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Enclosure E - DBNPS Unit 1, Technical Requirements Manual, Rev. 23
	ML22263A417
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Site:	Davis Besse
Issue date:	09/20/2022
From:	; Energy Harbor Nuclear Corp
To:	; Office of Nuclear Reactor Regulation
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ML22263A410	List: L-22-129, Submittal of the Updated Final Safety Analysis Report, Revision 34; ML22263A412; ML22263A416; ML22263A417; ML22263A418;
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	L-22-129
	Download: ML22263A417 (151)
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Enclosure E L-22-129 DBNPS, Unit No. 1 Technical Requirements Manual Rev. 23 (150 pages follow)

Davis-Besse Technical Requirements Manual

6.0 INTRODUCTION

8.6 CONTAINMENT SYSTEMS 7.0 USE AND APPLICATION 8.6.1 Combustible Gas Control - Hydrogen 7.1 Definitions Analyzers 7.2 Logical Connectors/Restoration Times 8.7 PLANT SYSTEMS 7.3 Failure to Meet a Technical Normal 8.7.1 Steam Generator Pressure/Temperature Condition (TNC) or Technical Verification Limitation Requirement (TVR) 8.7.2 Sealed Source Contamination 7.4 Frequency 8.7.3 Snubbers 8.1 REACTIVITY CONTROL SYSTEMS 8.7.4 Liquid Storage Tanks 8.1.1 Boration Systems Operating 8.7.5 Explosive Gas Mixture 8.1.2 Boration Systems Shutdown 8.7.6 Auxiliary Feedwater System 8.1.3 Rod Program 8.7.7 Motor Driven Feedwater Pump Lube Oil 8.2 Not Used Interlocks 8.3 INSTRUMENTATION 8.8 ELECTRICAL SYSTEMS 8.3.1 Reactor Protection System Instrumentation 8.8.1. AC Sources - Operating Parameters 8.8.2 SBODG Availability 8.3.2 Incore Detectors 8.9 REFUELING OPERATIONS 8.3.3 Seismic Instrumentation 8.9.1 Communications 8.3.4 Meteorological Instrumentation 8.9.2 Crane Travel - Fuel Handling Building 8.3.5 Safety Features Actuation System 8.9.3 Spent Fuel Assembly Storage Response Times 8.9.4 Fuel Handling Bridge 8.3.6 Waste Gas System Oxygen Monitoring 9.0 Not Used 8.3.7 Post Accident Monitoring (PAM) ADMINISTRATIVE CONTROLS Instrumentation 10.1 Not Used 8.3.8 EDG Loss of Power Start 10.2 ORGANIZATION 8.3.9 Not Used 10.2.1 Facility Staff 8.3.10 Source and Intermediate Range Overlap 10.3 Not Used 8.3.11 Steam and Feedwater Rupture Control System Instrumentation Parameters 10.4 PROCEDURES 8.3.12 Ultrasonic Flow Meter Instrumentation 10.4.1 Process Control Program Procedures 8.4 REACTOR COOLANT SYSTEM (RCS) 10.5 PROGRAMS AND MANUALS 8.4.1 Chemistry 10.5.1 Process Control Program (PCP) Changes 8.4.2 Pressurizer 10.5.2 In-Plant Rad Monitoring 8.4.3 Pressurizer Heater Interlock 10.6 REPORTING REQUIREMENTS 8.4.4 Reactor Coolant System Vents 10.6.1 Annual Radiological Environmental Operating Report 8.4.5 Pilot Operated Relief Valve (PORV) 10.6.2 Radioactive Effluent Release Report 8.4.6 ASME Code class 1, 2, and 3 Components 10.6.3 Core Operating Limits Report (COLR) 8.5 EMERGENCY CORE COOLING SYSTEM (ECCS) 10.6.4 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) 8.5.1 ECCS Subsystems - Operating 10.6.5 Post Accident Monitoring Report 8.5.2 ECCS Subsystems - Shutdown 10.6.6 Steam Generator Tube Inspection Report 8.5.3 Emergency Sump Debris 10.6.7 Remote Shutdown System Report Appendix A - Access Openings Required to be Closed to Ensure Shield Building Integrity

PAGE/REVISION INDEX - TECHNICAL REQUIREMENTS MANUAL Rev. Date of Rev. Date of Page No. Revision Page No. Revision a-1 23 03/10/22 8.3.5-4 3 08/12/09 a-2 23 03/10/22 8.3.5-5 3 08/12/09 TOC-i 7 10/14/10 8.3.5-6 3 08/12/09 TOC-ii 7 10/14/10 8.3.5-7 3 08/12/09 TOC-iii 7 10/14/10 8.3.5-8 3 08/12/09 6.0-1 0 12/13/08 B 8.3.5-1 0 12/13/08 6.0-2 0 12/13/08 8.3.6-1 0 12/13/08 7.1-1 0 12/13/08 8.3.6-2 0 12/13/08 7.2-1 7 10/14/10 B 8.3.6-1 0 12/13/08 7.3-1 21 01/10/20 8.3.7-1 7 10/14/10 7.4-1 0 12/13/08 8.3.7-2 7 10/14/10 8.1.1-1 7 10/14/10 B 8.3.7-1 0 12/13/08 8.1.1-2 7 10/14/10 8.3.8-1 3 08/12/09 8.1.1-3 7 10/14/10 8.3.8-2 3 08/12/09 8.1.1-4 7 10/14/10 B 8.3.8-1 0 12/13/08 B 8.1.1-1 9 09/07/11 8.3.10-1 1 12/13/08 8.1.2-1 3 08/12/09 B 8.3.10-1 1 12/13/08 8.1.2-2 3 08/12/09 8.3.11-1 3 08/12/09 8.1.2-3 3 08/12/09 8.3.11-2 3 08/12/09 B 8.1.2-1 9 09/07/11 8.3.11-3 3 08/12/09 8.1.3-1 17 04/15/16 B 8.3.11-1 0 12/13/08 8.1.3-2 17 04/15/16 8.3.12-1 0 12/13/08 B 8.1.3-1 1 12/13/08 B 8.3.12-1 0 12/13/08 8.3.1-1 3 08/12/09 8.4.1-1 13 06/14/13 8.3.1-2 3 08/12/09 8.4.1-2 13 06/14/13 8.3.1-3 3 08/12/09 8.4.1-3 13 06/14/13 B 8.3.1-1 0 12/13/08 8.4.1-4 13 06/14/13 8.3.2-1 0 12/13/08 B 8.4.1-1 0 12/13/08 8.3.2-2 0 12/13/08 8.4.2-1 0 12/13/08 B 8.3.2-1 0 12/13/08 8.4.2-2 0 12/13/08 8.3.3-1 16 04/13/16 B 8.4.2-1 0 12/13/08 8.3.3-2 16 04/13/16 8.4.3-1 0 12/13/08 8.3.3-3 16 04/13/16 8.4.3-2 0 12/13/08 8.3.3-4 16 04/13/16 B 8.4.3-1 0 12/13/08 B 8.3.3-1 16 04/13/16 8.4.4-1 7 10/14/10 8.3.4-1 4 09/14/09 8.4.4-2 7 10/14/10 8.3.4-2 4 09/14/09 B 8.4.4-1 0 12/13/08 8.3.4-3 4 09/14/09 8.4.5-1 6 07/14/10 B 8.3.4-1 4 09/14/09 8.4.5-2 6 07/14/10 8.3.5-1 3 08/12/09 B 8.4.5-1 0 12/13/08 8.3.5-2 3 08/12/09 8.4.6-1 19 10/03/18 8.3.5-3 3 08/12/09 8.4.6-2 19 10/03/18 DAVIS-BESSE NUCLEAR POWER STATION NUMBER 1 a-1 Revision 23

PAGE/REVISION INDEX - TECHNICAL REQUIREMENTS MANUAL Rev. Date of Rev. Date of Page No. Revision Page No. Revision B 8.4.6-1 19 10/03/18 8.8.2-2 7 10/14/10 8.5.1-1 0 12/13/08 B 8.8.2-1 0 12/13/08 8.5.1-2 0 12/13/08 8.9.1-1 0 12/13/08 B 8.5.1-1 0 12/13/08 B 8.9.1-1 0 12/13/08 8.5.2-1 1 12/13/08 8.9.2-1 18 07/07/17 8.5.2-2 1 12/13/08 B 8.9.2-1 18 07/07/17 B 8.5.2-1 0 12/13/08 8.9.3-1 7 10/14/10 8.5.3-1 5 03/30/10 8.9.3-2 7 10/14/10 B 8.5.3-1 5 03/30/10 B 8.9.3-1 0 12/13/08 8.6.1-1 0 12/13/08 8.9.4-1 0 12/13/08 8.6.1-2 0 12/13/08 8.9.4-2 0 12/13/08 B 8.6.1-1 0 12/13/08 B 8.9.4-1 0 12/13/08 8.7.1-1 0 12/13/08 10.2.1-1 11 12/21/12 B 8.7.1-1 0 12/13/08 10.4.1-1 5 03/30/10 8.7.2-1 21 01/10/20 10.5.1-1 5 03/30/10 8.7.2-2 21 01/10/20 10.5.2-1 5 03/30/10 B 8.7.2-1 21 01/10/20 10.6-1 5 03/30/10 8.7.3-1 15 12/16/15 10.6.1-1 5 03/30/10 8.7.3-2 15 12/16/15 10.6.2-1 5 03/30/10 8.7.3-3 15 12/16/15 10.6.3-1 5 03/30/10 8.7.3-4 15 12/16/15 10.6.4-1 5 03/30/10 8.7.3-5 15 12/16/15 10.6.5-1 5 03/30/10 8.7.3-6 15 12/16/15 10.6.6-1 5 03/30/10 B 8.7.3-1 15 12/16/15 10.6.7-1 5 03/30/10 B 8.7.3-2 15 12/16/15 A-1 0 12/13/08 8.7.4-1 0 12/13/08 8.7.4-2 0 12/13/08 B 8.7.4-1 0 12/13/08 8.7.5-1 0 12/13/08 8.7.5-2 0 12/13/08 B 8.7.5-1 0 12/13/08 8.7.6-1 8 05/05/11 8.7.6-2 8 05/05/11 B 8.7.6-1 7 10/14/10 8.7.7-1 0 12/13/08 B 8.7.7-1 0 12/13/08 8.8.1-1 23 03/10/22 8.8.1-2 23 03/10/22 B 8.8.1-1 23 03/10/22 8.8.2-1 7 10/14/10 DAVIS-BESSE NUCLEAR POWER STATION NUMBER 1 a-2 Revision 23

TABLE OF CONTENTS TECHNICAL REQUIREMENTS MANUAL

6.0 INTRODUCTION

7.0 USE AND APPLICATION ....................................................................................

7.1 Definitions................................................................................................. 7.1-1 7.2 Logical Connectors/Restoration Times..................................................... 7.2-1 7.3 Failure to Meet a Technical Normal Condition (TNC) or Technical Verification Requirement (TVR). .............................................. 7.3-1 7.4 Frequency ................................................................................................ 7.4-1 TECHNICAL NORMAL CONDITIONS (TNC) and TECHNICAL VERIFICATION REQUIREMENTS (TVR) .................................................................................................

8.1 REACTIVITY CONTROL SYSTEMS 8.1.1 Borated Water Sources Operating ........................................... 8.1.1-1 8.1.2 Borated Water Sources Shutdown........................................... 8.1.2-1 8.1.3 Rod Program ............................................................................... 8.1.3.1 8.2 Not Used 8.3 INSTRUMENTATION 8.3.1 Reactor Protection System Instrumentation Parameters.............. 8.3.1-1 8.3.2 Incore Detectors ........................................................................... 8.3.2-1 8.3.3 Seismic Instrumentation ............................................................... 8.3.3-1 8.3.4 Meteorological Instrumentation..................................................... 8.3.4-1 8.3.5 Safety Features Actuation System Response Times ................... 8.3.5-1 8.3.6 Waste Gas System Oxygen Monitoring........................................ 8.3.6-1 8.3.7 Post Accident Monitoring (PAM) Instrumentation ........................ 8.3.7-1 8.3.8 EDG Loss of Power Start.............................................................. 8.3.8-1 8.3.9 Not Used 8.3.10 Source and Intermediate Range Overlap ..................................... 8.3.10-1 8.3.11 Steam and Feedwater Rupture Control System Instrumentation Parameters .................................................................................. 8.3.11-1 8.3.12 Ultrasonic Flow Meter Instrumentation ........................................ 8.3.12-1 8.4 REACTOR COOLANT SYSTEM (RCS) 8.4.1 Chemistry...................................................................................... 8.4.1-1 8.4.2 Pressurizer.................................................................................... 8.4.2-1 8.4.3 Pressurizer Heater Interlock ........................................................ 8.4.3-1 8.4.4 Reactor Coolant System Vents..................................................... 8.4.4-1 8.4.5 Pilot Operated Relief Valve (PORV) ............................................ 8.4.5-1 8.4.6 ASME Code class 1, 2, and 3 Components ................................ 8.4.5-1 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 TOC-i Revision 7

TABLE OF CONTENTS TECHNICAL REQUIREMENTS MANUAL 8.5 EMERGENCY CORE COOLING SYSTEM (ECCS) 8.5.1 ECCS Subsystems - Operating ................................................... 8.5.1-1 8.5.2 ECCS Subsystems - Shutdown ................................................... 8.5.2-1 8.5.3 Emergency Sump Debris ............................................................. 8.5.3-1 8.6 CONTAINMENT SYSTEMS 8.6.1 Combustible Gas Control - Hydrogen Analyzers ......................... 8.6.1-1 8.7 PLANT SYSTEMS 8.7.1 Steam Generator Pressure/Temperature Limitation .................... 8.7.1-1 8.7.2 Sealed Source Contamination ..................................................... 8.7.2-1 8.7.3 Snubbers ..................................................................................... 8.7.3-1 8.7.4 Liquid Storage Tanks ................................................................... 8.7.4-1 8.7.5 Explosive Gas Mixture ................................................................. 8.7.5-1 8.7.6 Auxiliary Feedwater System ........................................................ 8.7.6-1 8.7.7 Motor Driven Feedwater Pump Lube Oil Interlocks ..................... 8.7.7-1 8.8 ELECTRICAL SYSTEMS 8.8.1. AC Sources - Operating ............................................................. 8.8.1-1 8.8.2 SBODG Availability....................................................................... 8.8.2-1 8.9 REFUELING OPERATIONS 8.9.1 Communications ........................................................................... 8.9.1-1 8.9.2 Crane Travel - Fuel Handling Building ......................................... 8.9.2-1 8.9.3 Spent Fuel Assembly Storage ...................................................... 8.9.3-1 8.9.4 Fuel Handling Bridge ................................................................... 8.9.4-1 9.0 Not Used ADMINISTRATIVE CONTROLS 10.1 Not Used 10.2 ORGANIZATION 10.2.1 Facility Staff ................................................................................. 10.2.1-1 10.3 Not Used 10.4 PROCEDURES 10.4.1 Process Control Program Procedures .......................................... 10.4.1-1 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 TOC-ii Revision 7

TABLE OF CONTENTS TECHNICAL REQUIREMENTS MANUAL 10.5 PROGRAMS AND MANUALS 10.5.1 Process Control Program (PCP) Changes ................................... 10.5.1-1 10.5.2 In-Plant Rad Monitoring ............................................................... 10.5.2-1 10.6 REPORTING REQUIREMENTS 10.6.1 Annual Radiological Environmental Operating Report ................. 10.6.1-1 10.6.2 Radioactive Effluent Release Report ........................................... 10.6.2-1 10.6.3 Core Operating Limits Report (COLR).......................................... 10.6.3-1 10.6.4 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) ................................................................... 10.6.4-1 10.6.5 Post Accident Monitoring Report ................................................. 10.6.5-1 10.6.6 Steam Generator Tube Inspection Report ................................... 10.6.6-1 10.6.7 Remote Shutdown System Report .............................................. 10.6.7-1 Appendix A Access Openings Required to be Closed to Ensure Shield Building Integrity ............................................................... A-1 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 TOC-iii Revision 7

TRM INTRODUCTION TRM 6.0

6.0 INTRODUCTION

6.1 BACKGROUND

Based on the NRC's Final Policy Statement on Technical Specification Improvements for nuclear power plants, and 10 CFR 50.36, "Technical Specifications, " as amended in the Final Rule published in the Federal Register dated July 13, 1995, certain requirements may be relocated from the Operating License Technical Specifications to other licensee-controlled documents. In an effort to centralize the requirements relocated from the Technical Specifications and to ensure the necessary administrative controls are applied to these requirements, these requirements have been relocated as "Technical Requirements" into the Davis-Besse Technical Requirements Manual (TRM).

