areas, AND All sealing devices (fire doors, fire windows, fire dampers, cable, piping and ventilation duct penetration seals) in fire rated assembly penetrations shall be OPERABLE.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Establish a continuous fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months fire rated assemblies watch on at least one side and/or sealing devices of the affected assembly.

inoperable. OR A.2.1 Verify fire detectors on at 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months least one side of the inoperable assembly are OPERABLE.

AND A.2.2 Establish fire watch patrol. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months AND Once per hour thereafter OR A.3 Complete an evaluation as Prior to terminating permitted by NRC RIS Required Action Al 2005-07 to institute or A2 required action(s)

McGuire Units 1 and 2 - 16.9.5-1 Revision 81

Fire Rated Assemblies 16.9.5 TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.5.1 Verify that each unlocked fire door without electrical 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months supervision is closed.

TR 16.9.5.2 Verify that each fire door with automatic hold-open and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months release mechanisms are free of obstructions.

TR 16.9.5.3 Verify that each locked closed fire door is closed. 7 days TR 16.9.5.4 Perform a TADOT of the Fire Door Supervision System 31 days for each electrically supervised fire door.

TR 16.9.5.5 Perform an inspection of the automatic hold-open, 6 months release and closing mechanisms and latches on each associated fire door.

TR 16.9.5.6 Perform a functional test of the automatic hold-open, 18 months release and closing mechanisms and latches on each associated fire door.

TR .16.9.5.7 Perform a visual inspection of the exposed surfaces of 18 months each required fire rated assembly.

TR 16.9.5.8 ------------------ NOTE -------------------

Samples shall be selected such that each damper will be inspected every 15 years.

Perform a visual inspection of 10% of all required fire 18 months windows, fire dampers, and associated hardware.

TR 16.9.5.9 ------------------- NOTES -----------------

1. If a seal is found inoperable, an additional 10% of each type of sealed penetration shall be inspected until a 10% sample with no inoperable seals is found.

2. Samples shall be selected such that each penetration seal will be inspected every 15 years.

Perform a visual inspection on 10% of each type of 18 months sealed penetration. I McGuire Units 1 and 2 16.9.5-2 Revision 81

Fire Rated Assemblies 16.9.5 BASES The functional integrity of the fire rated assemblies, including associated penetration seals, ensures that fires will be confined or adequately retarded so that the following criteria are achieved:

Fire will not damage redundant analyzed Post Fire Safe Shutdown equipment, Fire will not spread from the balance of plant to the Control Complex, Fire will not spread from non-safety related areas to safety related areas, and Fire will not spread from non-containment areas to containment areas.

The fire related assemblies and associated penetration seals are a passive element in the facility fire protection program and are subject to periodic inspections.

Fire rated assemblies, including associated penetration seals (fire doors, fire windows, fire dampers, cable, piping and ventilation duct penetration seals) are considered operable when the visually observed condition is not degraded to a point that the assembly cannot perform its intended function. For fire rated assemblies that are questionable, an evaluation shall be performed, using the Problem Investigation Process, to determine the cause of any identified abnormal change in appearance or abnormal degradation and the effects of this change on the ability of the fire rated assembly to perform its function. Based on the results of the investigation process, additional assemblies may be selected for inspection.

During periods of time when a fire rated assembly is not OPERABLE, either: (A 1) a continuous fire Watch is required to be maintained on at least one side of the affected barrier, or (A 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 the barrier is restored to OPERABLE status,or (A3) a licensee may choose to implement a different required action or combination of actions (e.g., additional administrative controls, operator briefings, temporary procedures, interim shutdown strategies, operator manual actions, temporary fire barriers, temporary detection or suppression systems). Such a change must be made to the approved Fire Protection Plan (FPP). However, the licensee must complete a documented evaluation of the impact of the proposed required action to the FPP. The evaluation must demonstrate that the required actions would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire. Any change to the FPP must maintain compliance with the GDC and 10CFR50.48(a).

The evaluation of the required action should incorporate risk insights regarding the location, quantity, and type of combustible material in the fire area; the presence of ignition sources and their likelihood of occurrence; the automatic fire suppression and the fire detection capability in the fire area; the manual fire suppression capability in the fire area; and the human error probability where applicable.

The expectation is to promptly complete the corrective action at the first available opportunity and eliminate the reliance on the required action.

This Selected Licensee Commitment is part of the McGuire Fire Protection Program and therefore subject to the provisions of McGuire Facility Operating License Conditions C.4 (Unit 1) and C.7 (Unit 2).

McGuire Units 1 and 2 16.9.5-3 Revision 81

Fire Rated Assemblies 16.

9.5 REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 9.5.1

2. McGuire Nuclear Station SER Supplement 2, Chapter 9.5.1 and Appendix D

3. McGuire Nuclear Station SER Supplement 5, Chapter 9.5.1 and Appendix B

4. McGuire Fire Protection Review, as revised

5. McGuire Nuclear Station SER Supplement 6, Chapter 9.5.1 and Appendix C

6. McGuire Nuclear Station Facility Operating Licenses, Unit 1 License Condition C.(4) and Unit 2 License Condition C.(7)

7. Door Schedules MC-1108-01, MC-1208-01-01, -02, -03, -04, -05 and -06.

8. Fire Plan Drawings MC-1384-07 series.

9. McGuire Nuclear Station UFSAR, Section 18.2.8, Fire Protection Program.

10. McGuire Licensing Renewal Commitments MCS-1274.00-00-0016, Section 4.13, Fire Protection Program.

11. NRC Regulatory Issue Summary 2005 Compensatory Measures to Satisfy the Fire Protection Program Requirements, April 19, 2005 McGuire Units 1 and 2 16.9.5-4 Revision 81

Fire Detection Instrumentation 16.9.6 16.9 AUXILIARY SYSTEMS 16.9.6 Fire Detection Instrumentation COMMITMENT The fire detection instrumentation for each fire detection zone shown in Table 16.9.6-1 shall be OPERABLE.

APPLICABILITY Whenever equipment protected by fire detection instrument is required to be OPERABLE.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more, but not A.1 Restore the inoperable 14 days more than half, of the instrument(s) to Function A fire detectors OPERABLE status.

in any zone inoperable.

B. More than half of the B. 1 Establish fire watch patrol 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Function A fire detectors to inspect zones outside AND in any zone inoperable, containment with inoperable instruments. Once per hour thereafter OR AND One or more Function B B.2.1 Establish a fire watch patrol 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months fire detectors inoperable, to inspect zones inside AND containment with OR inoperable instruments. Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter Two or more adjacent OR fire detectors inoperable.

B.2.2 Monitor containment air Once per hour OR temperature at the locations given in ITS Required Action and 3.6.5.1 or 3.6.5.2.

associated Completion Time of Condition A not met.

(continued)

McGuire Units 1 and 2 16.9.6-1 Revision 107

Fire Detection Instrumentation 16.9.6 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more annulus fire C.1 Perform a fire watch patrol 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months detectors inoperable, of the annulus.

AND C.2.1 Verify at least one adjacent Once per hour annulus fire detector zone thereafter is OPERABLE.

OR C.2.2 Perform a fire watch patrol Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of the annulus if no thereafter adjacent zone is OPERABLE.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.6.1 Verify the non-supervised circuits associated with 31 days detector alarms between the instrument and the control room are OPERABLE.

TR 16.9.6.2 Verify the NFPA Standard 72D supervised circuits 6 months supervision associated with detector alarms are OPERABLE.

TR 16.9.6.3 Perform a TADOT on fire detectors which are accessible 6 months during plant operation.

(continued)

McGuire Units 1 and 2 16.9.6-2 Revision 107

Fire Detection Instrumentation 16.9.6 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.9.6.4 Perform a TADOT on fire detectors which are not Prior to entering accessible during plant operation.> MODE 4 when the unit has been in MODE 5 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more, if testing has not been performed in previous 6 months TR 16.9.6.5 -------------------- NOTE -----------------

Different detectors shall be selected for each test.

Perform a TADOT on at least one detector on each 6 months signal initiating circuit for fixed temperature/rate of rise restorable spot type heat detectors which are accessible during plant operation.

TR 16.9.6.6 Perform a TADOT on fixed temperature/rate of rise Prior to entering restorable spot type heat detectors which are not MODE 4 when the accessible during plant operation. unit has been in MODE 5 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more, if testing has not been performed in previous 6 months TR 16.9.6.7 -------------------- NOTES-----------------

1. For each failure that occurs, two additional detectors shall be removed and tested.

2. Replacement of all fixed temperature/rate of rise non-restorable spot type heat detectors within population satisfies testing requirement.

Perform a functional test on at least 2% of the fixed 5 years temperature/rate of rise non-restorable spot type heat detectors.

TR 16.9.6.8 Replace all fixed temperature/rate of rise non-restorable 15 years spot type heat detectors.

McGuire Units 1 and 2 16.9.6-3 Revision 107

Fire Detection Instrumentation 16.9.6 TABLE 16.9.6-1 FIRE DETECTION INSTRUMENTATION"1 '

Detector Description Location Number of Number of Function*'

Zone Smoke Detectors Heat Detectors 1 Reactor Coolant Pump 1A RCP-1A 0 1 A 2 Reactor Coolant Pump 1B RCP-1B 0 1 A 3 Reactor Coolant Pump 1C RCP-1C 0 1 A 4 Reactor Coolant Pump 1D RCP-1D 0 1 A 5 Reactor Coolant Pump 2A RCP-2A 0 1 A 6 Reactor Coolant Pump 2B RCP-2B 0 1 A RCP-2C 0 1 A 7 Reactor Coolant Pump 2C 8 Reactor Coolant Pump 2D RCP-2D 0 1 A 29 Aux. Bldg. Vent Filter KK52-53 EL. 767 2 0 A 30 Elec. Pen. Rm. CC-51 EL. 767 9 0 A CC-51 EL. 750 10 0 A 31 Elec. Pen. Rm.

CC-51 EL. 733 11 0 A 32 Elec. Pen. Rm.

KK-59/60 EL. 767 2 0 A 33 Unit 2 Aux. Bldg. Vent. Filter CC-61 EL. 767 9 0 A 34 Unit 2 Elec. Penetration Room CC-61 EL. 750 10 0 A 35 Unit 2 Elec. Penetration Room CC-61 EL. 733 11 0 A 36 Unit 2 Elec. Penetration Room 37 Diesel Gen. 1A CC-43 EL. 733 0(0) 8(4) A(B) 38 Diesel Gen. 2A CC-69 EL. 733 0(0) 8(4) A(B) 39 Cable Room CC-55 EL. 750 6 5 A 40 Control Room CC-56 EL. 767 24 19 A 41 Swgr. Rm. IETA AA-49 EL. 750 9 0 A 42 Swgr. Rm. IETB AA-49 EL. 733 10 2 A 43 SWG. Room 2ETA AA-62 EL. 750 9 0 A 44 SWG. Room 2ETB AA-62 EL. 733 10 2 A 45A Battery Room EVCA CC-54 EL. 733 2 2 A 45B Battery Room EVCB CC-55 EL. 733 2 2 A McGuire Units 1 and 2 16.9.6-4 Revision 107

Fire Detection Instrumentation 16.9.6 TABLE 16.9.6-1 FIRE DETECTION INSTRUMENTATION(1 )

Detector Description Location Number of Number of Function72 7 Zone Smoke Detectors Heat Detectors 45C Battery Room EVCC CC-56 EL. 733 2 2 A 45D Battery Room EVCD CC-57 EL. 733 2 2 A 45G Battery Chg. Equip. & Pnl EVCA, EVCC CC-56 EL. 733 13 0 A 45H Battery Chg. Equip. & PnI EVCB, EVCD BB-56 EL. 733 12 0 A 50 Diesel Gen. 1B BB-43 EL. 733 0(0) 8(4) A(B) 51 Diesel Gen. 2B BB-69 EL. 733 0(0) 8(4) A(B) 52 Unit 2 Cable Room CC-57 EL. 750 6 5 A 61 Cont. Spray. Pump 1A/Corridor GG-55 EL. 695 2 2 A 62 Cont. Spray Pump 1 B/Cooridor GG-56 EL. 695 2 2 A 63 RHR Pump 1B FF-54 EL. 695 1 1 A 64 RHR Pump 1A GG-54 EL. 695 1 1 A 66 Cont. Spray Pump 2B/Corridor GG-56 EL. 695 2 2 A 67 Cont. Spray Pump 2A/Cooridor GG-57 EL. 695 2 2 A 68 RHR Pump 2A GG-58 EL. 695 1 1 A 69 RHR Pump 2B FF-58 EL. 695 1 1 A 70 Aux. FW Pumps BB-51 EL. 716 10(0) 80) A(B) 72 Mech. Pen. Rm./Cables JJ-51 EL. 716 4 4 A 73 Corridor/Cables HH-53 EL. 716 5 5 A 74 Sample Panel/Cables EE-55 EL. 716 5 5 A 75 Cent. Chg. Pump 1B JJ-55 EL. 716 2 2 A 76 Cent. Chg. Pump 1A JJ-55 EL. 716 2 2 A 77 PD Pump #1 JJ-54 EL. 716 2 2 A 78 Safety Injection Pump 1A HH-54 EL. 716 2 2 A 79 Safety Injection Pump 1B GG-54 EL. 716 2 2 A 80 Aisle/Cables GG-55 EL. 716 12 12 A 81 Aisle/Cables GG-57 EL. 716 10 10 A McGuire Units 1 and 2 16.9.6-5 Revision 107

Fire Detection Instrumentation 16.9.6 TABLE 16.9.6-1 FIRE DETECTION INSTRUMENTATION"1 '

Detector Description Location Number of Number of Function"'

Zone Smoke Detectors Heat Detectors 82 Cent. Chg. Pump 2B JJ-57 EL. 716 2 2 A 83 Cent. Chg. Pump 2A JJ-57 EL. 716 2 2 A 84 PD Pump #2 JJ-58 EL. 716 2 2 A 85 Safety Injection Pump 2A HH-58 EL. 716 2 2 A 86 Safety Injection Pump 2B GG-58 EL. 716 2 2 A 87 Aux. FW Pumps CC-60 EL. 716 10(0) 8(1) A(B) 88 Mech. Penetration Room/Cables JJ-61 EL. 716 4 4 A 90 Corridor/Cables NN-59 EL. 716 5 5 A 91 Corridor/Cables EE-53 EL. 733 4 4 A 92 Corridor/Cables JJ-51 EL. 733 6 6 A 93 Corridor/Cables NN-52 EL. 733 11 11 A 94 Aisle/Cables JJ-55 EL. 733 9 9 A 95 600V MCC 1EMXB - 1 EMXB3 FF-55 EL. 733 1 1 A 96 Cable Tray Access EE-55 EL. 733 1 1 A 97 Cable Tray Access EE-57 EL. 733 1 1 A 98 600V MCC 2EMXB - 2EMXB3 FF-57 EL. 733 1 1 A 99 Aisle/Cables JJ-57 EL. 733 9 9 A 100 Corridor/Cables NN-58 EL. 733 12 12 A 101 Corridor/Cables JJ-61 EL. 733 6 6 A 102 Corridor/Cables EE-59 EL. 733 4 4 A 103 Corridor/Cables MM-51 EL. 750 6 6 A 104 Hatch Area Cables LL-53 EL. 750 7 7 A 106 600V MCC 1 EMXA FF-54 EL. 750 2 2 A 107 600V MCC 2EMXA FF-57 EL. 750 3 3 A 108 Aisle/Cables JJ-55 EL. 750 14 14 A 109 Hatch Area Cables PP-57 EL. 750 15 15 A 110 Corridor/Cables PP-60 EL. 750 8 8 A 111 Corridor/Cables LL-59 EL. 750 6 6 A McGuire Units 1 and 2 16.9.6-6 Revision 107

Fire Detection Instrumentation 16.9.6 TABLE 16.9.6-1 FIRE DETECTION INSTRUMENTATION"1 '

Detector Description Location Number of Number of Function"2 '

Zone Smoke Detectors Heat Detectors 112 Aisle/Cables JJ-57 EL. 750 13 13 113 HVAC Equipment Area/Cables FF-56 EL. 767 8 8 114 Respiratory Equipment Room GG-54 EL. 767 1 1 115 Corridor/Cables JJ-54 EL. 767 13 13 116 HVAC Equipment Area/Cables NN-52 EL. 767 7 7 120 Environmental Lab PP-55 EL. 767 1 1 122 HVAC Equipment Area NN-59 EL. 767 7 7 123 Corridor/Cables JJ-57 EL. 767 14 14 125 Fuel Pool Area NN-62 EL. 778+10 19 14 127 Fuel Pool Area NN-50 EL. 731+ 6 18 14 128 Aisle/Cable EE-57 EL. 716 5 5 129 600V MCC 2EMXH KK-56 EL. 733 1 1 130 Cables/KF Pumps PP-52 EL. 750 4 4 131 Respiratory HH-56 EL. 767 5 5 134 RB Pipe Corridor-Unit 1 2150 -2700 0 5 135 RB Pipe Corridor-Unit 1 2700- 3150 0 5 136 RB Pipe Corridor-Unit 1 3150 - 00 0 6 137 RB Pipe Corridor-Unit 1 ^00 - 440 0 4 138 RB Pipe Corridor-Unit 1 440 - 900 0 4 139 RB Pipe Corridor-Unit 1 900 - 1260 0 4 140 RB Pipe Corridor-Unit 1 1260 - 1730 0 7 141 RB Below Oper. Floor-Unit 1 3290 - 3490 0 7 142 RB Below Oper. Floor-Unit 1 130 - 290 0 4 143 RB Below Oper. Floor-Unit 1 340 - 510 0 3 144 RB Below Oper. Floor-Unit 1 510- 1240 0 13 145 RB Below Oper. Floor-Unit 1 1240- 1430 0 3 146 RB Below Oper. Floor-Unit 1 1430- 1670 0 8 McGuire Units 1 and 2 16.9.6-7 Revision 107

Fire Detection Instrumentation 16.9.6, TABLE 16.9.6-1 FIRE DETECTION INSTRUMENTATION"1 '

Detector Description Location Number of Number of Function"'~

Zone Smoke Detectors .Heat Detectors 147 RB Below Oper. Floor-Unit I RCP - 1A Motor 0 5 148 RB Below Oper, Floor-Unit 1 RCP - 1 B Motor 0 2 149 RB Below Oper. Floor-Unit 1 RCP - lC Motor 0 4 150 RB Below Oper. Floor-Unit 1 RCP - 1D Motor 0 5 151 RB Below Oper. Floor-Unit 1 Purge Filter Bed 0 2 152 RB Below Oper. Floor-Unit 1 1700-1900, R20'-R35' 0 2 c153 RB Annulus - Unit 1 2930 -3310 10 10 ci 54 RB Annulus - Unit 1 3240 - 00 4 4 cl 55 RB Annulus - Unit 1 00 - 500 5 5 cl 56 RB Annulus - Unit 1 500 - 880 4 4 cl 57 RB Annulus - Unit 1 880 - 1230 24 24 cl 58 RB Annulus - Unit 1 1230- 1650 22 22 cl 59 RB Annulus - Unit I 3330 - 160 13 13 c160 RB Annulus - Unit I 160 - 540 23 23 c161 RB Annulus - Unit 1 1220 - 1800 16 16 c162. RB Annulus - Unit 1 1800 -2560 14 13 163 Unit 2 RB Pipe Corridor 2150 -2700 0 4 164 Unit 2 RB Pipe Corridor 2700- 3150 0 5 165 Unit 2 RB Pipe Corridor 3150 - 00 0 6 166 Unit 2 RB Pipe Corridor 00-440 0 4 167 Unit 2 RB Pipe Corridor 440 - 900 0 4 168 Unit 2 RB Pipe Corridor 90 - 1260 0 4 169 Unit 2 RB Pipe Corridor 1260- 1730 0 7 170 Unit 2 RB Below Oper. Floor 3290 - 3470 0 7 171 Unit 2 RB Below Oper. Floor 130 - 290 0 4 172 Unit 2 RB Below Oper. Floor 340 - 510 0 3 0 13 173 Unit 2 RB Below Oper. Floor 510- 1240 McGuire Units 1 and 2 16.9.6-8 Revision 107

Fire Detection Instrumentation 16.9.6 TABLE 16.9.6-1 FIRE DETECTION INSTRUMENTATION(1 )

Detector Description Location Number of Number of Function(2)

Zone Smoke Detectors Heat Detectors 174 Unit 2 RB Below Oper. Floor 1240- 1430 0 3 A 175 Unit 2 RB Below Oper. Floor 1430- 1670 0 8 A 176 Unit 2 RB Below Oper. Floor RCP - 2A Motor 0 4 A 177 Unit 2 RB Below Oper. Floor RCP - 2B Motor 0 3 A 178 Unit 2 RB Below Oper. Floor RCP - 2C Motor 0 3 A 179 Unit 2 RB Below Oper. Floor RCP - 2D Motor 0 5 A 180 Unit 2 RB Below Oper. Floor Purge Filter Bed 0 2 A 181 Unit 2 RB Below Oper. Floor 1700-1900, R20'-R35' 0 2 A d182 Unit 2 RB Annulus 2930 _3310 10 10 B d183 Unit 2 RB Annulus 3240 _ 00 4 4 B d184 Unit 2 RB Annulus 00 - 500 5 5 B d185 Unit 2 RB Annulus 500 - 880 4 4 B d186 Unit 2 RB Annulus 880 - 1230 24 24 B d187 Unit 2 RB Annulus 1230 - 1650 22 22 B d188 Unit 2 RB Annulus 3330 - 160 13 13 B d189 Unit 2 RB Annulus 160 - 540 23 23 B d190 Unit 2 RB Annulus 1220 - 1800 16 16 B d191 Unit 2 RB Annulus 1800 -2560 13 13 B 197 Mech. Pen. Rm./UHI Valves JJ-52 EL. 750 5 5 A 198 Mech. Pen. Rm./UHI Valves JJ-60 EL. 750 5 5 A 206 Control Room Control Board AA-56 EL. 767 20 5 A c153A RB Annulus - Unit 1 (Note 3) 00-3600 EL. 745 0 Note 5 B c153B RB Annulus - Unit 1 (Note 3) 00-3600 EL. 765 0 Note 5 B c153C RB Annulus - Unit 1 (Note 3) 00-3600 EL. 785 0 Note 5 B c153D RB Annulus - Unit 1 (Note 3) 00-3600 EL. 805 0 Note 5 B c153E RB Annulus - Unit 1 (Note 3) 00-3600 EL. 820 0 Note 5 B cl 53F RB Annulus - Unit 1 (Note 3) 00-3600 EL. 835 0 Note 5 B McGuire Units 1 and 2 16.9.6-9 Revision 107

Fire Detection Instrumentation 16.9.6 TABLE 16.9.6-1 FIRE DETECTION INSTRUMENTATION"1 '

Detector Description Location Number of Number of Function"2 '

Zone Smoke Detectors Heat Detectors d182A RB Annulus - Unit 2 (Note 4) 00-3600 EL. 745 0 Note 5 B d182B RB Annulus - Unit 2 (Note 4) 00-3600 EL. 765 0 Note 5 B d182C RB Annulus - Unit 2 (Note 4) 00-3600 EL. 785 0 Note 5 B d182D RB Annulus - Unit 2 (Note 4) 00-3600 EL. 805 0 Note 5 B d182E RB Annulus - Unit 2 (Note 4) 00-3600 EL. 820 0 Note 5 B d182F RB Annulus - Unit 2 (Note 4) 00-3600 EL. 835 0 Note 5 B NOTES:

1. The fire detection instruments located within containment are not required to be OPERABLE during the performance of Type A containment leakage rate tests.

2. Function A-- Early warning fire detection and notification only.

Function B: Actuation of fire suppression system and early warning and notification.

3. Upon implementation of NSM MG-12106/00, zones 153- 162 in RB Annulus - Unit 1 will be deleted and zones 153A - 153F will be active fire detection instrumentation.

4. Upon implementation of NSM MG-22106/00, zones 182- 191 in RB Annulus - Unit 2 will be deleted and zones 182A - 182F will be active fire detection instrumentation.

5. The fire detection instruments located in the RB Annulus are restorable, cable-type sensors which cover the entire 360 degrees of the annulus at each subzone elevation.

McGuire Units 1 and 2 16.9.6-10 Revision 107

Fire Detection Instrumentation 16.9.6 BASES Fire detection instrumentation is required to be operable at all times unless a complete evaluation has been made of the area protected by any particular instrument and all equipment in that area has been identified and determined not to be required operable.

This evaluation would have to consider not only mechanical equipment in the area but all piping, tubing, and cables that transit through the area.

OPERABILITY of the detection instrumentation ensures that both adequate warning capability is available for prompt detection of fires and that fire suppression systems, that are actuated by fire detectors, will discharge the extinguishing agent in a timely manner.

Prompt detection and suppression of fires will reduce the potential for damage to safety-related equipment and is an integral element in the overall facility fire protection program.

An inoperable detector is defined as: a) a fire alarm with no actual fire or b) a trouble alarm.

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 that are installed solely for early fire warning and notification. Consequently, the minimum number of OPERABLE fire detectors must be greater.

The loss of detection capability for the Fire Suppression Systems, actuated by fire detectors, represents a significant degradation of fire protection for any area. The establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation is restored to OPERABLE status. If fire detection capability is monitored by a local or remote panel, the fire watch patrol needs only check the panel to verify no loss in fire detection capability. Note that the MNS Fire Protection Safe Shutdown Review considers the annulus to be part of the containment building.

This selected licensee commitment is part of the McGuire Fire Protection Program and therefore subject to the provisions of McGuire Facility Operating License Conditions C.4 (Unit 1) and C.7 (Unit 2).

TR 16.9.6.7, "Perform a functional test on at least 2% of the fixed temperature/rate of rise non-restorable spot type heat detectors," is satisfied by either testing within population or replacement of all heat detectors as per TR 16.9.6.8.

TR 16.9.6.8, "Replace all fixed temperature/rate of rise non-restorable spot type heat detectors," purpose is compliance with NFPA 72-2002 Table 10.4.2.2 Device 13 (d) 3. This is applicable to all containment fire zones (Unit 1 zones 134 - 152, Unit 2 zones 163 - 181).

McGuire Units 1 and 2 16.9.6-11 Revision 107

Fire Detection Instrumentation 16.

9.6 REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 9.5.1

2. McGuire Nuclear Station SER Supplement 2, Chapter 9.5.1 and Appendix D

3. McGuire Nuclear Station SER Supplement 5, Chapter 9.5.1 and Appendix B

4. McGuire Fire Protection Review, as revised

5. McGuire Nuclear Station SER Supplement 6, Chapter 9.5.1 and Appendix C

6. NFPA Codes 72D and 72E

7. McGuire Nuclear Station Facility Operating Licenses, Unit 1 License Condition C.(4) and Unit 2 License Condition C.(7)

McGuire Units 1 and 2 16.9.6-12 Revision 107

Standby Shutdown System 16.9.7 16.9 AUXILIARY SYSTEMS - FIRE PROTECTION SYSTEMS 16.9.7 Standby Shutdown System COMMITMENT The Standby Shutdown System (SSS) shall be operable.

APPLICABILITY MODES 1, 2, and 3.

REMEDIAL ACTIONS

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

1. The SRO should ensure that security is notified 10 minutes prior to declaring the SSS inoperable. Immediately upon discovery of the SSS inoperability, Security must be notified to implement compensatory measures within 1.0 minutes of the discovery.

2. If inoperable SSS component is located inside containment, repairs shall be made at the first outage which permits containment access.

CONDITION REQUIRED ACTION COMPLETION TIME

-NOTE ---------- A.1 Verify the OPERABILITY of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Not applicable to the SSS fire detection and Diesel Generator or 24 V suppression systems in the Battery Bank and Charger. associated areas identified in Table16.9.7-1.

A. One or more required AND SSS components identified in Table A.2 Restore the component to 7 days 16.9.7-1 inoperable. OPERABLE status.

B. SSS Diesel Generator or B.1 Verify the OPERABILITY of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months 24 V Battery Bank and fire detection and Charger inoperable, suppression systems in the associated areas identified in Table16.9.7-1.

AND (continued)

McGuire Units 1 and 2 16.9.7-1 Revision 117

Standby Shutdown System 16.9.7 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. (continued). B.2 Verify offsite power and 1hour one emergency diesel generator OPERABLE.

AND B.3 Restore the component to 7days OPERABLE status.

C. Total Unidentified C.1 Declare the Standby Immediately LEAKAGE, Identified Makeup Pump inoperable.

LEAKAGE, and reactor coolant pump seal AND leakoff > 20 gpm.

