Text

Amends 73 & 66 to Licenses DPR-42 & DPR-60,respectively, Changing Administrative Tech Specs Per Generic Ltr 83-43 Re 10CFR50.73 & Revising Table of Contents
	ML20129E520
Person / Time
Site:	Prairie Island
Issue date:	06/25/1985
From:	Butcher E; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20129E525	List: ML20129E520; ML20129E534;
References
	GL-83-43, TAC-55445, TAC-55446, NUDOCS 8507160840
	Download: ML20129E520 (50)
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UNITED STATES NUCLEAR REGULATORY COMMISSION

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NORTHERN STATES POWER COMPANY DOCKET N0. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 73 License No. DPR-42 l

l 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Northern States Power Company (the licensee) dated July 11, 1984, as supplemented April 26, 1985, complies ~with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and i

regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health l

and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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l 8507160840 050625 PDR ADOCK 05000202 P

PDR

. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-42 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 73, are hereby in-corporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 3

h Edward J. Bu cher, Acting Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

l Changes to the Technical Specifications Date of Issuance: June 25, 1985 i

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UNITED STATES

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NUCLEAR REGULATORY COMMISSION O

tj WASHINGTON, D. C. 20555

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NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 66 License No. DPR-60 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Northern States Power Company (the licensee) dated July 11, 1984, as supplemented April 26, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set.forth in 10 CFR Chapter I; The facility will operate in conformity with the application, B.

the provisions of the Act, and the rules and regulations of the Comission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be inimical-to the comon defense and security or to the health and safety of the public; and l

E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.

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. 2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-60 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 66, are hereby in-corporated in the license. The licensee shall operate the facility in accordance with the Technical Specifica-tions.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 2

D Edward J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: June 25, 1985

C ATTACHMENT TO LICENSE AMENDMENTS NOS. 73 AND 66 TO FACILITY OPERATING LICENSE NOS. DPR-42 AND DPR-60 DOCKET NOS. 50-282 AND 50-306 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of changes.

Remove Insert Remove Insert i thru iv i thru x TS.1-1 TS.1-1 TS.4.11-2 TS.4.11-2 TS.1-6 TS.1-6 TS.4.12-5 TS.4.12-5 TS.3.1-12 TS.3.1-12 TS.4.13-1 TS.4.13-1 TS.3.3-3 TS.3.3-3 TS.4.13-2 TS.4-13-2 TS.3.6-2 TS.3.6-2 TS.4.13-3 TS.4.13-3 TS.3.6-5 TS.3.6-5 Table TS.4.17-1 Table TS.4.17-1 TS.3.8-1 TS.3.8-1 (page 1 of 2)

(page 1 of 2)

.TS.3.8-3 TS.3.8-3 Table TS.4.17-2 Table TS.4.17-2 TS.3.8-4 TS.3.8-4 (page 1 of 2)

(page 1 of 2)

TS.3.9-1

.TS.3.9-1 Table TS.6.1-1 Table TS.6.1-1 TS.3.9-3 TS.3.9-3 Figure TS.6.1-1 Figure TS.6.1-1 Table TS.3.9-1 Table TS.3.9-1 TS 6.2-1 TS.6.2-1 (page 1 of 2)

(page 1 of 2)

TS.6.2-3 TS.6.2-3 l

Table TS.3.9-2 Table TS.3.9-2 TS.6.2-5 TS.6.2-5 (page 1 of 2)

(page 1 of 2)

TS.6.2-6 TS.6.2-6 TS.3.12-1 TS.3.12-1 TS.6.7.3a TS.6.7-4 TS.6.7-4 Table TS.3.12-1 i

(8 pages)

TS.6.7-5 TS.6.7-5 TS.3.14-1 TS.3.14-1

.TS.6.7-6 TS.3.14-2 TS.3.14-2 TS.6.7-7 TS.3.14-3 TS.3.14-3 TS.6.7-8 TS.3.14-4 TS.3.14-4 TS.3.15-1 TS.3.15-1 Table TS.4.10-1 Table TS.4.10-1 (page 3 of 4)

(page 3 of 4) l i

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TS-1 TECHNICAL SPECIFICATIONS TABLE OF CONTENTS TS SECTION TITLE PAGE 1.0 DEFINITIONS TS.1-1 A. Reportable Event TS.1-1 B. Deleted C. Containment System Integrity TS.1-2 D. Degree of Instrumentation Redundancy TS.1-3 E. Instrumentation Surveillance TS.1-3 F. Safety Limits TS.1-3 G. Limiting Safety System Settings TS.1-4 H. Limiting Conditions for Operation TS.1-4 I. Operable TS.1-4 J. Power Operation TS.1-4 K. Protection Instrumentation and ' Logic TS.1-4 L. Quadrant Power Tilt M. Rated Power TS.1-5 TS.1-5 N. Reactor Critical TS.1-5 O. Refuel!ng Operation P. Shutdown TS.1-5 TS.1-5 Q. Thgrmal Power TS.1-6 R. Physics Tests TS.1-6 S. Startup Operation TS.1-6 T. Fire Suppression Water System TS.1-6 U. Minimum Pressurization Temperature (MPT)

TS.1-6 W. Process Control Program (PCP)

TS.1-7 X. Solidification TS.1-7 Z. Offsite Dose Calculation Manual (ODCM)

TS.1 A.A. Source Check TS.1-7 A.B. Gaseous Radwaste Treatment System TS.1-7 A.C. Ventilation Exhaust Treatment System TS.1-7 A.D. Purging TS.1-8 A.E. Venting TS.1-8 A.F. Members of the Public TS.1-8 A.G. Site Boundary TS.1-8 A.H. Unrestricted Areas TS.1-8 Prairie Island Unit 1 - Amendment No.

73 Prairie Island Unit 2 - Amendment No.

66

TS-11 TABLE OF CONTENTS (Continued)

TS SECTION TITLE PAGE 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGTS.2.1-1 2.1 Safety Limit, Reactor Core TS.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure TS.2.2-1 2.3 Limiting Safety System Settings, Protective Instrumentation TS.2.3-1 A. Protective Instrumentation Settings for Reactor TS.2.3-1 Trip

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B. Protective Instrumentation Settings for Reactor Trip Interlocks TS.2.3-3 C. Control Rod Withdrawal Stops TS.2.3-4 3.0 LIMITING CONDITIONS FOR OPERATION TS.3.1-1 3.1 Reactor Coolant System TS.3.1-1 A. Operational Components

1. Co31 ant Pumps TS.3.1-1

2. Steam Generators TS.3.1-1A

3. Requirements for Decay Heat Removal Below 350*F TS.3.1-1A

4. Pressurizer TS.3.1-2 o9 5.. Reactor Coolant Vent System TS.3.1-2A B. Heatup and Cooldown TS.3.1-4 C. Leakage TS.3.1-9,,