The TRM provides one location for relocated items in a consistent format. The Technical Requirements are formatted in a manner consistent with NRC Regulatory Issue Summary 2005-20: Revision To Guidance Formerly Contained in NRC Generic Letter 91-18, Information to Licensees Regarding Two NRC Inspection Manual Sections On Resolution of Degraded and Nonconforming Conditions and on Operability. Although many of the terms defined in the Technical Specifications apply within the TRM, the TRM contains additional Definitions which are specific to the TRM and not defined in the Technical Specification Definitions.

6.2 REGULATORY STATUS/REQUIREMENTS The requirements in the TRM are part of the licensing basis for the Davis-Besse Nuclear Power Station. Furthermore, the TRM is incorporated by reference in the Updated Safety Analysis Report (USAR) and is considered to be part of the USAR. Violations of the TRM requirements should be documented by the corrective action process. Deviations from the TRM will be screened for reportability in accordance with the corrective action process.

These controls are in place because the purpose of relocating the requirements for Technical Specifications is not to reduce the level of control on the items, but to provide flexibility for change under 10 CFR 50.59, Changes, Tests and Experiments.

Technical Requirements Manual Section 10.6, Reporting Requirements has been developed to provide a central location for various Technical Specification reports. These reports are not controlled or revised under the change process for the Technical Requirements Manual. The reports contained in Section 10.6 are revised and issued as required by Technical Specification Section 5.6.

6.3 CHANGES TO THE TRM Design modifications, procedure changes, license amendments, etc. have the potential to affect the TRM. If this occurs, the initiating department must follow the administrative controls in NOP-LP-4008, "Licensing Documents Change Process." This program requires that the TRM's Technical Requirements be considered in a manner similar to the USAR when evaluating changes. Changes to the TRM will be reported, as a minimum, to the NRC as part of the USAR update submittal in accordance with 10 CFR 50.71(e). Related 10 CFR 50.59 evaluations will be reported as part of the 10 CFR 50.59(d) report to the NRC.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 6.0-1 Revision 0

TRM INTRODUCTION TRM 6.0 6.4 TECHNICAL VERIFICATION REQUIREMENTS Each Verification Requirement shall be performed within the specified time interval with a maximum allowable extension not to exceed 25% of the specified Technical Verification Requirement interval.

The provisions of this requirement provide allowable tolerances for performing technical verification activities beyond those specified in the nominal Technical Verification Requirement interval. These tolerances are necessary to provide operational flexibility because of scheduling and performance considerations. The phrase "at least" associated with a Technical Verification Requirement frequency does not negate this allowable tolerance value and permits the performance of more frequent verification activities.

The allowable tolerance for performing verification activities is sufficiently restrictive to ensure that the reliability associated with the verification activity is not significantly degraded beyond that obtained from the nominal specified interval. It is not intended that the allowable tolerance be used as a convenience to repeatedly schedule the performance of verification requirements at the allowable tolerance limit.

The allowable tolerance for performing verification activities also provides flexibility to accommodate the length of a fuel cycle for Technical Verification Requirements that are specified to be performed at least once each 24 Months. It is the intent that 24 Month verification requirements be performed in a MODE consistent with safe plant operation.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 6.0-2 Revision 0

Definitions TRM 7.1 7.0 USE AND APPLICATION 7.1 Definitions

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

1. Definitions are defined in Section 1.1 of the Technical Specifications and are applicable throughout the Technical Requirements Manual (TRM) and Bases. Only definitions specific to the TRM will be defined in this section.

2. The defined terms of this section and the Technical Specifications (TS) appear in capitalized type and are applicable throughout the TRM and the TRM Bases.

3. When a term is defined in both the TS and the TRM, TRM definition takes precedence within the TRM and the TRM Bases.

Term Definition FUNCTIONAL A structure, system or component (SSC), shall be FUNCTIONALITY FUNCTIONAL or have FUNCTIONALITY when it is capable of performing its specified function(s) as set forth in the Current License Basis. FUNCTIONALITY does not apply to specified safety functions, but does apply to the ability of non-TS SSCs to perform other specified functions that have a necessary support function.

TECHNICAL NORMAL Specify minimum requirements for ensuring safe operation of CONDITIONS (TNC) the Unit. The Contingency Measures associated with a TNC state Nonconformances that typically describe the ways in which the requirements of the TNC can fail to be met.

Specified with each stated Nonconformance are Contingency Measures and Restoration Time(s).

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 7.1-1 Revision 0

Logical Connectors/Restoration Times TRM 7.2 7.0 USE AND APPLICATION 7.2 Logical Connectors/Restoration Times Logical Connectors are discussed in Section 1.2 of the Technical Specifications and are applicable throughout the Technical Requirements Manual and Bases.

Completion Times are discussed in Section 1.3 of the Technical Specifications and are applicable throughout the Technical Requirements Manual and Bases. Completion Times in the Technical Specifications are equivalent to Restoration Times in the Technical Requirements Manual.

When "Immediately" is used as a Restoration Time, the Contingency Measure should be pursued without delay in a controlled manner.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 7.2-1 Revision 7

Failure to Meet TNC or TVR TRM 7.3 7.0 USE AND APPLICATION 7.3 Failure to Meet a Technical Normal Condition (TNC) or Technical Verification Requirement (TVR).

When a TNC and the associated Contingency Measures are not met, an associated Contingency Measure is not provided, or if directed by the associated Contingency Measures, action shall be initiated immediately to communicate the situation to the Shift Manager and document the condition in accordance with the company corrective action program. The safety significance of the condition shall be evaluated per NOP-OP-1009 Operability Determinations and Functionality Assessments and appropriate corrective actions initiated, within the time frame determined by the Shift Manager that shall not exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months from the time of entry into TRM 7.3. The time frame for completion of the corrective actions shall be commensurate with the safety significance of the condition, consistent with the guidance of NOP-OP-1009.

Where corrective measures are completed that permit operation in accordance with the TNC or Contingency Measures, completion of the actions required by TRM 7.3 is not required.

When it is discovered that a TVR frequency (including the 1.25 times extension) has not been met, the equipment subject to the TVR is in a nonconforming condition. In this situation, a Condition Report shall be initiated and, if indicated, determination to evaluate the impact on plant safety shall be performed in a timely fashion and in accordance with plant procedures.

Actions should be taken to restore conformance with the TNCs / TVRs in a timely fashion.

If equipment has been removed from service or declared nonfunctional, it may be returned to service under administrative control to perform testing required to demonstrate its functionality.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 7.3-1 Revision 21

Frequency TRM 7.4 7.0 USE AND APPLICATION 7.4 Frequency Frequency is discussed in Section 1.4 of the Technical Specifications and is applicable throughout the Technical Requirements Manual and Bases, with the exception that Technical Verification Requirements are used in the place of Surveillance Requirements.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 7.4-1 Revision 0

Boration Systems - Operating TRM 8.1.1 8.1 REACTIVITY CONTROL SYSTEMS 8.1.1 Boration Systems - Operating TECHNICAL NORMAL CONDITIONS TNC 8.1.1 The Boration Systems shall be FUNCTIONAL consisting of the following:

a. A flow path from the concentrated FUNCTIONAL boric acid addition system (BAAS) via a FUNCTIONAL boric acid pump and a FUNCTIONAL makeup pump to the Reactor Coolant System (RCS);

AND

b. A flow path from the OPERABLE borated water storage tank via a FUNCTIONAL makeup pump to the RCS System.

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

Separate Makeup pumps are required to be FUNCTIONAL in MODES 1, 2 and 3, and in MODE 4 when RCS pressure is 150 psig.

A FUNCTIONAL decay heat removal (DHR) pump may be used in place of a makeup pump in MODE 4 when RCS pressure is < 150 psig.

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

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A One Boron System A.1 Restore the nonfunctional 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months flow path Nonfunctional. flow path to FUNCTIONAL status.

B. Contingency Measures B.1 Initiate action to evaluate Immediately and associated failure to meet TNC per Restoration Time of TRM Section 7.3.

Nonconformance A not met.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.1.1-1 Revision 7

Boration Systems - Operating TRM 8.1.1 CONTINGENCY MEASURES (continued)

NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME C. The boron injection flow C.1 Restore the nonfunctional 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months path from the borated flow path to FUNCTIONAL water storage tank is status.

Nonfunctional.

D. Contingency Measures D.1 Be in MODE 3 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and associated Restoration Time of AND Nonconformance C not met. D.2 Be in MODE 5 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.1.1-2 Revision 7

Boration Systems - Operating TRM 8.1.1 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.1.1.1 Verify BAAS solution temperature is > 105F. 7 days

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

8.1.1.2 If the 7 day verification falls during transfers of makeup water or dilute boron solutions (fluid source concentration of less than 5000 ppmB), the verification period may be extended up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the addition of dilute boron solution has been stopped for a period of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Verify the pipe temperature of the heat traced portion of 7 days the boron injection flow path from the concentrated boric acid storage system is 105°F.

8.1.1.3 Verify borated water volume of BAAS is in accordance 31 days with TRM Figure 8.1.1-1.

8.1.1.4 Verify the boron concentration in BAAS is > 7,875 ppm 31 days and < 13,125 ppm.

8.1.1.5 Verify each valve (manual, power operated or 31 days automatic) in the boron injection flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.1.1-3 Revision 7

Boration Systems - Operating TRM 8.1.1 Figure 8.1.1-1 Boric Acid Addition System Volume vs Boric Acid Concentration in Modes 1-4 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.1.1-4 Revision 7

Boration Systems - Operating TRM B 8.1.1 BASES 8.1.1 Boration Systems - Operating The boron injection system ensures that negative reactivity control is available during each mode of facility operation.

The boric acid addition system (BAAS) is one of the borated water sources for the boron injection system. The maximum boration capability requirement occurs from full power equilibrium xenon conditions and requires the equivalent of 12,200 gallons of 7875 ppm borated water from the BAAS, or borated water from the BWST at a volume and concentration as specified in Technical Specification 3.5.4. The minimum value for the BAAS of 12,200 gallons at a concentration of 7875 ppm boron is a lower value than that shown in TRM Figure 8.1.1-1 because the Bases value is the minimum required actual value, whereas TRM Figure 8.1.1-1 shows the minimum indicated value, which was conservatively increased to account for instrument and chemical analysis tolerance.

The components required for the boron injection function, depending on operating conditions, include (1) borated water sources, (2) makeup or DHR pumps, (3) separate flow paths, (4) boric acid pumps, and (5) associated heat tracing systems.

With the RCS average temperature above 200°F, a minimum of two separate and redundant boron injection systems are provided to ensure single functional capability in the event an assumed failure renders one of the systems nonfunctional. Allowable out-of-service periods ensure that minor component repair or corrective action may be completed without undue risk to overall facility safety from injection system failures during the repair period.

The boration capability of either system is sufficient to provide a SHUTDOWN MARGIN from all operating conditions of 1.0% k/k after xenon decay and cooldown to 200°F. The available borated water volume range and boron concentration range for the Boric Acid Addition System (BAAS), required to support this boration capability, are provided in the Updated Safety Analysis Report. The requirements relative to the Borated Water Storage Tank (BWST) are provided in Technical Specification 3.5.4.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.1.1-1 Revision 9

Boration Systems - Shutdown TRM 8.1.2 8.1 REACTIVITY CONTROL SYSTEMS 8.1.2 Boration Systems - Shutdown TECHNICAL NORMAL CONDITIONS TNC 8.1.2 The Boration Systems shall be FUNCTIONAL consisting of the following:

a. A flow path from the FUNCTIONAL boric acid addition system (BAAS) via a FUNCTIONAL boric acid pump and a FUNCTIONAL makeup pump to the Reactor Coolant System (RCS); or

b. A flow path from the borated water storage tank via a FUNCTIONAL makeup pump to the RCS.

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

The makeup pump is only required to be FUNCTIONAL in MODE 5 with the RCS pressure 150 psig.

A FUNCTIONAL decay heat removal (DHR) pump may be used in place of a makeup pump when RCS pressure is < 150 psig APPLICABILITY: MODES 5 and 6 CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Required Boration A.1 Suspend movement of Immediately System flow path irradiated fuel assemblies.

Nonfunctional.

AND A.2 Suspend operations involving Immediately positive reactivity additions.

AND A.3 Initiate actions to restore Immediately Boration System flow path to functional status.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.1.2-1 Revision 3

Boration Systems - Shutdown TRM 8.1.2 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.1.2.1 Verify BWST solution temperature > 35°F. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if the BWST is used as a borated water source and the outside air temperature is < 35°F 8.1.2.2 Verify boron concentration in BWST is > 2600 ppm. 7 days, if the BWST is used as a borated water source 8.1.2.3 -----------------------------NOTE------------------------------------

If the 7 day verification falls during transfers of makeup water or dilute boron solutions (fluid source concentration of less than 5000 ppmB), the verification period may be extended up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the addition of dilute boron solution has been stopped for a period of at least 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

Verify the pipe temperature of the heat traced portion of 7 days, when a flow the boron injection flow path is 105°F. path from the concentrated boric acid storage system is used 8.1.2.4 Verify BAAS solution temperature is > 105°F. 7 days, if the BAAS is used as borated source 8.1.2.5 Verify BWST water volume is > 3000 gallons 7 days, if the BWST is used as a borated water source 8.1.2.6 Verify available borated water volume > 900 gallons. 31 days, if the BAAS is used as borated source DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.1.2-2 Revision 3

Boration Systems - Shutdown TRM 8.1.2 TECHNICAL VERIFICATION REQUIREMENTS (continued)

TVR VERIFICATION FREQUENCY 8.1.2.7 Verify the boron concentration in BAAS is > 7,875 ppm 31 days, if the BAAS is and < 13,125 ppm. used as borated source 8.1.2.8 Verify each valve (manual, power operated or 31 days automatic) in the boron injection flow path that is not locked, sealed or otherwise secured in position, is in its correct position.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.1.2-3 Revision 3

Boration Systems - Shutdown TRM B 8.1.2 BASES 8.1.2 Boration Systems - Shutdown A description of the boration system and component requirements is provided in the Bases for TRM 8.1.1, Boration Systems - Operating.

With the RCS temperature below 200°F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting movement of irradiated fuel assemblies and positive reactivity changes in the event the single injection system becomes nonfunctional.

The boration capability required below 200°F is sufficient to provide a SHUTDOWN MARGIN of 1% k/k after xenon decay and cooldown from 200°F to 70°F. This condition requires either 900 gallons of 7875 ppm borated water from the BAAS or 3,000 gallons of 2600 ppm borated water from BWST.

The FUNCTIONALITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.1.2-1 Revision 9

Rod Program TRM 8.1.3 8.1 REACTIVITY CONTROL SYSTEMS 8.1.3 Rod Program TECHNICAL NORMAL CONDITIONS TNC 8.1.3 Each control rod assembly (safety, regulating and APSR) shall be programmed to operate in the core location and rod group specified in the CORE OPERATING LIMITS REPORT.

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

During the performance of PHYSICS TESTS in MODE 1, the requirements of TNC 8.1.3 may be suspended, if the requirements of Technical Specification 3.1.8 are in effect.

During the performance of PHYSICS TESTS, in MODE 2, the requirements of TNC 8.1.3 may be suspended, if the requirements of Technical Specification 3.1.9 are in effect.

APPLICABILITY: MODES 1 and 2.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Any control rod A.1 Be in MODE 3. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months assembly not programmed to operate as specified above.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.1.3.1 Verify all control rod assemblies are programmed to After the Control Rod operate in the specified core location and rod group by Drive Control System selection and actuation from the control room and (CRDCS) programming verification of movement of the proper rod as indicated is complete, or by both the absolute and relative position indicators. following maintenance or reprogramming within the CRDCS that could affect the rod or group assignment.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.1.3-1 Revision 17

Rod Program TRM 8.1.3 TECHNICAL VERIFICATION REQUIREMENTS (continued)

TVR VERIFICATION FREQUENCY 8.1.3.2 Verify that the specifically affected individual control rod After maintenance, assemblies are programmed to operate in the specified test, reconnection or core location and rod group by selection and actuation modification of power from the control room and verification of movement of or instrumentation the proper rod as indicated by both the absolute and cables from the control relative position indicators. rod drive control system to the control rod drive 8.1.3.3 Verify each control rod assembly cable has been After disconnection of properly matched and reconnected to the specified control rod assembly control rod drive. cable DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.1.3-2 Revision 17

Rod Program TRM 8.1.3 BASES None DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.1.3-1 Revision 1

Reactor Protection System Instrumentation Parameters TRM 8.3.1 8.3 INSTRUMENTATION 8.3.1 Reactor Protection System Instrumentation Parameters TECHNICAL NORMAL CONDITIONS TNC 8.3.1 The Reactor Protection System (RPS) instrumentation RPS RESPONSE TIMES listed in TRM Table 8.3.1-1 shall be maintained in the manner specified in Technical Specification 3.3.1.