C.2 Enter Condition A.

OR Total reactor coolant pump seal leakoff > 16.3 gpm.

OR Any reactor coolant pump No. 1 seal leakoff

> 4.0 gpm.

D. Lake Norman level D. Verify the "C"Fire 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months below 746 feet. Suppression Pump is OPERABLE (Unit 1 only).

E. Required Action A.2 and E.1 Prepare and submit a 30 days its associated Special Report to the NRC Completion Time not outlining the cause of the met. inoperability, corrective actions taken, and plans for restoring the SSS to OPERABLE status.

F. Required Action B.3 and F.1 Prepare and submit a 14 days its associated Special Report to the NRC Completion Time not outlining the extent of met. repairs required, schedule for completing repairs, and basis for continued operation.

McGuire Units 1 and 2 16.9.7-2 Revision 117

Standby Shutdown System 16.9.7 TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.7.1 Verify total Identified LEAKAGE, Unidentified LEAKAGE, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months and reactor coolant pump seal leakoff are < 20 gpm.

AND Verify total reactor coolant pump seal leakoff < 16.3 gpm.

AND Verify each reactor coolant pump No. 1 seal leakoff < 4.0 gpm.

TR 16.9.7.2 Verify the requirements for spent fuel water level in 7 days Surveillance Requirement 3.7.13.1 are met and the boron concentration in the spent fuel storage pool is within the limits specified in the COLR.

OR Verify the refueling water storage tank is capable of being aligned to the SSS standby makeup pump.

TR 16.9.7.3 Verify fuel oil level in the SSS diesel generator fuel 31 days storage tank is > 4.0 ft.

TR 16.9.7.4 Verify the SSS diesel generator starts from ambient 31 days conditions and operates for > 30 minutes at > 700 kW.

TR 16.9.7.5 Verify fuel oil properties of new and stored fuel oil for the In accordance with SSS diesel generator are tested in accordance with, and the Diesel Fuel Oil maintained within the limits of, the Diesel Fuel Oil Testing Testing Program Program.

TR 16.9.7.6 Verify the SSS diesel generator 24 V battery voltage is > 31 days 24 volts.

TR 16.9.7.7 Perform a CHANNEL CHECK of the SSS Instrumentation 31 days as required by Table 16.9.7-2.

TR 16.9.7.8 Verify the electrolyte level of each SSS 250/125 V battery 31 days bank is above the plates.

(continued)

McGuire Units 1 and 2 16.9.7-3 Revision 117

Standby Shutdown System 16.9.7 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.9.7.9 Verify the total battery terminal voltage of each SSS 31 days 250/125 V battery bank is > 258/129 V on float charge.

TR 16.9.7.10 Verify the average specific gravity of each SSS 250/125 92 days V battery bank is > 1.200.

TR 16.9.7.11 Verify the standby makeup pump's developed head and 92 days capacity is greater than or equal to that required by the Inservice Testing Supplemental Program.

TR 16.9.7.12 Verify the SSS diesel generator 24 V batteries and 18 months battery racks show no visual indication of physical damage or abnormal deterioration.

TR 16.9.7.13 Verify SSS diesel generator 24 V battery to battery and 18 months terminal connections are clean, tight, and free of corrosion.

TR 16.9.7.14 Perform a CHANNEL CALIBRATION of the SSS 18 months Instrumentation as required by Table 16.9.7-2.

TR 16.9.7.15 Perform inspection of SSS diesel generator in 18 months accordance with procedures prepared in conjunction with manufacturer's recommendations for class of service.

TR 16.9.7.16 Verify the SSS 250/125 V batteries, cell plates, and 18 months battery racks show no visual indication of physical damage or abnormal deterioration.

TR 16.9.7.17 Verify the SSS 250/125 V battery to battery and terminal 18 months connections are clean, tight, free of corrosion, and coated with anti-corrosion material.

TR 16.9.7.18 Verify the "C" solenoid to valve SA48ABC can be 18 months deenergized to provide steam supply to the turbine driven auxiliary feedwater pump.

TR 16.9.7.19 Verify the CA Storage Tank level is > 20 feet. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TR 16.9.7.20 Verify Lake Norman level is > 746 feet 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months McGuire Units 1 and 2 16.9.7-4 Revision 117

Standby Shutdown System 16.9.7 TABLE 16.9.7-1 STANDBY SHUTDOWN SYSTEM FIRE DETECTION & SUPPRESSION SYSTEMS VERIFICATION"1 )

INOPERABLE SSS COMPONENT FIRE DETECTION & SUPPRESSION SYSTEMS LOCATION EL 716 EL 733 EL 750 Control Battery Cable Turbine Driven Motor Driven Containment EE-KK EE-KK EE-KK Room Room Rooms AFW Pump AFW Pump SSS Diesel Generator(3) X X X X X X X X Note 2 SSS DG Starting 24 V Battery X X X X X X X X Note 2 Bank and Charger(3)

Standby Makeup Pump and X X X Water Supply SSS 250/125V Battery and X X X Note 2 Charger(3)

Turbine Driven AFW Pump and X Water Supplies(4)

Turbine Driven AFW Pump X Solenoid "C" Groundwater Drainage Sump X Pump A, SumpA AND Groundwater Drainage Sump Pump A, Sump B Fire Suppression Pump "C" (see X Condition D).

INSTRUMENTATION:

1. RCS Pressure X X X Note 2

2. Pressurizer Level X X X Note 2

3. SG Level X X X Note 2

4. Incore Temperature X X X Note 2

5. NC Wide Range Cold Leg X X X Note 2 Temperature NOTES:

1. If fire detection and/or suppression systems are inoperable, then the ACTION statement(s) of the applicable fire detection and/or suppression SLC shall be complied with.

2. Monitor containment air temperature at least once per hour at the locations specified in Technical Specification Surveillance Requirement 3.6.5.1 or 3.6.5.2, in lieu of verification of operability of systems inside containment.

3. With this component inoperable, then denoted areas of both units are affected.

4. Water supplies include the Auxiliary Feedwater Storage Tank (CAST) and Condenser Circulating Water (RC) System via valves CA-161C and 162C. Continuous vents at 1/2RN-1 065 and 1RN-1 066 support OPERABILITY of the RC source for Unit 1 only.

McGuire Units 1 and 2 16.9.7-5 Revision 117

Standby Shutdown System 16.9.7 TABLE 16.9.7-2 STANDBY SHUTDOWN SYSTEM INSTRUMENTATION TESTING REQUIREMENTS REQUIRED TESTING READOUT INSTRI JMENT CHANNELS REQUIREMENTS LOCATION

1. Reactor Coolant Pressure 1 TR 16.9.7.7 SSF Control TR 16.9.7.14 Panel

2. Pressurizer Leve .1 TR 16.9.7.7 SSF Control TR 16.9.7.14 Panel

3. Steam Generato r Level 1 per SG TR 16.9.7.7 SSF Control (Wide Range) TR 16.9.7.14 Panel

4. Incore Temperat ure 1 TR 16.9.7.7 SSF Control TR 16.9.7.14 Panel

5. Standby Makeup Pump Flow 1 TR 16.9.7.14 SSF Control Panel

6. NC Wide Range Cold Leg 2 TR 16.9.7.7 SSF Control Temperature TR 16.9.7.14 Panel McGuire Units 1 and 2 16.9.7-6 Revision 117

Standby Shutdown System 16.9.7 BASES The Standby Shutdown System (SSS) is designed to mitigate the consequences of certain postulated fire incidents, sabotage, or station blackout events by providing capability to maintain HOT STANDBY conditions and by controlling and monitoring vital systems from locations external to the main control room. The facility is credited with the ability to cope with a station black out (SBO) event of 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months duration. This capability is consistent with the requirements of 10 CFR Part 50, Appendix R and 10 CFR 50.63.

By design, the SSS is intended to respond to those low-probability events which render both the control room and automatic safety systems inoperable. Because of the low probability of occurrence of these events, the remedial actions rely on compensatory action, timely repair or return to operability and, if necessary, a justification for continued operation.

Because the SSS performs a redundant fire protection function, compensatory action during periods when the SSS is inoperable relies largely on assurance of the operability of fire detection and suppression systems. Table 16.9.7-1 establishes requirements for operability of fire detection and suppression systems.

Both A&D NC Cold Leg Wide Range Temperatures are required for SSS operability. This conclusion is based on NRC Correspondence during issuance of the original operating license.

The Source Range Wide Range Neutron Flux Instrumentation was installed at the SSS Control Panel as part of NRC review of this system in the early 1980s. The indication is not required for SSS operability, based on the NRCs response to Duke dated July 21, 1983.

Controls and power to the pressurizer heater banks are included for SSF events; however, they are not required for SSS operability. NRC Generic Letter 86-10 provides that conclusion.

The Testing Requirements ensure that the SSS systems and components are capable of performing their intended functions. The testing requirements were based largely on SSS Technical Specifications for the Catawba Nuclear Station, which was approved prior to the issuance of the fuel load license for Unit 1 of that plant. Also considered in the formulation of the testing requirements were existing McGuire Technical Specifications, such as those for the 1 E Diesel Generators, Refueling Water Storage Tank, Fire Protection & Detection Systems, and other Tech Specs which are related to the safe operation and/or shutdown of the plant.

The required level in the SSS diesel generator fuel storage tank ensures sufficient fuel for 3 1/2 days of uninterrupted operation. Per Appendix R requirements, the unit must be in cold shutdown within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of going to the SSF. The 3 11/2 day supply of fuel oil assures this capability.

Testing has demonstrated the ability of plant operations to start the SSF diesel within 10 minutes of the recognition of an SBO event, thus satisfying the intent of NUMARC 87-00 guidance. The SSF diesel generator has sufficient capacity and capability to operate equipment necessary to achieve and maintain safe shutdown conditions for a 4-hour SBO event.

McGuire Units 1 and 2 16.9.7-7 Revision 117

Standby Shutdown System 16.9.7 BASES (continued)

Fuel oil for the SSS diesel generator is tested and maintained in accordance with the same Diesel Fuel Oil Testing Program used for the 4kV emergency diesel generators (see Technical Specification 5.5.13, Surveillance Requirement 3.8.3.2 and associated Bases).

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 Standby Makeup Pump (SMP) functions as part of the SSF to provide makeup capacity to the reactor coolant system and cooling flow to the reactor coolant pump (RCP) seals.

The RCP seal leak-off flow is temperature dependent (i.e., the higher the temperature the higher the leak-off flow). During normal operation the RCP seals are supplied from the Centrifugal Charging Pump (CCP) drawing from the Volume Control Tank (VCT). During the SSF event, the SMP draws from the Spent Fuel Pool (SFP). During the SSF event there is no SFP cooling, so water injected into the RCP seals will have a higher temperature than during normal operation. The SMP is capable of providing a makeup capacity of 26 gpm.

The revised SLC limit of 20 gpm total accumulative leakage is based on a calculation that was performed by Westinghouse, indicating increased RCP seal leak-off at higher seal water temperatures, to relate the SSF event leakage of 26 gpm at elevated RCP seal temperatures. This more conservative limit will ensure that the SMP will be capable of providing makeup and seal cooling flow equal to or greater than total leakage during the SSF event, increased RCP seal leak-off flow due to heat-up of the SFP, and still provide a margin of safety. As a conservative measure, during normal power operation the total accumulative system leakage (unidentified + identified + RCP seal leak-off flows) shall be limited to 20 gpm. The Testing Requirement concerning the SMP 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 .

The additional requirement that total RCP seal leak-off flow be < 16.3 gpm resulted from a historical review of NRC correspondence that specified the SMP also provide for reactor coolant system makeup and boration in addition to RCP seal leakage requirements (Ref.

17). Calculations show that this upper limit for RCP seal leak-off provides sufficient margin to maintain the required unit conditions for a bounding SSS event.

Calculation MCC-1201.01-00-0053, Rev. 0, "MNS Units 1 & 2 Reactor Coolant Pump Response To Loss Of Seal Cooling," Sections 2 and 10 (Tab D, page 15) determined the elapsed time from loss of all seal cooling (loss of NV seal injection and loss of KC flow to the RCP thermal barrier heat exchanger) to when hot NC water entered the RCP No. 1 seal at varying seal leakoff rates. Chart interpolation determined that at a nominal No. 1 seal leakoff rate of 4 gpm, the seal would be at 2350 F in 6.4 minutes from loss of all seal cooling event initiation. Therefore, for a maximum No. 1 seal leakoff of 4 gpm and if the Operators are instructed to stop all 4 RCPs at 3 minutes into the scenario, 3.4 minutes remain for the RCP motors to coast down to a stop and no seal rotation would occur above the No. 1 seal trip setpoint (235TF) during loss of all seal cooling. The 4 gpm limit is conservative based on the guidance provided in Westinghouse WCAP-17100, Section 1.2.3.4, "Response during a Loss of All Seal Cooling," and Westinghouse Technical Bulletin TB-04-22, Revision 1, for RCP coast down times and time for hot NC system water to reach the No. 1 seal on loss of all seal cooling.

The Groundwater Drainage Sump Pump A, in the A (Unit 1) and B (Unit 2) sumps, can be controlled and powered from the SSF. These Sump Pumps remove accumulation of groundwater, Turbine driven AFW Pump drains, and other miscellaneous sources. For the McGuire Units 1 and 2 169.7-8 Revision 1 17

Standby Shutdown System 16.9.7 BASES (continued)

SSS to be OPERABLE, a minimum of one of these pumps must be OPERABLE. Credit is taken for the groundwater underdrain system to transport water from one sump to the other.

The turbine driven AFW pump can be controlled from the SSF and is utilized during an SSS event to maintain adequate secondary side heat removal. For the SSS to be OPERABLE, the turbine driven AFW pump must be OPERABLE. The turbine driven AFW pump water supply for the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SBO event is provided by the CA Storage Tank (CAST). The water supply for the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months fire event is initially provided by the CAST then later by manual alignment to the RC system via valves CA-161C and CA-162C. These valves are maintained closed and de-energized to prevent spurious actuation and air entrainment (Ref.

22). Adequate CAST inventory of 200,000 gallons (20 feet of level) is ensured by TR 16.9.7.19. For a fire event, an initial CAST inventory is needed to allow time to perform the manual alignment from the CAST to the RC system. For Unit 1 only, in the remote chance that the level of Lake Norman drops below 746 feet, air entrainment from the RC system cannot be prevented. In this case, adequate water supplies for a fire event are ensured by making up to the Unit 1 CAST from the Fire Suppression system using the "C" Fire Suppression pump.

The SSF is provided with its own 250/125 VDC power system which is independent from the normal 125 VDC and 120 VAC vital I&C power systems. The SSF batteries are charged by the SSF diesel generator and are available to power the SSF instruments and controls necessary to achieve and maintain hot standby conditions from the SSF control room following a station black out (SBO) event.

While the SSS 24 VDC battery charger is isolated for battery surveillance testing, the SSS Diesel Generator remains operable as long as the battery voltage is > 24 volts.

The SSS 125V batteries and battery chargers consist of three pairs SDSP1, SDSP2 and SDSS. Each pair consists of a battery and associated battery charger. Pair SDSS can be used to substitute for either pair SDSP1 or SDSP2. Only two of these pairs are required operable since pair SDSS is spare.

This selected licensee commitment is part of the McGuire Fire Protection Program and therefore subject to the provisions of McGuire Facility Operating License Conditions C.4 (Unit 1) and C.7 (Unit 2).

McGuire Units 1 and 2 16.9.7-9 Revision 117

Standby Shutdown System 16.

9.7 REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 9.5.1

2. McGuire Nuclear Station SER Supplement 2, Chapter 9.5.1 and Appendix D

3. McGuire Nuclear Station SER Supplement 5, Chapter 9.5.1 and Appendix B

4. McGuire Nuclear Station SER Supplement 6, Chapter 9.5.1 and Appendix C

5. McGuire Fire Protection Review, as revised

6. McGuire Fire Protection Safe Shutdown Review

7. IEEE 308-1974, Class 1E Power Systems

8. IEEE 450-1975, Maintenance Testing & Replacement of Large Lead Storage Batteries

9. OP1OIB16350104, Standby Shutdown Facility Diesel Operation

10. McGuire Nuclear Station Facility Operating Licenses, Unit 1 License Condition C.(4) and Unit 2 License Condition C.(7)

11. PIP 0-M-99-03926

12. PIP-M-01-3466

13. 10 CFR 50.63, Loss of All Alternating Current

14. Letter from H.B. Tucker to NRC, dated April 4, 1990, Requirements for Station Blackout.

15. Letter from H.B. Tucker to NRC, dated April 17, 1989, Requirements for Station Blackout.

16. McGuire Nuclear Station, Units 1 and 2, Safety Evaluation for Station Blackout (10CFR50.63), Dated February 19, 1992.

17. SAIC-91/1265, "Technical Evaluation Report, McGuire Nuclear Station, Station Blackout Evaluation," Dated December 10, 1991.

18. McGuire Nuclear Station UFSAR, Section 18.2.4, Chemistry Control Program.

19. MCS-1465.00-00-0019, "Plant Design Basis Specification For Station Blackout Rule,"

Rev. 3.

20. McGuire License Renewal Commitments MCS-1274.00-00-0016, Section 4.6, Chemistry Control Program.

21. PIP M-04-3317.

22. MCC-1223.42-00-0055, "Design Considerations and Bases for 1/2CA-161C and 1/2CA-162C Automatic Open Deletion Modifications MD101869 and MD201870."

McGuire Units 1 and 2 16.9.7-10 Revision 117

Groundwater Level Monitoring System 16.9.8 16.9 AUXILIARY SYSTEMS 16.9.8 Groundwater Level Monitoring System COMMITMENT a. The groundwater level monitors listed in Table 16.9.8-1 shall be OPERABLE.

b. The groundwater level shall be maintained within the limits of Table 16.9.8-1 APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Reactor A.1 Restore the inoperable 7 days Building or Diesel monitor to OPERABLE Building Groundwater status.

Level Monitors inoperable.

B Required Action and B. 1 Provide an alternate Immediately associated Completion method for determining the Time of Condition A not groundwater level for the met. monitored area.

AND B.2 Enter the inoperable Level 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Monitor(s) into the Corrective Action Program.

C. Reactor Building or C. 1 Restore the groundwater 7 days Diesel Building level to within limit.

groundwater level not within limit.

(continued)

McGuire Units 1 and 2 16.9.8-1 Revision 96

Groundwater Level Monitoring System 16.9.8 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Initiate an engineering Immediately associated Completion evaluation to determine the Time of Condition C not cause and provide met. corrective action.

AND D.2 Enter the groundwater 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months '

'level(s) not within limit into the Corrective Action Program.

E. Two Auxiliary Building E. 1 Enter the inoperable level 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Groundwater, Level Monitor(s) into the Monitors Inoperable. Corrective Action Program.

AND E.2 Restore at least one 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable monitor to OPERABLE status.

-OR E.3 Provide an alternate 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months method of determining the groundwater level for at least one of the affected monitored location(s).

F. Groundwater level at F.1 Reduce the groundwater 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months three or more Auxiliary level to within limit.

Building monitored locations not within limit.

McGuire Units 1 and 2 16.9.8-2 Revision 96

Groundwater Level Monitoring System 16.9.8 G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition F not AND met.

G.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.8.1 Verify the Auxiliary Building groundwater level within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months limits by absence of alarms, by visual observation of the monitor level gauge or by alternate methodology.

TR 16.9.8.2 Perform a COT on the Auxiliary Building groundwater 12 months level monitors listed in Table16.9.8-1.

(continued)

McGuire Units 1 and 2 16.9.8-3 Revision 96

Groundwater Level Monitoring System 16.9.8 TESTING REUIREMENTS (continued)

TEST FREQUENCY TR 16.9.8.3 Perform a CHANNEL CALIBRATION on the Auxiliary 12 months Building groundwater level monitors listed in Table 16.9.8-1.

TR 16.9.8.4 Perform a COT on the Reactor Building and Diesel 18 months Building groundwater level monitors listed in Table 16.9.8-1.

TR 16.9.8.5 Perform a CHANNEL CALIBRATION on the Reactor 18 months Building and Diesel Building groundwater level monitors listed in Table 16.9.8-1.

McGuire Units 1 and 2 16.9.8-4 Revision 96

Groundwater Level Monitoring System 16.9.8 TABLE 16.9.8-1 GROUNDWATER LEVEL MONITORS LOCATION and EXTERIOR/INTERIOR LEVEL LIMIT (MSL) APPLICABILITY INSTRUMENT Reactor Building Exterior 731 ft Unit 1 1WZLS5060 Diesel Building Interior 739 ft 2 in. Unit 1 AA-40, ELEV. 736' 1WZLP5080 Diesel Building Interior 739 ft 2 in. Unit 1 DD-42, ELEV. 736' 1WZLP5090 Reactor Building Exterior 736 ft Unit 2 2WZLS5060 Diesel Building Interior 739 ft 2 in. Unit 2 BB-72, ELEV. 736' 2WZLP5080 Diesel Building Interior 739 ft 2 in. Unit 2 DD-69, ELEV. 736' 2WZLP5090 Aux Bldg PP-51 Interior 731 ft Unit 1 & 2 1WZLP5100 Aux Bldg QQ-56 Interior 731 ft Unit 1 & 2 0WZLP51 10 Aux Bldg PP-61 Interior 731 ft Unit 1 & 2 2WZLP51 00 Aux Bldg West Wall Exterior 731 ft Unit 1 & 2 1WZLS5070 Aux Bldg East Wall Exterior 731 ft Unit 1 & 2 2WZLS5070 McGuire Units 1 and 2 16.9.8-5 Revision 96

Groundwater Level Monitoring System 16.9.8 BASES The Reactor and Auxiliary Building complex for McGuire incorporates a permanent groundwater dewatering system that is designed to preclude groundwater from rising above a structural distress level of elevation 737 ft. Mean Sea Level (MSL) for the Auxiliary Building, or 23 feet below the site grade level of 760 ft. MSL.

The groundwater drainage system incorporates a grid system beneath the Reactor and Auxiliary Building basemats, 3 sumps in the Auxiliary Building, each with pumps and level alarms, a peripheral exterior drain system and 11 groundwater level monitors.

Originally, McGuire incorporated all 11 level monitors as Tech Spec monitors. Subsequently, an analysis performed by Design Engineering demonstrated that the Reactor and Diesel Generator Buildings were designed to withstand groundwater stresses up to 760 ft. MSL. Therefore, a.

Tech Spec revision was sought and obtained (Amendment Nos. 93 and 74) that removed all but 5 of the Auxiliary Building level monitors from the Tech Specs. The other 6 Reactor and Diesel Building level monitors were placed in Chapter 16 of the UFSAR as Selected Licensee Commitments (SLC). These 6 monitors, having locations listed in SLC Table16.9.8-1, were retained in administrative procedures for the groundwater monitoring program and utilized as an indication of any localized groundwater increases that may be indicative of increase due to ruptured pipes and as an indication of a problem with the underground grid system. This commitment was made as part of the justification for relaxing the groundwater monitoring Tech Spec.

The Reactor Building level monitors are exterior monitors and their first alarm is in the "Hi" alarm at 730 ft MSL on Unit 1 and 736 ft MSL on Unit 2. This ensures an alarm at or below the Groundwater Alert level (731 ft MSL on Unit 1, 736 ft MSL on Unit 2) is reached for the Reactor Building. The Diesel Building level monitors have similar first alarms but are termed "Alert" alarms and should not be confused with "Alert Levels." These alert levels for the Reactor Buildings have no safety significance because the Reactor and Diesel Buildings are analyzed for hydrostatic loads up to grade (760 ft.) elevation. Under the requirements of this SLC, if one or more Reactor or Diesel Building groundwater level monitors becomes inoperable or not within its level limit for 7 days or more, the condition shall be entered into McGuire's 10 CFR 50, Appendix B, Criterion XVI program (Corrective Action Program) for cause evaluation, corrective action, and trending.

The Auxiliary Building level monitors were placed in SLC 16.9.8 when McGuire converted to Improved Tech Specs. The SLC limits for the Auxiliary Building are provided to ensure that groundwater levels will be monitored and prevented from rising to the potential failure limit for the McGuire Units 1 and 2 Auxiliary Buildings. This potential failure limit is based on engineering calculations that have determined that the Auxiliary Buildings are susceptible to overturning due to buoyancy at elevation 737 feet Mean Sea Level (MSL). Under the requirements of this SLC, if groundwater level exceeds elevation 731 feet MSL, (3 out of 5 SLC groundwater monitor alarms), and cannot be reduced in one (1) hour, McGuire must begin reducing Units 1 and 2 to Mode 5, Cold Shutdown.

Elevation 731 feet MSL is the action level of the five Auxiliary Building groundwater monitors listed in Tablel 6.9.8-1. The East Wall exterior monitor alarm at elevation 731 feet MSL is the Alert alarm. The other four (4) monitors are Hi-Hi alarms at elevation 731 feet MSL.

McGuire Units 1 and 2 16.9.8-6 Revision 96

Groundwater Level Monitoring System 16.9.8 BASES (continued)

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

As required by Operations procedures, any alarms on SLC groundwater monitors will also be investigated. Additionally, if three (3) out of the five (5) groundwater monitors alarm at levels below the action levels, Operations will contact Civil Engineering for investigation and resolution of the increased groundwater level.

If one or more of the 5 Auxiliary Building groundwater monitors is determined to be inoperable, the monitor(s) will be considered to be indicating above the 731'-0" MSL until repaired and returned to an operable' status or groundwater levels at the affected location(s) are determined to be within limits through alternate methods. Appropriate techniques shall be utilized to assure the accuracy of measurements taken through these alternate methods.

REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 2.4.13.

2. McGuire Nuclear Station UFSAR, Appendix 2B.

3. McGuire Nuclear Station UFSAR, Chapter 9.5.8.

4. McGuire Nuclear Station UFSAR, Appendix 2D, Chapter 5.1.1.

5. McGuire Nuclear Station UFSAR, Chapter 7.6.11.

6. McGuire Nuclear Station UFSAR, Figure 2.4.13-1.

7. OP/1/A/6100/101, Annunciator Response to Panel IAD-8

8. McGuire Nuclear Station SER Section 2.4.5 dated March 1, 1978.

9. PIP M-03-1377

10. PIP M-07-1139

11. NRC SER dated March 2, 1989 pursuant to McGuire License Amendment Request of January 27, 1988.

McGuire Units 1 and 2 16.9.8-7 Revision 96

Boration Systems - Flow Path (Operating)

. 16.9.9 16.9 AUXILIARY SYSTEMS 16.9.9 Boration Systems - Flow Path (Operating)

COMMITMENT Two of the following three boron injection flow paths shall be OPERABLE:

a. The flow path from a boric acid tank via a boric acid transfer pump and a charging pump to the reactor coolant system, and

b. Two flow paths from the refueling water storage tank via charging pumps to the reactor coolant system.

Note: An OPERABLE charging pump used to satisfy OPERABILITY requirements of one boration flow path may not be used to satisfy OPERABILITY requirements for a second boration flow path.

APPLICABILITY MODES 1, 2, and 3.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required boron A.1 Restore the required boron 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months injection flow path injection flow path to inoperable. OPERABLE status.

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

B.2 Borate to the SDM 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months requirements of Tech Spec 3.1.1.

AND 7 days B.3 Restore the required boron injection flow path to OPERABLE status.

(continued)

McGuire Units 1 and 2 16.9.9-1 Revision 101

Boration Systems - Flow Path (Operating) 16.9.9 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. Required Action and C. 1 Be in MODE 4. 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months associated Completion Time of Condition B not met.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.9.1 Verify the temperature of piping associated with the flow 7 days path from the boric acid storage tanks is > 651F when it is a required water source TR 16.9.9.2 Verify that each manual, power operated, or automatic 31 days valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position.

TR 16.9.9.3 Verify that each automatic valve in the flow path actuates 18 months to its correct position on a safety injection test signal.

TR 16.9.9.4 Verify that each charging pump's developed head at the In accordance with test flow point is greater than or equal to the required the Inservice developed head. Testing Program TR 16.9.9.5 Verify that the flow path from the boric acid tanks via a 18 months boric acid transfer pump and a charging pump delivers

> 30 gpm to the reactor coolant system.

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

In Modes 1, 2, and 3, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths McGuire Units 1 and 2 16.9.9-2 Revision 101

Boration Systems - Flow Path (Operating) 16.9.9 BASES (continued) inoperable. The boration capability of either flow path is sufficient to provide a SHUTDOWN MARGIN from expected operating conditions of 1.3% delta k/k after xenon decay and cooldown to 200 0 F. The maximum expected boration capability requirement occurs at EOL from full power equilibrium xenon conditions. Further discussion is provided in Bases for Shutdown Margin Requirements (Tech Spec 3.1.1 and 3.1.2).