D. Maximum Coolant Activity TS.3.1-11 E. Maximum Reactor Coolant Oxygen, Chloride TS.3.1-14 and Fluoride Concentration F. Minimum conditions for criticality TS.3.1-17 G. Minimum Conditions for RCS Temperature Less TS.3.1-19 Than MPT H. Primary Coolant System Pressure Isolation TS.3.1-21 Valves 3.2 Chemical and Volume Control System TS.3.2-1 3.3 Engineered Safety Features TS.3.3-1 A. Safety Injection and Residual Heat Removal TS.3.3-1 Systems B. Containment Cooling Systems TS.3.3-2 C. Component Cooling Water System TS.3.3-4 D. Cooling Water System TS.3.3-5 3.4 Steam and Power Conversion System TS.3.4-1 A.1 Safety and Relief Valves TS.3.4-1 A.2 Auxiliary Feed System TS.3.4-1 A.3 Steam Exclusion System TS.3.4-2 A.4 Radiochemistry TS.3.4-2 Prairie Island Unit 1 - Amendment No. 73 Prairie Island Unit 2 - Amendment No. 66 l

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TS-iii TABLE OF CONTENTS TS SECTION TITLE PAGE 3.5 Instrumentation System TS.3.5-1 3.6

-Containment System TS.3.6-1 A. Containment System Integrity TS.3.6-1 B. Containment Intctnal Pressure TS.3.6-3 C. Containment and Shield Building Air Temperature TS.3.6-3 D. Containment Shell Temperature TS.3.6-3 E. Emergency Air Treatment Systems TS.3.6-3A F. Electric Hydrogen Recombiners TS.3.6-3A 3.7 Auxiliary Electrical System TS.3.7-1 3.8 Refueling and Fuel Handling TS.3.8-1 D. Spent Fuel Pool Special Ventilation System TS. 3. 8-2A 3.9 Radioactive Effluents TS.3.9-1 A. Liquid Effluents TS.3.9-1 B. Gaseous Effluents TS.3.9-3 C. Solid Radioactive Waste TS.3.9-6 D. Dose from All Uranium Fuel Cycle Sources TS.3.9-7 E. Radioactive Liquid Effluent Monitoring TS.3.9-7 Instrumentation F. Radioactive Gaseous Effluent Monitoring TS.3.9-8 Igstrumentation 3.10 Control Rod and Power Distributien Limits TS.3.10-1 A. Shutdown Reactivity TS.3.10-1 B. Power Distribution Limits TS.3.10-1 C. Quadrant Pover Tilt Limits TS.3.10-4 i

D. Rod Insertion Limits

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E. Rod Misalignment Limitations TS.3.10-5 t

TS.3.10-6 F. Inoperable Rod Position Indicator Channels TS.3.10-6 G. Inoperable Rod Limitations TS.3.10-6*

H. Rod Drop Time TS.3.10-7 j

I. Monitor Inoperability Requirements TS.3.10-7 J. DNB Parameters TS.3.10-8 3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Snubbers TS.3.12-1 3.13 Control Room Air Treatment System TS.3.13-1 3.14 Fire Detection and Protection Systems TS.3.14-1 A. Fire Detection Instrumentation TS.3.14-1 B. Fire Suppression Water System TS.3.14-1 C. Spray and Sprinkler Systems TS.3.14-2 D. Carbon Dioxide System TS.3.14-3 E. Fire Hose Stations TS.3.14-3 F. Yard Hydrant Hose Houses TS.3.14-4 i

l G. Penetration Fire Barriers TS.3.14-4 3.15 Event Monitoring Instrumentation TS.3.15-1 A. Process Monitors TS.3.15-1 B. Radiation Monitors TS.3.15-1 Prairie Island Unit 1 - Amendment No. 73 Prairie Island. Unit 2 - Amendment No. 66

TS-iv d

TABLE OF CONTENTS TS SECTION

_ TITLE PAGE 4.0 SURVEILLANCE REQUIREMENTS TS.4.1-1 4.1 Operational Safety Review TS.4.1-1 4.2 Inservice Inspection and Testing of Pumps and TS.4.2-1 Valves Requirements A. Inspection Requirements TS.4.2-1 B. Corrective Measures C. Records TS.4. 2 -2 TS.4.2-3 4.3 Primary Coolant Systen Pressure Isolation Valves TS.4.3-1 4.4 Containment System Tests TS.4.4-1 A. Containment Leakage Tests TS.4.4-1 B. Emergency Charcoal Filter Systems TS.4.4-3 C. Containment Vacuum Breakers TS.4.4-4 D. Residual Heat Removal System TS.4.4-4 E. Containment-Isolation Valves TS.4.4-5 F. Post Accident Containment Ventilation System TS.4.4-5 G. Containment and Shield Building Air TS.4.4-5 Temperature H. Containment Shell Temperature TS.4.4-5 I. Electric Hydrogen Recombiners TS.4.4-5 4.5 Engin,eered Safe,ty Features TS.4.5-1 A. System Tests TS.4.5-1

1. Safety Injection System

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TS.4.5-1

2. Containment Spray System TS.4.5-1

3. Containment Fan Coolers TS.4.5-2

4. Component Cooling Water System TS.4.5-2

5. Cooling Water System TS.4.5-2 B. Component Tests TS.4.5-2

1. Pumps TS. 4. 5 2. Containment Fan Motors TS.4.5-3

3. Valves TS.4.5-3 4.6 Periodic Testing of Emergency Power System TS.4.6-1 A. Diesel Generators TS.4.6-1 B. Station Batteries TS.4.6-1A C. Pressurizer Heater Emergency Power. Supply TS.4.6-1A 4.7 Main Steam Stop Valves TS.4.7-1 4.8 Steam and Power Conversion Systems TS.4.8-1 A. Auxiliary Feedwater System TS.4.8-1 B. Power Operated Relief Valves TS.4.8-1 C. Steam Exclusion System TS.4.8-2 4.9 Reactivity Anomalies TS.4.9-1 4.10 Radiation Environmental Monitoring Program TS.4.10-1 A. Sample Collection and Analysis TS.4.10-1 B. Land Use Census TS.4.10-2 C. Interlaboratory Comparison Program TS.4.10-2 4.11 Radioactive Source Leakage Test TS.4.11-1 Prairie Island Unit 1 - Amendment No. 73 Prairie Island Unit 2 - Amendment No. 66 v-

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TS-v TABLE OF CONTENTS TS SECTION TITLE PAGE 4.12 Steam Generator Tube Surveillance TS.4.12-1 A. Steam Generator Sample Selection and TS.4.12-1 Inspection B. Steam Generator Tube Sample Selection TS.4.12-1 and Inspection C. Inspection Frequencies TS.4.12-3 D. Acceptance Criteria TS.4.12-4 E. Reports 4.13 Snubbers TS.4.12-5 TS.4.13-1 4.14 Control Room Air Treatment System Tests TS.4.14-1 4.15 Spent Fuel Pool Special Ventilation System TS.4.15-1 4.16 Fire Detection and Protection Systems TS.4.16-1 A. Fire Detection Instrumentation TS.4.16-1 B. Fire Suppression Water System TS.4.16-1 C. Spray and Sprinkler Systems TS.4.16-3 D. Carbon Dioxide System TS.4.16-3 E. Fire Hose Stations TS.4.16-3 F. Yard Hydrant Hose Houses TS.4.16-4 G. Penetration Fire Barriers TS.4.16-4 4.17 Radioactive Effluents Surveillance TS.4.17-1 A. Liquid Effluents TS.4.17-1 B. Gaseous Effluents TS.4.17-2 C. Solid Radioactive Waste TS.4.lf-4 D. Dose from All Uranium Fuel Cycle Sources TS.4.17-4 4.18 Reactor Coolant Vent System Paths TS.4.17-2 A. Vent Path Operability TS.4.17-2 B. System Flow Testing TS.4.17-2 l

Prairie Island Unit 1 - Amendment No. 69,73 Prairie Island Unit 2 - Amendment No. 63,66 t

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TS-vi TABLE OF CONTENTS TS SECTION TITLE PAGE 5.0 DESIGN FEATURES 5.1 Site TS.5.1-1 5.2 Containment System TS.S.1-1 A. ' Containment Structures TS.S.2-1 1.