AND The RPS instrumentation RPS SETPOINTS listed in TRM Table 8.3.1-2 shall be maintained in the manner specified in Technical Specification

3.3.1. APPLICABILITY

As specified in Technical Specification Table 3.3.1-1.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME NONE TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY NONE DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.1-1 Revision 3

Reactor Protection System Instrumentation Parameters TRM 8.3.1 Table 8.3.1-1 (page 1 of 1)

Reactor Protection System Instrumentation Response Times FUNCTION RESPONSE TIMES (b)

(seconds)

1. High Flux (a) < 0.266

2. RC High Temperature Not Applicable

3. RC High Pressure < 0.341

4. RC Low Pressure < 0.341

5. RC Pressure - Temperature - Constant Not applicable Temperature

6. Containment High Pressure Not applicable

7. High Flux / Number of Reactor Coolant < 0.631 Pumps On (a)(c)

8. Flux - Flux - Flow (a)

a. Variable Flow < 1.77

b. Constant Flow < 0.266 (a) Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the channel shall be measured from detector output or input of first electronic component in channel.

(b) Including sensor (except as noted), RPS instrument delay and the breaker delay.

(c) A delay time has been assumed for the Reactor Coolant Pump monitor in the determination of the response time of the High Flux / Number of Reactor Coolant Pumps On functional unit.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.1-2 Revision 3

Reactor Protection System Instrumentation Parameters TRM 8.3.1 Table 8.3.1-2 (page 1 of 1)

Reactor Protection System Instrumentation Trip Setpoints FUNCTION (Note 1) Setpoint 1.a High Flux Four Pump Limiting Trip Setpoint (Ultrasonic) 104.5875% FP Four Pump Limiting Trip Setpoint (Venturi) 102.9875% FP Three Pump Limiting Trip Setpoint (Ultrasonic or Venturi) 80.2875% FP Four Pump Normal Trip Setpoint (Note 2) (Ultrasonic) 104.5% FP Four Pump Normal Trip Setpoint (Note 2) (Venturi) 102.9% FP Three Pump Normal Trip Setpoint (Note 2) (Ultrasonic or 80.1% FP Venturi)

As-Found Acceptance Criteria (Note 3) [Previous As-Left - Current As-Found]

NTSP < 0.3125% Power As-Left Acceptance Criteria (Note 3) NTSP +/- 0.0875% Power

5. RC Pressure - Temperature Limiting Trip Setpoint (LTSP) 16.25 Tout - 7886.602 psig Nominal Trip Setpoint (NTSP) (Note 2) 16.25 Tout - 7885.5 psig As-Found Acceptance Criteria Band (Note 3) [Previous As-Left - Current As-found]

11.2 psi As-Left Setpoint Tolerance Band (Note 3) NTSP +/- 6.0 psi Note 1 Setpoint information is not provided for Tech Spec Table 3.3.1-1 Functions 2, 3, 4, 6, 7, 8, 9.

Note 2 Nominal Trip Setpoint is a value more conservative than the Limiting Trip Setpoint.

Conservative margin is added (subtracted) to the Limiting Trip Setpoint to generate the Nominal Trip Setpoint.

Note 3 Compliance with the As-Found Acceptance Criteria Band is determined by taking the absolute value of the difference between the As-Left value from the previous surveillance test and the As-Found value from the current surveillance test. This must be evaluated separate from compliance with the Technical Specification Allowable Value. (Applicable to Functional Unit 5 only).

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.1-3 Revision 3

Reactor Protection System Instrumentation Parameters TRM B 8.3.1 BASES 8.3.1 Reactor Protection System Instrumentation Parameters The measurement of response time at the specified frequencies provides assurance that the RPS action function associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

The measurement of trip setpoints at the specified frequencies provides assurance that the RPS function associated with each channel is in conformance with the trip requirements assumed in the safety analysis. The trip setpoint is established by addition (or subtraction depending on the conservative direction) of instrument uncertainties to the Analytical Limit (value used in the safety analysis).

This assurance is based on compliance with the methodology for establishment of nuclear safety related setpoints. The setpoint and acceptance criteria are established in compliance with Regulatory Guide 1.105, Setpoints for Safety-Related Instrumentation. The setpoint and acceptance criteria are established using Method 1 or Method 2 from Section 7 of ISA RP67.04.02-2000, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation. Reset of the setpoint within the assumed As-Left Setpoint Tolerance Band will provide assurance that the channel is in compliance with the methodology and the calculation establishing the setpoint. Additional assurance is provided by repeated, successful setpoint verification at the prescribed surveillance frequency.

Setpoints found outside of the prescribed values require evaluation to ensure the equipment is able to perform within the calculational values and to determine if the equipment is able to perform the intended protective function.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.1-1 Revision 0

Incore Detectors TRM 8.3.2 8.3 INSTRUMENTATION 8.3.2 Incore Detectors TECHNICAL NORMAL CONDITIONS TNC 8.3.2 The Incore Detection System shall be FUNCTIONAL as specified below:

a. > 75% of the Symmetric Incore Detectors in each core quadrant shall be FUNCTIONAL for QUADRANT POWER TILT measurements.

b. > 75% of all incore detectors in each core quadrant shall be FUNCTIONAL for AXIAL POWER IMBALANCE, FNH and FQ measurements.

APPLICABILITY: When the Incore Monitoring System is used for measurement of:

a. AXIAL POWER IMBALANCE;

b. QUADRANT POWER TILT;

c. FNH; or

d. FQ .

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Less than the specified A.1 Do not use the Incore Immediately number of incore Monitoring System for the detectors FUNCTIONAL. above applicable measurement.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.2-1 Revision 0

Incore Detectors TRM 8.3.2 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.3.2.1 Perform CHANNEL CHECK. Within 7 days prior to its use for measurement of the AXIAL POWER IMBALANCE or the QUADRANT POWER TILT AND 7 days thereafter 8.3.2.2 ------------------------------NOTE-----------------------------------

Channel calibration does not include neutron detectors.

Perform CHANNEL CALIBRATION. 24 months DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.2-2 Revision 0

Incore Detectors TRM B 8.3.2 BASES 8.3.2 Incore Detectors The FUNCTIONALITY of the incore detectors ensures that the measurements obtained from the Incore Monitoring System accurately represent the spatial neutron flux distribution of the reactor core.

Technical Specification 3.2.4, Quadrant Power Tilt becomes applicable in plant MODE 1 above 20% of Rated Thermal Power. This requires a determination of Quadrant Power Tilt at least once every 7 days, under Technical Specification SR 3.2.4.1. The channel check of the incore detector system (TRM TVR 8.3.2.1) must be performed within 7 days prior to this initial performance of TS SR 3.2.4.1, without the benefit of TRM 6.4.

Therefore, assuming the continued applicability of TS 3.2.4, for each subsequent performance of TRM TVR 8.3.2.1, the 25 percent allowable verification test interval extension of TRM 6.4 may be applied.

REFERENCES

1. NRC Letter Log No. 5382, dated December 3, 1998, to Centerior Service Company DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.2-1 Revision 0

Seismic Instrumentation TRM 8.3.3 8.3 INSTRUMENTATION 8.3.3 Seismic Instrumentation TECHNICAL NORMAL CONDITIONS TNC 8.3.3 The seismic monitoring instrumentation for each Location in TRM Table 8.3.3-1 shall be FUNCTIONAL.

APPLICABILITY: At all times.

CONTINGENCY MEASURES

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

Separate Nonconformance entry is allowed for each channel.

NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. -------------NOTE------------- A.1 Restore instrument to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Contingency Measures FUNCTIONAL status.

A.2, A.3, and A.4 shall be completed whenever AND Nonconformance A is entered. A.2 Perform TVR 8.3.3.3 and 5 days

TVR 8.3.3.4 One or more seismic AND monitoring instrumentation A.3 Analyze data retrieved from 10 days Nonfunctional due to instrument to determine the being actuated during a magnitude of the vibratory seismic event. ground motion.

AND A.4 Prepare and submit a special 10 days report to the Commission describing the magnitude, frequency, spectrum, and resultant effect upon the facility features important to safety.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.3-1 Revision 16

Seismic Instrumentation TRM 8.3.3 CONTINGENCY MEASURES (continued)

NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME B. One or more seismic B.1 Restore instrument to 30 days monitoring FUNCTIONAL status.

instrumentations Nonfunctional for reasons other than Nonconformance A.

C. Contingency Measures C.1 Initiate action to evaluate Immediately and associated failure to meet TNC per TRM Restoration Time of Section 7.3.

Nonconformance B not met.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.3-2 Revision 16

Seismic Instrumentation TRM 8.3.3 TECHNICAL VERIFICATION REQUIREMENTS

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

Refer to TRM Table 8.3.3-1 to determine which TVRs apply for each seismic monitoring instrumentation.

TVR VERIFICATION FREQUENCY 8.3.3.1 ---------------------------------NOTE------------------------------ 31 days Channel Check for Instrument 1.d shall include seismic trigger and cabinet room indication of trigger.

Perform CHANNEL CHECK for Instrument 1.

8.3.3.2 Perform CHANNEL FUNCTIONAL TEST for 184 days Instrument 1.

8.3.3.3 Perform CHANNEL CALIBRATION for 18 months Instruments 1.c, 1.d, and 2.

8.3.3.4 Perform CHANNEL CALIBRATION for 24 months Instruments 1.a and 1.b.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.3-3 Revision 16

Seismic Instrumentation TRM 8.3.3 Table 8.3.3-1 (page 1 of 1)

Seismic Monitoring Instrumentation INSTRUMENTS AND SENSOR LOCATIONS MINIMUM MEASUREMENT RANGE INSTRUMENT FUNCTIONAL

1. Strong Motion Triaxial Accelerographs

a. Containment Concrete Foundation, Elev. 565 (inside containment) 1 + 1g

b. Containment Interior Secondary Shield Wall Elev. 653 (inside containment) 1 + 1g

c. Auxiliary Building Basement Floor, Elev.

545 (outside containment) 1 + 1g

d. Station site - Minimum of 300 feet from containment vessel within the site 1(a) + 1g(b) boundary (outside containment)

2. Peak Recording Accelerometers

a. Shield Building Top, Minimum 1 + 1g Elev. 812

b. Auxiliary Building Roof, Elev. 660 1 + 1g

c. Control Room, Elev. 623 1 + 1g (a) Includes Seismic Trigger function with cabinet room indication (b) Seismic Trigger function characteristics:

a. Minimum Frequency Response Range: 1 - 10 Hz

b. Actuation Range: 0.005g - 0.02g DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.3-4 Revision 16

Seismic Instrumentation TRM B 8.3.3 BASES 8.3.3 Seismic Instrumentation The FUNCTIONALITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event so that the response of those features important to safety may be evaluated. This capability is required to permit comparison of the measured response to that used in the design basis for the facility. Item 1.d on Table 8.3.3-1 is credited with the required seismic trigger function because it is the only free-field sensor.

TVR 8.3.3.1 includes a note indicating the extent to which the seismic trigger and control room indication functions are required to be verified during channel checks of the strong motion triaxial accelerographs. Although all of these instruments are capable of triggering the system, a check of the trigger function is only required for instrument 1.d and is optional for the others.

Check of the trigger function should confirm that recording is initiated at all recorders and that appropriate indication of the triggered condition occurs in the cabinet room.

This instrumentation is consistent with the recommendations of Regulatory Guide 1.12 "Instrumentation for Earthquakes," April 1974.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.3-1 Revision 16

Meteorological Instrumentation TRM 8.3.4 8.3 INSTRUMENTATION 8.3.4 Meteorological Instrumentation TECHNICAL NORMAL CONDITIONS TNC 8.3.4 The meteorological monitoring instrumentation channels for each function shown in TRM Table 8.3.4-1 shall be FUNCTIONAL.

APPLICABILITY: At all times.

CONTINGENCY MEASURES

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

Separate entry is allowed for each instrumentation channel.

NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. One or more A.1 Restore channels to 7 days meteorological monitoring FUNCTIONAL status.

instrumentation channels Nonfunctional.

B. Contingency Measures B.1 Initiate action to evaluate Immediately and associated failure to meet TNC per TRM Restoration Time of Section 7.3.

Nonconformance A not met.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.4-1 Revision 4

Meteorological Instrumentation TRM 8.3.4 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.3.4.1 Perform CHANNEL CHECK of each required channel. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 8.3.4.2 ---------------------------------NOTE------------------------------

Wind direction and wind speed sensors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION of each required 184 days channel.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.4-2 Revision 4

Meteorological Instrumentation TRM 8.3.4 Table 8.3.4-1 (page 1 of 1)

Meteorological Monitoring Instrumentation FUNCTION LOCATION REQUIRED CHANNELS

1. Wind Speed

a. Nominal Elev. 612 1

b. Nominal Elev. 827 1

2. Wind Direction

a. Nominal Elev. 612 1

b. Nominal Elev. 827 1

3. Air Temperature - Delta T

a. Nominal Elev. 827 - 612 1 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.4-3 Revision 4

Meteorological Instrumentation TRM B 8.3.4 BASES 8.3.4 Meteorological Instrumentation The FUNCTIONALITY of the meteorological instrumentation ensures that sufficient meteorological data is available for estimating potential radiation doses to the public as a result of routine or accidental release or radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public. This instrumentation is consistent with the recommendations of Regulatory Guide 1.23 "Meteorological Programs in Support of Nuclear Power Plants," September 1980.

TVR 8.3.4.2 includes a Note. The Note indicates that the sensors for wind speed and wind direction are excluded from CHANNEL CALIBRATION. This note is necessary because the sensors are pre-calibrated by a vendor in offsite wind tunnel facilities. The vendor certified calibrations are valid for 5 years after the date of performance when stored in a temperature controlled storage, and are valid for 1 year in service.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.4-1 Revision 4

Safety Features Actuation System Response Times TRM 8.3.5 8.3 INSTRUMENTATION 8.3.5 Safety Features Actuation System Response Times TECHNICAL NORMAL CONDITIONS TNC 8.3.5 The Safety Features Actuation System (SFAS) instrumentation RESPONSE TIMES listed in TRM Table 8.3.5-1 shall be maintained in the manner specified in Technical Specification 3.3.5.

APPLICABILITY: As specified in Technical Specification Table 3.3.5-1.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME NONE TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY NONE DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.5-1 Revision 3

Safety Features Actuation System Response Times TRM 8.3.5 Table 8.3.5-1 (Page 1 of 7)

Safety Features System Instrumentation Response Times INITIATING SIGNAL AND FUNCTION RESPONSE TIME (seconds) 1 Manual

a. Fans

1. Emergency Vent Fan NA

2. Containment Cooler Fan NA

b. HV & AC Isolation Valves

1. ECCS Room NA

2. Emergency Ventilation NA

3. Containment Air Sample NA

4. Penetration Room Purge NA

c. Control Room HV & AC Units NA

d. High Pressure Injection

1. High Pressure Injection Pumps NA

2. High Pressure Injection Valves NA

e. Component Cooling Water

1. Component Cooling Water Pumps NA

2. Component Cooling Aux. Equip. Inlet Valves NA

3. Component Cooling to Makeup Pump NA Header Inlet Valve

f. Service Water System

1. Service Water Pumps NA

2. Service Water From Component Cooling NA Heat Exchanger Isolation Valves

g. Containment Spray Isolation Valves NA

h. Emergency Diesel Generator NA

i. Containment Isolation Valves

1. Vacuum Relief NA

2. Normal Sump NA

3. RCS Letdown Delay Coil Outlet NA

4. RCS Letdown High Temperature NA (1) Diesel generator starting and sequence loading delays included when applicable. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.5-2 Revision 3

Safety Features Actuation System Response Times TRM 8.3.5 Table 8.3.5-1 (page 2 of 7)

Safety Features System Instrumentation Response Times INITIATING SIGNAL AND FUNCTION RESPONSE TIME (seconds)

i. Containment Isolation Valves (cont'd)

5. Pressurizer Sample NA

6. Service Water to Cooling Water NA

7. Vent Header NA

8. Drain Tank NA

9. Core Flood Tank Vent NA

10. Core Flood Tank Fill NA

11. Steam Generator Sample NA 12 Quench Tank NA

13. Emergency Sump NA

14. RCP Seal Return NA

15. Air Systems NA

16. N2 System NA

17. Quench Tank Sample NA

18. RCP Seal Inlet NA

19. Core Flood Tank Sample NA

20. RCP Standpipe Demin Water NA Supply

21. Containment H2 Dilution Inlet NA

22. Containment H2 Dilution Outlet NA

j. BWST Outlet Valves NA

k. Low Pressure Injection

1. Decay Heat Pumps NA

2. Low Pressure Injection Valves NA

3. Decay Heat Pump Suction Valves NA

4. Decay Heat Cooler Outlet Valves NA

5. Decay Heat Cooler Bypass Valves NA

l. Containment Spray Pump NA (1) Diesel generator starting and sequence loading delays included when applicable. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.5-3 Revision 3

Safety Features Actuation System Response Times TRM 8.3.5 Table 8.3.5-1 (page 3 of 7)

Safety Features System Instrumentation Response Times INITIATING SIGNAL AND FUNCTION RESPONSE TIME (seconds)

m. Component Cooling Isolation Valves

1. Inlet to Containment NA

2. Outlet from Containment NA

3. Inlet to CRDM's NA

4. CRDM Booster Pump Suction NA

5. Component Cooling from Decay Heat NA Coolers

2. Containment Pressure - High

a. Fans

1. Emergency Vent Fans 25(1)

2. Containment Cooler Fans 45(1)

b. HV & AC Isolation Valves

1. ECCS Room 75(1)

2. Emergency Ventilation 75(1)

3. Containment Air Sample 30(1)

4. Penetration Room Purge 75(1)

c. Control Room HV & AC Units 10(1)

d. High Pressure Injection

1. High Pressure Injection Pumps 30(1)

2. High Pressure Injection Valves 30(1)

e. Component Cooling Water

1. Component Cooling Water Pumps 180(1)

2. Component Cooling Aux. Equip. Inlet 180(1)

Valves

3. Component Cooling to Makeup Pump 180(1)

Header Inlet Valve

4. Component Cooling from Decay Heat NA(1)

Cooler

f. Service Water System

1. Service Water Pumps 45(1)

2. Service Water From Component NA(1)

Cooling Heat Exchanger Isolation Valves

g. Containment Spray Isolation Valves 80(1)

h. Emergency Diesel Generator 15(1)

(1) Diesel generator starting and sequence loading delays included when applicable.

Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.5-4 Revision 3

Safety Features Actuation System Response Times TRM 8.3.5 Table 8.3.5-1 (page 4 of 7)

Safety Features System Instrumentation Response Times INITIATING SIGNAL AND FUNCTION RESPONSE TIME (seconds)

2. Containment Pressure - High (Continued)

i. Containment Isolation Valves

1. Vacuum Relief 30(1)

2. Normal Sump 30(1)

3. RCS Letdown Delay Coil Outlet 30(1)

4. RCS Letdown High Temperature 30(1)

5. Pressurizer Sample 45(1)

6. Service Water to Cooling Water 45(1)

7. Vent Header 15(1)

8. Drain Tank 15(1)

9. Core Flood Tank Vent 15(1)

10. Core Flood Tank Fill 15(1)

11. Steam Generator Sample 15(1)

12. Quench Tank 15(1)

13. Emergency Sump NA(1)

14. RCP Seal Return 45(1)

15. Air System 15(1)

16. N2 System 15(1)

17. Quench Tank Sample 35(1)

18. RCP Seal Inlet 17(1)

19. Core Flood Tank Sample 15(1)

20. RCP Standpipe Demin Water Supply 15(1)

21. Containment H2 Dilution Inlet 75(1)

22. Containment H2 Dilution Outlet 75(1)

j. BWST Outlet Valves NA(1)

k. Low Pressure Injection

1. Decay Heat Pumps 30(1)

2. Low Pressure Injection Valves NA(1)

3. Decay Heat Pump Suction Valves NA(1)

4. Decay Heat Cooler Outlet Valves NA(1)

5. Decay Heat Cooler Bypass Valves NA(1)

(1) Diesel generator starting and sequence loading delays included when applicable. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.5-5 Revision 3

Safety Features Actuation System Response Times TRM 8.3.5 Table 8.3.5-1 (page 5 of 7)

Safety Features System Instrumentation Response Times INITIATING SIGNAL AND FUNCTION RESPONSE TIME (seconds)

3. Containment Pressure--High-High

a. Containment Spray Pump 80(1)

b. Component Cooling Isolation Valves

1. Inlet to Containment 30(1)

2. Outlet from Containment 30(1)

3. Inlet to CRDM's 35(1)

4. CRDM Booster Pump Suction 35(1)

4. RCS Pressure-Low

a. Fans

1. Emergency Vent Fans 25(1)

2. Containment Cooler Fans 45(1)

b. HV & AC Isolation Valves

1. ECCS Room 75(1)

2. Emergency Ventilation 75(1)

3. Containment Air Sample 30(1)

4. Penetration Room Purge 75(1)

c. Control Room HV & AC Units 10(1)

d. High Pressure Injection

1. High Pressure Injection Pumps 30(1)

2. High Pressure Injection Valves 30(1)

e. Component Cooling Water

1. Component Cooling Water Pumps 180(1)

f. Service Water System

1. Service Water Pumps 45(1)

2. Service Water from Component Cooling NA(1)

Heat Exchanger Isolation Valves

g. Containment Spray Isolation Valves 80(1)

h. Emergency Diesel Generator 15(1)

(1) Diesel generator starting and sequence loading delays included when applicable. Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.5-6 Revision 3

Safety Features Actuation System Response Times TRM 8.3.5 Table 8.3.5-1 (page 6 of 7)

Safety Features System Instrumentation Response Times INITIATING SIGNAL AND FUNCTION RESPONSE TIME (seconds)

4. RCS Pressure-Low (continued)

i. Containment Isolation Valves

1. Vacuum Relief 30(1)

2. Normal Sump 30(1)

3. RCS Letdown Delay Coil Outlet 30(1)

4. RCS Letdown High Temperature 30(1)

5. Pressurizer Sample 45(1)

6. Service Water to Cooling Water 45(1)

7. Vent Header 15(1)

8. Drain Tank 15(1)

9. Core Flood Tank Vent 15(1)

10. Core Flood Tank Fill 15(1)

11. Steam Generator Sample 15(1)

12. Quench Tank 15(1)

13. Emergency Sump NA(1)

14. Air Systems 15(1)

15. N2 System 15(1)

16. Quench Tank Sample 35(1)

17. Core Flood Tank Sample 15(1)

18. RCP Standpipe Demin Water Supply 15(1)

19. Containment H2 Dilution Inlet 75(1)

20. Containment H2 Dilution Outlet 75(1)

j. BWST Outlet Valves NA(1)

(1) Diesel generator starting and sequence loading delays included when applicable.

Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.5-7 Revision 3

Safety Features Actuation System Response Times TRM 8.3.5 Table 8.3.5-1 (page 7 of 7)

Safety Features System Instrumentation Response Times INITIATING SIGNAL AND FUNCTION RESPONSE TIME (seconds)

5. RCS Pressure--Low-Low

a. Low Pressure Injection

1. Decay Heat Pumps 30(1)

2. Low Pressure Injection Valves NA(1)

3. Decay Heat Pump Suction Valves NA(1)

4. Decay Heat Cooler Outlet Valves NA(1)

5. Decay Heat Cooler Bypass Valves NA(1)

b. Component Cooling Isolation Valves

1. Auxiliary Equipment Inlet 90(1)

2. Inlet to Makeup Pump Header 90(1)

3. Component Cooling from Decay Heat NA(1)

Cooler

c. Containment Isolation Valves

1. RCP Seal Return 45(1)

2. RCP Seal Inlet 17(1)

(1) Diesel generator starting and sequence loading delays included when applicable.

Response time limit includes movement of valves and attainment of pump or blower discharge pressure.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.5-8 Revision 3

Safety Features Actuation System Response Times TRM B 8.3.5 BASES 8.3.5 Safety Features Actuation System Instrumentation The measurement of response time at the specified frequencies provides assurance that the SFAS action function associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.5-1 Revision 0

Waste Gas System Oxygen Monitoring TRM 8.3.6 8.3 INSTRUMENTATION 8.3.6 Waste Gas System Oxygen Monitoring TECHNICAL NORMAL CONDITIONS TNC 8.3.6 Waste Gas System Oxygen monitoring shall be FUNCTIONAL with its alarm setpoints set to ensure the limits of TRM 8.7.5 are not exceeded.

APPLICABILITY: During additions to the waste gas surge tank.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Both Waste gas system A.1 Declare the channel Immediately oxygen monitor alarm Nonfuntional and comply with setpoint less conservative Contingency Measures than required by B.1 and B.2.

TRM 8.7.5.

B. Both Waste gas system B.1 Additions to waste gas surge Every 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months during oxygen monitor tank may continue provided degassing Nonfunctional. another method for ascertaining oxygen AND concentrations, such as grab sample analysis, is 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during implemented to provide operations other than measurements. degassing AND B.2 Exert best efforts to return the 30 days waste gas system oxygen monitor to FUNCTIONAL status.

C. Contingency Measures C.1 Explain in the next Date of next and associated Radioactive Effluent Release Radioactive Effluent Restoration Time of Report why the Release Report Nonconformance B Nonconformance was not not met. corrected in a timely manner.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.6-1 Revision 0

Waste Gas System Oxygen Monitoring TRM 8.3.6 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.3.6.1 Perform CHANNEL CHECK. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months during additions to the waste gas surge tank 8.3.6.2 Perform CHANNEL CALIBRATION using standard gas 92 days samples containing a nominal:

a. 1 volume % oxygen, balance nitrogen; and

b. 4 volume % oxygen, balance nitrogen.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.6-2 Revision 0

Waste Gas System Oxygen Monitoring TRM B 8.3.6 BASES 8.3.6 Waste Gas System Oxygen Monitor The waste gas system oxygen monitor is provided to monitor oxygen concentration of gaseous radwaste being admitted to the waste gas surge tank. Oxygen concentration is monitored to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas treatment system is maintained below the flammability limits of hydrogen with oxygen.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.6-1 Revision 0

Post Accident Monitoring (PAM) Instrumentation TRM 8.3.7 8.3 INSTRUMENTATION 8.3.7 Post Accident Monitoring (PAM) Instrumentation TECHNICAL NORMAL CONDITIONS TNC 8.3.7 The PAM instrumentation for each Function in Table 8.3.7-1 shall be FUNCTIONAL.

APPLICABILITY: MODES 1, 2 and 3.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. One or more Functions A.1 Restore required channel to 30 days with one required FUNCTIONAL status.

channel Nonfunctional.

B. Contingency Measure B.1 Initiate action to evaluate Immediately and associated failure to meet TNC per Restoration Time of TRM Section 7.3.

Nonconformance A not met.

TECHNICAL VERIFICATION REQUIREMENTS

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

The below TVRs apply to each PAM instrumentation Function in Table 8.3.7-1.

TVR VERIFICATION FREQUENCY 8.3.7.1 Perform CHANNEL CHECK for each required 31 days instrumentation channel that is normally energized.

8.3.7.2 Perform CHANNEL CALIBRATION. 24 months DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.7-1 Revision 7

Post Accident Monitoring (PAM) Instrumentation TRM 8.3.7 Table 8.3.7-1 (page 1 of 1)

Post Accident Monitoring Instrumentation FUNCTION REQUIRED CHANNELS

1. RC System Subcooling Margin Monitor 1

2. PORV Position Indicator 1

3. PORV Block Valve Position Indicator 1

4. Pressurizer Safety Valve Position Indicator 1/valve

5. Containment Normal Sump Level 1 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.7-2 Revision 7

Post Accident Monitoring (PAM) Instrumentation TRM B 8.3.7 BASES 8.3.7 Post Accident Monitoring (PAM) Instrumentation The primary purpose of the PAM instrumentation is to display unit variables that provide information required by the control room operators during accident situations. This information provides the necessary support for the operator to take the manual actions for which no automatic control is provided and that are required for safety systems to accomplish their safety functions for Design Basis Events.

The FUNCTIONALITY of the accident monitoring instrumentation ensures that there is sufficient information available on selected unit parameters to monitor and to assess unit status and behavior following an accident.

The availability of accident monitoring instrumentation is important so that responses to corrective actions can be observed, and so that the need for and magnitude of further actions can be determined. These essential instruments are identified by UFSAR Section 7.13 (Ref. 1) addressing the recommendations of Regulatory Guide 1.97 (Ref. 2) as required by Supplement 1 to NUREG-0737 (Ref. 3).

Only those instruments monitoring Type A and Category 1 variables are required to be included in Technical Specifications. The instruments in this Technical Requirement did not meet the criterion for inclusion into Technical Specifications.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.7-1 Revision 0

EDG Loss of Power Start TRM 8.3.8 8.3 INSTRUMENTATION 8.3.8 EDG Loss of Power Start TECHNICAL NORMAL CONDITIONS TNC 8.3.8 The EDG Loss of Power Start (LOPS) instrumentation setpoints listed in TRM Table 8.3.8-1 shall be maintained in the manner specified in Technical Specification 3.3.8.

APPLICABILITY: As specified in Technical Specification 3.3.8.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME NONE TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY NONE DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.8-1 Revision 3

EDG Loss of Power Start TRM 8.3.8 Table 8.3.8-1 (page 1 of 1)

EDG Loss of Power Start Setpoints FUNCTIONS TRIP SETPOINTS Degraded Voltage Function 3734 Volts, +/- 7 Volts (Dropout) 3759 Volts, Max (Pickup) 7.5 +/- 0.2 Seconds (Delay)

Loss of Voltage Function 2429 Volts, +/- 7 Volts (Dropout) 2466 Volts, Max (Pickup) 0.5 +/- 0.05 Seconds (Delay)

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.8-2 Revision 3

EDG Loss of Power Start Instrumentation TRM B 8.3.8 BASES 8.3.8 EDG Loss of Power Start Instrumentation Compliance with TNC 8.3.8 provides assurance that the SFAS function associated with each channel is in conformance with the trip requirements assumed in the safety analysis. The trip setpoint is established by addition (or subtraction depending on the conservative direction) of instrumentation uncertainties to the Analytical Limit (value used in the AC Power System Analysis).

This assurance is based on compliance with the methodology for establishment of nuclear safety related setpoints. The setpoint and acceptance criteria are established in compliance with Regulatory Guide 1.105, Setpoints for Safety-Related Instrumentation. The setpoint and acceptance criteria are established using Method 2 from section 7 of ISA RP67.04.02-2000, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation.

Reset of the setpoint within the assumed Setpoint Tolerance Band will provide assurance that the channel is in compliance with the methodology and the calculation establishing the setpoint.

Additional assurance is provided by repeated, successful setpoint verification at the prescribed surveillance frequency.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.8-1 Revision 0

Source and Intermediate Range Overlap TRM 8.3.10 8.3 INSTRUMENTATION 8.3.10 Source and Intermediate Range Overlap TECHNICAL NORMAL CONDITIONS TNC 8.3.10 The requirements of TVR 8.3.10.1 shall be performed.

APPLICABILITY: When transitioning between the source range and intermediate range neutron flux instrumentation during a reactor startup.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. TVR 8.3.10.1 not met. A.1 Evaluate OPERABILITY Immediately requirements of Technical Specifications 3.3.9 and 3.3.10.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.3.10.1 Verify at least one decade overlap between Source and Each reactor startup Intermediate Range Monitors. during the transition between source and intermediate range monitors, if not verified in previous 7 days DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.10-1 Revision 1

Source and Intermediate Range Overlap TRM B 8.3.10 BASES 8.3.10 Source and Intermediate Range Overlap The overlap check requires an expectation of one decade of overlap when transitioning between the source range and intermediate range neutron flux instrumentation during a reactor startup.

During a power increase near the top scale for the source range monitors, an intermediate range monitor reading is expected with at least on decade overlap. Without such an overlap, the intermediate range monitors are considered inoperable unless it is clear that a source range monitor inoperability is responsible for the lack of the expected overlap.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.10-1 Revision 1

Steam And Feedwater Rupture Control System Instrumentation Parameters TRM 8.3.11 8.3 INSTRUMENTATION 8.3.11 Steam And Feedwater Rupture Control System Instrumentation Parameters TECHNICAL NORMAL CONDITIONS TNC 8.3.11 The Steam and Feedwater Rupture Control System (SFRCS) instrumentation RESPONSE TIMES listed in TRM Table 8.3.11-1 shall be maintained in the manner specified in Technical Specification 3.3.11.

AND The SFRCS Trip Setpoints listed in TRM Table 8.3.11-2 shall be maintained in the manner specified in Technical Specification 3.3.11.

APPLICABILITY: As specified in Technical Specification Table 3.3.11-1.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME NONE TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY NONE DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.11-1 Revision 3

Steam And Feedwater Rupture Control System Instrumentation Parameters TRM 8.3.11 Table 8.3.11-1 (page 1 of 1)

Steam and Feedwater Rupture Control System Response Time ACTUATED EQUIPMENT RESPONSE TIME (seconds)

1. Auxiliary Feed Pump 40

2. Main Steam Isolation Valves (1)

a. Main Steam Low Pressure Channels 6

b. Feedwater/Steam Generator High 6.5 Differential Pressure Channels

3. Main Feedwater Valves

a. Main Control 8

b. Startup Control 13

c. Stop Valve 16

4. Turbine Stop Valves (2) 1 (1) The response time is to be the time elapsed from the monitored variable exceeding the trip setpoint until the Main Steam Isolation Valve is fully closed.