REFERENCES

1. PIP M-07-03237

2. DPND-1552.63-0099, Rev. 0, "Required Boration Flow Paths in Mode 4"

3. NRC Issuance of Amendments 184/166, Improved Technical Specification conversion and relocations to SLC Manual.

McGuire Units 1 and 2 16.9.9-3 Revision 101

Boration Systems - Charging Pumps (Operating) 16.9.10 16.9 AUXILIARY SYSTEMS 16.9.10 Boration Systems - Charging Pumps (Operating)

(DELETED - COMBINED WITH 16.9.9)

McGuire Units 1 and 2 16.9.10-1 Revision 13

Borated Water Sources (Operating) 16.9.11 16.9 AUXILIARY SYSTEMS 16.9.11 Borated Water Sources (Operating)

COMMITMENT As a minimum, the following borated water source(s) shall be OPERABLE as required by SLC 16.9.9:

a. A boric acid tank (BAT) and,

b. The refueling water storage tank.

APPLICABILITY MODES 1, 2, and 3, MODE 4 with all RCS cold leg temperatures > 300°F.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required boric acid tank A.1 Restore the required boric 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable, acid tank to OPERABLE status.

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

B.2 Borate to the SDM 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months requirements of TS 3.1.1 AND B.3 Restore the required boric 7 days acid tank to OPERABLE status.

C. Required Action and C.1 Be in MODE 4 with any 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months associated Completion RCS cold leg temperature Time of Condition B not < 300 0 F.

met.

(continued)

McGuire Units 1 and 2 16.9.11-1 Revision 22

Borated Water Sources (Operating) 16.9.11 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Refueling water storage D.1 Enter the applicable Immediately tank inoperable. Conditions and Required Actions of LCO 3.5.4, "Refueling Water Storage Tank."

TESTING REQUIREMENTS TEST FREQUENCY, TR 16.9.11.1 Verify the refueling water storage tank solution 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months temperature is > 70°F and < 100OF when the outside air temperature is < 70OF or > 100 0 F.

TR 16.9.11.2 Verify the boron concentration of the required borated 7 days water source is within the limits specified in the COLR.

TR 16.9.11.3 Verify the borated water volume of the required borated 7 days water source is within the limits specified in the COLR.

TR 16.9.11.4 Verify the boric acid tank solution temperature is > 65 0 F 7 days when the boric acid storage tank is a required source.

BASES The borated water sources ensure that negative reactivity control is available during each mode of facility operation.

In Modes 1-3 and Mode 4 with all RCS cold leg temperatures above 300 '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. The boration capability of either borated water source, in association with a flow path and charging pump, is sufficient to provide a SDM from expected operating conditions of 1.3% delta k/k after xenon decay and cooldown.

The SLC commitment values are presented in the Core Operating Limits Report (COLR) as:

(1) the minimum boron concentrations and minimum volumes necessary to attain and BASES (continued)

McGuire Units 1 and 2 16.9.11-2 Revision 22

Borated Water Sources (Operating) 16.9.11 maintain SDM in the BAT or the refueling water storage tank, (2) the minimum contained volumes in the BAT or the refueling water storage tank, and (3) a curve specifying the minimum contained volume in the BAT near EOC. The minimum contained water volume is based on the required volume to maintain shutdown margin, an allowance for water not available because of discharge line location and additional margin. The additional margin term includes allowances for instrument uncertainty, vortexing and a margin term consisting of at least 5% of the volume necessary for SDM. The COLR specified volumes are volumes reserved for use during a cooldown, and in conjunction with the boron concentrations, satisfy SDM requirements during Modes 1-3 and Mode 4 with all RCS cold leg temperatures above 300 OF.

Boric Acid Tank Requirements for Maintaining SDM Required volume for maintaining SDM Presented in the COLR Unusable volume (to maintain full suction pipe) 4,199 gallons Additional margin 4,100 gallons Refueling Water Storage Tank Requirements for Maintaining SDM Required volume for maintaining SDM Presented in the COLR Unusable volume (to maintain full suction pipe) 16,000 gallons Additional margin 23,500 gallons The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 9.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

REFERENCES None McGuire Units 1 and 2 16.9.11-3 Revision 22

Boration Systems - Flow Path (Shutdown) 16.9.12 16.9 AUXILIARY SYSTEMS 16.9.12 Boration Systems - Flow Path (Shutdown).

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

a. a flow path from a boric acid tank via a boric acid transfer pump and a charging pump to the reactor coolant system if the boric acid storage tank in SLC 16.9.14 is OPERABLE, or

b. the flow path from the refueling water storage tank via a charging pump to the reactor coolant system ifthe refueling water storage tank in SLC 16.9.14 is OPERABLE.

Note: An OPERABLE safety injection pump (and associated suction from RWST and discharge flowpath to cold legs) may be used in lieu of the charging pump in (b.) during Modes 5 and 6 when seal injection is not needed.

APPLICABILITY MODES 4, 5, and 6.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required boron injection A.1 Suspend CORE Immediately flow path inoperable. ALTERATIONS.

OR AND Required boron injection A.2 Suspend positive reactivity Immediately flow path not capable of additions.

being powered from an emergency power source.

McGuire Units 1 and 2 16.9.12-1 Revision 101

Boration Systems - Flow Path (Shutdown) 16.9.12 TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.12.1 Verify the temperature of piping associated with the flow 7 days path is > 65°F when a flow path from the boric acid storage tank is used.

TR 16.9.12.2 Verify that the charging pump's or safety injection pump's In accordance with developed head at the test flow point is greater than or the Inservice equal to the required developed head. Testing Program TR 16.9.12.3 Verify that each manual, power operated, or automatic 31 days valve in the flow path that is not locked, sealed, or otherwise secured in position, is in the correct position.

BASES The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include: (1) borated water source, (2) charging pump, (3) separate flow path, (4) boric acid transfer pump, and (5) an emergency power supply from OPERABLE diesel generator. A safety injection pump with suction flow path from RWST and discharge flow path to RCS cold legs may also be used to perform boron injection functions during Modes 5 and 6.

In Modes 4, 5, and 6, one Boron Injection System is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single Boron Injection System becomes inoperable. Further discussion is provided in Bases for Shutdown Margin Requirements (Tech Spec 3.1.1 and 3.1.2). The Mode 4 shutdown margin requirements are mandated by Tech Spec Section 3.1 .1.

The capability was added to utilize a boration flow path from the RWST to the RCS cold legs via safety injection pump during Modes 5 and 6 as sufficient head is developed to borate the RCS at the LTOP actuation setpoint and below the applicable pressure limits of Tech Spec 3.4.2 (RCS P-T Limits).

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

The REMEDIAL ACTION statement requires suspension of all operations 'involving CORE ALTERATIONS or positive reactivity changes.' The intent is that specific evolutions or operations that involve positive reactivity changes (fuel movement, dilutions, control rod movements or sustained NC temperature changes adding positive reactivity) are McGuire Units 1 and 2 16.9.12-2 Revision 101

Boration Systems - Flow Path (Shutdown) 16.9.12 BASES (continued) discontinued if the conditions described above do not exist. There are operations (e.g.,

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

Operational or testing activities that result in NC temperature swings of 20 degrees F about an initial value have been judged not to constitute positive reactivity changes as described in this SLC when in MODE 5. There must be at least 500 ppm boron beyond the required Shutdown Boron Concentration for this interpretation to remain valid. This interpretation should not be used to establish sustained NC system heatups or cooldowns that result in sustained positive reactivity additions.

Limited Boron concentration changes are allowed for inventory control or testing provided SDM is maintained and Keff is <0.99. Operations are not permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM boron requirements.

REFERENCES

1. NSD-403: NGD Shutdown Risk Management

2. MSD-403: McGuire Shutdown Risk Management

3. Nuclear/Reactor Engineering Memo to File R.F.4.0.i, August 23, 1994 'NC Temperature Swings affect on Shutdown Margin'

4. PIPs M97-0601, M98-4643, M-07-03237

5. DPND-1552.63-0099, Rev. 0, "Required Boration Flow Paths in Mode 4"

6. NRC Issuance of Amendments 184/166, Improved Technical Specifications conversion and relocations to SLC Manual McGuire Units 1 and 2 16.9.12-3 Revision 101

Boration Systems - Charging Pumps (Shutdown) 16.9.13 16.9 AUXILIARY SYSTEMS 16.9.13 Boration Systems - Charging Pumps (Shutdown)

(DELETED-COMBINED WITH 16.9.12)

McGuire Units 1 and 2 16.9.13-1 Revision 13

Borated Water Sources (Shutdown) 16.9.14 16.9 AUXILIARY SYSTEMS 16.9.14 Borated Water Sources (Shutdown)

COMMITMENT One of the following borated water sources shall be OPERABLE:

a. A boric acid tank (BAT), or

b. The refueling water storage tank.

APPLICABILITY MODE 4 with any RCS cold leg temperature < 3001F, MODES 5 and 6.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Required borated water A.1 Suspend CORE Immediately source inoperable. ALTERATIONS.

AND A.2 Suspend positive reactivity Immediately additions.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.14.1 Verify the refueling water storage tank solution 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months temperature is > 70°F when the outside air temperature is < 70 0 F.

TR 16.9.14.2 Verify the boron concentration of the required borated 7 days water source is within the limits specified in the COLR.

(continued)

McGuire Units 1 and 2 16.9.14-1 Revision 22

Borated Water Sources (Shutdown) 16.9.14 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.9.14.3 Verify the borated water volume of the required borated 7 days water source is within the limits specified in the COLR.

TR 16.9.14.4 Verify the boric acid tank solution temperature is > 65°F 7 days when the boric acid storage tank is a required source.

BASES The borated water sources ensure that negative reactivity control is available during each mode of facility operation.

In Mode 4 with any RCS cold leg temperature below 300 °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 SDM of 1.3% delta k/k in Mode 4 and 1.0% delta k/k in Modes 5 and 6 after xenon decay and cooldown to 680 F.

The SLC commitment values are presented in the Core Operating Limits Report (COLR) as:

(1) the minimum boron concentrations and minimum volumes necessary to attain and maintain SDM in the boric acid tank or the refueling water storage tank, (2) the minimum contained volumes in the boric acid tank or the refueling water storage tank, and (3) a curve specifying the minimum contained volume in the boric acid tank near EOC. The minimum contained water volume is based on the required volume to maintain shutdown margin, an allowance for water not available because of discharge line location and additional margin.

The additional margin term includes allowances for instrument uncertainty, vortexing and a margin term consisting of at least 5% of the volume necessary for SDM. The COLR specified volumes and boron concentrations satisfy SDM requirements during Mode 4 with any RCS cold leg temperature below 300 OF and in Modes 5 and 6.

Boric Acid Tank Reauirements for Maintainina SDM Required volume for maintaining SDM Presented in the COLR Unusable volume (to maintain full suction pipe) 4,199 gallons Additional margin 4,100 gallons Refuelinq Water Storaqe Tank Requirements for Maintaininq SDM Required volume for maintaining SDM Presented in the COLR Unusable volume (to maintain full suction pipe) 16,000 gallons Additional margin 23,500 gallons McGuire Units 1 and 2 16.9.14-2 Revision 22

Borated Water Sources (Shutdown) 16.9.14 BASES (continued)

The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 9.5 for the solution recirculated within containment after a LOCA. This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

REFERENCES None McGuire Units 1 and 2 16.9.14-3 Revision 22

Snubbers 16.9.15 16.9 AUXILIARY SYSTEMS 16.9.15 Snubbers COMMITMENT All snubbers shall be OPERABLE.

-NOTE Snubbers installed on non-safety systems may be excluded from these requirements provided their failure or the failure of the system on which they are installed would not have an adverse affect on any safety-related system.

APPLICABILITY At all times for snubbers located on systems required OPERABLE.

REMEDIAL ACTIONS NOTE Conditions A, B, and C are applicable to "seismic snubbers" as defined in the BASES.

McGuire Units 1 and 2 16.9.15-1 Revision 116

Snubbers 16.9.15 CONDITION REQUIRED ACTION COMPLETION TIME A. ------- NOTE------- A. 1.1 Verify that at least one Immediately If the opposite train of AFW train (including a the associated system minimum set of supporting becomes inoperable for equipment required for its reasons not related to successful operation) not snubbers while in associated with the Condition A, exit inoperable snubber(s), or Condition A and enter alternative core cooling Condition C. method, is OPERABLE.

AND One or more seismic snubbers associated A.1.2 Verify the opposite train of Immediately with one train of a the associated system is multiple train system operable, if a multiple train inoperable for system.

maintenance or testing and the opposite train of AND the associated system is operable. A.1.3 Log the affected system(s) Immediately for tracking in TSAIL.

OR AND One or more seismic snubbers associated A.1.4 Enter the applicable 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months with a single train ACTIONS for the train or system inoperable for component associated with maintenance or testing. the inoperable snubber(s).

OR A.2 Declare the supported Immediately system inoperable, (continued)

McGuire Units 1 and 2 16.9.15-2 Revision 116

Snubbers 16.9.15 CONDITION REQUIRED ACTION COMPLETION TIME B. One or more seismic B.1.1 Verify that at least one Immediately snubbers associated AFW train (including a with more than one train minimum set of supporting of a multiple train system equipment required for its inoperable for successful operation) not maintenance or testing. associated with the inoperable snubber(s), or alternative core cooling method, is OPERABLE.

AND B.1.2 Enter the applicable 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ACTIONS for the train or component associated with the inoperable snubber.

OR B.2 Declare the supported Immediately system inoperable.

(continued)

McGuire Units 1 and 2 16.9.15-3 Revision 116

Snubbers 16.9.15 CONDITION REQUIRED ACTION COMPLETION TIME C. One or more seismic C.1.1 Verify that at least one Immediately snubbers associated AFW train (including a with one train of a minimum set of supporting multiple train system equipment required for its inoperable for successful operation) not maintenance or testing. associated with the inoperable snubber(s), or AND alternative core cooling method, is OPERABLE.

The opposite train of the associated system is AND inoperable for reasons not related to snubbers. C.1.2 -------- Note-------

Ifthe reason for the inoperability of either the snubber or the opposite train of the associated system is due to planned work, the risk assessment must be completed prior to starting work.

Initiate a qualitative risk Immediately assessment of the resulting configuration.

AND C.1.3 Enter the applicable 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months from failure ACTION for the train or to meet the component associated with COMMITMENT the inoperable snubber.

OR C.2 Declare the supported Immediately system inoperable D. Required Action and D. 1 Enter the applicable Immediately associated Completion ACTIONS for the train or Time of Condition A, B, component associated with or C not met. the inoperable snubber.

(continued)

McGuire Units 1 and 2 16.9.15-4 Revision 116

Snubbers 16.9.15 CONDITION REQUIRED ACTION COMPLETION TIME E. One or more snubbers E.1 Enter the applicable Immediately with any significant non- ACTIONS for any affected seismic loads inoperable system(s) and for maintenance or component(s) that are testing. determined to be inoperable.

OR E.2 Perform an engineering Prior to removing the evaluation to determine the snubber from effect of the inoperable service.

snubber on the operability of the associated system.

F. One or more snubbers F.1 Perform an engineering 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable due to actual evaluation to determine the failure of the snubber or effect of the inoperable failure to meet test snubber on the operability acceptance criteria, of the associated system.

McGuire Units 1 and 2 16.9.15-5 Revision 116

Snubbers 16.9.15 TESTING REQUIREMENTS

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

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

2. Snubbers which fail the visual inspection or the functional test acceptance criteria shall bE repaired or replaced. Replacement snubbers and snubbers which have repairs which might affect the functional test results shall be tested to meet the functional test criteria before installation in the unit. Mechanical snubbers shall have met the acceptance criteri.

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.

3. As used herein, type of snubber shall mean snubbers of the same design and manufacturer, irrespective of capacity.

TEST FREQUENCY TR 16.9.15.1 NOTES -----------------

1. Snubbers are categorized as inaccessible or accessible during reactor operation and may be inspected independently according to the schedule determined by Table 16.9.15-1.

2. The first inspection interval using Table 16.9.15-1 shall be based upon the previous inspection interval as established by the requirements in effect before Technical Specification amendment 126.

Perform a visual inspection for each category of snubber. In accordance with Table 16.9.15-1 TR 16.9.15.2 -NOTE In case of a severe dynamic event, mechanical snubbers in that system which experienced the event shall be inspected during the refueling outage to assure that the mechanical snubbers have freedom of movement and are not frozen up Perform an inspection to determine if there has been a 18 months severe dynamic event for systems which have the potential for a severe dynamic event.

(continued)

McGuire Units 1 and 2 16.9.15-6 Revision 116

Snubbers 16.9.15 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.9.15.3 ------------------

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

1. The large bore steam generator hydraulic snubbers shall be treated as a separate population for functional test purposes and are functional tested under Sample Plan 1.

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

Perform snubber functional testing on a representative 18 months sample of each type of snubber in accordance with one of the following three Sampling Plans:

1. Functionally test 10% of a type of snubber with an additional 10% tested for each functional testing failure, or

2. Functionally test a sample size and determine sample acceptance or continue testing using Figure 16.9.15-1, or

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

TR 16.9.15.4 ------------------ NOTE ------------------

The parts replacement shall be documented and the documentation shall be retained for the duration of the unit operating license.

Verify that the service life of hydraulic snubbers has not 18 months been exceeded or will not be exceeded prior to the next scheduled surveillance inspection.

McGuire Units 1 and 2 16.9.15-7 Revision 116

Snubbers 16.9.15 BASES This commitment contains requirements for plant snubbers. There are snubbers installed for seismic loads only (i.e., those loads induced by seismic events, "earthquakes") and snubbers that are installed for the combined effects of both seismic loads and non-seismic loads (i.e.,

those dynamic loads induced by operational events such as steamhammer, waterhammer, LOCAs, and pipe rupture). Thus for the purpose of this commitment, there are two categories of snubbers:

1) snubbers which have only seismic loads, and snubbers which have both seismic and non-seismic loadings, but Engineering has determined that the non-seismic loads are insignificant and do not effect the operability of the associated system. Since the seismic loads are those of significance, these snubbers are termed "Seismic Snubbers" in this commitment; and

2) snubbers which have both seismic and non-seismic loadings and Engineering has determined that the non-seismic loads are significant and do effect the operability of the associated system.

The Remedial Actions for each of these snubber categories are discussed below. Remedial Action F. 1 and the Testing Requirements of this commitment apply to both categories of snubbers. The programmatic requirements for the' visual inspection and functional testing of snubbers do not meet the criteria in 10 CFR 50.36(c)(2)(ii) for inclusion in the plant TS, and as such, are appropriate for control by this commitment and are the same for both categories of snubbers.

The snubber requirements of SLC 16.9.15 were originally located in the Technical Specifications. The Nuclear Regulatory Commission (NRC) authorized the use of these requirements, while located in Technical Specifications, as an acceptable alternative to the requirements of the ASME Code, 1989 Edition,Section XI, Article IWF-5000 (References 3, 4).

Any revision to these snubber visual inspection and functional test requirements shall consider the basis for the granted relief from the ASME Code requirements and any resulting requirement for NRC review and approval.

All snubbers are required OPERABLE to ensure that the structural integrity of the Reactor Coolant System and all other safety-related systems is maintained during and following a seismic or other event initiating dynamic loads. Snubbers excluded from this inspection program are those installed on nonsafety-related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on any safety-related system. Snubbers are classified and grouped by design and manufacturer but not by size. For example, mechanical snubbers utilizing the same design features of the 2 kip, 10 kip, and 100 kip capacity manufactured by Company "A" are of the same type. The same design mechanical snubbers manufactured by Company "B" for the purposes of this specification would be of a different type, as would hydraulic snubbers from either manufacturer.

Snubbers removed from service for any reason cannot be considered OPERABLE since it is not connected to the supported system or component.

McGuire Units 1 and 2 16.9.15-8 Revision 116

Snubbers 16.9.15 BASES (continued)

Seismic Snubbers Seismic snubbers are installed primarily to address loads resulting from a seismic event.

However, some seismic snubbers do have other non-seismic loads, but these other loads have been determined to have an insignificant effect on the operability of the associated system, as determined by Engineering. If used, TS LCO 3.0.8 contains the OPERABILITY requirements for seismic snubbers.

LCO 3.0.8 establishes conditions under which systems are considered to remain capable of performing their intended safety function when associated seismic snubbers are not capable of providing their associated support function(s). This LCO states that the supported system is not considered to be inoperable solely due to one or more of these snubbers not being capable of performing their associated support function(s). Thus, any affected supported LCO(s) are not required to be declared not met solely for this reason, if risk is assessed and managed.

This is appropriate because a limited length of time is allowed for inspection, testing, maintenance, or repair of one or more of these snubbers not capable of performing their associated support function(s), remedial actions are specified in this commitment, and the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the snubber(s) are not capable of performing their associated support function, and as applicable, due to the availability of the redundant train of the supported system.

If the allowed time expires and the seismic snubber(s) are unable to perform their associated support function(s), the affected supported system's LCO(s) must be declared not met and the Conditions and Required Actions entered in accordance with LCO 3.0.2.

Snubbers with Both Seismic and Significant Non-Seismic Loads If the affected snubber has more than one function, one of which MUST be seismic loads, then LCO 3.0.8 may be applied. However, there must be a deterministic analysis that demonstrates that the supported system can still perform its function for the non-seismic load(s). For example, if the affected snubber has support functions for both seismic loads and LOCA loads (i.e., blowdown loads), then only that LOCA load is considered deterministically to determine if the system is OPERABLE. If the supported TS system is OPERABLE for the non-seismic loads, then LCO 3.0.8 may be applied to the seismic loads. Otherwise, LCO 3.0.8 may not be applied and the OPERABILITY requirements are contained in this commitment.

Remedial Actions - A Remedial Action A applies when one or more seismic snubbers associated with one train of a multiple train system and the opposite train of the associated system is operable or associated with a single train system are inoperable for maintenance or testing, thus are not capable of providing their associated support function(s). This commitment allows up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the seismic snubber(s) before declaring the supported system inoperable, provided:

1) there is an immediate determination that at least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), or alternative core cooling method (e.g., feed and bleed, firewater system or "aggressive secondary cooldown" using the steam generators) is OPERABLE, 2) the opposite McGuire Units 1 and 2 16.9.15-9 Revision 116

Snubbers 16.9.15 BASES (continued) train of the supported system is OPERABLE, if applicable, and 3) the affected system is logged for tracking in TSAIL. The 72- hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the seismic snubber(s) are not capable of performing their associated support function, and due to the availability of the redundant train of the supported system (if applicable).

At the end of the specified 72-hour period the required seismic snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met.

Condition A is modified by a Note which specifies that if the opposite train of the associated system becomes inoperable for reasons not related to snubbers while in Condition A, Condition A can be exited and Condition C is entered.

If the provisions of LCO 3.0.8 are not entered, the supported system shall be declared inoperable immediately.

Remedial Action - B When one or more seismic snubber(s) are not capable of providing their associated support function(s) to more than one train of a multiple train supported system, this commitment allows.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to restore the seismic snubber(s) before declaring the supported system(s) inoperable, provided there is an immediate determination that at least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), or alternative core cooling method (e.g., feed and bleed, firewater system or "aggressive secondary cooldown" using the steam generators) is OPERABLE. The 12-hour Completion Time is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the seismic snubber(s) are not capable of performing their associated support function.

At the end of the specified 12-hour period the required seismic snubbers must be able to perform their associated support function(s), or the affected supported system(s) LCO(s) shall be declared not met.

If the provisions of LCO 3.0.8 are not entered, the supported system shall be declared inoperable immediately.

Remedial Action - C When one or more seismic snubbers are not capable of providing their associated support function(s) to one train of a multiple train supported system, and the opposite train of the supported system is inoperable for reasons not related to snubbers, this commitment allows up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore the seismic snubber(s) before declaring the supported system inoperable provided: 1) there is an immediate determination that at least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), or alternative core cooling method (e.g., feed and bleed, firewater system or "aggressive secondary cooldown" using the steam generators) is OPERABLE, and 2)

McGuire Units 1 and 2 16.9.15-10 Revision 116

Snubbers 16.9.15 BASES (continued) there is an immediate assessment of risk associated with the resulting configuration and the risk assessment is acceptable. The 72-hour Completion Time from failure to meet the COMMITMENT (in case Condition C is entered after exiting Condition A) is reasonable based on the low probability of a seismic event concurrent with an event that would require operation of the supported system occurring while the seismic snubber(s) are not capable of performing their associated support function, and due to an acceptable conclusion of the risk assessment.

At the end of the specified 72-hour period the required seismic snubbers must be able to perform their associated support function(s), or the affected supported system LCO(s) shall be declared not met.

If the provisions of LCO 3.0.8 are not entered, the supported system shall be declared inoperable immediately.

Risk Assessment and Management Remedial Action A, B, and C require that risk be assessed and managed. Industry and NRC guidance on the implementation of 10 CFR 50.65(a)(4) (the Maintenance Rule) does not address seismic risk. However, use of these commitments should be considered with respect to other plant maintenance activities, and integrated into the existing Maintenance Rule Process to the extent possible so that maintenance on any unaffected train or subsystem is properly controlled, and emergent issues are properly addressed. When using this commitment to remove seismic snubber(s) from an operable state, the risk assessment must ensure that at least one AFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), or alternative core cooling method (e.g., feed and bleed, firewater system or "aggressive secondary cooldown" using the steam generators) is OPERABLE. This risk assessment is tracked by use of the TSAIL program. The risk assessment need not be quantified, but may be a qualitative awareness of the vulnerability of systems and components when one or more snubbers are not able to perform their associated support function. Actions that could be taken include protection of other trains or subsystems for example.

Remedial Action - D If the Required Action and associated Completion Time of Condition A, B or C are not met, the applicable ACTIONS for the train(s) or components(s) associated with the inoperable seismic snubber shall be entered immediately.

Remedial Action - E Should one or more snubbers which have any significant non-seismic loads be inoperable for the purposes of maintenance or testing. OPERABILITY of the affected system(s) and component(s) must be determined and the applicable ACTIONS entered immediately. If there remains a reasonable assurance of OPERABILITY of the affected system(s) or component(s) with the condition of an inoperable snubber(s), then it is not necessary to enter the respective ACTIONS for inoperable system(s) and component(s).

Remediate Action - F McGuire Units 1 and 2 16.9.15-11 Revision 116

Snubbers 16.9.15 BASES (continued)

Should one or more snubbers (of either category) fail to meet testing acceptance criteria or be discovered in a condition where failure is apparent, an engineering evaluation is to be performed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , as described in "Functional Test Failure Analysis".

The snubber-testing program may remove snubbers from service and restore OPERABILITY of the snubber application by replacement with another like snubber. In this situation, if the removed snubber application by replacement with another like snubber. In this situation, if the removed snubber later fails to meet test acceptance criteria, Conditions A, B, C, and E are not applicable since the snubber component has no current required function; however, ACTION F.1 would be applicable. During the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months allowed to perform an engineering evaluation, or at any other time when conditions of the affected system(s) and component(s) are determined to no longer support a reasonable assurance of OPERABILITY, applicable ACTIONS shall be entered immediately.

Visual Inspections The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable snubbers found during an inspection. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

Visual inspections shall verify: (1) that there are no visible indications of damage or impaired OPERABILITY, and (2) attachments to the foundation or supporting structure are secure.

Snubbers which appear inoperable as a result of visual inspections shall be classified as unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that: (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type that may be generically susceptible; and (2) the affected snubber is functionally tested in the as found condition and determined OPERABLE. A hydraulic snubber found with the fluid port uncovered and all hydraulic snubbers found connected to an inoperable common reservoir shall be classified as unacceptable and may be reclassified acceptable by functionally testing each snubber starting with the piston in the as-found setting, extending the piston rod in the tension direction.

Refueling Outage 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 isolation 'valves) the main steam safety and power-operated relief valves and piping, auxiliary feedwater system, main steam supply to the auxiliary feedwater pump turbine, and the letdown and charging portion of the NV system shall be inspected to determine if there has been a severe dynamic event.

McGuire Units 1 and 2 16.9.15-12 Revision 116

Snubbers 16.9.15 In case of a severe dynamic event, mechanical snubbers in that system which experienced the event shall be inspected during the refueling outage to assure that the mechanical snubbers have freedom of movement and are not frozen up. The inspection shall consist of verifying BASES (continued) freedom of motion using one of the following: (1) manually induced snubber movement; (2) evaluation of in-place snubber piston setting; (3) stroking the mechanical snubber through its full range of travel. If one or more mechanical snubbers are found to be frozen up during this inspection, those snubbers shall be replaced or repaired before returning to power. The requirements of TR 16.9.15.1 are independent of the requirements of this item.