Containment Vessel TS.S.2-1 2.

Shield Building TS.S.2-1 TS.5.2-2 3.

Auxiliary Building Special Ventilation Zone TS.5.2-2 B.

Special Ventilation Systems TS.5.2-2 C.

Containment System Functional Design 5.3 Reactor TS.5.2-3 A.

Reactor Core TS.5.3-1 B.

Reactor Coolant System TS.S.3-1 C.

Protection Systems TS.5.3-1 5.4 Engineered Safety Features TS.S.3-1 5.5 Radioactive Waste Systems TS.S.4-1 A.

Accidental Releases TS.5.5-1 B.

Routine Releases TS.S.5-1 1.

Liquid Wastes TS.5.5-1 2.

Gaseous Wastes TS.5.5-1 3.

Solid Wastes TS.S.5-2 C.

TS.5.5-3 Process and Effluent Radiological Monitoring System 5.6 Fuel Handling TS.5.5-3 A.

Criticality Consideration TS.S.6-1 B.

Spent Fuel Storage Structure TS.5.6-1 C.

Fuel Handling TS.5.6-1 Spent Fuel Storage Capacity TS.S.6-2 D.

TS.5.6-3 Prairie Island Unit 1 - Amendment No.32, 73 Prairie Island Unit 2 - Amendment No. 26,66

TS-vii TABLE OF CONTENTS (Continued)

TS SECTION TITLE PAGE 6.0 ADMINISTRATIVE CONTROLS 6.1 Organization TS.6.1-1 TS.6.1-1 6.2 Review and Audit TS.6.2-1 A. Safety Audit Committee (SAC)

TS.6.2-1

1. Membership TS.6.2-1

2. Qualifications TS.6.2-1

3. Meeting Frequency

4. Quorum TS. 6. 2-2 TS.6.2-2

5. Responsibilities

6. Audit TS.6.2-2 TS.6.2-3

7. Authority

8. Records TS.6.2-4 TS.6.2-4

9. Procedures TS.6.2-4 B. Operations Committee (OC)

TS.6.2-5

1. Membership TS.6.2-5

2. Meeting Frequency

3. Quorum TS.6.2-5 TS.6.2-5

4. Responsib111 ties

5. Authority TS.6.2-5 6.. Records TS.6.2-6 TS.6.2-6

7. Procedures TS.6.2-6 6.3 Special Inspections and Audits TS.6.3-1 6.4 Safety Limit Violation TS.6.4-1 6.5 Plant Operating Procedures TS.6.5-1 A. Plant Operations TS.6.5-1 B. Radiological TS.6.5-1 C. Maintenance and Test TS.6.5-3 D. Process Control Program (PCP)

TS.6.5-3 *

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E. Offsite Dose Calculation Manual (ODCM)

TS.6.5-4 F. Temporary Changes to Procedures TS.6.5-4 6.6 Plant Operating Records TS.6.6-1 A. Records Retained for Five Years TS.6.6-1 B. Records R2tained for the Life of the Plant TS.6.6-1 6.7 Reporting Requirements TS.6.7-1 A. Routine Reports TS.6.7-1

1. Startup Report TS.6.7-1

2. Occupational Exposure Report TS.6.7-2

3. Monthly Operating Report TS.6.7-2

4. Steam Generator Tube Inservice Inspection TS.6.7-2

5. Semiannual Radioactive Ef fluent Release TS.6.7-3 Report

6. Annual Summaries of thteorological Data TS.6.7-3

7. Report of Safety and Relief Valve Failures TS.6.7-4 and Challenges B. Reportable Events TS.6.7-4 Prairie Isladd Unit 1 - Amendment No.73 Prairie Island Unit 2 - Amendment No.66 O

TS-viii TABLE OF CONTENIS (Continued)

TS SECTION TITLE PAGE C. Environmental Reports TS.6.7-4

1. Annual Radiation Environmental Monitoring TS.6.7-4 Reports

2. Environmental Special Reports TS.6.7-5

3. Other Environmental Reports TS.6.7-5 (non-radiological, non-aquatic)

D. Special Reports 6.8 Environmental Qualification TS.6.7-5 TS.6.8-1 s*

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Prairie Island Unit 1 - Amendment No. 46,73 i

i Prairie Island-Unit 2 - Amendment No. 40,66

TS-ixl TECHNICAL SPECIFICATIONS s

LIST OF TABLES TS TABLE TITLE 3.1-1 Unit 1 Reactor Vessel Toughness Data (Unirradiated) 3.1-2 Unit 2 Reactor Vessel Toughness Data (Unirradiated) 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2 Instrument Operating Conditions for Reactor Trip 3.5-3' instrument Operating Conditions for Emergency Cooling System 3.5-4 Instrument Operating conditions for Isolation Functions 3.5-5 Instrument Operating Conditions for Ventilation Systems 3.5-6 Instrument Operating Conditions for Auxiliary Electrical System 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Gaseous Effluent Monitoring Instrumentation 3.14-1 Safety Related Fire Detection Instruments 3.15-1 Event Monitoring Instrumentation - Process & Containment l

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3.15-2 Event Monitoring Instrumentation - Radiation 4.1-1 Minimum Frequencies for Checks, Calibrations and Test of Instrument Channels 4.1-2A Minimum Frequencies for Esaipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special laservice Inspection Requirements 4.4-1 Unit 1 and Unit 2 Penetration Designation for Leakage Tests 4.10-1 Radiation Environmental Monitoring Program (REMP)

Sample Collection and Analysis 4.10-2 REMP - Maximum Values for the Lower Limits of Detection 4.10-3 REMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Monitoring Instrumentation Surveillance Requirements 4.17-3 Radioactive Liquid Waste Scapling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program 5.5-1 Anticipated Annual Release of Radioactive Material. in Liquid Effluents from Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating l

Plant (Per Unit) l 6.1-1 Minimum Shift Crew Composition 1

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Unit 1 - Amendment No. 59, 59, 97 73 Unit 2 - Amendment No. 44, 53, 77., 66 l

TS-x l

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APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop Operation 3.1-1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1.2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 Effect of Fluence and Copper Content on Shift of RT Reactor Vessel Steels Exposed to 550 F TemperatureNDT ' #