(2) The response time is to be the time elapsed from the Main Steam Line Pressure Low trip condition until the Turbine Stop Valve is fully closed.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.11-2 Revision 3

Steam And Feedwater Rupture Control System Instrumentation Parameters TRM 8.3.11 Table 8.3.11-2 (page 1 of 1)

Steam and Feedwater Rupture Control System (SFRCS) Trip Setpoints FUNCTION

1. Main Steam Line Pressure - Low (3)

Limiting Trip Setpoint (LTSP) 625.65 psig Nominal Trip Setpoint (NTSP) 630 psig As-Found Acceptance Criteria Band NTSP +/- 14.0 psig As-Left Setpoint Tolerance Band NTSP +/- 10.0 psig

2. Feedwater / Steam Generator Differential Pressure - High (2, 3)

Limiting Trip Setpoint (LTSP) 132.90 psid Nominal Trip Setpoint (NTSP) 125.0 psid As-Found Acceptance Criteria Band NTSP +/- 10.0 psid As-Left Setpoint Tolerance Band NTSP +/- 7.14 psid

3. Steam Generator Level - Low (1, 3)

Limiting Trip Setpoint (LTSP) 23.30 inches Indicated Nominal Trip Setpoint (NTSP) 23.50 inches Indicated As-Found Acceptance Criteria Band NTSP +/- 0.25 inches As-Left Setpoint Tolerance Band NTSP +/- 0.135 inches

4. Loss of Reactor Coolant Pumps-Trip Setpoint - High 1384.6 amps Trip Setpoint - Low 106.5 amps (1) Steam Generator Level - Low Function references actual water level above the lower steam generator tubesheet, with this setpoint being consistent with the Allowable Value listed in the Technical Specifications.

(2) Differential Pressure is steam generator pressure minus feedwater pressure.

(3) Compliance with the As-Found Acceptance Criteria Band is determined by evaluating the current surveillance test value and comparing it to the As-Found Acceptance Criteria Band with respect to the Nominal Trip Setpoint. This must be evaluated separate from compliance with the Technical Specification Allowable Value.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.11-3 Revision 3

Steam And Feedwater Rupture Control System Instrumentation Parameters TRM B 8.3.11 BASES 8.3.11 Steam And Feedwater Rupture Control System Instrumentation The measurement of response time at the specified frequencies provides assurance that the SFRCS action function associated with each channel is completed within the time limit assumed in the safety analyses. No credit was taken in the analyses for those channels with response times indicated as not applicable.

Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or 2) utilizing replacement sensors with certified response times.

The SFRCS response time for the turbine stop valve closure is based on the combined response times of main steam line low pressure sensors, logic cabinet delay for main steam line low pressure signals and closure time of the turbine stop valves. This SFRCS response time ensures that the auxiliary feedwater to the unaffected steam generator will not be isolated due to a SFRCS low pressure trip during a main steam line break accident.

The measurement of trip setpoints at the specified frequencies provides assurance that the SFRCS function associated with each channel is in conformance with the trip requirements assumed in the safety analysis. The trip setpoint is established by addition (or subtraction depending on the conservative direction) of instrument uncertainties to the Analytical Limit (value used in the safety analysis).

This assurance is based on compliance with the methodology for establishment of nuclear safety related setpoints. The setpoint and acceptance criteria are established in compliance with Regulatory Guide 1.105, Setpoints for Safety-Related Instrumentation. The setpoint and acceptance criteria are established using Method 1 or Method 2 from Section 7 of ISA RP67.04.02-2000, Methodologies for the Determination of Setpoints for Nuclear Safety-Related Instrumentation. Reset of the setpoint within the assumed As-Left Setpoint Tolerance Band will provide assurance that the channel is in compliance with the methodology and the calculation establishing the setpoint. Additional assurance is provided by repeated, successful setpoint verification at the prescribed surveillance frequency.

Setpoints found outside of the prescribed values require evaluation to ensure the equipment is able to perform within the calculational values and to determine if the equipment is able to perform the intended protective function.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.11-1 Revision 0

Ultrasonic Flow Meter Instrumentation TRM 8.3.12 8.3 INSTRUMENTATION 8.3.12 Ultrasonic Flow Meter Instrumentation TECHNICAL NORMAL CONDITIONS TNC 8.3.12 Ultrasonic Flow Meter Instrumentation shall be FUNCTIONAL.

APPLICABILITY: MODE 1 when greater than 50% RATED THERMAL POWER.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Ultrasonic Flow Meter A.1 Restore Ultrasonic Flow Prior to the next instrumentation Meter to FUNCTIONAL required daily Nonfunctional. status. calorimetric heat balance measurement B. Contingency Measure and B.1 Comply with the applicable Immediately associated Restoration actions of Technical Time is not met. Specification 3.3.1.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.3.12.1 Perform CHANNEL CHECK for Ultrasonic Flow Meter 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months instrumentation.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.3.12-1 Revision 0

Ultrasonic Flow Meter Instrumentation TRM B 8.3.12 BASES 8.3.12 Ultrasonic Flow Meter instrumentation The LEFM includes a flow meter measurement section in each of the two main feedwater flow headers. Each measurement section consists of sixteen ultrasonic transducers. With any transducer nonfunctional, the Ultrasonic Flow Meter instrumentation system is considered nonfunctional.

The daily CHANNEL CHECK utilizes the on-line verification and self-diagnostic features of the LEFM to ensure the instrumentation is performing as designed.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.3.12-1 Revision 0

Chemistry TRM 8.4.1 8.4 REACTOR COOLANT SYSTEM 8.4.1 Chemistry TECHNICAL NORMAL CONDITIONS TNC 8.4.1 The Reactor Coolant System chemistry shall be maintained within the limits specified in TRM Table 8.4.1-1.

APPLICABILITY: At all times.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME

NOTE--------------- A.1 Restore the parameter to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Only applicable in MODES 1, within its Steady State Limit.

2, 3, or 4.

A. With any one or more chemistry parameter in excess of its Steady State Limit but within its Transient Limit.

NOTE--------------- B.1 Initiate action to evaluate Immediately Only applicable in MODES 1, failure to meet TNC per 2, 3, or 4. TRM Section 7.3.

B. Contingency Measures and associated Restoration Time of Nonconformance A not met OR With any one or more chemistry parameter in excess of its Transient Limit.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.1-1 Revision 13

Chemistry TRM 8.4.1 CONTINGENCY MEASURES (continued)

NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME

NOTE--------------- C.1 Reduce the Reactor Coolant Immediately Not applicable in MODES 1, System pressure to 500 2, 3, or 4. psig, if applicable.

AND C. With the concentration of either chloride or fluoride C.2 Initiate action to perform an Immediately in the Reactor Coolant engineering evaluation to System in excess of its determine the effects of the Steady State Limit for out-of-limit condition on the more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or in structural integrity of the excess of its Transient Reactor Coolant System.

Limit.

AND C.3 Determine that the Reactor Prior to increasing the Coolant System remains system pressure above acceptable for continued 500 psig or prior to operation. proceeding to MODE 4 D. ----------NOTE---------- D.1 ------------NOTE---------------

Not applicable in Applicable only when the MODES 1, 2, 3, 4, 5, ability to sample the RCS is or 6. restored Unable to determine Initiate action to perform Immediately limits of chloride and TVR 8.4.1.1 fluoride in the Reactor Coolant System due to the inability to sample the RCS.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.1-2 Revision 13

Chemistry TRM 8.4.1 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.4.1.1 Determine by analysis, the parameters listed in TRM 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Table 8.4.1-1 are within their specified limits.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.1-3 Revision 13

Chemistry TRM 8.4.1 TABLE 8.4.1-1 (page 1 of 1)

REACTOR COOLANT SYSTEM CHEMISTRY LIMITS PARAMETER STEADY STATE TRANSIENT 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 (1) Limit not applicable with Tavg 250 F.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.1-4 Revision 13

Chemistry TRM B 8.4.1 BASES 8.4.1 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduce the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits shown on TRM Table 8.4.1-1 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 verification requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.4.1-1 Revision 0

Pressurizer TRM 8.4.2 8.4 REACTOR COOLANT SYSTEM 8.4.2 Pressurizer TECHNICAL NORMAL CONDITIONS TNC 8.4.2 The pressurizer temperature shall be limited to:

a. A maximum heatup and cooldown of 100°F in any one hour period;

b. A maximum spray water temperature differential of 410°F and

c. A minimum temperature of 120°F when the pressurizer pressure is 625 psig.

APPLICABILITY: At all times.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Pressurizer temperature A.1 Restore the temperature to 30 minutes limits in excess of any of within limits.

the above limits.

AND A.2 Initiate action to perform an Immediately engineering evaluation to determine the effects of the out-of-limit condition on the fracture toughness properties of the pressurizer.

B. Pressurizer not B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months acceptable for continued operation. AND B.2 Reduce pressurizer pressure 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to < 500 psig.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.2-1 Revision 0

Pressurizer TRM 8.4.2 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.4.2.1 ---------------------------------NOTE------------------------------

Only required during heatup or cooldown operation.

Determine pressurizer temperature within limits. 30 minutes 8.4.2.2 ---------------------------------NOTE------------------------------

Only required during spray operation with pressurizer temperature 440°F.

Determine spray water temperature differential within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months limits.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.2-2 Revision 0

Pressurizer TRM B 8.4.2 BASES 8.4.2 Pressurizer The conditions, actions and verification requirements for the pressurizer temperature limits define the limitations on the pressurizer heatup and cooldown, spray water temperature differential, and minimum temperature when pressure is greater than 625 psig to assure that the pressurizer remains within the design criteria assumed for the pressurizer fatigue analysis. As discussed in Section 5.5.10 of the Davis-Besse UFSAR, the total stresses resulting from thermal expansion, pressure and mechanical and seismic loadings are considered in the design of the pressurizer. The total stresses expected in the pressurizer are within the maximum allowed by the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code,Section III.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.4.2-1 Revision 0

Pressurizer Heater Interlock TRM 8.4.3 8.4 REACTOR COOLANT SYSTEM 8.4.3 Pressurizer Heater Interlock TECHNICAL NORMAL CONDITIONS TNC 8.4.3 Two Pressurizer Heater Interlock channels shall be FUNCTIONAL.

APPLICABILITY: MODE 3 when either decay heat removal (DHR) isolation valve DH-11 or DH-12 is open.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. One or more Pressurizer A.1 Place nonfunctional Immediately Heater Interlock channels channel(s) in trip.

Nonfunctional.

OR A.2 ---------------NOTE--------------

Only applicable if RCS pressure < 328 psig.

Restore nonfunctional Prior to increasing RCS channel(s) to FUNCTIONAL pressure 328 psig status.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.3-1 Revision 0

Pressurizer Heater Interlock TRM 8.4.3 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.4.3.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 8.4.3.2 Perform CHANNEL CALIBRATION. The Allowable 24 months Value shall be < 328 psig and is referenced to the RCS Pressure instrumentation tap.

8.4.3.3 Verify Pressurizer Heater Interlock deenergizes the 24 months pressurizer heaters on a actual or simulated RCS pressure which is greater than the Allowable Value with either DH-11 or DH-12 open.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.3-2 Revision 0

Pressurizer Heater TRM B 8.4.3 BASES 8.4.3 Pressurizer Heaters Pressurizer Heater Interlock setpoint is based on preventing over-pressurization of the Decay Heat Removal System normal suction line piping. The value stated is the RCS pressure at the sensing instrument's tap. It has been adjusted to reflect the elevation difference between the sensor's location and the pipe of concern.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.4.3-1 Revision 0

Reactor Coolant System Vents TRM 8.4.4 8.4 REACTOR COOLANT SYSTEM 8.4.4 Reactor Coolant System Vents TECHNICAL NORMAL CONDITIONS TNC 8.4.4 The following Reactor Coolant System vent paths shall be FUNCTIONAL:

a. Reactor Coolant System Loop 1 with vent path through valves RC 4608A and RC 4608B;

b. Reactor Coolant System Loop 2 with vent path through valves RC 4610A and RC 4610B; and

c. Pressurizer with vent path through either valves RC11 and RC2A (PORV) or valves RC 239A and RC 200.

APPLICABILITY: MODES 1, 2, and 3.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. One vent path A.1 Restore the Nonfunctional 30 days Nonfunctional. vent path to FUNCTIONAL status.

B. Two or more vent paths B.1 Restore all but one 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Nonfunctional. Nonfunctional vent paths to FUNCTIONAL status.

C. Contingency Measure C.1 Initiate action to evaluate Immediately and associated failure to meet TNC per Restoration Time of TRM Section 7.3.

Nonconformance A or B not met.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.4-1 Revision 7

Reactor Coolant System Vents TRM 8.4.4 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.4.4.1 Verify all manual isolation valves in each required vent 24 months path are locked in the open position.

8.4.4.2 Cycle each valve in each required vent path through at 24 months least one complete cycle of full travel from the Control Room.

8.4.4.3 Verify flow through each required reactor coolant vent 24 months system vent paths.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.4-2 Revision 7

Reactor Coolant System Vents TRM B 8.4.4 BASES 8.4.4 High Point Vents The Reactor Coolant System high point vents are installed per NUREG-0737 item II.B.1 requirements. The functionality of the system ensures capability of venting steam or noncondensable gas bubbles in the reactor cooling system to restore natural circulation following a small break loss of coolant accident.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.4.4-1 Revision 0

Pressurizer PORV TRM 8.4.5 8.4 REACTOR COOLANT SYSTEM 8.4.5 Pressurizer Pilot Operated Relief Valve (PORV)

TECHNICAL NORMAL CONDITIONS TNC 8.4.5 The requirement of TVR 8.4.5.1 shall be performed with an Allowable Value of 2435 psig.

APPLICABILITY: As specified in Technical Specification 3.4.11.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. TVR 8.4.5.1 not met. A.1 Evaluate PORV Immediately OPERABILITY in accordance with Technical Specification 3.4.11.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.5-1 Revision 6

Pressurizer PORV TRM 8.4.5 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.4.5.1 Perform CHANNEL CALIBRATION of the PORV 24 months opening setpoint.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.5-2 Revision 6

Pilot Operated Relief Valve (PORV)

TRM B 8.4.5 BASES 8.4.5 Pilot Operated Relief Valve (PORV)

None.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.4.5-1 Revision 0

ASME Code Class 1, 2, and 3 Components TRM 8.4.6 8.4 REACTOR COOLANT SYSTEM 8.4.6 ASME Code Class 1, 2, and 3 Components TECHNICAL NORMAL CONDITIONS TNC 8.4.6 The structural integrity of ASME Code Class 1, 2, and 3 components shall be maintained in accordance with the Inservice Inspection Program.

APPLICABILITY: MODES 1, 2, 3, 4, 5, and 6.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Structural integrity of one A.1 Isolate the affected Prior to increasing the or more ASME Code component. RCS temperature to Class 1 not meeting the > 50F above the TNC. minimum temperature required by NDT considerations B. Structural integrity of one B.1 Isolate the affected Prior to increasing the or more ASME Code component. RCS temperature to Class 2 not meeting the > 200F TNC.

C. Structural integrity of one C.1 Isolate the affected Immediately or more ASME Code component.

Class 3 not meeting the TNC.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.6-1 Revision 19

ASME Code Class 1, 2, and 3 Components TRM 8.4.6 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.4.6.1 Verify the structural integrity of ASME Code Class 1, In accordance with 2, and 3 components. the Inservice Inspection Program DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.4.6-2 Revision 19

ASME Code Class 1, 2, and 3 Components TRM B 8.4.6 BASES 8.4.6 ASME Code Class 1, 2, and 3 Components The inspection and testing programs for ASME Code Class 1, 2, and 3 components, except the Steam Generator tubes, ensure that the structural integrity of these components will be maintained at an acceptable level throughout the life of the plant. To the extent possible, the inspection program for these components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

The reference to ASME Code Class 1, 2, and 3 Components in this Technical Requirements Manual (TRM) section pertains to components within the sites Inservice Inspection boundary (i.e., ASME Section XI, Code Class 1, 2, or 3.) This is not to be confused with other sections of the ASME Code. The Inservice Inspection boundary diagrams are depicted on the ISID2 series of drawings.

If containment air cooling (CAC) service water (SW) piping is ever isolated due to a breach of the piping in containment, that system is no longer acting as a penetration boundary. Because the CAC SW piping does not connect with either the reactor coolant system or the containment atmosphere, and post LOCA operation is for the system to be in service, CAC SW penetrations are not subject to testing under the Containment Leakage Rate Testing Program. Therefore, breached CAC SW piping in containment that is isolated would represent unquantifiable secondary containment bypass leakage for the associated CAC SW outlet penetration (LCO 3.6.3).

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.4.6-1 Revision 19

ECCS Subsystems - Operating TRM 8.5.1 8.5 EMERGENCY CORE COOLING SYSTEMS 8.5.1 ECCS Subsystems - Operating TECHNICAL NORMAL CONDITIONS TNC 8.5.1 The requirements of TVR 8.5.1.1 and 8.5.1.2 shall be performed.