Functional Testinq During the first refueling shutdown and at least once per refueling thereafter, a representative sample of snubbers shall be tested using one of the following sample plans. The large bore steam generator hydraulic snubbers shall be treated as a separate population for functional test purposes and are functional tested under Sample Plan 1. A 10% random sample from previously untested snubbers shall be tested at least once per refueling outage until the entire population has been tested. This testing cycle shall then begin anew. For each large bore steam generator hydraulic snubber that does not meet the functional test acceptance criteria, at least 10% of the remaining population of untested snubbers for that testing cycle shall be tested. The sample plan shall be selected prior to the test period and cannot be changed during the test period. The NRC shall be notified of the sample plan selected prior to the test period.

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

2. A representative sample of the required snubbers shall be functionally tested in accordance with Figurel6.9.15-1. "C" is the total number of snubbers found not meeting the acceptance requirements (failures). The cumulative number of snubbers tested is denoted by "N." Test results shall be plotted sequentially in the order of sample assignment (i.e.,

each snubber shall be plotted by its order in the random sample assignments, not by the order of testing). If at any time the point plotted falls in the "Accept region, testing of .

snubbers may be terminated. When the point plotted lies in the "Continue Testing" region, additional snubbers shall be tested until the point falls in the 'Accept" region, or all the required snubbers have been tested. Testing equipment failure during functional testing may invalidate that day's testing and allow that day's testing to resume anew at a later time, providing all snubbers tested with the failed equipment during the day of equipment failure are retested; or 3: An initial representative sample of fifty-five (55) snubbers shall be functionally tested. For each snubber which does not meet the functional test acceptance criteria, another sample of at least one-half the size of the initial sample shall be tested until the total number tested is equal to the initial sample size multiplied by the factor, 1 + C/2, where "C" is the number of snubbers found which do not meet the functional test acceptance criteria. This can be plotted using an "Accept" line which follows the equation N = 55(1 +

C/2). Each snubber should be plotted as soon as it is tested. If the point plotted falls on or McGuire Units 1 and 2 16.9.15-13 Revision 116

Snubbers 16.9.15 below the "Accept" line, testing may be discontinued, If the point plotted falls above the "Accept" line, testing must continue .unless all snubbers have been tested.

McGuire Units 1 and 2 16.9.15-14 Revision 1 16

Snubbers 16.9.15 BASES (continued)

The representative samples for the functional test sample plans shall be randomly selected from the required snubbers and reviewed before beginning the testing. The review shall ensure as far as practical that they are representative of the various configurations, operating environments, range of sizes, and capacities. Snubbers placed in the same locations as snubbers which failed the previous functional test shall be retested at the time of the next functional test but shall not be included in the sample plan. If during the functional testing, additional sampling is required due to failure of only one type of snubber, the functional testing results shall be reviewed at that time to determine if additional samples should be limited to the type of snubber which has failed the functional testing.

Figure 16.9.15-1 was developed using "Wald's Sequential Probability Ratio Plan" as described in "Quality Control and Industrial Statistics" by Acheson J. Duncan.

Permanent or other exemptions from the surveillance program for individual snubbers may be granted by the NRC if a justifiable basis for exemption is presented and, if applicable, snubber life destructive testing was performed to qualify the snubber for the applicable design conditions at either the completion of their fabrication or at a subsequent date.

Functional Test 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. Where required, 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.

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

McGuire Units 1 and 2 16.9.15-15 Revision 116

Snubbers 16.9.15 BASES (continued)

For the snubbers found inoperable, an engineering evaluation shall be performed on the components to which the inoperable snubbers are attached. The purpose of this engineering evaluation shall be to determine if the components to which the inoperable snubbers are attached were adversely affected by the inoperability of the snubbers in order to ensure that the component remains capable of meeting the designed service.

If any snubber selected for functional testing either fails to activate or fails to move, i.e., frozen-in-place, the cause will be evaluated and, if caused by manufacturer or design deficiency, all snubbers of the same type subject to the same defect shall be evaluated in a manner to ensure their OPERABILITY. This testing requirement shall be independent of the requirements stated in TR 16.9.15.3 for snubbers not meeting the functional acceptance criteria.

Service Life The expected service life for the various seals, seal materials, and applications shall be determined and established based on engineering information and the seals shall be replaced so that the expected service life will not be exceeded during a period when the snubber is required to be OPERABLE.

The service life of a snubber is established via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature area, etc.). The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.

McGuire Units 1 and 2 16.9.15-16 Revision 116

Snubbers 16.9.15 REFERENCES

1. Letter from M. S. Tuckman to NRC, Licensing Position Regarding Snubbers, May 20, 1999.

2. Letter from NRC to H.B. Barron, Licensing Position Regarding Snubbers, July 9, 1999.

3. Letter from H.B. Barron to NRC, Request for Relief 97-005, Snubber Inspections -

Performance and Schedule, December 17, 1997.

4. Letter from NRC to H.B. Barron, Relief Request for Snubber Visual examination and Functional Testing, May 27, 1998.

5. Letter from H.B. Barron to NRC, Request for Relief 01-004, June 1, 2001.

6. Letter from NRC to M.S Tuckman, Safety Evaluation of Relief Request No.01-004, Alternative for Snubber Examinations, January 30, 2002.

7. Letter from G.R. Peterson to NRC, Request for Relief 03-002, March 8, 2004.

8. Letter from G.R. Peterson to NRC, RAI Response, September 22, 2004.

9. Letter from NRC to G.R. Peterson, Safety Evaluation of Relief Request No.03-002, November 22, 2004.

10. Technical Specification Task Force (TSTF) 372-A, Revision 4, Addition of LCO 3.0.8.

Inoperability of Snubbers.

11. TSTF-IG-05-03, Rev 1, Technical Specifications Task Force Implementation Guidance for TSTF-372-A, Revision 4, Addition of LCO 3.0.8, Inoperability of Snubbers.

12. Nuclear System Directive 415, Operational Risk Management (Modes 1-3) per 10 CFR 50.65(a)(4).

13. Federal Register, 70FR23252, Notice of Availability of Model Application Concerning Technical Specification Improvement to Modify Requirements Regarding the Addition of Limiting Condition for Operation 3.0.8 on the Inoperability of Snubbers Using the Consolidated Line Item Improvement Process.

McGuire Units 1 and 2 16.9.15-17 Revision 116

Snubbers 16.9.15 TABLE 16.9.15-1 SNUBBER VISUAL INSPECTION INTERVAL NUMBER OF UNACCEPTABLE SNUBBERS Population or Category Column A Column A Column C (Notes 1, 2) Extended Interval Repeat Interval Reduced Interval (Notes 3, 6) (Notes 4, 6) (Notes 5, 6) 1 0 0 1 80 0 0 2 100 0 1 4 150 0 3 8 200 2 5 13 300 5 12 25 400 8 18 36 500 12 24 48 750 20 40 78

>1000 29 56 109 NOTES:

1. The next visual inspection interval for a snubber population or category size shall be determined based upon the previous inspection interval and the number of unacceptable snubbers found during that interval. Snubbers may be categorized, based upon their accessibility during power operation, as accessible or inaccessible. The categories may be examined separately or jointly. However, the licensee must make and document that decision before any inspection and shall use that decision as the basis upon which to determine the next inspection interval for that category.

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

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.

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

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.

6. The provisions of SLC 16.2.7 are applicable for all inspection intervals up to and including 48 months.

McGuire Units 1 and 2 16.9.15-18 Revision 116

Snubbers 16.9.15 10 9

8 7

C 6 5

CONTINUE TESTING 4

3 C =0O.055N- 2.007 2

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

FIGURE 16.9.15-1 SAMPLE PLAN 2 FOR SNUBBER FUNCTIONAL TEST McGuire Units 1 and 2 16.9.15-19 Revision 116

Area Temperature Monitoring 16.9.16

.16.9 AUXILIARY SYSTEMS 16.9.16 Area Temperature Monitoring COMMITMENT The temperature of each area shown in Table 16.9.16-1 shall be maintained within the limits indicated in Tablel6.9.16-1.

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

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIMI]

A. One or more areas, A.1 Initiate actions to restore Immediately except Diesel Generator the temperature within limit.

Rooms, exceeding temperature limit(s) AND shown in Table 16.9.16-

1. A.2 Declare equipment in the Immediately affected area inoperable.

B. Diesel Generator Room B.1 Initiate actions to restore Immediately temperature is > 125 0 F. Diesel Generator Room temperature to within limits.

AND B.2 Declare Diesel Generator Immediately inoperable.

C. Diesel Generator Room C.1 Initiate actions to restore Immediately temperature is < 55 0 F. Diesel Generator Room temperature to within AND limits.

Diesel Generator is not AND running.

C.2. Declare Diesel Generator Immediately inoperable.

McGuire Units 1 and 2 16.9.16-1 Revision 111

Area Temperature Monitoring 16.9.16 REMEDIAL ACTIONS (continued)

TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.16.1 Verify temperature in each of the areas shown in Table 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 16.9.16-1 is within limits.

McGuire Units 1 and 2 16.9.16-2 Revision 111

Area Temperature Monitoring 16.9.16 TABLE 16.9.16-1 AREA TEMPERATURE MONITORING AREA TEMPERATURE LIMIT (-F) i -

1. Containment Spray Pump Rooms 145

2. Miscellaneous Terminal Cabinets

a. TB208-209 (Turbine Building Unit 1) 150

b. TB496 (Fuel Building Unit 1) 150

c. TB1208-1209 (Turbine Building Unit 2) 150

d. TB1496 (Fuel Building Unit 2) 150

3. Residual Heat Removal Pump Rooms 145

4. Diesel Generator Rooms >55 and < 125

5. Spent Fuel Pool Cooling Pump Room 145 McGuire Units 1 and 2 16.9.16-3 Revision 111

Area Temperature Monitoring 16.9.16 BASES The area temperature limitations ensure that safety-related equipment will not be subjected to temperatures in excess of their environmental qualification temperatures. Exposure to excessive temperatures may degrade equipment and can cause a loss of its OPERABILITY. The Diesel Generator Room temperature is measured by taking an average of the Battery Area and Control Panel Area thermocouples. This is indicated on an OAC point called Room Average Temperature. The Diesel Generator Room temperature limit of 125 0 F is based on the temperature qualification of the Diesel Generator Battery Chargers. The Diesel Generator Room temperature limit of 55 0 F is based on the currently analyzed ambient area temperature for various diesel generator systems. The low temperature limit applies only when the Diesel Generator is not running to support standby readiness. The OPERABILITY evaluation associated with the restoration of the Diesel Generator to OPERABLE status will consider the effect of the off-limit temperature on the associated Diesel Generator SSC's.

Miscellaneous Terminal Cabinets TB208 (TB1208) and TB209 (TB1209) contain circuits associated with the function of closing the main feedwater control valves CF17, 20, 23 and 32 and main feedwater bypass valves CF104, 105, 106 and 107 on safety injection, Hi Hi S/G level, and Lo Tavg coincident with Rx trip. These cabinets also include relays used to forward the signal for main feedwater pump trip on turbine trip. TB208 is A train and TB209 is B train.

Miscellaneous Terminal Cabinets TB208 and TB1208 also contain the AMSAC inputs from main feedwater control valves CF17, 20, 23 and 32 limit switches.

Miscellaneous Terminal Cabinets TB496 (TB1496) contain circuits for IASV5080 which closes on phase A isolation to separate the air reservoir for the upper Containment Air lock from the VI system.

The maximum temperatures allowed to prevent failure of relays and fuses located in cabinets TB208, 209, 496, 1208, 1209, and 1496 are based upon manufacturer's recommendations and environmental qualification summary data.

Each air handling fan coil unit that is located in a Containment Spray (NS) pump rooms and Residual Heat Removal (ND) pump rooms provide an essential support function to the operability of the associated NS and ND pump motor. Should one of these air handling units become degraded, the operability of the affected train of the NS and ND system shall be evaluated per Technical Specification requirements in addition to requirements of SLC 16.9.16.

Although the Spent Fuel Pool Cooling (KF) pumps and motors are not contained within a Technical Specification, or subject to Operability, their function is vitally important to preventing boiling of the Spent Fuel pool. As such KF pump room temperature problems should be expeditiously resolved or alternate pool cooling would need to be implemented as committed in UFSAR section 9.1.3.

McGuire Units 1 and 2 16.9.16-4 Revision 111

Area Temperature Monitoring 16.9.16 REFERENCES

1. MCC-1 211.00-00-0004, Diesel Generator Ventilation Calculation.

2. MCC-1 381.05-00-0313, ND Pump Motor Upper Thrust Bearing Environmental Qualification.

3. MCC-1381.05-00-0316, KF Motor Stator Thermal Life Environmental Qualification.

4. MCC-1381.05-00-0333, NS Pump Motor Lower Bearing and Oil Environmental Qualification.

5. MCC- 1240.03-00-0001, McGuire Plant Environmental Parameters (PEP) Manual

6. MCC-1223.24-00-0065, ND, NS, and KF Pump Motor Cooler Operability Evaluation

7. MCTC-1579-VD.SO01-01, Diesel Generator Room Air Temperature.

8. PIPs M-93-0004, M-94-0013, M-00-1248, M-02-0015, M-04-0074, and M-03-1309.

McGuire Units 1 and 2 16.9.16-5 Revision 111

Refueling Operations - Decay Time 16.9.17 16.9 AUXILIARY SYSTEMS 16.9.17 Refueling Operations - Decay Time COMMITMENT The reactor shall be subcritical format least 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

APPLICABILITY During movement of irradiated fuel in the reactor vessel.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor is subcritical for A.1 Suspend all operations Immediately

< 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . involving movement of irradiated fuel in reactor vessel.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.17.1 Verify reactor has been subcritical for > 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months by Prior to movement of verification of date and time of subcriticality. irradiated fuel in reactor vessel BASES The minimum requirement for reactor subcriticality prior to movement of irradiated fuel assemblies in the reactor vessel ensures that sufficient time has elapsed to allow the radioactive decay of the short-lived fission products. This decay time is consistent with the assumptions used in the accident analyses.

This SLC is limited to irradiated fuel movement within the reactor vessel. Movement of irradiated fuel assemblies from containment to the spent fuel pool is controlled through the Total Core Offloading process. A decay heat calculation may need to be performed to determine when irradiated fuel can be moved to the spent fuel pool following subcriticality.

McGuire Units 1 and 2 16.9.17-1 Revision 86

Refueling Operations - Decay Time 16.9.17 REFERENCES

1. MCC 1227.00-00-0097, AST Dose Analysis of Fuel Handling Accidents..

2. AST Tech Spec Amendment Nos. 236/218 dated December 22, 2006.

McGuire Units 1 and 2 16.9.17-2 Revision 86

Refueling Operations - Communications 16.9.18 16.9 AUXILIARY SYSTEMS 16.9.18 Refueling Operations - Communications COMMITMENT Direct communications shall be maintained between the control room and personnel at the refueling station.

APPLICABILITY During CORE ALTERATIONS.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION, COMPLETION TIME A. Direct communications A.1 Suspend CORE Immediately between control room ALTERATIONS.

and refueling station personnel cannot be maintained.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.18.1 Demonstrate direct communications between control Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior room and personnel at refueling station. to start of CORE ALTERATIONS AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

'thereafter McGuire Units 1 and 2 16.9.18-1 Revision 0

Refueling Operations - Communications 16.9.18 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.

REFERENCES None McGuire Units 1 and 2 16.9.18-2 Revision 0

Refueling Operations - Manipulator Crane 16.9.19 16.9 AUXILIARY SYSTEMS 16.9.19 Refueling Operations - Manipulator Crane COMMITMENT The reactor building manipulator crane and an auxiliary hoist shall be used for movement of fuel assemblies or control rods and shall be OPERABLE with:

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

1. A minimum capacity of 3250 pounds, and

2. An overload cutoff limit < 2900 pounds.

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

1. A minimum capacity of 1000 pounds, and

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

APPLICABILITY During movement of fuel assemblies and control rods within the reactor vessel.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor building crane A.1 Suspend use of inoperable Immediately or auxiliary hoist reactor building crane or inoperable, auxiliary hoist from operations involving movement of fuel assemblies and control rods within the reactor vessel.

McGuire Units 1 and 2 16.9.19-1 Revision 102

Refueling Operations - Manipulator Crane 16.9.19 TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.19.1 Perform load test of > 3250 pounds and demonstrate Within 30 days automatic load cutoff at < 2900 pounds on each prior to the start of manipulator crane used for movement of fuel assemblies movement of fuel within the reactor vessel, assemblies within the reactor vessel TR 16.9.19.2 Perform a load test of >1000 pounds on each auxiliary Within 30 days hoist and associated load indicator used for movement of prior to the start of control rods or control rod drag load testing within the movement of reactor vessel, control rods or control rod drag load testing BASES The OPERABILITY requirements for the manipulator cranes ensure that: (1) manipulator cranes will be used for movement of drive rods and fuel assemblies, (2) each crane has sufficient load capacity to lift a drive rod or fuel assembly, and (3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

REFERENCES None McGuire Units 1 and 2 16.9.19-2 Revision 102

Crane Travel, - Spent Fuel Storage Pool Building 16.9.20 16.9 AUXILIARY SYSTEMS 16.9.20 Crane Travel - Spent Fuel Storage Pool Building COMMITMENT .The following requirements shall be met:

a. Loads in excess of 3000 pounds shall be prohibited from travel over fuel assemblies in the storage pool,

b. Spent fuel casks shall be carried along the path outlined in Figure 16.9.20-1 in the fuel pit and fuel pool area, and

c. The requirements of LCO 3.8.2 shall be met whenever loads are moved over the spent fuel storage pool.

Spent fuel pool weir gates may be moved over the stored fuel provided the decay time is > 17.5 days since last being part of a core at power.

APPLICABILITY With fuel assemblies in the storage pool.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Place the crane load in a Immediately safe condition and suspend movement of loads over the spent fuel pool.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.20.1 Verify weight of each load, other than a fuel assembly Prior to moving the and control rod, is < 3000 pounds. load over fuel assemblies McGuire Units 1 and 2 16.9.20-1 Revision 8

Crane Travel - Spent Fuel Storage Pool Building 16.9.20 NOTE: Cask may move as shown in the east or west directions inside the Cask Pit once the cask has completely cleared the Cask Pit north wall.

N Required Path of CL Spent Fuel Cask +

6" (Either Side)

Figure 16.9.20-1 REQUIRED PATH FOR MOVEMENT OF SPENT FUEL CASKS McGuire Units 1 and 2 16.9.20-2 Revision 8

Crane Travel - Spent Fuel Storage Pool Building 16.9.20 BASES The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the-storage pool ensures that in the event this load is dropped: (1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array. This assumption is consistent with the activity release assumed in the accident analysis. The requirement for following the load path shown in Figure 16.9.20-1 assumes that the cask can not fall into the spent fuel pool.

REFERENCES None McGuire Units 1and 2 16.9.20-3 Revision 8

Water Level - Spent Fuel Storage Pool 16.9.21 16.9 AUXILIARY SYSTEMS 16.9.21 Water Level - Spent Fuel Storage Pool COMMITMENT At least 23 ft of water shall be maintained over the top of irradiated fuel assemblies seated in the storage racks.

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

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel storage pool A.1 Suspend all movement of Immediately water level not within fuel assemblies and crane limit. operations with loads in the fuel storage area.

AND A.2 Restore spent fuel storage 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months pool water level to within limit.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.21.1 Verify the water level in the spent fuel storage pool is 7 days

> 23 ft over the top of irradiated fuel assemblies seated in the storage racks.

McGuire Units 1 and 2 16&9.21-1 Revision 0

Water Level - Spent Fuel Storage Pool 16.9.21 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 accident analysis.

REFERENCES None McGuire Units 1 and 2 16.9.21-2 Revision 0

SGRVS 16.9.22 16.9 AUXILIARY SYSTEMS 16.9.22 Switchgear Room Ventilation System (SGRVS)

COMMITMENT 1. Two trains of SGRVS for each switchgear room shall be FUNCTIONAL. I

2. Temperature in each switchgear room shall be < 90 'F.

APPLICABILITY Whenever the specified equipment in the switchgear room is required to be OPERABLE.

REMEDIAL ACTIONS NOTE------

. NOTE-------- ........................

Separate condition entry is allowed for each switchgear room.

CONDITION REQUIRED ACTION COMPLETION TIME A. One train of SGRVS non- A.1 Restore SGRVS to 30 days functional. FUNCTIONAL status.

B. Two trains of SGRVS non- B.1 Restore one SGRVS to 7 days functional. FUNCTIONAL status.

AND B.2 Verify switchgear room Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months temperature is < 90 0 F.

C. Switchgear room C. 1 Restore temperature to < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months temperature > 90 0 F. 90 0 F.

AND C.2 Initiate an engineering Immediately evaluation to determine the effect of the off-limit temperature on the affected equipment.

McGuire Units 1 and 2 16.9.22-1 Revision 109

SGRVS 16.9.22 TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.22.1 Verify one train of SGRVS FUNCTIONAL and in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months service.

TR 16.9.22.2 Verify each switchgear room is < 90'F. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months BASES The SGRVS (see Figure 16.9.22-1) provides cooling for the four essential switchgear rooms 1ETA, 1 ETB, 2ETA, and 2ETB. Each switchgear room SGRVS has two redundant trains that provide cooling through shared ductwork. Each train consists of an air handling unit (AHU) and isolation dampers. Each AHU contains a pre-filter, water cooling coils and a fan. Air conditioning is provided through circulation of chilled water in the AHU water cooling coils.

Temperature control for each switchgear room is affected by a thermostat sensing the return air of each AHU and controlling a 3-way chilled water control valve. The air is cooled to a desired temperature by exchanging heat with the chilled water. The design basis for the SGRVS is to maintain the environment in the switchgear room within an acceptable limit for the operation of unit controls.

Each train of the SGRVS is capable of maintaining the temperature in the switchgear room to less than or equal to 90 0 F. This temperature is limited by the Agastat timers in panels 1ATC23, 1ATC24, 2ATC23, and 2ATC24 located in the switchgear rooms. This temperature limit is provided to assure that the equipment in the room will have an acceptable service life; therefore, it will not affect the switchgear OPERABILITY. When the room temperature limit is exceeded, alternate cooling method can be used to return the temperature to within limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

If both trains of SGRVS are secured, switchgear room temperature shall be verified less than or equal to 90°F once per two hours.

REFERENCES

1. PIR 0-M91-0114, PIP M99-1819, PIP M99-4473, PIP MOO-1604, PIP M05-5880

2. Letter dated 11/20/91, S.C. Shealy to R.R. Weidler

3. UFSAR 6.4

4. UFSAR 7.6.10 McGuire Units 1 and 2 16.9.22-2 Revision 109

SGRVS 16.9.22 Figure 16.9.22-1 SWITCHGEAR AHUs w0 C!n McGuire Units 1 and 2 16.9.22-3 Revision 109

CRAVS 16.9.23 16.9 AUXILIARY SYSTEMS 16.9.23 Control Room Area Ventilation System (CRAVS)

COMMITMENT 1. Two trains of CRAVS shall be OPERABLE.

2. Temperature in areas listed in Table 16.9.23-1 and Table 16.9.23-2 shall be < specified limits.

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

REMEDIAL ACTIONS

,J CONDITION REQUIRED ACTION COMPLETION TIME A. One control room area air A.1 Restore CRA-AHU to 30 days handling unit (CRA-AHU) OPERABLE status.

inoperable.

B. Two CRA-AHUs B.1 Restore one CRA-AHU to 7 days inoperable. OPERABLE status.

AND B.2 Verify temperature in Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months areas listed in Table 16.9.23-1 is < specified limits.

C. One battery room exhaust C. 1 Restore BR-XF to 30 days fan (BR-XF) inoperable. OPERABLE status.

(Continued)

McGuire Units 1 and 2 16.9.23-1 Revision 88

CRAVS 16.9.23 REMEDIAL ACTIONS (Continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Two BR-XFs inoperable. D.1 Lock open all BR-XF Immediately check dampers.

AND D.2 Restore at least one BR- 7 days XF to OPERABLE status with all check dampers unlocked/restored.

AND D.3.1 Verify temperature in Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Battery Room No. 701 is

< 950 F.

OR D.3.2 Verify temperature in Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months each battery room listed in Table 16.9.23-2 is <

104 0 F.

AND D.4.1 Verify total flow through Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months battery rooms is > 770 cfm.

OR D.4.2 Verify hydrogen Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months concentration in each battery room listed in Table 16.9.23-2 is < 2%.

(Continued)

McGuire Units 1 and 2 16.9.23-2 Revision 88

CRAVS 16.9.23 REMEDIAL ACTIONS (Continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Room temperature in Table E. 1 Restore temperature to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months 16.9.23-1 or Table 16.9.23- within limit.

2, except temperature in Battery Room No. 701, not AND met.

E.2 Initiate an engineering Immediately evaluation to determine the effect of the off-limit temperature on the affected equipment.

F. Temperature in Battery F. 1 Verify temperature in Once per 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Room No. 701 not met. each battery room listed in Table 16.9.23-2 is <

104 0F.

G. Total flow through battery G.1 Restore total flow or 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months rooms or hydrogen hydrogen concentration concentration in battery to within limit.

room not met.

AND G.2 Suspend all battery Immediately equalize charging.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.9.23.1 Verify one CRA-AHU and one BR-XF OPERABLE 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in service.

McGuire Units 1 and 2 16.9.23-3 Revision 88

CRAVS 16.9.23 BASES The CRAVS (see Figure 16.9.23-1) provides cooling for the electrical penetration rooms, battery rooms, motor control center (MCC) rooms, cable rooms, restricted instrument shop, instrument storage room, and mechanical equipment room. The restricted instrument shop and instrument storage room temperature limits are not required by this SLC since they do not contain equipment vital to the operation of the plant. The CRAVS has two redundant trains. Each train consists of a non-safety control room area outside air fan (CRA-OAF), an air handling unit (AHU) and a battery room exhaust fan (BR-XF). The CRA-OAFs are not required to maintain operability of the CRAVS trains. An AHU of the CRAVS contains a pre-filter, water cooling coils and a fan. Air conditioning is provided through circulation of chilled water in the AHU water cooling coils. Temperature control for the CRAVS is affected by a room thermostat located in each of 24 zones. The air is cooled to a desired temperature by exchanging heat with the chilled water. The supply air to the battery rooms is returned via BR-XFs. These fans assist cooling by exhausting more air than is supplied. These fans also prevent hydrogen concentration from increasing to a 2% volume limit. Total flow of at least 770 cfm through the battery rooms is adequate to maintain hydrogen concentration less than or equal to 2% volume based on minimum supply flow to maintain temperature of 104 0 F. Hydrogen is emitted during discharging, float operation and especially during equalize charging of the batteries.

Each train of the CRAVS is capable of maintaining the temperature in the rooms to less than or equal to 104 0 F. This temperature is based on the maximum allowable temperature for continuous duty rating for equipment and instrumentation found in the areas served by CRAVS. The 95°F limit for Battery Room 701 in Table 16.9.23-1 is an administrative limit while the 1040 F limit for Battery Rooms 706 through 711 in Table 16.9.23-2 is the actual equipment duty rating limit. This temperature limit is provided to assure that the equipment in the room will have an acceptable service life; therefore, it will not affect the battery or MCC OPERABILITY. When the room temperature limit or hydrogen concentration limit is exceeded, alternate cooling method or hydrogen purging method can be used to return the temperature or hydrogen concentration to within limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The Batteries capacities can also be affected at a minimum temperature of 60 degrees.

Technical Specification Surveillance Requirement 3.8.6.3 verifies the average electrolyte temperature remains equal to or above 60 degrees. In addition, it has been shown by calculation MCC-1 211.00-00-00042 that the CRAVS cannot drive the Battery room's temperature below 60 degrees. Therefore, there is no need for minimum temperature requirement within this SLC If both CRA-AHUs or both BR-XFs are secured, verify temperature in Table 16.9.23-1 is less than or equal to the specified limits once per two hours.

McGuire Units 1 and 2 16.9.23-4 Revision 88

CRAVS 16.9.23 BASES (Continued)

Table 16.9.23-1 ROOM DESCRIPTION ROOM MAX.