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3.1-4 Fast Neutron Fluence (E >l MeV) as a Function of Full Power Service' Life 3.1-5 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 pCi/ gram DOSE EQUIVALENT I-131 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents 3.10-1 Required Shutdown Reactivity Vs Reactor Boron Concentration 3.10-2 Control Bank Insertion Limits 3.10-3 Insertion Limits 100 Step overlap with One Bottomed Rod 3.10-4 Insertion Limits 100 Step Overlap with One Inoperable Rod 3.10-5 Hot Channel Factor Normalized Operating Envelope 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10-7 Normalized Exposure Dependent Function BU(E ) for Exxon Nuclear Company Fuel 3

3.10-8 y(Z) as a Function of Core Height 4.4-1 Shield Building Design In-Leakage Rate 6.1-1 NSP Corporation Organization Relationship to On Site Operating Organizations 6.1.2 Prairie Island Nuclear Generating Plant Functioval Organization for On-Site Operating Group Prairie Island Unit 1 - Amendment No. 52, 59, 66, 70, 73 Prairie Island Unit 2 - Amendment No. 46, 55, 60, 6(, 66

TS.1-1 1.0 DEFINITIONS The succeeding infrequently used terms are explicitly defined so that a uniform interpretation of the specifications may be achieved.

A.

Reportable Event A Reportable Event shall be any plant occurrence or event which must be reported, per 10 CFR 50.73, requiring written reports td the Commission.

B.

Deleted l

Prairie Island Unit 1 - Amendment No. 9,73 Prairie Island Unit 2 - Amendment No. 4,66

TS.1-6 3.

Refueling Shutdown A reactor is in the refueling shutdown condition when a refueling operation is scheduled, Keff is equal to or less than 0.95, and the reactor coolant average temperature is less than 140 F.

Q.

Thermal Power Thermal power of a unit is the total heat transferred from the reactor core to the coolant.

R.

Physics Tests Physics tests are those conducted to measure fundamental character-istics of the core and related instrumentation.

Physics tests are conducted such that the core power is sufficiently reduced to allow for the perturbation due to the test and therefore avoid exceeding power distribution limits in Specification 3.10.B.

Low power physics tests are run at reactor powers less than 5% of rated power.

S.

Startup Operation The process of heating up a reactor above 200 F, making it critical, and bringing it up to power operation.

T.

Fire Suppression Water System The fire suppression water system consists of:

Water sources; pumps; and distribution piping with associated sectionalizing isolation valves.

Such valves include yard-hydrant valves, and the first valve ahead of the water flow' alarm device on.each sprinkler, hose standpipe, or spray system riser.

U.

Minimum Pressurization Temperature (MPT)

Reactor coolar.t system temperature below which reactor coolant system pressure is limited by Figures TS.3.1-1 and TS.3.1-2, Reactor Coolant System Heatup and Cooldown Limitations.

Prairie Island Unit 1 - Amendment No. 39, 73 l

Prairie Island Unit 2 - Amendment No. 33, 66

C TS.3.1-12 4.

If a reactor is at or above cold shutdown:

(a) With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E_

microcuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.1-2B until the specific activity of the primary coolant is restored to within its limits.

A special report shall be submitted to the Commission within 30 days. This report shall contain the.results of the specific activity analyses together with the' following information:

1.

Reactor power history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded.

2.

Fuel burnup by core region, 3.

Clean-up flow history starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, 4.

History of de-gassing operations, if any, starting 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to the first sample in which the limit was exceeded, and 5.

The time duration when the specific activity of the primary coolant exceeded 1 0 microcurie per gram DOSE EQUIVALENT I-131.

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The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the site boundary will not exceed an appropriately small_ fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1 0 GPM.

The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations.

These values are conservative in that specific site parameters of the Prairie Island site, such as site boundary location and meteorological conditions, were not considered in this evaluation.

Specification 3.1.D.2, permitting power operation to continue for limited time periods with the primary coolant's specific activity greater than 1 0 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure TS.3.1-5, accommodates possible iodine spiking phenomenon which may occur following changes in thermal power. Operation with specific activity levels exceed-ing 10 microcuries/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure TS.3 1-5 must be restricted to no more than 800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months per year (approximately 10 percent of the unit's yearly operating time) since the activity levels allowed by Prairie Island Unit 1 - Amendment No. 5/7, 73 Prairie Island Unit 2 - Amendment No. AAI, 66

TSc3.3-3 4

The spray additive tank contains not less than 2590 gallons of c.

solution with a sodium hydroxide concentration of 9% to 11% by weight inclusive.

d.

Manual valves in the above systems that could (if improperly positioned) reduce spray flow below that assumed for accident analysis, shall be blocked and tagged in the proper position. During power opera-tion, changes in valve position will be under direct administrative control.

Automatic valves, interlocks, ducts, dampers, controls and piping '

e.

associated with the above components and required for accident condi-tions are operabic.

3 i

f.

The following motor-operated valve conditions shall exist:

(1) The Unit 1 operation, containment spray system motor-operated valves MV32096 and MV32097 shall be closed and shall have the 1

motor congrol center supply breakers open.

i (2) For Unit 2 operation, containment spray system motor-operat'ed valves MV32108 and MV32109 shall be closed and shall have the motor control center supply breakers open.

5 2.

During startup operation or power operation, any one of the following m

conditions of inoperability may exist for each unit provided startup

.- ~

operation is discontinued until operability is restored. The reactor shall be placed in the hot shutdown condition if during power operation operability is not restored within the time specified. The reactor shall be placed in the cold shutdown condition if operability is not restored within an additional 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

One fan cooler unit or one duct for a fan cooler unit may be out of a.

service for a period not to exdeed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Prior to initiating repairs and once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter, both containment spray pumps and the remaining three fan cooler units shall be demonstrated to be operable.

b.

One containment spray pump may be out of service for a period not to exceed 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The remaining containment spray pump and the four fan units shall be demonstrated to be operable before initiating repairs and once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

1 Prairie Island Unit 1 - Amendment No. 65, 73 Prairie Island Unit 2 - Amendment No. 59, 66 b

e n-

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mv.--.,-,,,,w

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1 l

TS.3.6-2 4.

Positivs reac'tivity changes shall not oe made by boron dilution when containment system integrity is not intact unless the more restrictive of the following reactivity conditions is met: K rg is < 0.95 or the e

boron concentration > 2000 ppm.

The vacuum breaker system shall be considered operable for containment 5.

system integrity when both valves in each of two vacuum breakers, including actuating and power circuits, are operable or when one vacuum breaker is daily demonstrated as operable and the other has been inoperable for no more than 7 days under conditions for which contain-ment integrity is required.

6.

Automatic containment isolation valves listed in Table TS.4.4-1 shall be considered operable for containment system integrity when all auto-matic isolation valves, including actuation circuits, for each penetra-tion are operable or the inoperable valve is deactivated in the closed position, or at least one valve in each penetration having an inoperable valve is locked closed.

7.

a.