APPLICABILITY: MODES 1, 2 and 3.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. TNC 8.5.1 not met. A.1 Comply with the applicable Immediately ACTIONS of Technical Specification 3.5.2.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.5.1.1 ---------------------------------NOTE--------------------------------

Only the subsystem(s) directly affected by the flowpath modification needs to be tested in accordance with this Verification Requirement.

Each HPI and LPI ECCS subsystem shall be Following completion of demonstrated OPERABLE by performing a flow test, a modification to the during shutdown, to verify the injection phase flow rate subsystem flowpath and flow distribution (for HPI only) meet or exceed the that could alter the LOCA requirements. subsystem flow characteristics DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.5.1-1 Revision 0

ECCS Subsystems - Operating TRM 8.5.1 TECHNICAL VERIFICATION REQUIREMENTS (continued)

TVR VERIFICATION FREQUENCY 8.5.1.2

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

The inspection port on the watertight enclosure may be opened without requiring performance of the vacuum leakage rate test, to perform inspections. After use, the inspection port must be verified as closed in its correct position. Provisions of TS 3.0.3 are not applicable during these inspections.

Each ECCS subsystem shall be demonstrated 24 months OPERABLE by performing a vacuum leakage rate test of the watertight enclosure for valves DH-11 and DH-12 AND that assures the motor operators on valves DH-11 and DH-12 will not be flooded for at least 7 days following a After each opening of LOCA. the watertight enclosure AND After any maintenance on or modification to the watertight enclosure which could affect its integrity DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.5.1-2 Revision 0

ECCS Subsystems - Operating TRM 8.5.1 BASES 8.5.1 ECCS Subsystems - Operating The verification requirement for flow and flow distribution (HPI only) testing provides assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) ensure an amount of ECCS flow that is equal to or greater than the flow assumed in the ECCS-LOCA analyses, and (3) ensure proper flow distribution between HPI injection points, in accordance with the assumptions used in the ECCS-LOCA analyses.

The frequency of Technical Verification Requirement 8.5.1.1 ensures that changes in system performance are detected and verified not to degrade the subsystem's ability to provide the flows that are required for accident mitigation. The HPI and LPI pumps are monitored in accordance with other surveillance requirements that specifically measure pump performance.

Therefore, this surveillance does not apply to subsystem modifications that are limited to only the pumps.

The intent of Technical Verification Requirement 8.5.1.1 is to verify that the subsystem flow characteristics have not been unacceptably altered by modifications that could affect the resistance of the subsystem flowpath. Taken together, the pump verifications and this verification ensure that the LOCA analyses remain valid.

Technical Verification Requirement 8.5.1.1 requires verification of flow rate and flow distribution (HPI only) for the injection phase. This may be accomplished by testing in an alternate system lineup (e.g., RCS recirculation) and verifying equivalent flow rates by calculation, as long as the affected portion of the flowpath is in the tested flowpath.

Decay Heat Removal System valves DH-11 and DH-12 are located in an area that would be flooded following a LOCA. These valves are located in a watertight enclosure to ensure their functionality up to seven days following a LOCA. Verification Requirements are provided to verify the acceptable leak tightness of this enclosure. An inspection port is located on this watertight enclosure, which is typically used for performing inspections inside the enclosure.

During the vacuum leakage rate test, the inspection port is in a closed position and subject to the test. This inspection port may be subsequently opened for use in viewing inside the enclosure. Opening this inspection port will not require performance of the vacuum leakage rate test because of the design of the closure fitting, which will preclude leakage under LOCA conditions, when properly installed. Proper installation includes independent verification.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.5.1-1 Revision 0

ECCS Subsystems - Shutdown TRM 8.5.2 8.5 EMERGENCY CORE COOLING SYSTEMS 8.5.2 ECCS Subsystems - Shutdown TECHNICAL NORMAL CONDITIONS TNC 8.5.2 The requirements of TVR 8.5.2.1 and 8.5.2.2 shall be performed.

APPLICABILITY: MODE 4.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. TNC 8.5.2 not met. A.1 Comply with the applicable Immediately ACTIONS of Technical Specification 3.5.3.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.5.2.1 ---------------------------------NOTE--------------------------------

Only the subsystem(s) directly affected by the flowpath modification needs to be tested in accordance with this Verification Requirement.

Each LPI ECCS subsystem shall be demonstrated Following completion of OPERABLE by performing a flow test, during a modification to the shutdown, to verify the injection phase flow rate meets subsystem flowpath or exceeds the LOCA requirements. that could alter the subsystem flow characteristics DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.5.2-1 Revision 1

ECCS Subsystems - Shutdown TRM 8.5.2 TECHNICAL VERIFICATION REQUIREMENTS (continued)

TVR VERIFICATION FREQUENCY 8.5.2.2 ---------------------------------NOTE--------------------------------

The inspection port on the watertight enclosure may be opened without requiring performance of the vacuum leakage rate test, to perform inspections. After use, the inspection port must be verified as closed in its correct position. Provisions of TS 3.0.3 are not applicable during these inspections.

Each LPI ECCS subsystem shall be demonstrated 24 months OPERABLE by performing a vacuum leakage rate test of the watertight enclosure for valves DH-11 AND and DH-12 that assures the motor operators on valves DH-11 and DH-12 will not be flooded for at After each opening of least 7 days following a LOCA. the watertight enclosure AND After any maintenance on or modification to the watertight enclosure which could affect its integrity DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.5.2-2 Revision 1

ECCS Subsystems - Shutdown TRM 8.5.2 BASES 8.5.2 ECCS Subsystems - Shutdown The verification requirement for flow testing provides assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to:

(1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, and (2) ensure an amount of ECCS flow that is equal to or greater than the flow assumed in the ECCS-LOCA analyses.

The frequency of Technical Verification Requirement 8.5.2.1 ensures that changes in system performance are detected and verified not to degrade the subsystem's ability to provide the flows that are required for accident mitigation. The LPI pumps are monitored in accordance with other surveillance requirements that specifically measure pump performance. Therefore, Technical Verification Requirement 8.5.2.1 does not apply to subsystem modifications that are limited to only the pumps.

The intent of Technical Verification Requirement 8.5.2.1 is to verify that the subsystem flow characteristics have not been unacceptably altered by modifications that could affect the resistance of the subsystem flowpath. Taken together, the pump surveillances and this verification ensure that the LOCA analyses remain valid.

Technical Verification Requirement 8.5.2.1 requires verification of flow rate for the injection phase. This may be accomplished by testing in an alternate system lineup (e.g.,

RCS recirculation) and verifying equivalent flow rates by calculation, as long as the affected portion of the flowpath is in the tested flowpath.

Decay Heat Removal System valves DH-11 and DH-12 are located in an area that would be flooded following a LOCA. These valves are located in a watertight enclosure to ensure their operability up to seven days following a LOCA. Verification Requirements are provided to verify the acceptable leak tightness of this enclosure. An inspection port is located on this watertight enclosure, which is typically used for performing inspections inside the enclosure. During the vacuum leakage rate test, the inspection port is in a closed position and subject to the test. This inspection port may be subsequently opened for use in viewing inside the enclosure. Opening this inspection port will not require performance of the vacuum leakage rate test because of the design of the closure fitting, which will preclude leakage under LOCA conditions, when properly installed. Proper installation includes independent verification.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.5.2-1 Revision 0

Emergency Sump Debris TRM 8.5.3 8.5 EMERGENCY CORE COOLING SYSTEMS 8.5.3 Emergency Sump Debris TECHNICAL NORMAL CONDITIONS TNC 8.5.3 The requirements of TVR 8.5.3.1 and TVR 8.5.3.2 shall be met.

APPLICABILITY: MODES 1, 2, 3 and 4.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. TNC 8.5.3 not met. A.1 Evaluate OPERABILITY of Immediately ECCS per Technical Specification 3.5.2 and 3.5.3.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.5.3.1 Perform a visual inspection of all accessible Prior to establishing areas of the containment to verify that no loose Containment debris (rags, trash, clothing, etc) is present in OPERABILITY the containment which could be transported to the containment emergency sump and cause restriction of the pump suction during LOCA conditions.

8.5.3.2 Perform a visual inspection of all areas of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while work is containment affected by an entry to verify that ongoing no loose debris (rags, trash, clothing, etc) is present in the containment which could be AND transported to the containment emergency sump and cause restriction of the pump suction During final exit after during LOCA conditions. completion of work (containment closeout) when containment OPERABILITY is established DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.5.3-1 Revision 5

Emergency Sump Debris TRM B 8.5.3 BASES 8.5.3 Emergency Sump Debris None DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.5.3-1 Revision 5

Combustible Gas Control - Hydrogen Analyzers TRM 8.6.1 8.6 CONTAINMENT SYSTEMS 8.6.1 Combustible Gas Control - Hydrogen Analyzers TECHNICAL NORMAL CONDITIONS TNC 8.6.1 Two independent containment hydrogen analyzers shall be FUNCTIONAL.

APPLICABILITY: MODES 1 and 2.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. One hydrogen analyzer A.1 Restore the non-functional 30 days Nonfunctional. analyzer to FUNCTIONAL status.

B. Both hydrogen analyzers B.1 Restore one analyzer to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Nonfunctional. FUNCTIONAL status.

C. Contingency Measures C.1 Initiate action to evaluate Immediately and associated failure to meet TNC per TRM Restoration Time of Section 7.3.

Nonconformance A or B not met.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.6.1-1 Revision 0

Combustible Gas Control - Hydrogen Analyzers TRM 8.6.1 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.6.1.1 Perform a CHANNEL CHECK. 31 days 8.6.1.2 Perform a CHANNEL CALIBRATION using sample 46 days on a gases containing: STAGGERED TEST BASIS

a. 0 volume % hydrogen, balance nitrogen; and

b. 2.0 to 3.0 volume % hydrogen, balance nitrogen.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.6.1-2 Revision 0

Combustible Gas Control - Hydrogen Analyzers TRM 8.6.1 BASES 8.6.1 Combustible Gas Control - Hydrogen Analyzers Two redundant hydrogen analyzers are available to determine the content of hydrogen within the containment vessel. The hydrogen analyzers provide diagnostic capability for beyond design-basis accidents.

A rule change to 10 CFR 50 dated September 16, 2003 (68 FR 54123) eliminated the requirement that the hydrogen analyzers be safety-related components, and allowed their requirements to be relocated from the Technical Specifications.Section III.D of the final rule (68 FR 54127) categorized the hydrogen monitoring system as Category 3 of Regulatory Guide 1.97 because the monitors are required to diagnose the course of significant beyond design-basis accidents.Section III.D further stated that Category 3 applies to high-quality, off-the-shelf backup and diagnostic instrumentation.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.6.1-1 Revision 0

Steam Generator Pressure / Temperature Limitation TRM 8.7.1 8.7 PLANT SYSTEMS 8.7.1 Steam Generator Pressure / Temperature Limitation TECHNICAL NORMAL CONDITIONS TNC 8.7.1 The temperature of the secondary coolant in the steam generators shall be

> 110°F when the pressure of the secondary coolant in the steam generator is > 237 psig.

APPLICABILITY: At all times.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Requirements of TNC not A.1 Reduce the steam 30 minutes met. generator pressure of the applicable side to

< 237 psig.

AND A.2 Determine by engineering Prior to increasing evaluation the effects of the steam generator overpressurization on the pressure > 237 psig structural integrity of the steam generator and that the steam generator remains acceptable for continued operation.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.7.1.1 ---------------------------------NOTE--------------------------------

Only required when the secondary pressure in the steam generator is > 237 psig and Tavg is < 200°F Verify temperature of secondary coolant in each steam 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months generator is > 110°F.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.1-1 Revision 0

Steam Generator Pressure / Temperature Limitation TRM 8.7.1 BASES 8.7.1 Steam Generator Pressure / Temperature Limitation The limitation on steam generator pressure and temperature ensures that the pressure induced stresses in the steam generators do not exceed the maximum allowable fracture toughness stress limits. The limitations of 110°F and 237 psig are based on a steam generator RTndt of 40°F and are sufficient to prevent brittle fracture.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.7.1-1 Revision 0

Sealed Source Contamination TRM 8.7.2 8.7 PLANT SYSTEMS 8.7.2 Sealed Source Contamination TECHNICAL NORMAL CONDITIONS TNC 8.7.2 Each sealed source containing radioactive material > 100 Ci of beta and/or gamma emitting material or > 5 Ci of alpha emitting material, shall be free of > 0.005 Ci of removable contamination.

APPLICABILITY: At all times.

CONTINGENCY MEASURES

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

Separate Nonconformance entry is allowed for each sealed source.

NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. One or more sealed A.1 Withdraw the sealed source Immediately sources with removable from use.

contamination not within limits. AND A.2.1 Initiate action to Immediately decontaminate and repair the sealed source.

OR A.2.2 Initiate action to dispose of Immediately the sealed source in accordance with NRC Regulations.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.2-1 Revision 21

Sealed Source Contamination TRM 8.7.2 TECHNICAL VERIFICATION REQUIREMENTS

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

1. The TVRs shall be performed by company personnel or other personnel specifically authorized by the NRC or Agreement State.

2. The test method used shall have a detection sensitivity of < 0.005 Ci per test sample.

TVR VERIFICATION FREQUENCY 8.7.2.1 ---------------------------------NOTE--------------------------------

Startup sources and fission detectors previously subjected to core flux are excluded.

Perform leakage and contamination testing on each 184 days sealed source in use containing radioactive materials with a half-life > 30 days (excluding Hydrogen 3) and in any form other than gas.

8.7.2.2 Perform leakage and contamination testing for each Prior to placing in use sealed source and fission detector not in use. or transferring to another licensee, if not performed within the previous 184 days 8.7.2.3 Perform leakage and contamination testing on each Prior to placing in use sealed source and fission detector not in use that was received without a certificate indicating the last test date.

8.7.2.4 Perform leakage and contamination testing on each Once within 31 days sealed startup source and fission detector. prior to being subjected to core flux or installed in the core AND Following repair or maintenance to the sealed source 8.7.2.5 Submit report to NRC for sealed source or fission 12 months detector leakage tests revealing the presence of 0.005 Ci of removable contamination.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.2-2 Revision 21

Sealed Source Contamination TRM B 8.7.2 BASES 8.7.2 Sealed Source Contamination The limitations on removable contamination for sources requiring leak testing, including alpha emitters, are based on 10 CFR 70.39(c) limits for plutonium. This limitation will ensure that leakage from by-product, source and special nuclear material sources will not exceed allowable intake values.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.7.2-1 Revision 21

Snubbers TRM 8.7.3 8.7 PLANT SYSTEMS 8.7.3 Snubbers TECHNICAL NORMAL CONDITIONS TNC 8.7.3 Each safety related snubber shall perform its associated support function(s).

APPLICABILITY: MODES 1, 2, 3, and 4, MODES 5 and 6 for snubbers located on systems required OPERABLE in those MODES.

CONTINGENCY MEASURES

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

Separate entry is allowed for each snubber.

NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. One or more required A.1 Declare the supported Immediately snubbers unable to system or train LCO(s) not perform their associated met.

support function.

OR A.2.1 Enter LCO 3.0.8. Immediately AND A.2.2 Verify at least one train (or Immediately subsystem) of systems supported by the inoperable snubbers would remain capable of performing their required safety or support function for postulated design loads other than seismic loads.

AND DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.3-1 Revision 15

Snubbers TRM 8.7.3 CONTINGENCY MEASURES (continued)

NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. (continued) A.2.3 ---------NOTE------------------

Only applicable if LCO 3.0.8.a is used in MODES 1,2,3 or 4.

Verify at least one EFW train Immediately not associated with the inoperable snubber is available.

AND A.2.4 ---------NOTE------------------

Only applicable if LCO 3.0.8.b is used in MODES 1,2,3 or 4.

Verify at least one EFW train Immediately not associated with the inoperable snubber(s), or some alternative means of core cooling is available.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.7.3.1 Perform snubber inservice inspections in accordance to In accordance with Inservice Testing Program. Table 8.7.3-1 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.3-2 Revision 15

Snubbers TRM 8.7.3 Table 8.7.3-1 (page 1 of 4)

Snubber Inservice Testing Program Safety-related snubbers are listed in the latest revision of applicable verification test procedure(s). Snubbers may be added to, or removed from, safety-related systems and their assigned groups without a License Amendment.

In accordance with 10 CFR 50.55a(b)(3)(v)B with Conditions and Clarifications - the snubber population, as defined in ASME OM Code, 2004 Edition with OMa 2005 and OMb 2006 Addenda, Section IST, Subsection ISTA, article ISTA-1100(c) - complies with the requirements for Examination and Testing of the ASME OM Code, 2004 Edition with OMa-2006 and OMb-2006 Addenda, Section IST, Subsections ISTA and ISTD, and Code Case OMN-13. The Snubber IST Program complies with the requirements set forth in the Snubber Program Plan and the Snubber Program Procedure.