NO. ELEV. (ft.) TEMP. (OF) 926 ELECTRICAL PENETRATION 767 104 928 ELECTRICAL PENETRATION 767 104 808 MCCs 1 EMXA AND 1EMXA-1 750 104 722 MCCs 1EMXB, 1EMXB-1, 1EMXB-2 & 1EMXB-3 733 104 821 MCCs 2EMXA, 2EMXA-1 & 1EMXH 750 104 724 MCCs 2EMXB, 2EMXB-1, 2EMXB-2 & 2EMXB-3 733 104 723A MCC 2EMXH 733 104 801 CABLE 750 104 801C CABLE 750 104 933 MECHANICAL EQUIPMENT 767 104 701 BATTERY 733 95 Table 16.9.23-2 ROOM DESCRIPTION ROOM MAX.

NO. ELEV. (ft.) TEMP. (OF) 707 BATTERY EVCA 733 104 708 BATTERY EVCB 733 104 710 BATTERY EVCC 733 104 711 BATTERY EVCD 733 104 706 BATTERY CXA 733 104 709 BATTERY CXB 733 104 REFERENCES

1. PIR 0-M91-0114, PIP M99-1819, PIP M99-4473, PIP M00-1604, PIP M-04-3742

2. UFSAR 6.4

3. UFSAR 7.6.10

4. MCC-1211.00-00-00042

5. MCS-1578.VC-00-0001 McGuire Units 1 and 2 16.9.23-5 Revision 88

CRAVS 16.9.23 Figure 16.9.23-1 SIMPLIFIED CONTROL AREA VENTILATION SYSTEM

I I- w.

I- m =) i f" UjjI: U) -

<z-7 I IR°° III 0 a. < CRAVS--

Ir E03 BAIT CABLE)

EL. Pr4ET ROOM RETURN OCRAVS bz

'uozý-C)O-i ný00 r McGuire Units 1 and 2 16.9.23-6 Revision 88

16.9.24-16.9 AUXILIARY SYSTEMS 16.9.24 Not Used McGuire Units 1 and 2 16.9.24-1 Revision 74

Refueling Operations - Containment Equipment Hatch

,16.9.25 16.9 AUXILIARY SYSTEMS 16.9.25 Refueling Operations - Containment Equipment Hatch COMMITMENT The containment equipment hatch shall be closed and held in place by a minimum of four bolts during movement of non-recently irradiated fuel assemblies within containment.

APPLICABILITY During movement of non-recently irradiated fuel assemblies within containment.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A.1 Suspend movement of Immediately non-recently irradiated fuel assemblies within containment.

TESTING REQUIREMENTS None BASES The Selective Alternative Source Term (AST) Technical Specification (TS) License Amendment allowed relaxation of containment closure requirements during movement of non-recently irradiated fuel assemblies by revising the Applicability requirements of TS 3.9.4, Containment Penetrations. Non-recently irradiated fuel is defined as fuel that has not occupied a critical reactor core within the last 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

The re-analysis of the Fuel Handling Accidents (FHAs) using AST methodology determined that Control Room, Exclusion Area Boundary (EAB), and Low Population Zone (LPZ) doses remained within regulatory limits without containment closure. After a FHA, it is advisable to close containment to further limit doses and prevent an uncontrolled airborne release. But, due to the present design of the containment equipment hatch, the hatch cannot be closed in a timely fashion without exposing workers to significant doses. All other smaller containment openings including the personnel air locks can be closed safely.

Therefore, until a safe and efficient means of closing or blocking the opening of the containment equipment hatch is developed, the containment equipment hatch shall McGuire Units 1 and 2 16.9.25-1 Revision 87

Refueling Operations - Containment Equipment Hatch 16.9.25 BASES (Continued) remain closed during movement of non-recently irradiated fuel assemblies within containment. Good engineering practice dictates that the bolts required by this SLC Commitment be approximately equally spaced.

The control of the movement of heavy loads within containment to preclude a FHA is provided by station procedures as specified by NUREG-0612..

REFERENCES

1. AST License Amendment Request dated December 20, 2005

2. AST Tech Spec Amendment Nos. 236/218 and NRC Safety Evaluation dated December 22, 2006

3. MCC 1227.00-00-0097, AST Dose Analysis of the Fuel Handling and Weir Gate Drop Accidents

4. NSD 403 Rev. 16, Shutdown Risk Management McGuire Units 1 and 2 16.9.25-2 Revision 87

Steam Generator Pressure/Temperature Limitation 16.10.1 16.10 STEAM AND POWER CONVERSION 16.10.1 Steam Generator Pressure/Temperature Limitation COMMITMENT Temperatures of both reactor and secondary coolants in the steam generators shall be maintained in accordance with Table 16.10.1-1.

NOTE: If steam generator level is < 10% WR, the secondary coolant temperature limit is not applicable.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Commitment not met. A. 1 Reduce steam generator 30 minutes pressure of the applicable side to within specified limits.

AND A.2 Perform an engineering Prior to increasing evaluation to determine the SG pressure above effect of overpressurization the specified limits.

on the structural integrity of the steam generator and determine the steam generator remains acceptable for continued operation.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.10.1.1 ------------------- NOTE -----------------

Only required to be performed when the temperature of either the reactor or secondary coolant is < 70 IF.

Verify the pressure in each side of the steam generator is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months less than the specified limits.

McGuire Units 1 and 2 16.10.1-1 Revision 56

Steam Generator Pressure/Temperature Limitation 16.10.1 Table 16.10.1-1 TEMPERATURE PRESSURE LIMIT Reactor and secondary coolant temperature No limitations by this SLC.

>700 F Lowest reactor or secondary coolant Primary pressure < 400 psig.

temperature > 55 and < 700 F Secondary pressure < 200 psig.

Lowest reactor or secondary coolant Primary pressure < 200 psig.

temperature < 550 F Secondary pressure < 200 psig.

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 steam generator P/T limits based on a steam generator RTNDT of 0 OF and are sufficient to prevent brittle fracture.

When steam generator WR level is less than 10%, the secondary coolant temperature indications are not valid. Due to close thermal coupling of temperatures at the tube sheet, primary system temperature should be used.

REFERENCES PIP M02-1502 MCC-1223.03-00-0049 MGMM-14512 and MGMM-14516 McGuire Units 1 and 2 16.110.1-2 Revision 56

Liquid Effluents - Concentration 16.11.1 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.1 Liquid Effluents - Concentration COMMITMENT The concentration of radioactive material released in liquid effluents to UNRESTRICTED AREAS (see Figure 16.11.1-1) shall be limited:

a. For radionuclides other than dissolved or entrained noble gases, 10 times the effluent concentrations specified in 10 CFR Part 20, Appendix B, Table 2, Column 2, and

b. For dissolved or entrained noble gases, the concentration shall be limited-to 2 x 10-4 microCurie/ml total activity.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of A. 1 Restore the concentration Immediately radioactive material to within limits.

released in liquid effluents to UNRESTRICTED AREAS not within limits.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.1.1 ------------------- NOTE-----------------

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.

Sample and analyze radioactive liquid wastes according According to to Table 16.11.1-1. Table 16.11.1-1 McGuire Units 1 and 2 16.11.1-1 Revision 112

Liquid Effluents - Concentration 16.11.1 TABLE 16.11.1-1 (Page 1 of 3)

RADIOACTIVE LIQUID WASTE SAMPLING AND ANALYSIS PROGRAM LIQUID RELEASE SAMPLING MINIMUM TYPE OF LOWER LIMIT TYPE FREQUENCY ANALYSIS ACTIVITY OF DETECTION FREQUENCY ANALYSIS (LLD) microCi/ml (1)

1. Batch Waste P P Principal 7 Release Tanks Each Batch Each Batch Gamma 5x10.

(WMT and Emitters(6)

RMT)(4) 1-131 1x10-6 P M Dissolved and One Batch/M Entrained 1xl0-Gases (Gamma emitters)(7)

P M H-3 1x10s Each Batch Composite(2)

Gross Alpha lx10,7 P Q Sr-89, Sr-90 5x10-8 Each Batch Composite(2)

2. Continuous Continuous(3) W Principal 7 Releases Composite(3 ) Gamma 5x10 (VUCDT Emitters (6) discharge, CWWTS outlet and Turbine Building Sump to RC)()

1-131 1x, 0-6 M M Dissolved and Grab Sample Entrained 1x10 5 Gases (Gamma emitters)(7)

Continuous(3) M H-3 1x10-1 Composite (3)

Gross Alpha 1x10 7 Continuous(3) Q Sr-89, Sr-90 5x10-"

Composite(3)

McGuire Units 1 and 2 16.11.1-2 Revision 112

Liquid Effluents - Concentration 16.11.1 TABLE 16.11.1-1 (Page 2 of 3)

NOTES:

(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 = 4.66 sb E

V - 2.22 x 106 - Y- exp (-AAI)

Where:

LLD is the "a priori" lower limit of detection as defined above (as microCurie per unit mass or volume),

Sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per disintegration),

V is the sample size (in units of mass or volume),

2.22 x 106 is the number of disintegrations per minute per microCurie, Y is the fractional radiochemical yield (when applicable),

X is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

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.

(2) A composite sample is one in which the quantity of liquid sampled is proportional to the quantity of liquid waste discharged and in which the method of sampling employed results in a specimen which is representative of the liquids released.

McGuire Units 1 and 2 16.11.1-3 Revision 112

Liquid Effluents - Concentration 16.11.1 TABLE 16.11.1-1 (Page 3 of 3)

(3) To be representative of the quantities and concentrations of radioactive materials in liquid effluents, samples shall be collected continuously or intermittently 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.

(4) A batch release is the discharge of liquid-wastes of a discrete volume. Prior to sampling for analyses, each batch shall be isolated and thoroughly mixed to assure representative sampling.

(5) A continuous release is the discharge of liquid wastes of a nondiscrete volume; e.g.,

from a volume of system that has an input flow during the continuous release.

(6) The principal gamma emitters for which the LLD specification applies exclusively are 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 5x10 6 microCi/ml. This list does not mean that only these nuclides are to be detected and reported. Other peaks which are measurable and identifiable, together with the above nuclides, shall be identified and reported in the Annual Radioactive Effluent Release Report.

(7) The principal gas gamma emitters for which the LLD specification applies are Xe-1 33 and Xe-135. These are the reference nuclides in Regulatory Guide 1.21.

McGuire Units 1 and 2 16.11.1-4 Revision 112

Liquid Effluents - Concentration 16.11.1 FIGURE 16.11.1-1 SITE BOUNDARY / EXCLUSION AREA BOUNDARY McGuire Units 1 and 2 16.11.1-5 Revision 112

Liquid Effluents - Concentration 16.11.1 BASES This commitment is provided to ensure that the concentration of radioactive materials released in liquid waste effluents to UNRESTRICTED AREAS will be less than 10 times the effluent concentration levels specified in 10 CFR Part 20, Appendix B, Table 2, Column 2.

This limitation provides additional assurance that the levels of radioactive materials in bodies of water in UNRESTRICTED AREAS will result in exposures within: (1) the Section II.A design objectives of Appendix 1, 10 CFR Part 50, to a MEMBER OF THE PUBLIC, and (2) the limits of 10 CFR Part 20.1301 to the population. The concentration limit for dissolved or entrained noble gases is based upon the assumption that Xe-1 35 is the controlling radioisotope and its EC in air (submersion) was converted to an equivalent concentration in water using the methods described in International Commission on Radiological Protection (ICRP) Publication 2. This commitment applies to the release of liquid effluents from all reactors at the site.

The basic requirements for the Selected Licensee Commitments concerning effluents from nuclear power reactors are stated in 10CFR50.36a. 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 10CFR20.106 (new 10CFR20.1301). 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 10CFR20.106 which references Appendix B, Table II concentrations (MPCs). These referenced concentrations are specific values which relate to an annual dose of 500 mrem. It is further indicated in 10CFR50.36a 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 10CFR50, Appendix I.

As stated in the Introduction to Appendix B of the new 10CFR20, the 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 10CFR50, 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/McGuire/Oconee has demonstrated that the use of the concentration values associated with the old 10CFR20.106 as SLC limits has resulted in calculated maximum individual doses to members of the public that are small percentages of the limits of 10CFR50, Appendix I. Therefore, the use of concentration values which correspond to an annual dose of 500 mrem should not have a negative impact on the ability to continue to operate within the limits of 10CFR50 Appendix I and 40CFR1 90.

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 10CFR20, 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.

BASES (continued)

McGuire Units 1 and 2 16.11.1-6 Revision 112

Liquid Effluents - Concentration 16.11.1 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 10CFR20, Appendix B, Table 2, Column 2 to apply at all times. The multiplier of ten is proposed because the annual dose of 500 mrem, upon which the concentrations in the old 10CFR20, Appendix B, Table II, Column 2 are based, is a factor of ten higher than the annual dose of 50 mrem, upon which the concentrations in the new 10CFR20, Appendix B, Table 2, Column 2, are based. Compliance with the limits of the new 10CFR20.1301 will be demonstrated by operating within the limits of 10CFR50, Appendix I and 40CFR190.

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," 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).

REFERENCES

1. McGuire Nuclear Station Offsite Dose Calculation Manual (ODCM)

2. International Commission on Radiological Protection (ICRP) Publication 2 McGuire Units 1 and 2 16.11.1-7 Revision 112

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.2 Radioactive Liquid Effluent Monitoring Instrumentation COMMITMENT The radioactive liquid effluent monitoring instrumentation channels shown in Table 16.11.2-1 shall be OPERABLE with their Alarm/Trip Setpoints set to ensure that the limits of SLC 16.11.1 are not exceeded.

AND The Alarm/Trip Setpoints of these channels shall be determined and adjusted in accordance with the methodology and parameters in the OFFSITE DOSE CALCULATION MANUAL (ODCM).

APPLICABILITY As shown in Table 16.11.2-1.

REMEDIAL ACTIONS

NOTE----

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more radioactive A.1 Suspend the release of Immediately liquid effluent monitoring radioactive liquid effluents channels Alarm/Trip monitored by the affected setpoint less channel.

conservative than required. OR A.2 Declare the channel Immediately inoperable.

OR A.3 Adjust setpoint to within Immediately limit.

B. One or more radioactive B.1 Enter the Remedial Action Immediately liquid effluent monitoring specified in Table 16.11.2-instrument channels 1 for the channel(s).

inoperable.

(continued)

McGuire Units 1 and 2 16.11.2-1 Revision 84

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One channel inoperable. I C.1.1 Analyze two independent Prior to initiating a samples per TR 16.11.1.1. release AND C.1.2 Perform independent Prior to initiating a verification of the discharge release line valving.

AND C. 1.3.1 Perform independent Prior to initiating a verification of manual release portion of the computer input for the release rate calculations performed by computer.

OR C.1.3.2 Perform independent Prior to initiating a verification of entire release release rate calculations for calculations performed manually.

AND C.1.4 Restore channel to 14 days OPERABLE status.

OR C.2 Suspend the release of Immediately radioactive effluents via this pathway.

(continued)

McGuire Units 1 and 2 16.11.2-2 Revision 84

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 REMEDIAL ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME D. One or more channels D.1 Obtain grab samples from Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable, the effluent pathway. during releases.

AND D.2 Perform an analysis of To meet LLD grab samples for requirements per radioactivity. Table 16.11.1-1.

AND D.3 Restore the channel to 30 days OPERABLE status.

E. One or more flow rate E. 1 --------- NOTE--------

measurement channels Pump performance curves inoperable, generated in place may be used to estimate flow.

Estimate the flow rate of Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the release. during releases AND E.2 Restore the channel to 30 days OPERABLE status.

F. RC minimum flow F. 1 Verify that the numberof Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months interlock inoperable, pumps providing dilution is during releases greater than or equal to the number of pumps required.

AND F.2 Restore the channel to 30 days OPERABLE status.

(continued)

McGuire Units 1 and 2 16.11.2-3 Revision 84

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 Explain why the In the next associated Completion inoperability was not scheduled Annual Time of Condition C, D, corrected within the Radioactive Effluent E or F not met. specified Completion Time Release Report in the Annual Radioactive Effluent Release Report.

McGuire Units 1 and 2 16.11.2-4 Revision 84

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2, TESTING REQUIREMENTS

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

Refer to Table 16.11.2-1 to determine which TRs apply for each Radioactive Liquid Effluent Monitoring channel.

TEST FREQUENCY TR 16.11.2.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TR 16.11.2.2 ------------------ NOTE -------------------

The CHANNEL CHECK shall consist of verifying indication of flow.

Perform CHANNEL CHECK. Every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during periods of release TR 16.11.2.3 Perform SOURCE CHECK. Prior to each release TR 16.11.2.4 Perform SOURCE CHECK. 31 days TR 16.11.2.5 ------------------- NOTES -----------------

1. For Instrument 1, the COT shall also demonstrate that automatic isolation of the pathway occurs if the instrument indicates measured levels above the Alarm/Trip Setpoint.

2. For Instruments 1 and 2, the COT shall also demonstrate that control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm/Trip Setpoint; circuit failure and, a downscale failure.

Perform CHANNEL OPERATIONAL TEST. 92 days TR 16.11.2.6 Perform a CHANNEL CALIBRATION. 18 months (continued)

McGuire Units 1 and 2 16.11.2-5 Revision 84

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.11.2.7 --------------------

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

The initial CHANNEL CALIBRATION shall be performed using standards certified by the National Institute of Standards and Technology (NIST) or using standards obtained from suppliers that participate in measurement assurance activities with NIST. 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.

Perform a CHANNEL CALIBRATION 24 months McGuire Units 1 and 2 16.11.2-6 Revision 84

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2 TABLE 16.11.2-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION INSTRUMENT MINIMUM REMEDIAL APPLICABILITY TESTING CHANNELS ACTION REQUIREMENTS OPERABLE

1. Radioactivity Monitors Providing Alarm And Automatic Termination of Release

a. Waste Liquid Effluent Line (EMF-49) 1 per station A, C, G During liquid TR 16.11.2.1 effluent releases TR 16.11.2.3 TR 16.11.2.5 TR 16.11.2.7 b- EMF-49 Minimum Flow Device 1 per station C, G During liquid TR 16.11.2.5 (2) effluent releases TR 16.11.2.7

c. Containment Ventilation Unit Condensate 1 A, D, G At all times TR 16.11.2.1 Line (EMF-44) TR 16.11.2.4 TR 16.11.2.5 TR 16.11.2.7

d. EMF-44 Minimum Flow Device 1 D, G At all times TR 16.11.2.5 (2) TR 16.11.2.7

2. Radioactivity Monitors Providing Alarm But Not Automatic Termination of Release

a. Conventional Waste Water Treatment 1 A, D, G At all times TR 16.11.2.1 Line or Turbine Building Sump to RC (EMF-31) TR 16.11.2.4 TR 16.11.2.5 TR 16.11.2.7

b. EMF-31 Minimum Flow Device 1 D, G At all times TR 16.11.2.5 (2) TR 16.11.2.7

3. Continuous Composite Samplers

a. Containment Ventilation Unit Condensate 1 D, G At all times TR 16.11.2.2 Line TR 16.11.2.5 TR 16.11.2.6

b. Conventional Waste Water Treatment Line 1 per station D, G At all times TR 16.11.2.2 TR 16.11.2.5 TR 16.11.2.6

c. Turbine Building Sump to RC D, G At all times TR 16.11.2.2 TR 16.11.2.6 (Continued)

McGuire Units 1 and 2 16.11.2-7 Revision 84

Radioactive Liquid Effluent Monitoring Instrumentation 16.11.2

4. Flow Rate Measurement Devices

a. Waste Liquid Effluent Line 1 per station E, G During liquid TR 16.11.2.2 effluent releases TR 16.11.2.5 TR 16.11.2.6

b. Containment Ventilation Unit Condensate 1 E, G At all times TR 16.11.2.2 Line TR 16.11.2.5 TR 16.11.2.6

c. Conventional Waste Water Treatment Line 1 per station E, G At all times TR 16.11.2.2 TR 16.11.2.5 TR 16.11.2.6

d. Turbine Building Sump to RC 1 E, G At all times TR 16.11.2.2 TR 16.11.2.6

5. RC Minimum Flow Interlock (1) 1 per station F, G At all times TR 16.11.2.5 NOTES:

1. Minimum flow dilution is assured by an interlock which terminates waste liquid release if the number of RC pumps running falls below the number of pumps required for dilution. The required number of RCpumps for dilution is determined per station procedures.

2. Radioactivity Monitor (EMF) shall not be declared operable unless both the EMF and the associated EMF's Minimum Flow Device are rendered operable.

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 potential releases of liquid effluents. The minimum flow devices for EMFs listed in Table 16.11.2-1 are required to provide assurance of representative sampling during actual or potential releases of liquid effluents. An interlock between the EMF's minimum flow device and its associated flow rate measurement device disables the remove alarm during non-release timeframes for the purpose of the control room black board annunciator criteria that disable expected alarms. An EMF flow rate measurement device measures total flow of the effluent while the EMF minimum flow device measures the sample flow rate through the EMF. The Alarm/Trip Setpoints of 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 stated in SLC 16.11.1. The OPERABILITY and use of this instrumentation is consistent with the requirements of General Design Criteria 60, 63, and 64 of Appendix A to 10 CFR Part 50. The Turbine Building Sump to RC Discharge Flow Measurement and Sampler Devices are for monitoring only and do not alarm or have any controls that require a COT.

REFERENCES

1. McGuire Nuclear Station Offsite Dose Calculation Manual (ODCM)

2. 10 CFR Part 50, Appendix A McGuire Units 1 and 2 16.11.2-8 Revision 84

Dose - Liquid Effluents 16.11.3 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.3 Dose - Liquid Effluents 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 Figurel6.11.1-1) shall be limited:

a. During any calendar quarter, to < 1.5 mrem to the total body and to < 5 mrem to any organ, and

b. During any calendar year, to < 3 mrem to the total body and to

< 10 mrem to any organ.

APPLICABILITY At all times.

REMEDIAL ACTIONS

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

Enter applicable Conditions and Required Actions of SLC 16.11.12, "Total Dose," when the limits of this SLC are exceeded by twice the specified limit.

CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated dose from ---------------- NOTE-------

release of radioactive The Special Report shall include materials in liquid the results of radiological analyses effluents exceeding of the drinking water source, and above limits, the radiological impact on finished drinking water supplies with regard to the requirements of 40 CFR 141, Safe Drinking Water Act, as applicable.

A. 1 Prepare and submit a 30 days Special Report to the NRC which identifies the causes for exceeding the limits, corrective actions taken to reduce releases, and actions taken to ensure that subsequent releases are within limits.

McGuire Units 1 and 2 16.11.3-1 .Revision 0

Dose - Liquid Effluents 16.11.3 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.3.1 Determine cumulative dose contributions from liquid 31 days effluents for current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM.

BASES This commitment is provided to implement the requirements of Sections II.A, III.A and IV.A of Appendix 1,10 CFR Part 50. The commitment implements the guides set forth in Section II.A of Appendix I. 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 I to assure that the releases of radioactive material in liquid effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable." Also, for fresh water sites with 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 concentrations in the finished drinking water that are in excess of the requirements of 40 CFR Part 141. These requirements are applicable only if the drinking water supply is~taken from the river 3 miles downstream of the plant discharge.

The dose calculation methodology and parameters in the ODCM implement the requirements in Section III.A of Appendix I 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 equations specified in the ODCM for calculating the doses due to the actual release rates of radioactive materials in liquid 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 the Purpose of Evaluating Compliance with 10 CFR Part 50, Appendix I," Revision 1, October 1977 and Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluent's from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

This commitment applies to the release of liquid effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system in accordance with the guidance given in NUREG-0133, Chapter 3.1.

McGuire Units 1 and 2 16.11.3-2 Revision 0

Dose - Liquid Effluents 16.

11.3 REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 40 CFR Part 141, Safe Drinking Water Act

3. 10 CFR Part 50, Appendix I

4. 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.

5. Regulatory Guide 1.113, "Estimating Aquatic Dispersion of Effluents from Accidental and Routine Reactor Releases for the Purpose of Implementing Appendix I," April 1977.

McGuire Units 1 and 2 16.11.3-3 Revision 0

Liquid Radwaste Treatment System 16.11.4 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.4 Liquid Radwaste Treatment System 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 projected doses due to the liquid effluent from each unit to UNRESTRICTED AREAS (see Figurel 6.11.1-1) would exceed 0.06 mrem to the total body or 0.2 mrem to any organ in a 31 day period.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Radioactive liquid waste A. 1 Prepare and submit a 30 days being discharged without Special Report to the NRC treatment and in excess which identifies the reasons of above limits, liquid radwaste was discharged without AND treatment, identification of inoperable equipment and Any portion of Liquid reasons for inoperability, Radwaste Treatment corrective actions taken to System not in operation. restore the equipment to OPERABLE status, and actions taken to prevent recurrence.

McGuire Units 1 and 2 16.11.4-1 Revision 0

Liquid Radwaste Treatment System 16.11.4 TESTING REQUIREMENTS Ik Iv#"r-r

II I--.........................------------------...

The Liquid Radwaste Treatment System shall be demonstrated OPERABLE by meeting SLC 16.11.1 and16.11.3.

TEST FREQUENCY TR 16.11.4.1 Project liquid release doses from each unit to 31 days UNRESTRICTED AREAS, in accordance with the methodology and parameters in the ODCM, when water systems are being released without being processed by its radwaste treatment system.

BASES 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 specification 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 II.D of Appendix I to 10 CFR Part 50. The specified limits governing 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 II.A of Appendix 1, 10 CFR Part 50, for liquid effluents.

This commitment applies to the release of liquid effluents from each reactor at the site. For units with shared Radwaste Treatment Systems, the liquid effluents from the shared system are to be proportioned among the units sharing that system in accordance with the guidance given in NUREG-0133, Chapter 3.1.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50

3. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.4-2 Revision 0

Chemical Treatment Ponds 16.11.5 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.5 Chemical Treatment Ponds COMMITMENT' The quantity of radioactive material contained in each chemical treatment pond shall be limited by the following expression (excluding tritium and dissolved or entrained noble gases):

264 _Z A1 <1.0 V . (C, x 10)

Where:

A j = pond inventory limit for single radionuclide "j", in Curies Cj = 10 CFR 20, Appendix B, Table 2, Column 2, concentration for single radionuclide "j", microCuries/ml; 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 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Quantity of radioactive A.1 Suspend all additions of Immediately material in any of the radioactive material to the chemical treatment pond.

ponds exceeding above limit. AND A.2 Initiate corrective action to Immediately reduce the pond contents to within limits.

McGuire Units 1 and 2 16.11.5-1 Revision 0

Chemical Treatment Ponds 16.11.5 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.5.1 Verify quantity of radioactive material in each batch of Prior to each slurry (powdex resin) to be transferred to chemical transfer treatment ponds is within limits by analyzing a representative sample of the slurry. Each batch to be transferred to the chemical treatment ponds is limited by:

SQ. < 6.0 x 1O5 pCi/gm S(Cj x 10) 10Ci/ml 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 SLC 16.11.5 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 4 to Appendix B of 10 CFR Part 20 applies.

The batch limits of slurry to the chemical treatment ponds assure that radioactive material in the slurry transferred to the CTP are "as low as is reasonably achievable" in accordance with 10 CFR Part 50.36a. The expression in SLC 16.11.5 assures no batch of slurry 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 I, Section ILA, total body dose level to the instantaneous whole body dose rate limitation, or that:

Y,c 3 mremn/ yr

< / = 0.006 j (C 1 x 10) 500 mrem / yr Where:

C. = Radioactive 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.

McGuire Units 1 and 2 16.11.5-2 Revision 0

Chemical Treatment Ponds 16.11.5 BASES (continued)

For the design of filter/demineralizers using powder resin, the slurry wash volume and the weight of resin used per batch is fixed by the cell surface area, and the slurry volume to resin weight ratio is constant at 100 ml/gram of wet, drained resin with a moisture content of approximately 55 to 60% (bulk density of about 58 pounds per cubic feet). Therefore, Z ,- < 0.006, and j (C, x 10) j (C X 10) (102 mI/gm) (106 pCi/,uCi)

(C Qj <6.0x 1 0 5 pCi/gm (C x 10) 10Ci/m/

Where:

Q, = concentration of radioactive materials in wet, drained slurry (powdex resin) for 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 picoCuries/gram. Estimates of the Sr-89 and Sr-90 batch concentration shall be included based on the most recent monthly composite analysis (within 3 months); and C = 10 CFR 20, Appendix B, Table 2, Column 2, concentration for single radionuclide "j", in microCuries/milliliter.

The batch limits provide assurance that activity input to the chemical treatment ponds will be minimized, and a means of identifying radioactive material in the inventory limitation of SLC 16.11.5.