The 36-inch containment purge system double gasketed blind flanges shall be installed whenever the reactor is above cold shutdown.

b.

The 18-inch containment inservice purge system double gasketed blind flanges shall be installed whenever the reactor is above cold shut-down except as noted below.

The inservice purge system may be operated above cold shutdown when c.

required for safe plant operation if the following conditions are met:

1.

The debris screens are installed on the supply and exhaust ducts in containment.

2.

Both valves shall satisfactorily pass a local leak rate test prior to use.

3.

The two automatic primary containment isolation valves and the automatic shield building ventilation damper in each duct that penetrates containment shall be operable, including instruments and controls associated with them.

4.

The blind flanges (i.e., 42B (53 in Unit 2) and 43A (52 in Unit 2) shall be reinstalled and satisfactorily pass a local leak rate test, each time after the in-service purge system is used.

8.

During maintenance, construction and testing activities, containment integrity is considered intact if the auxiliary building special vent zone boundary is opened intermittently, provided such openings are under direct administrative control and can be reduced to less than 10 square feet within 6 minutes following an accident.

Unit 1 - Amendment No. 67,73 Unit 2 - Amendment No. $7/j 66

TS.3.6-5 6

This specification also prevents positive insertion of reactivity whenever containment integrity is not maintained if such addition would violate the respective shutdown margins.

The boron concentra-tion must be maintained at > 2000 ppm and Keff must be < 0.95 if the containment system is to be disabled with the vessel open.

The 2 psig limit on internal pressure provides adequate margin between the maximum internal pressure of 46 psig and the peak resulting from the postulated Design Basis Accident.(2gecident pressure

/

The containment vessel is designed for 0.8 psi internal vacuum, the '

/

occurrence of which will be prevented by redundant vacuum breaker systems.

The containment has a nil ductility transition temperature of 0 F.

above NDIT during power operation when con *.ainmen accident (3)(6) is based on an initial shield building annu temperature of 60 F and an initial containment vessel air temperature of 104 F.

The calculated period following LOCA for which the shield building annulus pressure is positive, and the calculated off-site doses are sensitive to this initial air temperature difference.

accidentanalysis(g).specified 44 F tem erature difference is consistent with the The The initial testing of inleakage into the shield building and the auxiliary building special ventilation zone (ABSVZ) has resulted in greater specified inleakage (Figure TS 4.4-1, change No.1) and the necessity to deenergize the turbine building exhaust fans in order to achieve a negative pressure in the ABSVZ (TS 3.6.A.10).

The staff's changing allowable containment leconservative calculation of doses for thes offset the increased leakage.(5) ak rate from 0.5% to 0.25%/ day would High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers for all emergency air treatment systems.

The Charcoal adsorbers are installed to reduce the potential release of radio-iodine to the environment.

The in-place test results should indicate a HEPA filter leakage of less than 1% through DOP testing and a charcoal adsorber leakage of less than 1% through halogenated hydrocarbon testing.

The laboratory carbon sample test resu~'s should indicate a radioactive methyl iodide removal efficiency of at least 90% under test conditions which are more severe than accident conditions.

The satisfactory completion of these i

j Prairie Island Unit 1 - Amendment No. 11,73 4

Prairie Island Unit 2 - Amendment No. ZZ,66 i

i L_

TS.'3.8-1 3.8 REFUELING AND FUEL HANDLING Applicability Applies to operating limitations during fuel-handling and refueling operations.

Obj ec tives To ensure that no incident could occur during fuel handling and refueling opera-tions that would affect public health and safety.

Specification A.

During refueling operations the following conditions shall be satisfied:

1.

The equipment hatch and at least one door in each personnel air lock shall be closed.

In addition, at least one isolation valve chall be operable or locked closed in each line which penetrates the containment and provides a direct path from montainment atmosphere to the outside.

2.

Radiation levels in fuel handling areas, the containment and the spent fuel storage pool areas shall be monitored continuously.

3.

The core suberitical neutron flux shall be continuo'usly monitored by at least two neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containment, which.are in service whenever core geometry is being changed.

not being changed, at least one' neutron flux monitor shall be in service.When core 4.

During reactor vessel head removal and while loading and unloading fuel from the reactor, the boron concentration of the reactor coolant system and the refueling canal shall be sufficient to ensure that the more restric-tive of the following reactivity conditions is met:

K gf 1 0.95 or the boron concentration > 2000 ppm.

e verified by chemical analysis daily.. The required boron concentration shall be 5.

During movement of fuel assemblies or control rods out of the reactor vessel, at least 23 feet of water shall be maintained above the reactor vessel flange.

fuel assemblies or control rods and at least once every day while cavity is flooded.

6.

At least one residual heat removal pump shall be operable and running.

The pump may be shut down for up to one hour to facilitate movement of fuel or core components.

7.

If the water level above the top of the reactor vessel flange is less than 20 feet, except for control rod unlatching / latching operations or upper internals removal / replacement, operable.

both residual heat removal loops shall be 8.

If Specification 3.8.A.6 or 3.8.A.7 cannot be satisfied, all fuel handling operations in containment shall be suspended, the containment integrity requirements of Specification 3.8.A.1 shall be satisfied, and no reduction in reactor coolant boron concentration shall be made.

Prairie Island Unit 1 - Amendment No. 47, 63, Prairie Island Unit 2 - Amendment No. Al, 51, 73 66

TS.3.8-3 Basis The equipment and general procedures to be utilized during refueling are discussed in the FSAR.

Detailed instructions, the precautions specified above, and the design of the fuel handling equipment incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling o public health and safety.(1)perations that would result in a hazard to Whenever changes are not being made in core geometry, one flux conitor is sufficient. This permits maintenance of the instrumentation.

Continuous monitoring of radiation levels (B. above) and neutron flux provides immediate indication of an unsafe condition.

The residual heat removal pump is used to maintain a uniform boron concentration.

Under rodded and unrodded conditions, the Keff of the reactor must be < 0.95 and the boron concentration must be > 2000 ppm as indicated in A.4.

Periodic checks of refueling water boron concentration insure that proper shutdown margin is maintained.

A.9 above allows the control room operator to inform the manipulator ope.ator of any impending unsafe condition detected from the main control board indicators during fuel movement.

No movement of fuel in the reactor is permitted until the reactor has been suberitical for at least 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to permit decay of the fission products in the fuel.

analysis.(2) The delay time is consistent with the fuel handling accident l

The spent fuel assemblies will be loaded into the spent fuel cask for ship-ment to a reprocessing plant af ter sufficient decay of fission products.

In loading the cask into a carrier, there is a potential drop of 66 feet.(4)

The cask will not be loaded onto the carrier for shipment prior to a 3-month storage period.

At this time, the radioactivity has decayed so that a release of fission products from all fuel assemblies in the cask would result in off-site doses less than 10 CFR Part 100. It is assumed, for this dose analysis, that 12 assemblies rupture after storage for 90 days. Other assumptions are the same as those used in the dropped fuel assembly accident in the SER, Section 15.

The resultant doses at the site boundary are 94 Rems to the thyroid and 1 Rem whole body.