A. Visual Inspection Program

1. General Requirements At least once per inspection interval, each group of snubbers in use in the plant shall be visually inspected in accordance with Section A.2 and A.3. Visual inspections may be performed with binoculars, or other visual support devices, for those snubbers that are difficult to access and where required to keep exposure as low as reasonably achievable. Response to failures shall be in accordance with Section A.4.

2. Inspection Interval The inspection interval may be applied on the basis of snubber groups. The snubber groups may be established based on physical characteristics and accessibility.

Inaccessible snubbers are defined as those located: (a) inside containment, (b) in high radiation exposure zones, or (c) in areas where accessibility is limited by physical constraints such as the need for scaffolding.

Each of the groups may be inspected independently according to the schedule determined by ASME OM Code Case OMN-13, not to exceed 120 months.

3. Acceptance Criteria A snubber shall be considered able to perform its associated support function as a result of visual inspection if: (1) there is no visible indication of damage or inoperability, and (2) attachments to the foundation or supporting structure are secure.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.3-3 Revision 15

Snubbers TRM 8.7.3 Table 8.7.3-1 (page 2 of 4)

Snubber Inservice Testing Program

4. Response to Failures For each snubber unit which does not meet the visual inspection acceptance criteria of Section A.3:

a. Determine the snubber is able to perform its associated support function by functionally testing the snubber in the as-found condition per Section B, unless the (hydraulic) snubber was determined Nonfunctional because the fluid port was found uncovered; and

b. Perform an evaluation to determine the cause of the unacceptability.

5. Transient Event Inspection An inspection shall be performed of all hydraulic and mechanical snubbers attached to sections of systems that have experienced unexpected, potentially damaging transients as determined from a review of operational data. The affected snubber(s) and system(s) shall be reviewed and any appropriate corrective action taken. In addition to satisfying the visual inspection acceptance criteria, freedom-of-motion of mechanical snubbers shall be verified using at least one of the following: (1) manually induced snubber movement; or (2) evaluation of in-place snubber piston setting; or (3) stroking the mechanical snubber through its full range of travel.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.3-4 Revision 15

Snubbers TRM 8.7.3 Table 8.7.3-1 (page 3 of 4)

Snubber Inservice Testing Program B. Functional Test Program

1. General Requirements At least once per inspection interval, a representative sample of each group of snubbers in use in the plant shall be functionally tested in accordance with Section B.2 and B.3.

Response to the failures shall be in accordance with Section B.4.

For all snubbers, functional testing shall consist of either bench testing or in-place testing.

2. Inspection Interval and Sample Criteria The snubbers may be categorized into groups based on physical characteristics and accessibility. Each group may be tested independently from the standpoint of performing additional tests if failures are discovered. The snubbers (if any) attached to the Steam Generators and Reactor Coolant Pumps shall be considered their own group.

The inspection interval for functional testing shall be every fuel cycle, and may not begin any earlier than 60 days before a scheduled refueling outage.

Snubbers which are scheduled for removal for seal maintenance may be included in the test sample prior to any maintenance on the snubber.

The representative sample shall consist of at least 10 percent (rounded off to next highest integer) of each group of snubbers in use in the Plant.

3. Acceptance Criteria For hydraulic snubbers (either inplace testing or bench testing), the test shall verify that:

a. Snubber piston will allow the hydraulic fluid to "bypass" from one side of the piston to the other to assure unrestrained action is achieved within the specified range of velocity or acceleration in both tension and compression.

b. When the snubber is subjected to a movement which creates a load condition that exceeds the specified range of velocity or acceleration, the hydraulic fluid is trapped in one end of the snubber causing suppression of that movement.

c. Snubber release rate or bleed rate, where required, occurs in compression and tension.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.3-5 Revision 15

Snubbers TRM 8.7.3 Table 8.7.3-1 (page 4 of 4)

Snubber Inservice Testing Program For mechanical snubber in place and bench testing, the test shall verify that:

a. The force that initiates free movement of the snubber rod in either tension or compression is less than the specified maximum drag force.

b. Activation (restraining action) is achieved in both tension and compression within the specified range.

4. Response to Failures For each snubber failure per Section B.3:

a. Declare the supported system or train LCO(s) not met or enter Tech Spec 3.0.8 and TNC 8.7.3; and

b. Within the specified inspection interval, functionally test an additional sample of at least 5 percent of the snubber units from the group that the Nonfunctional snubber unit is in.

The functional testing of an additional sample of at least 5 percent from the Nonfunctional snubbers group is required for each snubber unit determined to be Nonfunctional in subsequent functional tests, or until all snubbers in that group have been tested; and

d. The cause of the snubber failure will be evaluated and, if caused by a manufacturing or design deficiency, all snubbers of the same or similar design subject to the same defect shall be functionally tested within the current inspection interval.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.3-6 Revision 15

Snubbers TRM B 8.7.3 BASES 8.7.3 Snubbers The following basis has been left largely intact as a historical reference that also complies with the current ASME OM Code snubber requirements. The historical snubber bases for Davis-Besse has been based in the Technical Specifications and/or Technical Requirements manual under approved relief requests. The 4th 10 year Interval for Inservice Inspection (ISI) and Inservice Testing (IST) began on September 21st, 2012 and will expire on September 20th, 2022.

During this time interval, the snubber program will comply with all requirements of the 2004 edition of the ASME OM Code with the 2005 OMa and 2006 OMb Addenda. This requires a Snubber Program Plan and Snubber Program Procedure. Both of these documents are available in the document management system and contain the additional detail for implementatioin of the Davis-Besse Snubber Program in accordance with the ASME OM Code.

All safety-related snubbers are required to meet their associated support function(s) to ensure the structural integrity of the reactor coolant system and all other safety-related systems is maintained during and following a dynamic event. Snubbers excluded from this inspection program are those installed on safety-related systems for loads other than dynamic or 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 during a dynamic event.

A Nonfuntional Snubber is defined as:

1. For Visual test

a. The fluid no longer is supplied to the valve block, or b Mounting pins are disengaged from the snubber.

c. Attachment to foundation or supporting structure is not secure.

2. For Functional test:

a. The snubber (excluding and anchors, i.e., pin-to-pin) does not meet specified test criteria.

The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems. Therefore, the required inspection interval is determined by the number of nonfunctional snubbers found during and inspection, the total population or group size for each snubber type, and the previous inspection interval. Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection.

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

When the cause of the rejection of a snubber is clearly established and remedied for that snubber and for any other snubbers that may be generically susceptible, and verified by functional testing, that snubber may be exempted from being counted as nonfunctional.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.7.3-1 Revision 15

Snubbers TRM B 8.7.3 BASES 8.7.3 Snubbers (Cont)

When a snubber is found not meeting its associated support function through a visual inspection or functional test, entry into Tech Spec 3.0.8 and TNC 8.7.3 is required and an engineering evaluation is performed as part of the Corrective Action Program response, in addition to the determination of the snubber mode of failure, in order to determine if any safety-related component or system has been adversely affected by the nonfunctional snubber.

To provide assurance of snubber functional reliability, a representative sample of the installed snubbers will be functionally tested every fuel cycle. Observed failures of these sample snubbers shall require functional testing of additional units. When a snubber is found not meeting its associated support function due to failure to lock up or failure to move (i.e., frozen in place), the cause will be evaluated for further action or testing.

In cases where the case of failure has been identified, additional snubbers that have a high probability for the same type of failure or are being used in the same application that caused the failure shall be tested. This requirement increases the probability of locating snubbers not meeting its associated support function without testing 100% of the snubbers.

Hydraulic snubbers and mechanical snubbers may each be treated as a different entity for the above programs.

Contingency Measures A.2.2, A.2.3, and A.2.4 are described below. Those verifications are required to satisfy regulatory commitment O21964.

Every time the provisions of LCO 3.0.8 are used, verification of at least one train (or subsystem) of systems supported by the inoperable snubbers would remain capable of performing their required safety or support functions for postulated design loads other than seismic loads is required. LCO 3.0.8 does not apply to non-seismic snubbers. In addition, a record of the design function of the inoperable snubber (i.e., seismic vs. non-seismic), implementation of any restrictions, and the associated plant configuration shall be available on a recoverable basis for NRC inspection.

When LCO 3.0.8.a is used in MODES 1,2,3 or 4, at least one EFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), must be available.

When LCO 3.0.8.b is used in MODES 1,2,3 or 4, at least one EFW train (including a minimum set of supporting equipment required for its successful operation) not associated with the inoperable snubber(s), or some alternative means of core cooling (e.g., feed and bleed, fire water system or "aggressive secondary cooldown" using the steam generators) must be available.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.7.3-2 Revision 15

Liquid Storage Tanks TRM 8.7.4 8.7 PLANT SYSTEMS 8.7.4 Liquid Storage Tanks TECHNICAL NORMAL CONDITIONS TNC 8.7.4 The quantity of radioactivity contained in each outdoor liquid storage tank that is not surrounded by liners, dikes, or walls, capable of holding the tank contents and that does not have tank overflows and surrounding area drains connected to the Liquid Radwaste Treatment System, shall be within the limit of Technical Specification 5.5.11.b.

APPLICABILITY: At all times.

CONTINGENCY MEASURES

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

Separate Nonconformance entry is allowed for each unprotected outdoor liquid storage tank.

NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. The quantity of A.1 Suspend addition of Immediately radioactivity contained in radioactive material to the any unprotected outdoor tank.

liquid storage tank not within limit. AND A.2 Reduce tank contents to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months within the limit.

AND A.3 Describe the event leading In the next Radioactive to this condition. Effluent Release Report DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.4-1 Revision 0

Liquid Storage Tanks TRM 8.7.4 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.7.4.1 Verify the quantity of radioactivity contained in each 7 days when unprotected outdoor liquid storage tank is within the radioactive materials limit by analyzing a representative sample of the tank are being added to the contents. tank DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.4-2 Revision 0

Liquid Storage Tanks TRM B 8.7.4 BASES 8.7.4 Liquid Storage Tanks Restricting the quantity of radioactive material contained in the specified tanks provides assurance that in the event of an uncontrolled release of the tank contents, the resulting concentrations would be less than the limits of 10 CFR Part 20, Appendix B, Table 2, Column 2, at the nearest potable water supply and the nearest surface water supply in an unrestricted area.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.7.4-1 Revision 0

Explosive Gas Mixture TRM 8.7.5 8.7 PLANT SYSTEMS 8.7.5 Explosive Gas Mixture TECHNICAL NORMAL CONDITIONS TNC 8.7.5 The concentration of oxygen in the Waste Gas System shall be limited to 2% by volume when the hydrogen concentration exceeds 4% by volume.

APPLICABILITY: At all times.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Oxygen concentration in A.1 Reduce Waste Gas System 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months the Waste Gas System oxygen concentration to

> 2% by volume and within the limit.

4% by volume.

B. Contingency Measure B.1 Suspend additions of waste Immediately and associated gases to the system.

Restoration Time of Nonconformance A not AND met.

B.2 Initiate action to reduce the Immediately OR oxygen concentration to 2% by volume.

Oxygen concentration contained in the waste gas system is > 4% by volume, and Hydrogen is

> 4% by volume.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.5-1 Revision 0

Explosive Gas Mixture TRM 8.7.5 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.7.5.1 Verify the oxygen concentration contained in the Waste Continuously Gas System is within the limit by monitoring the waste gases in the Waste Gas System.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.5-2 Revision 0

Explosive Gas Mixture TRM B 8.7.5 BASES 8.7.5 Explosive Gas Mixture This specification is provided to ensure that the concentration of potentially explosive gas mixtures contained in the waste gas system is maintained below the flammability limits of hydrogen with oxygen. Maintaining the concentration of hydrogen or 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.

TVR 8.7.5.1, which requires continuous monitoring of the Waste Gas System, is performed by the instrumentation covered in TNC 8.3.6, Waste Gas System Oxygen Monitoring. The contingency measures in TNC 8.3.6 address loss of monitoring.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.7.5-1 Revision 0

Auxiliary Feedwater System TRM 8.7.6 8.7 PLANT SYSTEMS 8.7.6 Auxiliary Feedwater System TECHNICAL NORMAL CONDITIONS TNC 8.7.6 The following Auxiliary Feedwater System features shall be FUNCTIONAL or in the required condition:

a. Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks;

b. Auxiliary Feed Pump Suction Pressure Interlocks; and

c. CW 196, CW 197, FW 32, FW 91, and FW 106 in the closed position.

APPLICABILITY: MODES 1, 2, and 3.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Auxiliary Feed Pump A.1 Restore Auxiliary Feed Pump 7 days Turbine Inlet Steam Turbine Inlet Steam Pressure Pressure Interlock Interlock to FUNCTIONAL Nonfunctional. status.

B. Contingency Measure B.1 Initiate action to evaluate Immediately and associated failure to meet TNC per Restoration Time of TRM Section 7.3.

Nonconformance A not met.

C. TNC 8.7.6 not met for C.1 Evaluate OPERABILITY of Immediately reasons other than associated AFW train per Nonconformance A. Technical Specification 3.7.5.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.6-1 Revision 8

Auxiliary Feedwater System TRM 8.7.6 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.7.6.1 Not Used 8.7.6.2 ----------------------------NOTE---------------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after steam line pressure is > 275 psig.

Perform CHANNEL FUNCTIONAL TEST on the 31 days Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlock.

8.7.6.3 Perform CHANNEL FUNCTIONAL TEST on the 31 days Auxiliary Feed Pump Suction Pressure Interlocks.

8.7.6.4 Not Used 8.7.6.5 Verify valves CW 196, CW 197, FW 32, FW 91, and 31 days FW 106 are in the closed position.

8.7.6.6 Perform CHANNEL CALIBRATION on the Auxiliary 24 months Feed Pump Turbine Inlet Steam Pressure Interlocks.

8.7.6.7 Perform CHANNEL CALIBRATION on the Auxiliary 12 months Feed Pump Low Suction Pressure Interlocks.

8.7.6.8 Perform CHANNEL CALIBRATION on the Auxiliary 24 months Feed Pump Low-Low Suction Pressure Interlocks.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.6-2 Revision 8

Auxiliary Feedwater System TRM B 8.7.6 BASES 8.7.6 Auxiliary Feedwater System Verification of the turbine plant cooling water valves (CW 196 and CW 197), the startup feedwater pump suction valves (FW 32 and FW 91), and the startup feedwater pump discharge valve (FW 106) in the closed position is required to address the concerns associated with potential pipe failures in the auxiliary feedwater pump rooms, that could occur during operation of the startup feedwater pump.

The FUNCTIONALITY of the Auxiliary Feed Pump Turbine Inlet Steam Pressure Interlocks is required only for high energy line break concerns and does not affect Auxiliary Feedwater System OPERABILlTY. However, an additional feature is provided for the Train 2 interlocks. A separate relay prevents actuation of the Train 2 interlocks until after a 25 second time delay.

This feature provides protection for the specific scenario of a steam line break on OTSG 1, concurrent with a loss of offsite power and a single failure of AFP 1. Failure of this time delay relay (a non Tech Spec support feature) should be addressed under Nonconformance C (Ref. 1).

The Service Water System is the safety-related secondary source of the water and must be available for the associated Auxiliary Feedwater System train to be OPERABLE. The transfer is initiated upon detection of a low suction pressure at the suction of the auxiliary feedwater pumps by suction pressure interlock switches. These pressure switches, upon sensing low suction pressure, will automatically transfer the suction of the auxiliary feedwater pumps to the Service Water System. On a sustained low-low suction pressure, additional Auxiliary Feedwater Pump Suction Pressure Interlocks will operate to close the steam supply valves to protect the turbine driven auxiliary feedwater pumps from cavitation. Both the low and the low-low suction Auxiliary Feed Pump Suction Pressure Interlocks are non Tech Spec support features that are required for OPERABILITY of the associated auxiliary feedwater train.

REFERENCES 1. UFSAR, Section 9.2.7.3 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.7.6-1 Revision 7

MDFP Lube Oil Interlocks TRM 8.7.7 8.7 PLANT SYSTEMS 8.7.7 Motor Driven Feedwater Pump (MDFP) Lube Oil Interlocks TECHNICAL NORMAL CONDITIONS TNC 8.7.7 The requirements of TVR 8.7.7.1 shall be performed.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. TNC 8.7.7.1 not met. A.1 Evaluate OPERABILITY of Immediately MDFP train per Technical Specification 3.7.5.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.7.7.1 Verify proper operation of the Motor Driven Feedwater 24 months Pump lube oil interlocks.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.7.7-1 Revision 0

MDFP Lube Oil Interlocks TRM B 8.7.7 BASES 8.7.7 None DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.7.7-1 Revision 0

A.C. Sources - Operating TRM 8.8.1 8.8 ELECTRICAL POWER SYSTEMS 8.8.1 AC Sources - Operating TECHNICAL NORMAL CONDITIONS TNC 8.8.1 a. The requirements of TVR 8.8.1.1 and TVR 8.8.1.2 shall be performed.