The basic requirements for the Selected Licensee Commitments concerning effluents from nuclear power reactors are stated in 10CFR50.36a. 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 10CFR20.106 (new 10CFR20.1301). 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 10CFR20.106 which references Appendix B, Table II concentrations- (MPCs). These referenced concentrations are specific values which relate to an annual dose of 500 mrem. It is further indicated in 10CFR50.36a 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 10CFR50, Appendix I.

McGuire Units 1 and 2 16.11.5-3 Revision 0

Chemical Treatment Ponds 16.11.5 BASES (continued)

As stated in the Introduction to Appendix B of the new 10CFR20, the 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 10CFR50, 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/McGuire/Oconee has demonstrated that the use of the concentration values associated with the old 10CFR20.106 as SLC limits has resulted in calculated maximum individual doses to members of the public that are small percentages of the limits of 10CFR50, Appendix I. Therefore, the use of concentration values which correspond to an annual dose of 500 mrem should not have a negative impact on the ability to continue to operate within the limits of 10CFR50, Appendix I and 40CFR190.

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 10CFR20, 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 10CFR20, Appendix B, Table 2, Column 2 to apply at all times. The multiplier of ten is proposed because the annual dose of 500 mrem, upon which the concentrations in the old 10CFR2O, Appendix B, Table II, Column 2 are based, is a factor of ten higher than the annual dose of 50 mrem, upon which the concentrations in the new 10CFR20, Appendix B, Table 2, Column 2, are based. Compliance with the limits of the new 10CFR20.1301 will be demonstrated by operating within the limits of 10CFR50, Appendix I and 40CFR190.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR 20, Appendix B

3. 10 CFR 50, Appendix I, Section IL.A

4. 10CFR20

5. 10CFR50.36a McGuire Units 1 and 2 16.11.5-4 Revision 0

Dose Rate - Gaseous Effluents 16.11.6 16.11 RADIOLOGICAL EFFLUENT CONTROL 16.11.6 Dose Rate - Gaseous Effluents COMMITMENT ,The dose rate due to radioactive materials released in gaseous effluents from the site to areas at and beyond the SITE BOUNDARY (see Figurel6.11.1-1) shall be limited to the following:

a. For noble gases: < 500 mrem/yr to the whole body and < 3000 mrem/yr to the skin, and

b. For Iodine- 131 and 133, for tritium, and for all radioactive materials in particulate form with half-lives greater than 8 days:

< 1500 mrem/yr to any organ.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Dose rate not within A.1 Restore the release rate to Immediately limit, within limits.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.6.1 Verify dose rates due to noble gases in gaseous effluents In accordance with are within limits in accordance with the methodology and the ODCM parameters in the ODCM.

TR 16.11.6.2 Verify dose rates due to radioactive materials, other than In accordance with noble gases, in gaseous effluents are within limits in Table 16.11.6-1 accordance with the methodology and parameters in the ODCM by obtaining representative samples and performing analyses in accordance with Table16.11.6-1.

McGuire Units 1 and 2 16.11.6-1 Revision 112

Dose Rate - Gaseous Effluents 16.11.6 TABLE 16.11.6-1 (Page 1 of 4)

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Gaseous Release Type Sampling Frequency Minimum Analysis Type of Activity Analysis Lower Limit of Frequency Detection (LLD)(')

(PCi/ml) 1.Waste Gas Storage Tanks P P Each Tank Each Tank Principal Gas Gamma Emitters(6) 1x10 4 Grab Sample

2. Containment Purge P P Each PURGE Each PURGE Principal Gas Gamma Emitters(6 ) 1x104 Grab Sample M H-3 lx10i0 4

3. Unit Vent W~z' W Principal Gas Gamma Emittersb) 1x10-Grab Sample H-3 1x10-5 4

4.a. Radwaste Facility Vent W W Principal Gas Gamma Emittersýb) 1x10

b. Waste Handling Building Grab Sample 1_X1_0-b

c. Equipment Staging Building H-3 1x10b

5. Unit Vents Continuoust W(13 1-131 lx10-12 Charcoal Sample 1-133 lx10i l" Continuous/ 1 W/b Principal Gamma Emitters(") 1xl0 1 1 Particulate (1-131, Others)

Sample Continuous*b) M Gross Alpha') lxl0-11 Composite Particulate Sample Continuous1 ý' Q Sr-89, Sr-90 lxl0-11 Composite Particulate Sample McGuire Units 1 and 2 16.11.6-2 Revision 112

Dose Rate - Gaseous Effluents 16.11.6 TABLE 16.11.6-1 (Page 2 of 4)

RADIOACTIVE GASEOUS WASTE SAMPLING AND ANALYSIS PROGRAM Gaseous Release Type Sampling Frequency Minimum Analysis Type of Activity Analysis Lower Limit of Frequency Detection (LLD)(1 )

(PCi/ml)

6. All Release Types as listed in 4 above. Continuousto) W{b 1-131 1x1012 Charcoal Sample 1-133 lx10"1 u ContinuousI' WId Principal Gamma Emitters") lxi 01 Particulate (1-131, Others)

Sample Continuous (b M Gross Alpha`" lxl 01l Composite Particulate Sample Continuous (b Q Sr-89, Sr-90 1xl0 11 Composite Particulate Sample McGuire Units 1 and 2 16.11.6-3 Revision 112

Dose Rate - Gaseous Effluents 16.11.6 TABLE 16.11.6-1 (Page 3 of 4)

NOTES:

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 = 4.66sh, E. V *2.22 x 106 .Y * "exp (-,AAt)

Where:

LLD = the "a priori" lower limit of detection as defined above (as microCurie per unit mass or volume);

sb = the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate. (as counts per minute);

E = the counting efficiency (as counts per disintegration);

V = the sample size (in units of mass or volume);

2.22 x106 = the number of disintegrations per minute per microCurie; Y = the fractional radiochemical yield (when applicable);

k = the radioactive decay constant for the particular radionuclide; and At = the elapsed time between midpoint of sample collection and time of counting (for plant effluents, not environmental samples).

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.

McGuire Units 1 and 2 16.11.6-4 Revision 112

Dose Rate - Gaseous Effluents 16.11.6 TABLE 16.11.6-1 (Page 4 of 4)

NOTES:

2. 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 refueling canal is flooded.

3. Not used.

4. Not used.

5. 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.

6. The principal gamma emitters for which the LLD specification applies include the following radionuclides: Kr-87, Kr-88, Xe-1 33, Xe-133m, Xe135, and Xe-1 38 in noble gas releases and Mn-54, Fe-59, Co-58, Co-60, Zn-65, Mo-99, 1-131, Cs-134, Cs-137, and Ce-141 in iodine and particulate releases. The LLD for Ce-144 is 5x10-9 microCi/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 reported in the Annual Radioactive Effluent Release Report.

7. The composite filter(s) will be analyzed for alpha activity by analyzing the filter media used during the collection period.

8. Samples shall be changed at least once per 7 days and analyses shall be completed to meet LLD after changing, or after removal from sampler. If the particulate and charcoal sample frequency is changed to a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months frequency the corresponding LLDs may be increased by a factor of 10 (i.e., LLD for 1-131 from 1 x 10-12 to 1 x 101 microCi/ml).

McGuire Units 1 and 2 16.11.6-5 Revision 112

Dose Rate- Gaseous Effluents 16.11.6 BASES Specific 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, and 3000 mrem/year to the skin from noble gases, and 1500 mrem/year to any organ from Iodine 131, Iodine 133, tritium, and all radionuclides in particulate form with half-lives greater than eight days. This commitment applies to the release of gaseous effluents from all reactors, at the site. The Exclusion Area Boundary (Site Boundary) is set as the boundary for gaseous effluent release limits. The Exclusion Area Boundary (EAB) is formed by a 2500 ft radius centered on the Reactor Buildings' centerlines as shown on Figure 16.11 .1-1.

The basic requirements for the Selected Licensee Commitments concerning effluents from nuclear power reactors are stated in 10CFR50.36a. 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 10CFR20.106 (new 10CFR20.1301). 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 10CFR20.106 which references Appendix B, Table II concentrations (MPCs). These referenced concentrations are specific values which relate to an annual dose of 500 mrem. It is further indicated in 10CFR50.36a 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 10CFR50, Appendix I.

As stated in the Introduction to Appendix B of the new 10CFR20, the effluent concentration (EC) limits given in Appendix B, Table 2, Column 1, are based on an annual dose of 50 mrem for isotopes for which inhalation or ingestion is limiting or 100 mrem for isotopes for which submersion (noble gases) is limiting. Since release concentrations corresponding to limiting dose rates of less than or equal to 500 mrem/year to the whole body, 3000 mrem/year to the skin from noble gases, and 1500 mrem/year to any organ from Iodine 131, Iodine 133, tritium and for 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 effluents to assure that the limits of 10CFR50, Appendix I and 40CFR190 are not likely to be exceeded, it should not be necessary to restrict the operational flexibility by incorporating the EC value for isotopes based on ingestion/inhalation (50 mrem/year) or for isotopes with the EC based on submersion (100 mrem/year).

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 10CFR20, Appendix B, Table 2, Column 1, relate to a dose of 50 or 100 mrem in a year. When applied on an instantaneous basis, this corresponds to a dose rate of either 50 or 100 mrem/year. These low values are impractical upon which to base effluent monitor setpoint calculations for many 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 16.11.6 will be maintained at the current dose rate limit for noble gases of 500 mrem/year to the whole body and 3000 mrem/year to the skin, for Iodine 131, Iodine 133, tritium and all radionuclides in particulate form with half-lives greater than eight days an instantaneous dose rate limit of 1500 mrem/year to any organ.

McGuire Units 1 and 2 16.11.6-6 Revision 112

Dose Rate - Gaseous Effluents 16.11.6 BASES (continued)

Compliance with the limits of the new 1 0CFR20.1301 will be demonstrated by operating within the limits of 10CFR50, Appendix I and 40CFR190. Operational history at Catawba/McGuire/Oconee has demonstrated that the use of the dose rate values listed above (i.e. 500 mrem/year, 3000 mrem/year and 1500 mrem/year) as SLC limits has resulted in calculated maximum individual doses to members of the public that are small percentages of the limits of 10CFR50, Appendix I and 40CFR190.

The required detection capabilities for radioactive materials 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 (1968), and Hartwell, J. K.

"Detection Limits for Radioanalytical Counting Techniques," Atlantic Richfield Hanford Company Report ARH-SA-215 (June 1975).

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 20, Appendix B

3. 10 CFR Part 20

4. 10CFRPart50 McGuire Units 1 and 2 16.11.6-7 Revision 112

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.7 Radioactive Gaseous Effluent Monitoring Instrumentation COMMITMENT The radioactive gaseous effluent monitoring instrumentation channels shown in Table 16.11.7-1 shall be OPERABLE with Alarm/Trip Setpoints set to ensure that the limits of SLC 16.11.6 are not exceeded.

AND The Alarm/Trip setpoints shall be determined and adjusted in accordance with the methodology and parameters in the ODCM.

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

Brief periods of routine sampling (not to exceed 15 minutes) do not make the instrumentation inoperable.

APPLICABILITY As shown in Table 16.11.7-1.

REMEDIAL ACTIONS

NOTE----

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more radioactive A.1 Suspend the release of Immediately gaseous effluent radioactive gaseous monitoring channels effluents monitored by the Alarm/Trip setpoint less affected channel.

conservative than required. OR A.2 Declare the channel Immediately inoperable.

OR A.3 Adjust setpoint to within Immediately.

limit.

(continued)

McGuire Units 1 and 2 16.11.7-1 Revision 84

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more radioactive B.1 Enter the Remedial Action Immediately gaseous effluent specified in Table 16.11.7-1 monitoring instrument for the channel(s).

channels inoperable.

C. One channel C.1.1 Analyze two independent Prior to initiating a inoperable.

samples of the tank release contents.

AND C.1.2 Perform independent Prior to initiating a verification of the discharge release valve lineup.

AND C.1.3.1 Perform independent Prior to initiating a verification of manual release portion of the computer input for the release rate calculations performed by computer.

OR C.1.3.2 Perform independent Prior to initiating a verification of entire release release rate calculations for calculations performed manually.

AND C.1.4 Restore channel to 14 days OPERABLE status.

OR C.2 Suspend the release of Immediately radioactive effluents via this pathway.

(continued)

McGuire Units 1 and 2 16.11.7-2 Revision 84

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. One or more flow;rate D.1 Estimate the flow rate of Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months measurement channels the release. during releases inoperable.

AND D.2 Restore the channel to 30 days OPERABLE status.

E. One or more noble gas E.1 Obtain grab samples from Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months activity monitor channels the effluent pathway. during releases inoperable.

AND E.2 Perform an analysis of grab To meet LLD samples for radioactivity, requirements per Table 16.11.6-1 AND E.3 Restore the channel to 30 days OPERABLE status.

F. Noble gas activity F.1 Suspend PURGING or Immediately monitor providing VENTING of radioactive automatic termination of effluents via this pathway.

release inoperable.

G. One or more sampler G.1 Perform sampling with Continuously channels inoperable, auxiliary sampling equipment as required by Table 16.11.6-1.

AND G.2 Restore the channel to 30 days OPERABLE status.

(continued)

McGuire Units 1 and 2 16.11.7-3 Revision 84

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 REMEDIAL ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. One or more Sampler H. 1 Verify flow through the Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Minimum Flow Device sampling apparatus. during releases Channels inoperable.

AND H.2 Restore the channel to 30 days OPERABLE status.

1. Required Action and 1.1 Explain why the In the next associated Completion inoperability was not scheduled Annual Time of Condition C, D, corrected within the Radioactive Effluent E, F, G, or H not met. specified Completion Time Release Report in the Annual Radioactive Effluent Release Report.

TESTING REQUIREMENTS

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

Refer to Table 16.11.7-1 to determine which TRs apply for each Radioactive Gaseous Effluent Monitoring channel.

TEST FREQUENCY TR 16.11.7.1 Perform CHANNEL CHECK. Prior to each release TR 16.11.7.2 ------------------ NOTE --------------------- Prior to each The SOURCE CHECK for these channels shall be the release qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity or a simulated source of radioactivity such as a light emitting diode.

Perform SOURCE CHECK.

TR 16.11.7.3 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TR 16.11.7.4 Perform CHANNEL CHECK. 7 days (continued)

McGuire Units 1 and.2 16.11.7-4 Revision 84

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 TESTING REQUIREMENTS (continued)

TEST FREQUENCY TR 16.11.7.5 ------------------ NOTE -------------------

The SOURCE CHECK for these channels shall be the qualitative assessment of channel response when the channel sensor is exposed to a source of increased radioactivity or a simulated source of radioactivity such as a light emitting diode.

Perform SOURCE CHECK. 31 days TR 16.11.7.6 ------------------- NOTES-----------------

1. For noble gas activity monitors providing automatic termination of release, the COT shall also demonstrate that automatic isolation of the pathway occurs if the instrument indicates measured levels above the Alarm/Trip Setpoint.

2. For all noble gas activity monitors, the COT shall also demonstrate that control room alarm annunciation occurs if the instrument indicates measured levels above the Alarm/Trip Setpoint; circuit failure and, a downscale failure.

Perform CHANNEL OPERATIONAL TEST. 92 days TR 16.11.7.7 ---------------- NOTE --------------------

For all noble gas activity monitors, the initial CHANNEL CALIBRATION shall be performed using standards certified by the National Institute of Standards and Technology (NIST) or using standards obtained from suppliers that participate in measurement assurance activities with NIST. 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.

18 months Perform a CHANNEL CALIBRATION.

McGuire Units 1 and 2 16.11.7-5 Revision 84

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 TABLE 16.11.7-1 (Page 1 of 3)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENTS MINIMUM REMEDIAL APPLICABILITY TESTING CHANNELS ACTION REQUIREMENTS OPERABLE

1. WASTE GAS HOLDUP SYSTEM

a. Noble Gas Activity Monitor - 1 per station A, C, I During gas effluent TR 16.11.7.1 Providing Alarm and Automatic releases. TR 16.11.7.2 Termination of Release (Low Range- TR 16.11.7.6 EMF-50 or 1EMF-36, low-range) TR 16.11.7.7

b. Effluent System Flow Rate 1 per station D, I At all times except TR 16.11.7.3 Measuring Device when isolation TR 16.11.7.6 valve is closed & TR 16.11.7.7 locked.

2. Condenser Evacuation System - Noble 1 A, E, I When air ejectors TR 16.11.7.3 Gas Activity Monitor (EMF-33) are operable. TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

3. Vent System

a. Noble Gas Activity Monitor (Low 1 A, E, I At all times. TR 16.11.7.3 Range - EMF-36) TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

b. Iodine Sampler 1 G, I At all times, except TR 16.11.7.4 during routine sampling.

c. Particulate Sampler (EMF-35) 1 G, I At all times, except TR 16.11.7.4 during routine sampling.

d. Unit Vent Flow Rate Monitor 1 D, I At all.times. TR 16.11.7.3 (Totalizer) TR 16.11.7.6 TR 16.11.7.7

e. Iodine Sampler Minimum Flow 1 H,I At all times, except TR 16.11.7.3 Device during routine TR 16.11.7.6 sampling. TR 16.11.7.7

f. Particulate Sampler Minimum Flow 1 G,I At all times, except TR 16.11.7.3 Device (1) during routine TR 16.11.7.6 sampling. TR 16.11.7.7

4. Containment Purge System - Noble Gas A, F, I Modes 1 through 6, TR 16.11.7.2 Activity Monitor - Providing Alarm and except when TR 16.11.7.3 Automatic Termination of Release (Low isolation valve is TR 16.11.7.6 Range - EMF-39) closed & locked. TR 16.11.7.7 (continued)

McGuire Units 1 and 2 16.11.7-6 Revision 84

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 TABLE 16.11.7-1 (Page 2 of 3)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENTS MINIMUM REMEDIAL APPLICABILITY TESTING CHANNELS ACTION REQUIREMENTS OPERABLE

5. Auxiliary Building Ventilation System - 1 A, E, I At all times. TR 16.11.7.3 Noble Gas Activity Monitor (EMF-41 or TR 16.11.7.5 EMF-36) TR 16.11.7.6 TR 16.11.7.7

6. Fuel Storage Area Ventilation System - 1 A, E, I At all times. TR 16.11.7.3 Noble Gas Activity Monitor (EMF-42 or TR 16.11.7.5 EMF-36) TR 16.11.7.6 TR 16.11.7.7

7. Contaminated Parts Warehouse Ventilation System

a. Noble Gas Activity Monitor (EMF-53) 1 per station A, E, I During gaseous TR 16.11.7.3 effluent releases. TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

b. Flow Rate Monitor 1 per station D, I During gaseous TR 16.11.7.3 effluent releases. TR 16.11.7.6 TR 16.11.7.7

c. EMF-53 Sampler Minimum Flow 1 per station H,I During gaseous TR 16.11.7.3 Device (1) effluent releases. TR 16.11.7.6 TR 16.11.7.7

8. Radwaste Facility Ventilation System

a. Noble Gas Activity Monitor (EMF-52) 1 per station A, E, I During gaseous TR 16.11.7.3 effluent releases. TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

b. Flow Rate Monitor 1 per station D, I During gaseous TR 16.11.7.3 effluent releases. TR 16.11.7.6 TR 16.11.7.7

c. EMF-52 Sampler Minimum Flow 1 per station H, I During gaseous TR 16.11.7.3 Device (1) effluent releases. TR 16.11.7.6 TR 16.11.7.7 (continued)

McGuire Units 1 and 2 16.11.7-7 Revision 84

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 TABLE 16.11.7-1 (Page 3 of 3)

RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION INSTRUMENTS MINIMUM REMEDIAL APPLICABILITY TESTING CHANNELS ACTION REQUIREMENTS OPERABLE

9. Equipment Staging Building Ventilation System

a. Noble Gas Activity Monitor (EMF-59) 1 per station A, E, I During gaseous TR 16.11.7.3 effluent releases. TR 16.11.7.5 TR 16.11.7.6 TR 16.11.7.7

b. Flow Rate Monitor 1 per station D, I During gaseous TR 16.11.7.3 effluent releases. TR 16.11.7.6 TR 16.11.7.7

c. EMF-59 Sampler Minimum Flow 1 per station H, I During gaseous TR 16.11.7.3 Device (1) effluent releases. TR 16.11.7.6 TR 16.11.7.7

10. Containment Air Release and Addition 1 A, E, I At all times except- TR 16.11.7.3 System - Noble Gas Activity Monitor when isolation TR 16.11.7.5 (EMF-39L or EMF-36L) valve is closed & TR 16.11.7.6 locked. TR 16.11.7.7 NOTES:

1. Radioactivity monitor (EMF) shall not be declared OPERABLE unless both the EMF and the associated EMF's Minimum Flow Device are rendered OPERABLE.

McGuire Units 1 and 2 16.11.7-8 Revision 84

Radioactive Gaseous Effluent Monitoring Instrumentation 16.11.7 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 instrumentation consists of monitoring and sampling instrumentation. Monitors provide continuous display of process parameters with appropriate alarms and trip setpoints established. Samplers collect a portion of the desired process for subsequent laboratory analysis, and do not have alarm/trip capability. Samplers and the analysis program provide a method to assure that long term effluent release quantities do not exceed the requirements of SLC 16.11.6.

Monitors provide assurance that instantaneous effluent releases do not exceed the requirements of SLC 16.11.6. The minimum flow devices for EMFs listed in Table 16.11.7-1 are required to provide assurance of representative sampling during actual or potential releases of gaseous effluents. The flow rate monitor quantifies the total gaseous effluent (both non-radioactive and radioactive) released to the environment.

During routine sampling, instrumentation may be turned off for short periods of time (not to exceed 15 minutes) in order to meet analysis requirements of SLC 16.11.6. This is considered to be a normal operable function of the equipment. 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 stated in SLC 16.11.6. 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.

REFERENCES

1. McGuire Nuclear Station, Offsite Dose Calculation Manual

2. 10 CFR Part 50, Appendix A McGuire Units 1 and 2 16.11.7-9 Revision 84

Noble Gases 16.11.8 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.8 Noble Gases COMMITMENT Air dose due to noble gases released in gaseous effluents, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure16.11.1-1) shall be limited to the following:

a. During any calendar quarter: Less than or equal to 5 mrad 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.

REMEDIAL ACTIONS

NO r-.1)-----------------------------------------------------

Enter applicable Conditions and Required Actions of SLC 16.11.12, "Total Dose," when the limits of this SLC are exceeded by twice the specified limit.

CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated air dose from A.1 Prepare and submit a 30 days radioactive noble gases Special Report to the NRC in gaseous effluents which identifies the causes exceeding any of above for exceeding the limits, limits, corrective actions taken to reduce releases, and actions taken to ensure that subsequent releases are within limits.

McGuire Units 1 and 2 16.11.8-1 Revision 0

Noble Gases 16.11.8 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.8.1 Determine cumulative dose contributions from noble 31 days gases in gaseous effluents for current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM.

BASES This commitment is provided to implement the requirements of Sections II.B, IIli.A and IV.A of Appendix 1,10 CFR Part 50. The Limiting Condition for Operation implements the guides set forth in Section II.B of Appendix I.

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 I to assure that the releases of radioactive material in gaseous effluents to UNRESTRICTED AREAS will be kept "as low as is reasonably achievable."

The TESTING REQUIREMENTS implement the requirements in Section III.A of Appendix I 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 under-estimated.

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 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.

The ODCM equations provided for determining the air doses at and beyond the SITE BOUNDARY are based upon the historical average atmospheric conditions.

This commitment applies at all times to the release of gaseous effluents from each reactor at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are to be proportioned among the units sharing that system in accordance with the guidance given in NUREG-01 33, Chapter 3.1.

McGuire Units 1 and 2 16.11.8-2 Revision 0

Noble Gases 16.

11.8 REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.8-3 Revision 0

Dose - Iodine-131 and 133, Tritium and Radioactive Materials in Particulate Form 16.11.9 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.9 Dose - Iodine-131 and 133, Tritium and Radioactive Materials in Particulate Form COMMITMENT The dose to a MEMBER OF THE PUBLIC from Iodine-131 and 133, tritium, and all radioactive materials 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 16.11.1-1) 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.

APPLICABILITY *At all times.

REMEDIAL ACTIONS

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

Enter applicable Conditions and Required Actions of SLC 16.11.12, "Total Dose," when the limits of this SLC are exceeded by twice the specified limit.

.CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated dose from A.1 Prepare and submit a 30 days the release of Iodine 131 Special Report to the NRC and 133, tritium, and which identifies the causes radioactive materials in for exceeding the limits, particulate form with corrective actions taken to half-lives greater than 8 reduce releases, and days in gaseous actions taken to ens~ure effluents exceeding any that subsequent releases of the above limits, are within limits.

McGuire Units 1 and 2 16.11.9-1 Revision 0

Dose - Iodine-131 and 133, Tritium and Radioactive Materials in Particulate Form 16.11.9 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.9.1 Determine cumulative dose contributions for Iodine 131 31 days and 133, tritium, and radioactive material in particulate form with half lives greater than 8 days in gaseous effluents for current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM.

BASES This commitment is provided to implement the requirements-of Sections- II.C, III.A and IV.A of Appendix 1,10 CFR Part 50. The Limiting Conditions for Operation are the guides set forth in Section II.C of Appendix I.

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 I 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 TESTING REQUIREMENTS implement the requirements in Section III.A of Appendix I 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 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 specifications for Iodine-1 31 and 133, tritium, and radionuclides in particulate form with half-lives 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 these calculations were: (1) individual inhalation of airborne 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.

McGuire Units 1 and 2 16.11.9-2 Revision 0

Dose - Iodine-131 and 133, Tritium and Radioactive Materials in Particulate Form 16.11.9 BASES (continued)

This commitment applies at all times to the release of gaseous effluents from each reactor at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are to be proportioned among the units sharing that system in accordance with the guidance given in NUREG 0133, Chapter 3.1.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.9-3 Revision 0

Gaseous Radwaste Treatment System 16.11.10 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.10 Gaseous Radwaste Treatment System COMMITMENT The VENTILATION EXHAUST TREATMENT and WASTE GAS HOLDUP SYSTEMS shall be OPERABLE and appropriate portions of these systems shall be used to reduce releases of radioactivity when the projected doses in 31 days due to gaseous effluent releases, from each unit, to areas at and beyond the SITE BOUNDARY (see Figure 16.11.1-1) would exceed:

a. 0.2 mrad to air from gamma radiation, or

b. 0.4 mrad to air from beta radiation, or

c. 0.3 mrem to any organ of a MEMBER OF THE PUBLIC.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Radioactive gases being A.1 Prepare and submit a 30 days discharged without Special Report to the NRC treatment and in excess which identifies inoperable of above limits, equipment and reasons for inoperability, actions taken to restore the equipment to OPERABLE status, and actions taken to prevent recurrence.

McGuire Units 1 and 2 16.11.10-1 Revision 0

Gaseous Radwaste Treatment System 16.11.10 TESTING REQUIREMENTS

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

The installed Gaseous Radwaste Treatment System shall be demonstrated OPERABLE by meeting SLC 16.11.6, 16.11.8 and16.11.9.

TEST FREQUENCY TR 16.11.10.1 Project gaseous release doses from each unit to areas 31 days at and beyond the SITE BOUNDARY, in accordance with the methodology and parameters in the ODCM, when gaseous systems are being released without being processed by its radwaste treatment system.

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 radioactive materials in gaseous effluents will be kept "as low as is reasonably achievable."

This commitment implements the requirements of 19 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and the design objectives given in Section II.D of Appendix I to 10 CFR Part 50. The specified limits governing the use of appropriate portions of the systems were specified as a suitable fraction of the dose design objectives set forth in Section II.B and II.C of Appendix 1,10 CFR Part 50, for gaseous effluents.

This commitment applies at all times to the release of radioactive materials in gaseous effluents from each unit at the site. For units with shared Radwaste Treatment Systems, the gaseous effluents from the shared system are to be proportioned among the units sharing that system in accordance with NUREG-0133, Chapter 3.1.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50, Appendix I

3. 10CFRPart50 McGuire Units 1 and 2 16.11.10-2 Revision 0

Solid Radioactive Waste 16.11.11 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.11 Solid Radioactive Waste COMMITMENT Radioactive wastes shall be processed and packaged to ensure compliance with the applicable requirements of 10 CFR Part 20, 10CFR Part 61, 10 CFR Part 71, and State regulations governing the transportation and disposal of radioactive wastes.