The Spent Fuel Pool Special Ventilation System (3) is a safeguards system which l

maintains a negative pressure in the spent fuel enclosure upon detection of high area radiation.

The Spent Fuel Pool Normal Ventilation System is auto-matically isolated and exhaust air is drawn through filter modules containing a roughing filter, particulate filter, and a charcoal filter before discharge to the environment via one of the Shield Building exhaust stacks.

Two completely redundant trains are provided. The exhr.ust fan and filter of each train are shared with the corresponding train of the Containment In-service Purge System.

High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the iodine adsorbers in Prairie Island Unit 1 - Amendment No. 17, 25, 73 Prairie Island Unit 2 - Amendment No. II,19,66 l

l l

l

TS.3.8-4 C

l each SFPSVS filter train.

The charcoal adsorbers are installed to reduce the potential release of radiciodine to the environment.

The in-place test results should indicate a hEPA filter leakage of less than 1% through DOP testing and a charcoal adsorber leakage of less than 1% through halogenated hydrocarbon testing.

The laboratory carbon sample test results should indicate a radio-active methyl iodide removal efficiency of at least 90% under test conditions which are more severe than accident conditions.

The satisfactory completion of these periodic tests combined with the qualification testing conducted on new filters and adsorber provide a high level of assurance that the emergency air treatment systems will perform as predicted in the accident analyses.

During movement of irradiated fuel assemblies or control rods, a water-level of 23 feet is maintained to provide sufficient shielding.

The water level may be lowered to the top of the RCCA drive shafts for latching and unlatching.

internals removal / replacement.The water level may also be lowered below 20 feet for up The bases for these allowances are (1) the refueling cavity pool has sufficient level to allow time to initiate repairs or emergency procedures to cool the core, internals removal / replacement the level is closely monitored because the(

activity uses this level as a reference point, (3) the time spent at this level is minimal.

1 References (1) FSAR Section 9.5.2 (2) FSAR Section 14.2.1 (3) FSAR Section 9.6 (4) FSAR Page 9.5-20a Prairie Island Unit 1 - Amendment No. 25, 47, 63, 73 Prairie Island Unic 2 - Amendment No. If, 41, 57, 66 1

~'

TS.3.9-1 3.9 RADIOACTIVE EFFLUENTS Applicability Applies at all times to the liquid and gaseou's radioactive effluents from the plant and the solidification and packaging for shipment of solid radioactive waste.

Objective To implement the requirements of 10CFR20,100FR71,10CFR50 Section 50.36a, Appendix A and Appendix I to 10CFR50, 40CFR141, and 40CFR190 pertaining to radioactive effluents.

Specifications A. Liquid Efflgents 1.

Concentration The concentration of liquid radioactive material released at a.

any time from the site (Figure 3.9-1) shall be limited to the q

concentrations specified in 10 CFR Part 20, Appendix B. Table II, Column 2, for radionuclides other than dissolved or entrained noble gases.

For dissolved or entrained ngble #

gases, the concentration shall be limited to 2 x 10 pei/mi total activity.

~

b.

When the concentration of radioactive material in liquid released from the site exceeds the limits in (a) above, immediately restore the concentration within acceptable limits.

2.

Dose a.

The dose or dose commitment to an individual from radioactive materials in liquid effluents released from the site (Figure 3.9-1) shall be limited:

l 1.

During any calendar quarter to $3.0 mrem to the total body and tog 10 mrem to any organ, and 2.

During any calendar year to$6 mrem to the total body and to $20 mrem to any organ.

Unit 1 - Amendment No. 11,//),73

Unit 2 - Amendment No. 5,A/A/,66

i TS.3.9-3 4

b.

With the quantity of radioactive material in any of the above listed tanks exceeding the limit in (a) above, immediately suspend all additions of radioactive materials to the tank and within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months reduce the tank contents to within the limit.

B. Gaseous Effluents

1. Dose Rate

s. The dose rate at any time due to radioactive materials released in gaseous ef fluents from the site (Figure 3.9-2) shall be l

limited to the following values:

1. The dose rate limit for noble gases shall be $500 arem/ year to the total body and $3000 arem/ year to the skin, and

2. The dose rate limit for I-131, tritium, and radioactive particulates with half-lives greater than eight days shall be $ 1500 area / year to any organ

b. With the dose rate (s) exceeding the limits in (a) above, immediately decrease the release rate to within acceptable

limits,

2. Dose from Noble Cases

a. The air dose in unrestricted areas due to noble gases released in gaseous effluents from the site (Figure 3.9-2) shall be limited to the following values:

1.

During any calendar quarter, to 5 10 mrad for gamma radiation and $ 20 arad for beta radiation, and 2.

During any calendar year, to $ 0 mrad for gamma radiation 2

and s,40 mrad for beta radiation.

b. With the calculated air dose from radioactive noble gases in gaseous effluent exceeding any of the above limits, within 30 days submit to the Commission a special report which identifies the cause(s) for exceeding the limit (s) and defines the corrective action (s) taken to reduce the releases and the proposed corrective actions to be taken to assure the subsequent releases will be within the above limits.

73 Unit 1 - Amendment No. 11,4/$.00 Unit 2 - Amendment No. 5, $d/,

d TABLE TS.3.9-1 (Pg 1 of 2)

TABLE TS.3.9-1 RADIOACTIVE' LIQUID EFFLUENT MONITORING INSTRL^tENTATION H1NIMUtt CRANNELS

-INSTRUMENT OPERABLE APPLICABILITY ACTION 1.

Gross Radioactivity Monitors Providing Automatic Termination of Release a.

Liquid Radwaste Effluent Line 1

During releases 1

b.

Steam Generator Blowdown 1/ Unit During releases 2

Effluent Line 2.

Flow Rate Measurement Devices Liquid Radwaste Effluent Line 1

During releases 4

a.

requiring throt-tling of flow b.

Steam Generator Blowdown Flow 1/ Gen During releases 4

3.

Continuous Composite Samplers a.

Each Turbine Building 1/ Unit During releases 3

Sump Effluent Line 4.

Discharge Canal Monitor 1

At all times 3

5.

Tank Level Monitor Condensate Storage Tanks 1/ Unit When tanks 5

a.

are in use b.

Temporary Dutdoor Tanks 1/ Tank Uhen tanks 5

Holding Radioactive Liquid are in use 6.

Discharge Canal Flow System NA At all times (Daily determination and following changes in flow)

Unit 1 - Amendment No. II,/FM, 66 73 Unit 2 - Amendment No. 5, 8/g,

TABLE TS.3.9-2 (Pg 1 of 2)

TABLE TS.3.9-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM CHANNELS INSTR'JMENT OPERABLE APPLICABILITY ACTION 1.

Waste Gas Holdup System 2

During system 2

Explosive Gas operation (Oxygen) Monitors 2.

Effluent Relense Points (Unit No. 1 Reactor Bldg, Unit No. 1 Aux. Blds, Unit No. 2 Reactor Bldg, Unit No. 2 Aux Bldg, Spent Fuel Pool, Radwaste Bldg) a.