AND

b. The switchyard shall not be in a single point vulnerable configuration.

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

CONTINGENCY MEASURES Nonconformance Contingency Measures Restoration Time A. TVR 8.8.1.1 or A.1 Initiate action to evaluate Immediately TVR 8.8.1.2 not met. failure to meet TNC per TRM Section 7.3.

AND Immediately A.2 Evaluate EDG(s) for OPERABILITY requirements of Technical Specification 3.8.1.

B. Switchyard in a single B.1 Enter Technical Specification Immediately point vulnerable 3.8.1 Condition for one configuration. inoperable offsite circuit.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.8.1-1 Revision 23

A.C. Sources - Operating TRM 8.8.1 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.8.1.1 For each EDG verify that the auto-connected loads do 24 months on a not exceed the 2000 hour0.0231 days 0.556 hours 0.00331 weeks 7.61e-4 months rating of 2838 kw. STAGGERED TEST BASIS 8.8.1.2 For each EDG, perform inspection in accordance with ------------NOTE-----------

procedures prepared in conjunction with its Extension allowance manufacturers recommendations or revision 5 of the per TRM 6.4 is not ESI - EMD Optimized Maintenance Program. allowed.

30 months DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.8.1-2 Revision 23

A.C. Sources - Operating TRM B 8.8.1 BASES 8.8.1 AC Sources - Operating Technical Verification Requirement 8.8.1.1 ensures that the emergency diesel generators are capable of supplying all automatically connected loads.

NRC Log Number 5668, dated May 31, 2000 provides guidance relative to the operability of the offsite A.C. electrical power sources. Switchyard equipment can be removed from service or switchyard breakers can be opened that leaves the remaining switchyard equipment vulnerable to a single point failure that would result in a loss of offsite power. These configurations do not satisfy GDC 17 requirements. In these cases, the switchyard is considered to be in a Vulnerable Configuration that does not satisfy TNC 8.8.1.b. TS 3.8.1 Condition A must be entered and the appropriate actions taken as specified.

Whenever switchyard components are out of service, the resulting configuration must be evaluated to determine if a vulnerable switchyard configuration exists. The 345 kV system operating procedure provides additional details.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.8.1-1 Revision 23

SBODG Availability TRM 8.8.2 8.8 ELECTRICAL POWER SYSTEMS 8.8.2 Station Blackout Diesel Generator (SBODG) Availability TECHNICAL NORMAL CONDITIONS TNC 8.8.2 The requirements of TVR 8.8.2.1 and TVR 8.8.2.2 shall be performed.

APPLICABILITY: MODES 1, 2, 3, and 4, with an emergency diesel generator (EDG) removed from service for preventive maintenance activities of greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. SBODG Nonfunctional. A.1 Initiate action to evaluate Immediately failure to meet TNC per TRM Section 7.3.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.8.2.1 Verify the SBODG is capable of connection to the 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to essential bus associated with an emergency diesel removing an generator removed from service for preventive emergency diesel maintenance. generator from service for preventive maintenance of greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.8.2-1 Revision 7

SBODG Availability TRM 8.8.2 TVR VERIFICATION FREQUENCY 8.8.2.2 Verify performance of SBODG test DB-SC-04271 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months prior to within the previous 30 days. removing emergency diesel generator from service for preventive maintenance of greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.8.2-2 Revision 7

SBDG Availability TRM B 8.8.2 BASES 8.8.2 SBODG Availability The Contingency Measures provide verification that the Alternate A. C. (AAC) power source, the Station Blackout Diesel Generator, is functional and capable of being connected to the safety bus associated with the inoperable Emergency Diesel Generator. These actions are consistent with the NRC criteria for ensuring that the probability of a core damage accident given a Station Blackout event is not significantly increased due to the performance of Emergency Diesel Generator preventive maintenance of greater than 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months during power operations.

These actions are applicable only when an Emergency Diesel Generator becomes inoperable for the performance of preventive maintenance. (Reference NRC Safety Evaluation for License Amendment 206, dated February 26, 1996)

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.8.2-1 Revision 0

Communications TRM 8.9.1 8.9 REFUELING OPERATIONS 8.9.1 Communications TECHNICAL NORMAL CONDITIONS TNC 8.9.1 Direct communications shall be maintained between the control room and personnel at the refueling station.

APPLICABILITY: During movement of irradiated fuel assemblies in containment.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. Direct communications A.1 Suspend movement of Immediately between the control room irradiated fuel assemblies.

and personnel at the refueling station AND not maintained.

A.2 Suspend operations involving Immediately positive reactivity additions that could result in loss of required SDM or boron concentration.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.9.1.1 Verify direct communications between the control room 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months prior to the start and personnel at the refueling station. of movement of irradiated fuel assemblies in containment AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.9.1-1 Revision 0

Communications TRM 8.9.1 BASES 8.9.1 Communications The requirements for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity condition during movement of irradiated fuel assemblies in containment.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.9.1-1 Revision 0

Crane Travel - Fuel Handling Building TRM 8.9.2 8.9 REFUELING OPERATIONS 8.9.2 Crane Travel - Fuel Handling Building TECHNICAL NORMAL CONDITIONS TNC 8.9.2 Loads > 2430 pounds shall be prohibited from travel over fuel assemblies in the spent fuel pool.

APPLICABILITY: With fuel assemblies in the spent fuel pool.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. TNC 8.9.2 not met. A.1 Place the crane load in a Immediately safe condition.

TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.9.2.1 Verify the weight of each load, other than a fuel Prior to moving each assembly, is 2430 pounds. load over fuel assemblies in the spent fuel pool DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.9.2-1 Revision 18

Crane Travel - Fuel Handling Building TRM 8.9.2 BASES 8.9.2 Crane Travel - Fuel Handling Building The restriction on movement of loads in excess of the nominal weight of a fuel assembly in a failed fuel container over other fuel assemblies in the spent fuel pool ensures that in the event this load is dropped (1) the activity release will not exceed the source term assumed in the design basis fuel handling accident for outside containment, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.

The restriction is applicable to the movement of heavy loads over fuel assemblies in the spent fuel pool. The restriction does not apply to movement of heavy loads such as a Dry Shielded Canister top shield plug over fuel assemblies in the Cask Pit while using the single-failure-proof main hoist on the spent fuel cask crane as part of a single-failure-proof handling system where heavy load drops do not have to be postulated or evaluated (Ref.1).

REFERENCES 1. UFSAR, Section 9.1.5.2.2B DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.9.2-1 Revision 18

Spent Fuel Assembly Storage TRM 8.9.3 8.9 REFUELING OPERATIONS 8.9.3 Spent Fuel Assembly Storage TECHNICAL NORMAL CONDITIONS TNC 8.9.3 The following limits apply to spent fuel assembly storage:

a. The heat generation rate of each spent fuel assembly stored in the spent fuel pool shall be 80,209 watts, and the heat generation rate per heat transfer surface area of assembly cladding shall be

< 445 watts/ft2; and

b. The total decay heat load of stored spent fuel assemblies following a discharge of fuel assemblies to the spent fuel pool shall be

< 30.15 x 106 BTU/hr.

APPLICABILITY: Whenever fuel assemblies are stored in the spent fuel pool.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. A fuel assembly A.1 Initiate action to evaluate Immediately exceeding the heat failure to meet TNC per generation rate limit or TRM Section 7.3.

heat generation rate per heat transfer surface area limit is stored in the spent fuel pool.

B. The total decay heat load B.1 Initiate action to evaluate Immediately of stored spent fuel failure to meet TNC per assemblies following a TRM Section 7.3.

discharge of fuel assemblies to the spent fuel pool exceeds the limit.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.9.3-1 Revision 7

Spent Fuel Assembly Storage TRM 8.9.3 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.9.3.1 Verify by administrative means that the fuel assembly Prior to storing the fuel heat generation rate and heat generation rate per heat assembly in the spent transfer surface area are within limits. fuel pool 8.9.3.2 Verify by administrative means that the total decay heat Prior to discharging load of stored spent fuel assemblies following a fuel assemblies to the discharge of fuel assemblies to the spent fuel pool is spent fuel pool within the limit.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.9.3-2 Revision 7

Spent Fuel Assembly Storage TRM 8.9.3 BASES 8.9.3 Spent Fuel Assembly Storage The restriction on the heat generation rate and the heat generation rate per heat transfer surface area of assembly cladding of each stored spent fuel assembly is consistent with the thermal-hydraulic analyses.

The restrictions on the total decay heat load of stored spent fuel assemblies are consistent with the thermal-hydraulic analyses.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.9.3-1 Revision 0

Fuel Handling Bridge TRM 8.9.4 8.9 REFUELING OPERATIONS 8.9.4 Fuel Handling Bridge TECHNICAL NORMAL CONDITIONS TNC 8.9.4 The control rod hoist and fuel assembly hoist of the fuel handling bridge shall be used for movement of control rods or fuel assemblies, and shall be FUNCTIONAL with:

a. The control rod hoist having:

1. A capacity of 3000 pounds; and

2. An overload cutoff limit of 2650 pounds.

b. The fuel assembly hoist having:

1. A capacity of 3000 pounds; and

2. An overload cutoff limit of 2700 pounds.

APPLICABILITY: During movement of control rods or fuel assemblies within the reactor pressure vessel.

CONTINGENCY MEASURES NONCONFORMANCE CONTINGENCY MEASURES RESTORATION TIME A. One or more hoists A.1 Suspend use of any Immediately Nonfunctional. nonfunctional hoist from operations involving the movement of control rods or fuel assemblies within the reactor pressure vessel.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.9.4-1 Revision 0

Fuel Handling Bridge TRM 8.9.4 TECHNICAL VERIFICATION REQUIREMENTS TVR VERIFICATION FREQUENCY 8.9.4.1 For each control rod hoist used for movement of Once within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months control rods or fuel assemblies within the reactor prior to start of pressure vessel, perform a hoist load test of movement of control 3000 pounds and verify an automatic overload cutoff rods or fuel assemblies when the control rod hoist load exceeds 2650 pounds. within the reactor pressure vessel 8.9.4.2 For each fuel assembly hoist used for movement of Once within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months control rods or fuel assemblies within the reactor prior to start of pressure vessel, perform a hoist load test of movement of control 3000 pounds and verify an automatic overload cutoff rods or fuel assemblies when the fuel assembly hoist load exceeds 2700 within the reactor pounds. pressure vessel DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 8.9.4-2 Revision 0

Fuel Handling Bridge TRM 8.9.4 BASES 8.9.4 Fuel Handling Bridge The FUNCTIONALITY requirements of the hoist bridges used for movement of fuel assemblies ensures that: 1) fuel handling bridges will be used for movement of control rods and fuel assemblies, 2) each hoist has sufficient load capacity to lift a fuel element, and 3) the core internals and pressure vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 B 8.9.4-1 Revision 0

Facility Staff TRM 10.2.1 10.0 ADMINISTRATIVE CONTROLS 10.2 ORGANIZATION 10.2.1 Facility Staff Each on duty shift shall be composed of at least the following minimum shift crew composition:

APPLICABLE MODES LICENSE CATEGORY 1, 2, 3, and 4 5 and 6 Senior Operating License 2 1(1)

Shift Technical Advisor 1 Operating License 2 1 (1) Does not include the licensed Senior Operator or Senior Operator Limited to Fuel Handling.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.2.1-1 Revision 11

Process Control Program Procedures TRM 10.4.1 10.0 ADMINISTRATIVE CONTROLS 10.4 PROCEDURES 10.4.1 Process Control Program Procedures Written procedures shall be established, implemented and maintained covering process control program activities.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.4.1-1 Revision 5

Process Control Program (PCP) Changes TRM 10.5.1 10.0 ADMINISTRATIVE CONTROLS 10.5 PROGRAMS AND MAUAL 10.5.1 Process Control Program (PCP) Changes

a. Changes to the PCP shall be documented, and records of reviews performed, shall be retained as required by the USAR, Chapter 17, "Quality Assurance Program".

This documentation shall contain:

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

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

b. Changes to the PCP shall become effective after review and acceptance by the PORC and the approval of the plant manager.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.5.1-1 Revision 5

In-Plant Radiation Monitoring TRM 10.5.2 10.0 ADMINISTRATIVE CONTROLS 10.5 PROGRAMS AND MANUALS 10.5.2 In-Plant Radiation Monitoring A program shall be provided which will ensure the capability to accurately determine the airborne iodine concentration in vital areas under accident conditions. This program shall include the following:

1. Training of personnel;

2. Procedures for monitoring; and

3. Provisions for maintenance of sampling and analysis equipment.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.5.2-1 Revision 5

REPORTING REQUIREMENTS TRM 10.6 10.0 ADMINISTRATIVE CONTROLS 10.6 REPORTING REQUIREMENTS Technical Requirements Manual Section 10.6, Reporting Requirements has been developed to provide a central location for various Technical Specification reports. These reports are not controlled or revised under the change process for the Technical Requirements Manual. The documents contained in Section 10.6 are revised and issued as required by Technical Specification Section 5.6.

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.6-1 Revision 5

Annual Radiological Environmental Operating Report TRM 10.6.1 10.0 ADMINISTRATIVE CONTROLS 10.6 REPORTING REQUIREMENTS 10.6.1 Annual Radiological Environmental Operating Report DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.6.1-1 Revision 5

Radioactive Effluent Release Report TRM 10.6.2 10.0 ADMINISTRATIVE CONTROLS 10.6 REPORTING REQUIREMENTS 10.6.2 Radioactive Effluent Release Report DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.6.2-1 Revision 5

Core Operating Limits Report (COLR)

TRM 10.6.3 10.0 ADMINISTRATIVE CONTROLS 10.6 REPORTING REQUIREMENTS 10.6.3 Core Operating Limits Report (COLR)

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.6.3-1 Revision 5

Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

TRM 10.6.4 10.0 ADMINISTRATIVE CONTROLS 10.6 REPORTING REQUIREMENTS 10.6.4 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.6.4-1 Revision 5

Post Accident Monitoring Report TRM 10.6.5 10.0 ADMINISTRATIVE CONTROLS 10.6 REPORTING REQUIREMENTS 10.6.5 Post Accident Monitoring Report DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.6.5-1 Revision 5

Steam Generator Tube Inspection Report TRM 10.6.6 10.0 ADMINISTRATIVE CONTROLS 10.6 REPORTING REQUIREMENTS 10.6.6 Steam Generator Tube Inspection Report DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.6.6-1 Revision 5

Remote Shutdown System Report TRM 10.6.7 10.0 ADMINISTRATIVE CONTROLS 10.6 REPORTING REQUIREMENTS 10.6.7 Remote Shutdown System Report DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 10.6.7-1 Revision 5

Station Emergency Ventilation System Boundary TRM Appendix A Table A-1 (page 1 of 1)

Access Openings Required to be Closed to Ensure Shield Building Area Negative Pressure Boundary is Intact (as required by Technical Specification 3.7.12)

Air Tight Doors Door No. Description Elevation (ft) 100 Access Door from the No. 1 ECCS Pump Room (Room 105) to 545 Pipe Tunnel 101 104A Access Door from Stair AB-3 to the No.1 ECCS Pump Room 555 (Room 105) 105 Access Door from Passge 110A to the area above the Decay 555 Heat Coolers 107 Access Door from the No.2 ECCS Pump Room (Room 115) to 555 the Miscellaneous Waste Monitor Tank and Pump Room (Room114) 108 Access Door from the No.2 ECCS Pump Room (Room 115) to 565 the Detergent Waste Drain Tank and Pump Room (Room 125) 201-A Access Door from Corridor 209 to the No.1 Mechanical 565 Penetration Room (Room 208) 204 Access Door from Passage 227 to the Makeup Pump Room 565 (Room 225) 205 Access Door from Passage 227 to the No.2 Mechanical 565 Penetration Room (Room 236) 307 Access Door from Corridor 304 to the No.3 Mechanical 585 Penetration Room (Room 303) 308 Access Door from Corridor 304 to the No.4 Mechanical 585 Penetration Room (Room 314)

Blowout Panels Total No. Location Elevation (ft) 1 No. 2 Mechanical Penetration Room (Room 236) 565 6 No. 3 Mechanical Penetration Room (Room 303) 585 6 No. 4 Mechanical Penetration Room (Room 314) 585 DAVIS-BESSE NUCLEAR POWER STATION UNIT NUMBER 1 A.1 Revision 0

ML22263A417

Contents

Text

6.0 INTRODUCTION

6.0 INTRODUCTION

6.0 INTRODUCTION

6.1 BACKGROUND

3.3.1. APPLICABILITY

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ML22263A417

Text

6.0 INTRODUCTION

6.0 INTRODUCTION

6.0 INTRODUCTION

6.1 BACKGROUND

3.3.1. APPLICABILITY

Navigation menu

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