The Solid Radwaste System or an approved alternative process shall be used in accordance with a PROCESS CONTROL PROGRAM (PCP) for the solidification of liquid or wet radioactive wastes or the dewatering of wet radioactive wastes to be shipped for direct disposal at a 10CFR61 licensed disposal site. Wastes shipped for off site processing in accordance with the processor's specifications and transportation requirements are not required to be solidified or dewatered to meet disposal requirements.

The PCP describes administrative and operational controls used for the solidification of liquid or wet solid radioactive wastes in order to meet applicable 10CFR61 waste form requirements.

The PCP describes the administrative and operational controls used for the dewatering of wet radioactive wastes to meet 10CFR61 free standing water requirements. /

The process parameters used in establishing the PCP shall be based on demonstrated processing of actual or simulated liquid or wet solid wastes and must adequately verify that the final product of solidification or dewatering meets all applicable Federal, State and disposal site requirements.

APPLICABILITY At all times.

McGuire Units 1 and 2 16.11.11-1 Revision 41

Solid Radioactive Waste 16.11.11 REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Applicable regulatory A.1 Suspend shipments of Immediately requirements for defectively packaged solid solidified or dewatered radioactive wastes from the wastes are not site.

satisified.

AND A.2 Initiate action to correct the PROCESS CONTROL Prior to next PROGRAM, procedures, or shipment for disposal solid waste equipment as of solidified or necessary to prevent dewatered wastes.

recurrence.

B. A solidification test as B. 1 Suspend solidification of the Immediately described in the PCP batch under test and follow fails to verify PCP guidance for test Solidification. failures.

B.2 Once a subsequent test Prior to next verifies Solidification, solidification for solidification of the batch shipment of waste may then be resumed as for disposal at a directed by the PCP. The 10CFR61 disposal PCP shall be modified as site.

required to assure Solidification of subsequent batches of waste (continued)

McGuire Units 1 and 2 16.11.11-2 Revision 41

Solid Radioactive Waste 16.11.11 REMEDIAL ACTIONS (continued)

C. With solidification or C.1 Reprocess the waste in Prior to shipment for dewatering for accordance with PCP disposal of the disposal not requirements. inadequately processed performed in waste that requires accordance with the OR solidification of dewatering PROCESS CONTROL C.2 Follow PCP or procedure PROGRAM. quidance for alternative free standing liquid verification to ensure the waste in each container meets disposal requirements and take appropriate administrative action to prevent recurrence.

D. With the solid waste D.1 Restore the equipment In a time frame that equipment incapable to OPERABLE status or supports the of meeting SLC provide for alternative COMMITMENT section of 16.11.11 or not in capability to process SLC 16.11.11 service wastes as necessary to satisfy all applicable disposal requirements TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.11.1 The Process Control Program shall be used to verify the Every tenth batch Solidification of at least one representative test of each type of specimens from at least every tenth batch of each type radioactive waste of radioactive waste to be solidified for disposal at a to be solidified.

10CFR61 disposal site per the COMMITMENT of this SLC.

McGuire Units 1 and 2 16.11.11-3 Revision 41

Solid Radioactive Waste 16.11.11 BASES:

This commitment implements the requirements of 10 CFR Part 50.36a, General Design Criterion 60 of Appendix A to 10 CFR Part 50 and requirements to use a Process Control Program to meet applicable 10CFR61 waste form criteria for solidified and dewatered radioactive wastes.

REFERENCES:

1. 10CFR Part 50, "Domistic Licensing of Production and Utilization Facilities"

2. 10 CFR Part 50, Appendix A

3. 10CFR20, "Standards for Protection Against Radiation"

4. 10CFR61, "Licensing Requirements for Land Disposal of Radioactive Waste

5. 10CFR71, "Packaging and Transportation of Radioactive Materials"

6. DPCo Process Control Program Manual

7. NRC Generic Letter 84-12, "Compliance With 10 CFR Part 61 And Implementation Of the Radiological Effulent Technical Specifications (Rets) and Attendant Process Control Program (PCP)"

8. NRC Generic Letter 89-01, "Implementation of Programmatic Controls for Radiological Effulent Technical Specifications In the Administrative Controls Section of the Technical Specifications and the Relocation of Procedural Details of Rets to the Offsite Dose Calculation Manual or to the Process Control Program" McGuire Units 1 and 2 16.11.11-4 Revision 41

Total Dose 16.11.12 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 radiation, from uranium fuel cycle sources shall.be limited to < 25 mrem to the whole body or any organ, except the thyroid, which shall be limited to < 75 mrem.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Calculated doses from A.1 Verify, by calculation, the Immediately releases exceeding cumulative dose from direct twice the specified limits radiation contributions, the of SLC 16.11.3, 16.11.8 ISFSI, outside storage or 16.11.9. tanks, and radioactivity releases are within the total dose limit.

AND A.2 ---------- NOTE------------

Only required to be performed if the total dose limit is exceeded.

Prepare and submit a 30 days Special Report to the NRC which identifies corrective actions to be taken to reduce subsequent releases to prevent recurrence and schedule for achieving conformance with specified limits.

McGuire Units 1 and 2 16.11.12-1 Revision 67

Total Dose 16.11.12 TESTING REQUIREMENTS

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

Cumulative dose contributions from liquid and gaseous effluents shall be determined in accordance with SLC 16.11.3, 16.11.8 and16.11.9, and in accordance with the methodology and parameters specified in the ODCM.

TEST FREQUENCY TR 16.11.12.1 Determine cumulative dose contributions from direct When calculated radiation from the units, the ISFSI, and from radwaste doses from storage tanks in accordance with the methodology and effluent releases parameters specified in the ODCM. exceeds twice the limits of SLCs 16.11.3, 16.11.8 or 16.11.9 BASES This commitment is provided to meet the dose limitations of 40 CFR Part 190 that have been incorporated into 10 CFR Part 20 by 46 FR 18525. The specification requires the preparation and submittal of a Special Report whenever the calculated doses from plant generated radioactive effluents and direct radiation 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. 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 10 CFR Part 50, Appendix I, and if direct radiation doses from the units and outside storage tanks are kept small.

This Special Report, as defined in 10 CFR Part 20.2203(a)(4), 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 material involved, and the cause of the exposure levels or concentrations. 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 Part 190 limits. For the purposes of the Special Report, it may be assumed that the 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 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 corrected), in McGuire Units 1 and 2 16.11.12-2 Revision 67

Total Dose 16.11.12 BASES (continued) accordance with the provisions of 40 CFR Part 190.11 and 10 CFR Part 20.2203(a)(4), is considered to be a timely request and fulfills the requirements of 40 CFR Part 190 and a variance is granted until NRC staff action is completed. The variance only relates to the limits of 40 CFR Part 190, and does not apply in any way to the other requirements for dose limitation of 10 CFR Part 20, as addressed in SLCs 16.11.1 and 16.11.6.

An individual is not considered a MEMBER OF THE PUBLIC during any period in which he/she is engaged in carrying out any operation that is part of the nuclear fuel cycle.

REFERENCES

1. McGuire Nuclear Station, Offsite Dose Calculation Manual

2. 10 CFR Part 20

3. 40 CFR Part 190

4. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.12-3 Revision 67

Radiological Environmental Monitoring Program 16.11.13 16.11 RADIOLOGICAL EFFLUENT MONITORING 16.11.13 Radiological Environmental Monitoring Program COMMITMENT The Radiological Environmental Monitoring Program shall be conducted as specified in Table 16.11.13-1.

APPLICABILITY . At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Radiological A.1 Identify the reasons for not Within the next Environmental conducting the program as scheduled Annual Monitoring Program not required and the plans for Radiological being conducted as preventing a recurrence in Environmental specified in Table the Annual Radiological Operating Report 16.11.13-1. Environmental Operating Report.

B. Radioactivity level of B.1 Prepare and submit a 30 days environmental sampling Special Report that defines medium at a specified the corrective actions to be location in excess of taken to reduce radioactive reporting limits of Table effluents so that the 16.11.13-2. 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.

(continued)

McGuire Units 1 and 2 16.11.13-1 Revision 91

Radiological Environmental Monitoring Program 16.11.13 REMEDIAL ACTIONS (continued CONDITION REQUIRED ACTION COMPLETION TIME C. Milk or fresh leafy C. 1 ---------- NOTE -------

vegetable samples Specific locations from unavailable from one or which samples were more required sample unavailable may be deleted locations. from the program.

Revise the Radiological 30 days Environmental Monitoring Program to identify locations for obtaining replacement samples.

AND C.2 Identify the cause of the Within the next unavailability of samples scheduled Annual and identify new location(s) Radioactive Effluent for obtaining replacement Release Report samples in the next Annual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.13.1 ------------------ NOTES-----------------

The maximum values for the lower limits of detection shall be as specified in Table16.11.13-3.

The radiological environmental monitoring samples shall In accordance with be collected from the locations given in the table and Table 16.11.13-1 figure in the ODCM and shall be analyzed pursuant to the requirements of Tables16.11.13-1.

McGuire Units 1 and 2 16.11.13-2 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 1 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY PATHWAY AND/OR AND SAMPLE LOCATIONS(') COLLECTION OF ANALYSIS SAMPLE FREQUENCY

1. Direct Radiation(2 ) Forty routine monitoring stations either with Quarterly Gamma dose quarterly.

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 sector in the general area of the SITE BOUNDARY; An outer ring of stations, one in each meteorological sector in the 6- to 8-km range from the site; and The balance of the stations placed in special interest areas such as population centers, nearby residences, schools, and in one or two areas to serve as control stations.

(continued)

McGuire Units 1 and 2 16.11.13-3 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 2 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY PATHWAY AND/OR AND SAMPLE LOCATIONS(1 ) COLLECTION OF ANALYSIS SAMPLE FREQUENCY

2. Airborne Samples from five locations: Continuous sampler Radioiodine Canister:

Radioiodine and operation with sample 1-131 analysis weekly.

Particulates Three samples from close to the three SITE collection weekly, or BOUNDARY locations, in different sectors, of more frequently if Particulate Sampler:

the highest calculated annual average required by dust Gross beta radioactivity ground level D/Q. loading, analysis following filter change(4); Gamma isotopic One sample from the vicinity of a community analysis(5) of composite (by having the highest calculated annual average location quarterly).

ground level D/Q.

One sample from a control location, as for example 15-30 km distant and in the least prevalent wind direction(3 ).

3. Waterborne One sample upstream. Composite sample Gamma isotope analysis(5)

a. Surface(6 ) One sample downstream. over 1-month period(7 ). monthly. Composite for tritium analysis quarterly.

b. Ground Samples from one or two sources only if Quarterly Gamma isotopic(5) and likely to be affected(8) tritium analysis quarterly.

(continued)

McGuire Units 1 and 2 16.11.13-4 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 3 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY PATHWAY AND/OR AND SAMPLE LOCATIONS(1 ) COLLECTION OF ANALYSIS SAMPLE FREQUENCY

c. Drinking One sample of each of one to three of the Composite sample 1-131 analysis on each nearest water supplies that could be affected over 2-week period~z) composite when the dose by its discharge. when 1-131 analysis is calculated for the performed; monthly consumption of the water is One sample from a control location, composite otherwise. greater than 1 mrem per year(9). Composite for gross beta and gamma isotopic analyses monthly.

Composite for tritium analysis quarterly.

d. Sediment from One sample from downstream area with Semiannually Gamma isotopic analysis(5) the shoreline existing or potential recreational value, semiannually.

4. Ingestion Samples from milking animals in three Semimonthly when Gamma isotopic(5) and 1-131

a. Milk locations within 5-km distance having the animals are on analysis semimonthly when highest dose potential. If there are none, pasture; monthly at animals are on pasture; then one sample from milking animals in other times. monthly at other times.

each of three areas between 5 to 8 km distant where doses are calculated to be greater than 1 mrem per year(9 .

One sample from milking animals at a control location 15 to 30 km distant and in the least prevalent wind direction.

(continued)

McGuire Units 1 and 2 16.11.13-5 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 4 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM EXPOSURE NUMBER OF REPRESENTATIVE SAMPLES SAMPLING AND TYPE AND FREQUENCY PATHWAY AND/OR AND SAMPLE LOCATIONS(1 ) COLLECTION OF ANALYSIS SAMPLE FREQUENCY

b. Fish and One sample each commercially and Sample in season, or Gamma isotopic analysis(5)

Invertebrates recreationally important species in vicinity of semiannually if they on edible portions plant discharge area. are not seasonal One sample of same species in areas not influenced by plant discharge.

c. Food Products One sample of each principal class of food At time of harvest(1 °) Gamma isotopic analyses(5) products from any area that is irrigated by on edible portion.

water in which liquid plant wastes have been discharged.

Samples of three different kinds of broad leaf Monthly, when Gamma isotopic(5) and 1-131 vegetation grown nearest each of two available, analysis.

different offsite locations of highest predicted annual average ground level D/Q if milk sampling is not performed.

One sample of each of the similar broad leaf Monthly, when Gamma isotopic(5) and 1-131 vegetation grown 15 to 30 km distant in the available, analysis.

least prevalent wind direction if milk sampling is not performed.

McGuire Units 1 and 2 16.11.13-6 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 5 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NOTES:

1. Specific parameters of distance and direction sector from the centerline of one reactor, and additional description where pertinent, shall be provided for each and every sample location in Table 16.11.13-1 in a table and figure(s) in the ODCM. Refer to NUREG-0133, "Preparation of Radiological Effluent 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 hazardous conditions, seasonal unavailability, malfunction of automatic sampling equipment and other legitimate reasons. If specimens are unobtainable due to sampling equipment malfunction, every effort shall be made to complete corrective action prior to the end of the next sampling period. All deviations from the sampling schedule shall be documented in the Annual Radiological Environmental Operating Report. It is recognized that, at times, it may not be possible or practical to continue to obtain samples of the media of choice at the most desired location or time. In these instances suitable alternative media and locations may be chosen for the particular pathway in question and appropriate substitutions made within 30 days in the Radiological Environmental Monitoring Program. In lieu of an Licensee Event Report, identify the cause of the unavailability of samples for that pathway and identify the new locations(s) for obtaining replacement samples in the next Annual Radioactive Effluent Release Report and also include in the report a revised figure(s) and table for the ODCM reflecting the new location(s).

2. One or more instruments, 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 badges shall not be used as dosimeters for measuring direct radiation. The forty 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 sections will be over water so that the number of dosimeters may be reduced accordingly. The frequency of analysis or readout for TLD systems will depend upon the characteristics of the specific system used and should be selected to obtain optimum dose information with minimal fading.

3. The purpose of'this sample is to obtain background information. If it is not practical to establish control locations in accordance with the distance and wind direction criteria, other sites that provide valid background data may be substituted.

4. Airborne particulate sample filters shall be analyzed for gross beta radioactivity 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or more after sampling to allow for radon and thoron daughter decay. If gross beta activity in air particulate samples is greater than ten times the yearly mean of control samples, gamma isotopic analysis shall be performed on the individual samples.

McGuire Units 1 and 2 16.11.13-7 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-1 (Page 6 of 6)

RADIOLOGICAL ENVIRONMENTAL MONITORING PROGRAM NOTES (continued):

5. Gamma isotopic analysis means the identification and quantification of gamma-emitting radionuclides that may be attributable to the effluents from the facility.

6. The "upstream sample" shall 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.

7. A composite sample is one in which the quantity (aliquot) of liquid sampled is proportional to the quantity of flowing liquid and in which the method of sampling employed results in a specimen that is representative of the liquid flow. In this program composite sample aliquots shall be collected at time intervals that are very short (e.g.,

hourly) relative to the compositing period (e.g., monthly) in order to assure obtaining a representative sample.

8. 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.

9. The dose shall be calculated for the maximum organ and age group, using the methodology and parameters in the ODCM.

10. 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 tuborous and root food products.

McGuire Units 1 and 2 16.11.13-8 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-2 (Page 1 of 1)

REPORTING LEVELS FOR RADIOACTIVITY CONCENTRATIONS IN ENVIRONMENTAL SAMPLES REPORTING LEVELS AIRBOURNE FISH MILK MILK BROAD LEAF ANALYSIS WATER (pCi/I) PARTICULATE OR (pCi/kg, wet) (pCi/I) VEGETATION GASES (pCi/m 3 ) (pCi/kg, wet) 20,000{1 N/A N/A N/A N/A H-3 Mn-54 1,000 N/A 30,000 N/A N/A Fe-59 400 N/A 10,000 N/A N/A Co-58 1,000 N/A 30,000 N/A N/A Co-60 300 N/A 10,000 N/A N/A Zn-65 300 N/A 20,000 N/A N/A Zr-Nb-95 400 N/A N/A N/A N/A 1-131 2 0.9 N/A 3 100 Cs-134 30 10 1,000 60 1,000 Cs-137 50 20 2,000 70 2,000 Ba-La-140 200 N/A N/A 300 N/A NOTES:

1. For drinking water samples. This is 40 CFR Part 141 value. If no drinking water pathway exists, a value of 30,000 pCi/I may be used.

McGuire Units 1 and 2 16.11.13-9 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-3 (Page 1 of 3)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD) (1)(2)(3)

ANALYSIS WATER AIRBORNE FISH MILK BROAD LEAF SEDIMENT (pCi/I) PARTICULATE (pCi/kg, wet) (pCi/I) VEGETATION (pCi/kg, dry)

OR GASES (pCi/kg, wet)

(pCi/m 3)

Gross Beta 4 0.01 N/A N/A N/A N/A H-3 2000* N/A N/A N/A N/A N/A Mn-54 15 N/A 130 N/A N/A N/A Fe-59 30 N/A 260 N/A N/A N/A Co-58, 60 15 N/A 130 N/A N/A N/A Zn-65 30 N/A 260 N/A N/A N/A Zr-95 15 N/A N/A N/A N/A N/A Nb-95 15 N/A N/A N/A N/A N/A 1-131 1(4) 0.07 N/A 1 60 N/A Cs-134 15 0.05 130 15 60 150 Cs-137 18 0.06 150 18 80 180 Ba-140 15 N/A N/A 15 N/A N/A La-140 15 N/A N/A 15 N/A N/A

If no drinking water pathway exists, a value of 3000 pCi/I may be used.

McGuire Units 1 and 2 16.11.13-10 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-3 (Page 2 of 3)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)

NOTES:

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 5% probability of falsely concluding that a blank observation represents a "real" signal.

For a particular measurement system (which may include radiochemical separation):

LLD = 4.66 sb E. V 2.22- Y -exp (-2 At)

Where:

LLD is the "a priori" lower limit of detection as defined above (as picoCurie per unit mass or volume),

sb is the standard deviation of the background counting rate or of the counting rate of a blank sample as appropriate (as counts per minute),

E is the counting efficiency (as counts per disintegration),

V is the sample size (in units of mass or volume),

2.22 is the number of disintegrations per minute per picoCurie, Y is the fractional radiochemical yield (when applicable),

k is the radioactive decay constant for the particular radionuclide, and At is the elapsed time between sample collection (or end of the sample collection period) and time of counting (for environmental samples, not plant effluent samples).

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. 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 circumstances may render these LLDs unachievable. In such cases, the contributing factors shall be identified and described in the Annual Radiological Environmental Operating Report.

McGuire Units 1 and 2 16.11.13-11 Revision 91

Radiological Environmental Monitoring Program 16.11.13 TABLE 16.11.13-3 (Page 3 of 3)

MAXIMUM VALUES FOR THE LOWER LIMITS OF DETECTION (LLD)

NOTES (continued):

2. 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.

3. Required detection capabilities for thermoluminescent dosimeters used for environmental measurements shall be in accordance with the recommendations of Regulatory Guide 4.13.

4. LLD for drinking water samples. If no drinking water pathway exists, the LLD of gamma isotopic analysis may be used.

McGuire Units 1 and 2 16.11.13-12 Revision 91

Radiological Environmental Monitoring Program 16.11.13 BASES The Radiological Environmental Monitoring Program is established to monitor the radiation and radionuclides in the environs of the plant. The program provides representative measurements of radioactivity in the highest potential exposure pathways, and verification of the accuracy of the effluent monitoring program and modeling of environmental exposure pathways. The program is contained in SLC 16.11.13 - 16.11.16 and conforms to the guidance of Appendix I to 10 CFR Part 50.

The program includes the following:

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

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

3. Participation in an Interlaboratory Comparison Program to ensure that independent checks on the precision and accuracy of the measurements of radioactive materials in environmental sample matrices are performed as part of the quality assurance program for environmental monitoring.

The portion of 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 station operation. This monitoring program implementsSection IV.B.2 of Appendix I 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 operation. 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 lower limits of detection (LLDs). The LLDs required by Table 16.11.13-3 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 a posteriori (after the fact) limit for a particular measurement.

With the level of radioactivity in an environmental sampling medium at a specified location exceeding the reporting levels of Table 16.11.13-3 when averaged over any calendar quarter, in lieu of a Licensee Event Report, prepare and submit to the Commission within 30 days a Special Report that defines the corrective actions to be McGuire Units 1 and 2 16.11.13-13 Revision 91

Radiological Environmental Monitoring Program 16.11.13 BASES (continued) 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 SLCs 16.11.6, 16.11.8, and 16.11.9. When more than one of the radionuclides in Table 16.11.13-2 are detected in the sampling medium, this report shall be submitted if:

concentration (1) + concentration (2) +. ...... > 1.0 limit level (1) limit level (2)

When radionuclides other than those in Table 16.11.13-2 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 is equal to or greater than the calendar year limits of SLCs 16.11.6, 16.11.8 and 16.11.9. This report is not required if the measured level of radioactivity was not the result of plant effluents; however, in such an event, the condition shall be reported and described in the Annual Radiological Environmental Operating Report. The methodology and parameters used to estimate the potential annual dose to a MEMBER OF THE PUBLIC shall be indicated in this report.

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).

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.13-14 Revision 91

Land Use Census 16.11.14 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:

a. the nearest milk animal,

b. the nearest residence, and

c. the nearest garden of greater than 50 m 2 (500 ft 2) producing broad leaf vegetation.

For elevated releases as defined in Regulatory Guide 1.111, Revision 1, July 1977, the land use census shall identify within a distance of 5 km (3 miles) the location in each of the 16 meteorological sectors of:

a. all milk animals, and

b. all gardens of greater than 50 m 2 producing broad leaf vegetation.

-NOTE Broad leaf vegetation sampling of three different kinds of vegetation may be performed at the SITE BOUNDARY in each of two different direction sectors with the highest predicted D/Qs in lieu of the garden census. Specifications for broad leaf vegetation sampling in Table 16.11.13-1 4c shall be followed, including analysis of control samples.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Location(s) identified A.1 Identify the new location in In next scheduled which yields a calculated the Annual Radioactive Annual Radioactive dose/dose commitment Effluent Release Report. Effluent Release greater than values Report currently calculated in SLC 16.11.9.

(continued)

McGuire Units 1 and 2 16.11.14-1 Revision 21

Land Use Census 16.11.14 REMEDIAL ACTIONS (continued)

B. Location(s) identified B.1 Add the new location to the 30 days which yields a Radiological Environmental calculated dose or dose Monitoring Program.

commitment (via same exposure pathway) 20% AND greater than at a location from which samples are B.2 --------- NOTES------

currently being obtaihed If samples cannot be in accordance with SLC obtained, an explanation of 16.11.13. why samples are not obtainable (substitute representative locations if possible) shall be included.

Identify the new In the next location(s), revised figures scheduled Annual and tables for the ODCM, Radiological Release in the next Annual Report Radiological Release Report.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.14.1 ---------------- NOTE ----------------------------------

The results of the land use census shall be included in the Annual Radiological Environmental Operating Report.

Conduct a land use census during the growing season 12 months using the information which will provide the best results such as a door-to-door survey, aerial survey, or consultation with local agricultural authorities.

McGuire Units 1 and 2 16.11.14-2 Revision 21

Land Use Census 16.11.14 BASES This commitment is provided to ensure that changes in the use of areas at and beyond the SITE BOUNDARY are identified and that modifications to the Radiological Environmental Monitoring Program are made if required by the results of this census. The best information from the door-to-door survey, from aerial survey, or from consulting with local agricultural authorities shall be used. This census satisfies the requirements of Section IV.B.3 of 2 Appendix I to 10 CFR Part 50. Restricting 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 quantity (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 vegetation (i.e., similar to lettuce and cabbage), and (2) a vegetation yield of 2 kg/M 2.

With a land use census identifying a location(s) which 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 to the Radiological Environmental Monitoring Program. The sampling location(s),

excluding the control station location, having the lowest calculated dose or dose commitment (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.

REFERENCES

1. McGuire Nuclear Station, Off site Dose Calculation Manual

2. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.14-3 Revision 21

Interlaboratory Comparison Program 16.11.15 1.6.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.15 Interlaboratory Comparison Program COMMITMENT Analyses shall be performed on radioactive materials, supplied as part of an Interlaboratory Comparison Program (ICP), that correspond to samples required by SLC 16.11.13.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Analyses not being A.1 Report corrective actions In next scheduled performed as required. taken to prevent recurrence Annual Radiological in the Annual Radiological Environmental Environmental Operating Operating Report Report.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.15.1 Report a summary of the results of the Interlaboratory 12 months Comparison Program in the Annual Radiological Environmental Operating Report.

McGuire Units 1 and 2 16.11.15-1 Revision 21

Interlaboratory Comparison Program 16.11.15 BASES This requirement for participation in an Interlaboratory Comparison Program is provided to ensure that independent checks on the precision and accuracy of the measurements of radioactive material in 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.

The Interlaboratory Comparison Program (ICP) shall be described in the Annual Radiological Environmental Operating Report.

REFERENCES

1. 10 CFR Part 50, Appendix I McGuire Units 1 and 2 16.11.15-2 Revision 21

Annual Radiological Environmental Operating Report 16.11.16 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.16 Annual Radiological Environmental Operating Report COMMITMENT Routine Annual Radiological Environmental Operating Reports covering the operation of the unit during the previous calendar year shall be submitted prior to May 15 of each year.

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 pre-operational studies, with operational controls as appropriate, and with previous environmental surveillance reports, and an assessment of the observed impacts of the plant operation on the environment. The reports shall also include the results of land use censuses required by SLC 16.11.14.

The Annual Radiological Environmental Operating Reports shall include the results 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 (one map shall cover stations near the site boundary; a second shall include the more distant stations);

the results of licensee participation in the Interlaboratory Comparison Program, required by SLC 16.11.15;

a discussion of all deviations from the sampling schedule of Table 16.11.13-1; and McGuire Units 1 and 2 16.11.16-1 Revision 1

Annual Radiological Environmental Operating Report 16.11.16 COMMITMENT (continued)

a discussion of all analyses in which the LLD required by Table 16.11.13-3 was not achievable.

A single submittal may be made for a multiple unit station..

APPLICABILITY At all times.

REMEDIAL ACTIONS None TESTING REQUIREMENTS None BASES None REFERENCES

1. Technical Specification 5.6.2 McGuire Units 1 and 2 16.11.16-2 Revision 1

Radioactive Effluent Release Reports 16.11.17 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.17 Radioactive Effluent Release Reports COMMITMENT Routine Radioactive Effluent Release Reports covering the operation of the unit during the previous calendar year of operation shall be submitted before May 1 of each year.

The Radioactive Effluent Release Reports shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit as outlined in Regulatory Guide 1.21, "Measuring, Evaluating, and Reporting Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid and Gaseous Effluents from Light-Water-Cooled Nuclear Power Plants," Revision 1, June 1974, with data summarized on a quarterly basis following the format of Appendix B thereof.

The Radioactive Effluent Release Report shall include an annual summary of hourly meteorological data collected over the previous calendar 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. This same report shall include 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. A five year average of representative onsite meteorological data shall be used in the gaseous effluent dose pathway calculations. Dispersion factors (X/Qs) and deposition factors (D/Qs) shall be generated using the computer code XOQDOQ (NUREG/CR-2919) which implements NRC Regulatory Guide 1.111.

The meteorological conditions concurrent with the time of release shall be reviewed annually to determine if the five-year average values should be revised. 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 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.

McGuire Units 1 and 2 16.11.17-1 Revision 118

Radioactive Effluent Release Reports 16.11.17 COMMITMENT (continued)

The Radioactive Effluent Release Reports shall include the following information for each type of solid waste shipped offsite or disposed of in the site landfill 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 agents (media)).

The Radioactive Effluent Release Reports shall include a list and description of unplanned releases from the site to UNRESTRICTED AREAS of radioactive materials in gaseous and liquid effluents made during the reporting period.

The Radioactive Effluent Release Reports shall include any changes made during 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.