Noble Gas Activity 1

During releases 4, 5, 7 Monitor

b.

Iodine Sampler 1

During releases 3

Cartridge c.

Particulate Sampler 1

During releases 3

7 Filter d.

Sampler Flow 1

During releases 1

Integrator l

3.

Air Ejector Noble Gas 1

During power 6

Monitors (Each Unit) operation

Noble gas activity monitors providing automatic termination of releases (except the Radwaste Building which has no automatic isolation function).

l Unit 1 - Amendment No. f/$, 66 73 l

Unit 2 - Amendment No. yg/,

l l

TS.3.12-1 3.12 SNUBBERS Applicability Applies to the operability of safety related snubbers.

Objective To define those conditions of snubber operability necessary to assure safe reactor cperation.

Specification A.

Except as permitted below, all safety related snubbers shall be l

operable above Cold Shutdown. Snubbers may be inoperable in Cold Shutdown and Refueling Shutdown whenever the supported system is not required to be Operable.

B.

With one or more snubbers made or found to be inoperable for any reason when Operability is required, within 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s:

1.

Replace or restore the inoperable snubbers to Operable status and perform an engineering evaluation per Specification 4.13.E on the supported component (s), or 2.

Declare the supported system inoperable and take the action required by the Technical Specifications for inoperability of that system.

Basis All snubbers are required to be Operable above Cold Shutdown to ensure that the structural integrity of the reactor coolant system and all other safety related systems is maintained during and following a seismic or other event initiating dynamic loads.

Prairie Island Unit 1 - Amendment No. 14, 56,73 Prairie Island Unit 2 - Amendment No. 8, 44, 66

v V

TS.3.14-1 3.14 FIRE DETECTION AND PROTECTION SYSTEMS Applicability Applies to instrumentation and plant systems used for fire detection protection of the nuclear safety-related structures, systems, and component and of the plant.

Objective To insure that to nuclear safety are protected from fire damage.the structures, s important

_S_ pecification A.

Fire Detection Instrumentation 1.

Except as specified below, the minimum fire det on whenever equipment in that fire detection zone is required to be operable.

Fire detection instruments located within containment are not required to be operable during the performance of Type A containment leakage rate tests.

2.

If Specification 3.14.A.1 cannot be met:

Within one hour, establish a fire watch patrol to inspect the a.

zone with the inoperable instruments at least once per hour.

Fire sones located inside primary containment are exempt from this requirement when contaisusent integrity is required.

b.

Restore the inoperable instruments to operable status within 14 days or submit a special report to the Commission within for restoring the instruments to operable status 30 l

B.

Fire Suppression Water System 1.

be operable at all times with:Except as specified in 3.14.B.2 or 3.14.B The following pumps, including automatic initiation logic, operable a.

and capable of delivering at of 108 psig.

least 2000 gym at a discharge pressure 1.

Diesel-driven fire pump 2.

Motor-driven fire pump 3.

Screen wash pump Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 26, A9 73 Amendment No. 20, pp,,66

TS.3.14-2 b.

An operable flow path capable of taking suction from the river and transferring the water through distribution piping with operable sectionalizing control or isolation valves to the yard hydrant valves and the first valve ahead of each deluge valve, hose station, or sprinkler system required to be operable.

c 2.

With one or two of the pumps required by Specification 3.14.B.l.a inoperable, restore the inoperable equipment to operable status within seven days or provide a special report to the Commission within 30 days outlining the plans and procedures to be used to provide for the loss of redundancy in the Fire Suppression Water With an inoperable pump, perform the surveillance required system.

by Specification 4.16.B.2.

3.

With the fire suppression water system otherwise inoperable:

Establish a backup Fire Suppression Water System within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

a.

i b.

Provide a special report to the Commission within 30 days out-l lining 3phe actions taken and the plans and schedule for restor-3

~

ing the system to operable status.

If Specification 3.14.B.3.a cannot be met, the reactors shall c.

be placed in hot standby within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown I

within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

T C.

Spray and Sprinkler Systems s

1.

Whenever equipment protected by the following spray and sprinkler systems is required to be operable, the spray and sprinkler system shall be operable:

J a.

Auxiliary Feed Pump Room WP-10 4

b.

Diesel Generator Areas PA-1 i

Unit No. 1 Electrical Penetration Area PA-3 c.

d.

Unit No. 1 Electrical Penetration Area PA-4 Unit'No. 2 Electrical Penetration Area PA-6 e.

f.

Unit No. 2 Electrical Penetration Area PA-7 g.

Screenhouse PA-9 j

2.

If Specification 3.14.C.1 cannot be met, a continuous fire watch with backup fire suppression equipment shall be established within one hour.

Restore inoperable spray and sprinkler systems to operable status within 14 days or submit a special report to the Commission within 30 days outlining the cause of inoperability and the plans for restoring the system to operable status.

Prairie Island Unit 1 - Amendment No. 26, 49,73 Prairie Island Unit 2 - Amendment No. 26, 43,66

TS.3.14-3 D.

Carbon Dioxide System 1.

Except as specified in 3.14.D.3 below, the CO2 system protecting the relay and cable spreading room area shall be operable with a minimum level of 60% in the CO2 storage tank.

2.

During those periods when the relay and cable spreading room area is normally occupied, automatic initiation of the CO2 system may be bypassed.

During those periods when the area is normally unoccupied, the CO2 system shall be capable of automatic initiation unless there are personnel actually in the area.

3.

If specification 3.14.D.1 cannot be met, a continuous fire watch with backup fire suppression equipment shall be stationed in the relay and cable spreading room within one hour. Restore the system to operable status within 14 days or submit a special report to the Commission within 30 days outlining the cause of inoperability and the plans for restoring the system to operable status.

F.

Fire Hose Stations 1.

Whenever equipment protected by hose stations in the following areas is required to be operable, the hose station (s) protecting that area shall be operable:

a.

Diesel Generator Rooms b.

Safety Related Switchgear Rooms Safety Related Areas of Screenhouse c.

d.

Auxiliary Building e.

Control Room f.

Relay & Cable Spreading Room g.

Battery Rooms h.

Auxiliary Feed Pump Room 2.

If Specification 3.14.E.1 cannot be met, within one hour hoses supplied from operable hose stations shall be made available for routing to each area with an inoperable hose station.

Restore the inopetable hose station (s) to Operable status within 14 days or submit a special report to the Commission within 30 days outlining the cause of the inoperability and the plans and schedule for restoring the stations to Operable status.

Prairie Island Unit 1 - Amendment No. 26, 49,73 Prairie Island Unit 2 - Amendment No. 26,43,66

6 TS.3.14-4 F.

Yard Hydrant Hose Houses 1.

Whenever equipment in the following buildings is required to be operable, the yard hydrant hose houses in the main yard loop adjacent to each building shall be operable:

Unit No. 1 Reactor Building a.

b.

Unit No. 2 Reactor Building Turbine Building c.

d.

Auxiliary Building e.

Screen House 2.

If Specification 3.14.F.1 cannot be met, within one hour have sufficient additional lengths of 2-1/2 inch diameter hose located in adjacent operable yard hydrant hose house (s) to provide service to the unprotected area (s).