The Radioactive Effluent Release Reports shall also identify any licensee initiated major changes to the Radioactive Waste Systems (liquid, gaseous, and solid). Otherwise, this information may be included in the annual UFSAR update. The discussion of each change shall contain:

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

b. Sufficient detailed information to totally support the reason for the change without benefit of additional or supplemental information;

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

d. 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; McGuire Units 1 and 2 16.11.17-2 Revision 118

Radioactive Effluent Release Reports 16.11.17 COMMITMENT (continued)

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

f. 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;

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

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

A single submittal may be made for a multiple unit station. The submittal should combine those sections that are common to all units at the station; however, for units with separate Radwaste Systems, the submittal shall specify the releases of radioactive material from each unit.

APPLICABILITY At all times REMEDIAL ACTIONS None TESTING REQUIREMENTS None BASES None REFERENCES

1. Technical Specification 5.6.3 McGuire Units 1 and 2 16.11.17-3 Revision 118

Liquid Holdup Tanks 16.11.18 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.18 Liquid Holdup Tanks COMMITMENT The quantity of radioactive material contained in each unprotected outdoor radwaste tank shall be limited to < 10 Curies, excluding, tritium and dissolved or entrained noble gases.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Quantity of radioactive A.1 Suspend all additions of Immediately material in tank not radioactive material to the within limit, tank.

AND A.2 Reduce the tank contents 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to within limit.,

AND A.3 Describe the events Within the next leading to this condition in scheduled Annual the next Annual Radioactive Effluent Radioactive Effluent Release Report Release Report.

TESTING REQUIREMENTS TEST FREQUENCY TR .16.11.18.1 Verify the quantity of radioactive material contained in 7 days unprotected outdoor radwaste tanks is within limits by analyzing a representative sample of the tank's contents when radioactive materials are being added to the tank..

McGuire Units 1 and 2 16.11.18-1 Revision 0

Liquid Holdup Tanks 16.11.18 BASES The tanks applicable to this SLC include all those outdoor radwaste tanks that are not surrounded by liners, dikes, or walls capable of holding the tank contents and that do not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System.

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

REFERENCES None McGuire Units 1 and 2 16.11.18-2 Revision 0

Explosive Gas Mixture 16.11.19 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.19 Explosive Gas Mixture COMMITMENT The concentration of oxygen in the WASTE GAS HOLDUP SYSTEM shall be limited to < 2% by volume whenever the hydrogen concentration exceeds 4% by volume.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Concentration of oxygen A.1 Reduce oxygen 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months in the WASTE GAS concentration to within HOLDUP SYSTEM limits.

> 2% but <4% by volume.

B. Concentration of oxygen B.1 Suspend all additions of Immediately in the WASTE GAS waste gases to the system.

HOLDUP SYSTEM

> 4% and hydrogen AND concentration > 4% by volume. B.2 Reduce the concentration Immediately of oxygen to < 4% by volume.

AND B.3 Reduce oxygen 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months concentration to within limits.

McGuire Units 1 and 2 16.11.19-1 Revision 0

Explosive Gas Mixture 16.11.19 TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.19.1 Verify the concentrations of hydrogen and oxygen in the During WASTE WASTE GAS HOLDUP SYSTEM is within limits by GAS HOLDUP monitoring waste gases in the WASTE GAS HOLDUP SYSTEM SYSTEM with the hydrogen and oxygen monitors operation required by SLC 16.7.8.

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

REFERENCES None McGuire Units 1 and 2 16.11.19-2 Revision 0

Gas Storage Tanks 16.11.20 16.11 RADIOLOGICAL EFFLUENT CONTROLS 16.11.20 Gas Storage Tanks COMMITMENT The quantity of radioactivity contained in each gas storage tank shall be limited < 49,000 Curies noble gases (considered as Xe-133).

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Quantity of radioactive A.A Suspend all additions of Immediately material in tank not radioactive material to the within limit, tank.

AND A.2 Reduce the tank contents 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months to within limit.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.11.20.1 Verify the quantity of radioactive material contained in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months each gas storage tank is within limit when radioactive materials are being added to the tank.

McGuire Units 1 and 2 16.11.20-1 Revision 0

Gas Storage Tanks 16.11.20 BASES This SLC considers postulated radioactive releases due to a waste gas system leak or failure, and limits the quantity of radioactivity in each pressurized gas storage tank in the WASTE GAS HOLDUP SYSTEM to assure that a release would be substantially below the dose guideline values of 10 CFR Part 100 for a postulated event.

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

REFERENCES None McGuire Units 1 and 2 16.11.20-2 Revision 0

In-Plant Iodine Monitoring 16.12.1 16.12 RADIATION PROTECTION 16.12.1 In-Plant Iodine Monitoring COMMITMENT A program shall be established, implemented, and maintained which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

a. Training of personnel,

b. Procedures for monitoring, and

c. Provisions for maintenance of sampling and analysis equipment.

APPLICABILITY At all times.

REMEDIAL ACTIONS None TESTING REQUIREMENTS None BASES This commitment is provided to ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions.

REFERENCES

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

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

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

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

McGuire Units 1 and 2 16.12.1-1 Revision 0

Sealed Source Contamination 16.12.2 16.12 RADIATION PROTECTION 16.12.2 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 CONDITION REQUIRED ACTION COMPLETION TIME A. Sealed source having A.1 Withdraw the sealed Immediately removable source from use.

contamination in excess of the above limits. AND A.2.1 Decontaminate and repair Immediately the sealed source.

OR A.2.2 Dispose of the sealed Immediately source in accordance with NRC Regulations AND A.3 Prepare and submit an 12 months annual report to the NRC for sealed sources or fission detectors that exceed the limits.

McGuire Units 1 and 2 16.12.2-1 Revision 0

Sealed Source Contamination 16.12.2 TESTING REQUIREMENTS

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

Testing shall be performed by the Licensee or other persons specifically authorized by NRC or an Agreement State.

TEST FREQUENCY TR 16.12.2.1 ------------------

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

Only applicable for sources containing radioactive materials with half-lives greater than 30 days (excluding H3) and in any form other than gas.

Each category of sealed source that is in use (excluding 6 months startup sources and fission detectors previously subjected to core flux) shall be tested for leakage and/or contamination with a detection sensitivity of at least 0.005 microCurie per test sample.

TR 16.12.2.2 -NOTE Sealed sources and fission detectors transferred without a certificate indicating the last test date shall be tested prior to being placed in use.

Each category of sealed source that is not in use Prior to use or (excluding startup sources and fission detectors transfer to another previously subjected to core flux) shall be tested for licensee unless leakage and/or contamination with a detection sensitivity tested in the of at least 0.005 microCurie per test sample. previous 6 months TR 16.12.2.3 Each sealed startup source or fission detector shall be 31 days prior to tested for leakage and/or contamination with a detection being subjected to sensitivity of at least 0.005 microCurie per test sample. core flux or installation in the core AND Following repair or maintenance to the source McGuire Units 1 and 2 16.12.2-2 Revision 0

Sealed Source Contamination 16.12.2 BASES The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) 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 Surveillance Requirements commensurate with the probability of damage to a source in that group.

Those sources which are frequently handled are required to be tested more often than those which are not. Sealed sources which are continuously enclosed within a shielded mechanism (i.e., sealed sources within radiation monitoring or boron measuring devices) are considered to be stored and need not be tested unless they are removed from the shielded mechanism.

REFERENCES None McGuire Units 1 and 2 16.12.2-3 Revision 0

Fire Brigade 16.13.1 16.13 CONDUCT OF OPERATIONS 16.13.1 Fire Brigade COMMITMENT A site Fire Brigade of at least five members shall be maintained onsite.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Fire Brigade composition A.1 Initiate action to fill required Immediately requirements not met. positions.

AND A.2 Restore minimum fire 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months brigade composition.

TESTING REQUIREMENTS TEST FREQUENCY TR 16.13.1.1 Verify fire brigade composition. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months McGuire Units 1 and 2 16.13.1-1 Revision 51

Fire Brigade 16.13.1 BASES The primary purpose of the Fire Protection Program is to minimize both the probability and consequences of postulated fires. Despite designed active and passive Fire Protection Systems installed throughout the plant, a properly trained and equipped fire brigade organization of at least five members is needed to provide immediate response to fires that may occur at the site.

A fire brigade leader is required by the McGuire operating licenses and Fire Protection Program. The individual fulfilling this position shall:

Have sufficient training or knowledge of plant safety related systems to understand the effects of a fire and fire suppression systems on safe shutdown capability,

" Be unavailable for other activities when directing the fire brigade,

Be a licensed RO or SRO who is qualified to be a fire brigade leader.

The Fire Brigade requirement is met by using personnel from Operations and SPOC. Four (4) personnel from Operations are required (including the Fire Brigade Leader) and the other (1) person is from SPOC.

The Fire Brigade shall not include members of the minimum shift crew necessary for safe shutdown of the unit and any personnel required for other essential functions during a fire emergency.

Fire Brigade equipment and training conform to the recommendations of the National Fire Protection Association, Appendix A to Branch Technical Position 9.5-1 and supplemental NRC Staff guidelines.

This selected licensee commitment is part of the McGuire Fire Protection Program and therefore subject to the provisions of McGuire Facility Operating License Conditions C.4 (Unit 1) and C.7 (Unit 2).

REFERENCES

1. McGuire Nuclear Station UFSAR, Chapter 13.2.

2. McGuire Nuclear Station, SER Supplement 2, Chapter 9.5.1 and Appendix D.

3. McGuire Nuclear Station, SER Supplement 5, Chapter 9.5.1 and Appendix B.

4. McGuire Fire Protection Review, as revised.

5. McGuire Nuclear Station, SER Supplement 6, Chapter 9.5.1 and Appendix C.

6. McGuire Nuclear Station Facility Operating Licenses, Unit 1 License Condition C.(4) and Unit 2 License Condition C.(7)

McGuire Units 1 and 2 16.13.1-2 Revision 51

16.13.2 16.13 CONDUCT OF OPERATIONS 16.13.2 Not Used McGuire Units 1 and 2 16.13.2-1 Revision 75

16.13.3 16.13 CONDUCT OF OPERATIONS 16.13.3 Not Used McGuire Units 1 and 2 16.13.3-1 Revision 75

Minimum Station Staffing Requirements 16.13.4 16.13 CONDUCT OF OPERATIONS 16.13.4 Minimum Station Staffing Requirements COMMITMENT Minimum station staffing shall be as indicated in Table 16.13.4-1.

APPLICABILITY At all times.

REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Minimum station A.1 Initiate actions to fill Immediately staffing not met. required positions.

AND A.2 Restore minimum 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months station staffing.

TESTING REQUIREMENTS None BASES This SLC represents a consolidation of staffing requirements from a number of different regulatory requirements. The specific requirements that must be met at all times are described in the text below. The station must meet all of the regulatory requirements at all times. There is some interaction between the requirements associated with (1) the ability to mitigate design basis accidents and events, (2) fire events and (3) the emergency plan. The station must be able to demonstrate that the emergency plan is staffed as stated in the required emergency plan. During any design basis accident or event (including the design basis fire event) individuals from the emergency plan may fill mitigation roles for that purpose. However, the site must be able to demonstrate that no single individual would be called upon to perform conflicting tasks at any point in time during a design basis accident or event.

The requirements of this SLC consolidate McGuire station staffing requirements into one document. This SLC includes the station staff requirements of the McGuire Facility Operating Licenses, Technical Specification (TS) 5.2.2, 10 CFR 50.54(m), applicable Operations Management Procedures (OMPs), Nuclear System Directive (NSD) 112, "Fire Brigade Organization, Training and Responsibilities," the McGuire Fire Protection Program, the McGuire Emergency Plan, and SLC 16.13. 1, "Fire Brigade." The total requirement for each position was obtained by summing the various individual requirements for that position.

The bases for the numbers in the first column of Table 16.13.4-1 are as follows:

McGuire Units 1 and 2 16.13.4-1 Revision 58

Minimum Station Staffing Requirements 16.13.4 1 OSM (active SRO) - Required by 10 CFR 50.54(m)(2)(ii) and implemented via OMP.

1 STA (active or inactive SRO) - Required by TS 5.2.2g and implemented via OMP. Note that old TS (pre-Improved TS) Table 6.2-1, which implemented the requirements of NUREG-0737, "Clarification of TMI Action Plan Requirements," did not require an STA on shift when both units were in MODE 5, 6, or defueled. Table 16.13.4-1 is more restrictive in that it requires an STA on shift at all times.

2 SROs (active SRO) - Required by 10 CFR 50.54(m)(2)(i). Per TS 5.2.2b and 10 CFR 50.54(m)(2)(iii), at least 1 SRO must be in the control room. Implemented by OMP.

4 ROs - Required by Tabletop Review of Abnormal Procedures, Memo to File dated 5/8/01.

Implemented via OMP.

3 NLOs - Required by TS 5.2.2a and Section B, Figure B-1 of the Emergency Plan and implemented via OMP and NSD.

3 NLOs - Required by the Fire Protection Program and implemented via NSD and OMP.

1 Chemistry Technician (ERO) - Required by Section B, Figure B-1 of the Emergency Plan.

Implemented via EP Group Manual Section 1.1. Any technician who is qualified may be credited towards fulfilling the ERO requirement.

3 Radiation Protection Technicians (2 technicians and 1 off-site dose assessor) (ERO) -

Required by Section B, Figure B-1 of the Emergency Plan. Implemented via EP Group Manual Section 1.1. 1 technician is required by TS 5.2.2d and may be counted towards fulfilling the ERO requirement. Any technician who is qualified may be credited towards fulfilling the ERO requirement. In the event of a fire, the technician will respond to the fire for radiological monitoring purposes until directed otherwise.

1 Mechanical Maintenance Technician (ERO) - Required by Section B, Figure B-1 of the Emergency Plan. Implemented via EP Group Manual Section 1.1. Any technician who is fire brigade qualified may be credited towards fulfilling the ERO requirement and the fire brigade requirement. In the event of a fire, either the Mechanical Maintenance Technician or an Instrument and Electrical Technician will respond to the fire until directed otherwise.

2 Instrumentation and Electrical Technicians (ERO) - Required by Section B, Figure B-1 of the Emergency Plan. Implemented via EP Group Manual Section 1.1. Any technician who is fire brigade qualified may be credited towards fulfilling the ERO requirement and the fire brigade requirement. In the event of a fire, either the Mechanical Maintenance Technician or an Instrument and Electrical Technician will respond to the fire until directed otherwise.

2 MERT (ERO) - Required by Section B, Figure B-1 of the Emergency Plan. Implemented via EP Group Manual Section 1.1. Any technician who is qualified may be credited towards fulfilling the ERO requirement. In the event of a fire, the technician will respond to the fire for security purposes until directed otherwise.

BASES (con't)

McGuire Units 1 and 2 16.13.4-2 Revision 58

Minimum Station Staffing Requirements 16.13.4 Fire Brigade - The primary purpose of the Fire Protection Program is to minimize both the probability and consequence of postulated fires. Despite designed active and passive fire protection systems installed throughout the plant, a properly trained and equipped fire brigade organization of at least 5 members is required to provide immediate response to fires that may occur at the site. The fire brigade requirement is met by using personnel from Operations and SPOC. 4 personnel from Operations are required (including the fire brigade leader) and the other 1 person is from SPOC.

Fire Brigade Leader- Required by the McGuire Facility Operating Licenses and Fire Protection Program and implemented via NSD and OMP. The individual fulfilling this position shall be a SRO or RO who is qualified to be a fire brigade leader. This individual functions as the fire brigade leader and is not available for other activities when directing the fire brigade. The fire brigade leader shall have sufficient training in or knowledge of plant safety related systems to understand the effects of a fire and fire suppression systems on safe shutdown capability.

Minimum station staffing totals for the SRO, RO, and NLO positions in Table 16.13.4.1 are a

-\ function of the number of units in MODES 1-4. The totals for the remaining positions in Table 16.13.4.1 are not a function of the operational MODES of the units.

10 CFR 50.54(m)(2)(i) requires 2 SROs when both units are in MODES 1-4, 2 SROs when one unit is in MODES 1-4, and 1 SRO when no unit is in MODES 1-4.

10 CFR 50.54(m)(2)(i) requires 3 ROs when both units are in MODES 1-4,3 ROs when one unit is in MODES 1-4, and 2 ROs when no unit is in MODES 1-4.

TS 5.2.2 a requires 3 NLOs when both units are in MODES 1-4, 3 NLOs when one unit is in MODES 1-4, and 2 NLOs when no unit is in MODES 1-4.

The 2-hour remedial action for restoring minimum station staffing levels is consistent with TS 5.2.2c and 5.2.2d, which allow 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

McGuire Units 1 and 2 16.13.4-3 Revision 58

Minimum Station Staffing Requirements 16.13.4 TABLE 16.13.4-1 MINIMUM STATION STAFFING REQUIREMENTS Title or Expertise Both Units in One Unit in Both Units in MODES 1-4 MODES 1-4 MODES 5,6,or No MODE Operations Shift Manager 1 1 1 (OSM)

Shift Technical Advisor 1 1 1 (STA)

Senior Reactor Operator 2 2 1 (SRO) (Notes 1,2, 3)

Reactor Operator (RO) 4 4 3 (Notes 1,4)

Non-Licensed Operator 6 6 5 (NLO)

Chemistry Technician 1 1 1 Radiation Protection 3 3 3 Mechanical Maintenance 1 1 1 Technician Instrumentation and 2 2 2 Electrical Technician Medical Emergency 2 2 2 Response Team (MERT)

Security Personnel Per Security Plan Fire Brigade Per SLC 16.13.1 McGuire Units 1 and 2 16.13.4-4 Revision 58

Minimum Station Staffing Requirements 16.13.4 TABLE 16.13.4-1 MINIMUM STATION STAFFING REQUIREMENTS (con't)

Note 1: Either a SRO (active or inactive) or a RO may be designated as the fire brigade leader. The totals for the appropriate position shall be increased by one, depending upon which position is being used to fulfill the role of fire brigade leader.

Note 2: In addition to these requirements, during CORE ALTERATIONS (including fuel loading or transfer), a SRO shall be present to directly supervise the activity. During this time, no other duties shall be assigned to this person.

Note 3: With any unit in MODES 1-4, a SRO shall be present in the control room at all times.

Note 4: For each fueled unit, a RO shall be present at the controls at all times.

McGuire Units 1 and 2 16.13.4-5 Revision 58

Minimum Station Staffing Requirements 16.

13.4 REFERENCES

1. McGuire Facility Operating Licenses for Units 1 and 2, NPF-9 and NPF-17.

2. McGuire TS 5.2.2, "Unit Staff'.

3. 10 CFR 50.54(m).

4. OMP 2-2, "Conduct of Operations."

5. NSD 112, "Fire Brigade Organization, Training and Responsibilities."

6. McGuire Emergency Plan.

7. SLC 16.13-1, "Fire Brigade."

8. NUREG-0800, "Standard Review Plan for the Review of Safety Analysis Reports For Nuclear Power Plants, LWR Edition," Section 9.5.1C3.

9. EP Group Manual Section 1.1, "Emergency Organization".

10. "Tabletop Review of AP/1 &2/A/5500/24 (Loss of Plant Control Due to Fire Or Sabotage)" Memo to File dated May 8, 2001.

11. H. B. Barron letter to NRC Document Control Desk dated March 28, 2000, Emergency Plan Revision of Table B-I.

12. H. B. Barron letter to NRC Document Control Desk dated September 7, 2000, Emergency Plan Table B-1.

13. Frank Rinaldi to H. B. Barron dated October 3, 2000, McGuire Nuclear Station, Units 1 and 2 - Revision to Station Emergency Plan (TAC Nos. MA8908 and MA8909).

14. NRC Information Notice 95-48, "Results of Shift Staffing Study."

15. NRC Information Notice 91-77, "Shift Staffing at Nuclear Power Plants."

16. NRC SECY 93-184, "Shift Staffing at Nuclear Power Plants."

McGuire Units 1 and 2 16.13.4-6 Revision 58

Startup Reports 16.14.1 16.14 TESTING 16.14.1 Startup Reports COMMITMENT The following report shall be submitted in accordance with 10CFR50.4:

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

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

3. Startup Reports shall be submitted within: (1) 90 days following completion of the STARTUP test program, or (2) 90 days following resumption or commencement of commercial POWER OPERATION, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of STARTUP test program, and resumption or commencement of commercial operation),

supplementary reports shall be submitted at least every 3 months until all three events have been completed.

APPLICABILITY At all times.

REMEDIAL ACTION None TESTING REQUIREMENTS Not Applicable.

McGuire Units 1 and 2 16.14.1-1 Revision 0

Startup Reports 16.14.1 BASES This commitment is to satisfy the requirements of 10CFR50.4.

REFERENCES

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

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

McGuire Units 1 and 2 16.14.1-2 Revision 0

Reactor Trip Breaker and SSPS Logic Train Out of Service Commitments 16.14.2 16.14 TESTING 16.14.2 Reactor Trip Breaker and Solid State Protection System (SSPS) Logic Train Out of Service Commitments COMMITMENTS: Risk-significant plant configurations shall not be entered when a reactor trip breaker or a SSPS logic train (Technical Specifications 3.3.1 and 3.3.2) is inoperable for maintenance:

1. To preserve ATWS mitigation capability, activities that degrade the ability of auxiliary feedwater system (AFWS), reactor coolant system (RCS) pressure relief systems (pressurizer PORVs and safety valves), ATWS mitigating system actuation circuitry (AMSAC), or turbine trip shall not be scheduled.when a reactor trip breaker or a SSPS logic train is inoperable for maintenance.

2. To preserve LOCA mitigation capability, one complete Emergency Core Cooling System (ECCS) train that can be actuated automatically must be maintained when a SSPS logic train is inoperable for maintenance.

3. To preserve reactor trip and safeguards actuation capability, activities that cause master relays or slave relays in the available train to be unavailable and activities that cause analog channels to be unavailable shall not be scheduled when a reactor trip breaker or a SSPS logic train is inoperable for maintenance.

4. Activities in electrical systems (e.g., AC and DC power) and cooling systems (e.g., essential service water and component cooling water) that support the systems or functions listed in Commitments 1,2 and 3 above shall not be scheduled when a reactor trip breal~er or SSPS logic train is inoperable for maintenance. That is, one complete train of a function noted above must be available.

5. To preserve capabilities to prevent large early releases, activities that degrade the ability of the containment spray systems, air return fans, and ice condenser shall not be scheduled when a SSPS logic train is inoperable for maintenance.

APPLICABILITY: When a reactor trip breaker or a SSPS logic train is required to be operable per Technical Specification 3.3.1 or 3.3.2.

McGuire Units 1 and 2 16.14.2-1 Revision 104

Reactor Trip Breaker and SSPS Logic Train Out of Service Commitments 16.14.2 REMEDIAL ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Risk-significant plant A. 1 Assess and manage the As soon as possible configurations. identified resulting risk in accordance in the COMMITMENTS with the Maintenance Rule are encountered during program, 10 CFR maintenance of a reactor 50.65(a)(4).

trip breaker or a SSPS logic train. AND A.2 Exit risk-significant plant As soon as possible configurations.

TESTING REQUIREMENTS None BASES As specified in the NRC SERs dated December 30, 2008 and March 9, 2009 for License Amendments 248/228 and License Amendments 250/230, respectively, revising Technical Specifications 3.3.1 and 3.3.2, the commitments listed in this SLC are Tier 2 commitments.

Tier 2 identifies and evaluates any potential risk-significant plant equipment outage configurations that could result if equipment, in addition to that associated with the proposed application, is taken out of service simultaneously, or if other risk-significant operational factors, such as concurrent system or equipment testing, are also involved. The purpose of this evaluation is to ensure that appropriate restrictions are in place such that risk-significant plant equipment outage configurations will not occur when equipment associated with the proposed Completion Time is implemented.

Even though this SLC is designed to prevent entering risk-significant plant configurations, it is also applicable when risk-significant plant configurations are encountered during maintenance of a reactor trip breaker or a SSPS logic train.

Commitment 1 is associated with preserving ATWS mitigation capability. The phrase, "activities that degrade the ability", refers to those activities that make the listed mitigation systems, functions, or equipment, unable (i.e., unavailable) to perform their required functions.

Commitment 2 is associated with preserving LOCA mitigation capability. The phrase, "must be maintained", means the listed mitigation train function must be able (i.e., available) to perform the required function.

McGuire Units 1 and 2 16.14.2-2 Revision 104

Reactor Trip Breaker and SSPS Logic Train Out of Service Commitments 16.14.2 Commitment 3 is associated with preserving reactor trip and safeguards actuation capability on the train opposite the reactor trip breaker train or SSPS logic train that is inoperable for maintenance. The phrase, "activities that cause master relays or slave relays in the available train to be unavailable", refers to activities that make the master relay(s) or slave relay(s) in the train opposite the reactor trip breaker train or SSPS logic train that is inoperable for maintenance, unable (i.e., unavailable) to perform their required functions.

The phrase, "activities that cause analog channels to be unavailable", refers to activities that cause any of the Reactor Protection System (RPS)/Engineered Safety Features Actuation System (ESFAS) analog channels that are inputs to SSPS to be unable (i.e., unavailable) to perform their required functions while either train of SSPS or any reactor trip breaker is inoperable for maintenance.

Commitment 4 is associated with activities in support systems that directly affect the availability of the systems and equipment listed in Commitments 1, 2 and 3. The systems and equipment listed in Commitments 1, 2 and 3,must be able (i.e., available) to perform their required functions. To ensure these systems and equipment are available, the required electrical systems (e.g., AC and DC power) and cooling systems that support these systems and equipment must also be available (i.e., able) to perform their required functions.

Commitment 5 is associated with preserving capabilities to prevent large early releases. The phrase, "activities that degrade the ability", refers to those activities that make the listed mitigation systems or equipment unable (i.e., unavailable) to perform their required functions.

The systems/functions and their associated TS/SLC involved in these commitments are:

Reactor Trip Breakers- TS 3.3.1, Functions 17 and 18 (Commitments 1, 2, 3, 4, 5)

SSPS Logic Train -TS 3.3.1 and TS 3.3.2 (Commitments 1, 2, 3, 4, 5)

AFWS-TS 3.7.5 (Commitment 1)

RCS Pressure Relief Systems (Pressurizer PORVs and Safety Valves) - TS 3.4.10, TS 3.4.11, and SLC 16.5.6 (Commitment 1)

AMSAC-SLC 16.7.1 (Commitment 1)

Turbine Trip - TS 3.3.2, Function 5 (Commitment 1)

ECCS Train - TS 3.5.2 and TS 3.5.3 (Commitment 2)

SSPS Master Relays, Slave Relays, and Analog Channels - TS 3.3.1 and TS 3.3.2 (Commitment 3)

Electrical Systems (AC and DC) -TS 3.8.1, TS 3.8.3, TS 3.8.4, TS 3.8.6, TS 3.8.7, TS 3.8.9, and SLC 16.8.3 (Commitment 4)

Cooling Systems (e.g., Essential Service Water (TS 3.7.7, TS 3.7.8) and Component Cooling Water (TS 3.7.6)) (Commitment 4)

Containment Spray System - TS 3.6.6 (Commitment 5)

McGuire Units 1 and 2 16.14.2-3 Revision 104

Reactor Trip Breaker and SSPS Logic Train Out of Service Commitments 16.14.2

Containment Air Return Fans - TS 3.6.11 (Commitment 5)

Ice Condenser - TS 3.6.12 and TS 3.6.13 (Commitment 5)

REFERENCES

1. NRC SER dated 12/30/08, Issuance of Amendments 248/228.

2. NRC SER dated 3/9/09, Issuance of Amendments 250/230.

3. WCAP-1 5376-P-A, Rev. 1, Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times.

4. WCAP-14333-P-A, Rev. 1, Probabilistic Risk Analysis of the RPS and ESFAS Test Times and Completion Times.

5. 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants.

McGuire Units 1 and 2 16.14.2-4 Revision 104

ML103560105

Contents

Text

Subject:

Attachment:

16.1 INTRODUCTION

16.1 INTRODUCTION

5.1 REFERENCES

5.5 REFERENCES

8.2 REFERENCES

9.2 REFERENCES

9.4 REFERENCES

9.5 REFERENCES

9.6 REFERENCES

9.7 REFERENCES

11.3 REFERENCES

11.8 REFERENCES

REFERENCES:

13.4 REFERENCES

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ML103560105

Text

Subject:

Attachment:

16.1 INTRODUCTION

16.1 INTRODUCTION

5.1 REFERENCES

5.5 REFERENCES

8.2 REFERENCES

9.2 REFERENCES

9.4 REFERENCES

9.5 REFERENCES

9.6 REFERENCES

9.7 REFERENCES

11.3 REFERENCES

11.8 REFERENCES

REFERENCES:

13.4 REFERENCES

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