Restore the yard hydrant hose house (s) to Operable status within 14 days or submit a special report to the Commission within 30 days outlining the cause of the inoperability and the plans and l

schedule for restoring the houses to Operable status.

G.

Penetration Fire Barriers 1

1.

All penetration fire barriers in fire area boundaries protecting equipment required to be operable shall be operable.

2.

If Specification 3.14.G.1 cannot be met, a continuous fire watch shall be established on at least one side of the affected penetra-tion (s) within one hour.

Restore the inoperable penetration fire barriers to Operable status within 14 days or submit a special report to the Commission within l

30 days outlining the cause of the inoperability and the plans and schedule for restoring the barriers to Operable status.

l l

Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 26, AS, 7g, 73 Amendment No. 26, f), $$, 66 i

i

TS.3.15-1 3.5 EVENT MONITORING INSTRUMENTATION Applicability Applies to plant instrumentation which does not perform a protective function, but which provides information to monitor and assess important parameters during and following an accident.

Objectives To ensure that sufficient information is available to operators to determine the effects of and determine the course of an accident to t extent required to carry out required manual actions.

Specification l

A.

Process Monitors I

1.

The event monitoring instrumentation channels specified in Table TS.3.15-1 shall be Operable.

2.

With the number of Operable event monitoring instrumentation channels less than the Required Total Number of Channels shown on Table TS.3.15-1, either restore the inoperable channels to Operable status within seven days, or be in at least Hot Shut-down within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

3.

With the number of Operable event monitoring instrumentation channels less than the Minimum Channels Operable requirements of Table TS.3.15-1, either restore the minimum number of channels to Operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or be in at least Hot Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B.

Radiation Monitors 1.

The event monitoring instrumentation channels specified in Table TS.3.15-2 shall be Operable.

2.

With the number of Operable event monitoring instrumentation channels less than the Required Total Number of Channels shown on Table TS.3.15-2, either restore the inoperable channels to j

operable status within seven days, or prepare and submit a special report to the Commission within 30 days outlining the action taken, the cause of the inoperability, the plans and the schedule for restoring the system to Operable status.

3.

With the number of Operable event monitoring instrumentation channels less than the Minimum Channels Operable requirement of Table TS.3.15-2, initiate the preplanned alternate method of monitoring the appropriate parameters in addition to submitting the report required in 2 above.

Prairie Island Unit 1 - Amendment No. 46, 63, 73 Prairie Island Unit 2 - Amendment No. 46, 51, 66

TABLE TS.4.10-1 (Page 3 of 4)

PRAIRIE ISLAND NUCLEAR GENERATING PLANT

.,22 RADIATION ENVIRONMENTAL MONITORING PROGRAM 40 SAMPLE COLLECTION AND ANALYSIS ww Number of Samples EE Exposure Pathway and

- and/or Sample Sample Locations **

Sampling and Type and Frequency

_ Collection Frequency of Analysis 3.

WATERBORNE EE (Continued) n n wr d.

Sediment from One sample upstream of i:

shoreline plant, one sample down-Semiannually Gamma isotopic analysis I

gg 5e stream of plant, and one of each sample E. E.

from shoreline of recreational area.

EE 4.

INGESTION

.F.F a.

Milk One sample from dairy farm having highest D/Q. one Monthly or biweekly Gamma isotopic and I-131 I

ww Ow

' sample from each of three if animals are on analysis of each sample pasture dairy farms calculated to gg have doses from I-131 >

1 mrem /yr, and one sample from 10-20 miles I

b.

Fish and One sample of one game Invertebrates specie of fish located Semiannually Gamma isotopic analysis i

upstream and downstream on each sample (edible of the plant site portion only on fish) me l

One sample of Invertebrates

?D i

upstream and downstream of on t-a t'8 1

the plant site La H of I

i

- Sample locations are given on the figure and table in the ODCM

?

t sw vo i

I i

l

TS.4.11-2 D.

Tests resulting in 0.005 microcuries or more of removable contamination on the test sample shall be reported to the Commission on an annual basis.

E.

Plant operating records shall be made as follows:

1.

An inventory of licensed radioactive materials in possession shall be maintained current at all times.

2.

The following records shall be retained for 2 years:

Test results in microcuries, for tests performed a.

pursuant to TS 4.11.

b.

Record of annual physical inventory verifying accountability of sources on record.

Bases Licensee's program, facili. ties, personnel, and procedures for safe storage, handling, and use of sealed sources containing radioactive materials is. described in FSAR Section 11.4 The surveillance program described in this specification is a part of licensee's program to detect and control contamination of areas in the plant by such radioactive materials.

Small quantities of byproduct materials are exempt for licensing by 10 CFR 30.18 and therefore are exempt from leakage tests in this specification.

Inhalation or ingestion of such small quantities of byproduct materials from a sealed source veuld l

result in less than one maximum permissible body burden for total body irradiation.

Sources containing less than 0.1 microcurie of plutonium are exempt from leakage tests by 10 CFR 70.39(c) and therefore such quantities of special nuclear materials (including alpha emitters) are exempt from leakage tests in this specification.

The acceptance criteria of less than 0.005 microcuries on the test sample is also based on 10 CFR 70.39(c).

l Prairie Island Unit 1 - Amendment No. 12, 73 l

Prairie Island Unit 2 - Amendment No. 6, 66 i

TS.4.12-5 2.

The steam generator shall be determined OPERABLE after completing the corresponding actions (plug all tubes exceeding the plugging limit and all tubes containing through-wall cracks) required by Table TS.4.12-1.

l E.

Reports 1.

Following each in-service inspection of steam generator tubes, if there are any tubes requiring plugging, the number of tubes plugged in each steam generator shall be submitted in a special report to the Commission within 15 days.

l 2.

Results of steam generator tube inspections which fall into Category C-3 require notification to the Commission prior to resumption of plant operation, and reporting as a special report to the Commission within 30 days. This special report shall proviqg a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

4 Prairie Island Unit 1 - Amendment No. 31, 11,73 Prairie Island Unit 2 - Amendment No. 25, 66

TS.4.13-1 4.13 SNUBBERS Applicability Applies to periodic testing and surveillance requirements of s'af,ety related hydraulic snubbers.

Objective

' Ac u A

\\

To verify the integrity and operability of hydraulic snubbers.

Specification The following surveillance requirements apply to all safety related snubbers. These requirements augment Section XI of the ASME Code.

the inspections required by A.

Visual Inspection of snubbers shall be conducted in accordance with the following schedule:

No. of Snubbers Found. _

Next Required Inoperable per Inspection Period Inspection Period 0

[

18 months 2 25%

1 1

12 months i 25%

2, 6 months 2 25%

3,4 124 days 2 25%

5,6,7 62 days 2 25%

8 or more 31 days 2 25%

The required inspection interval shall not be lengthened more than one step at a time.

s Snubbers may be categorized in two groups, " accessible" or

" inaccessible" based on their accessibility for inspection during reactor operation.

These two groups may be inspected independently according to the above schedule.

B.

Visual inspections shali verify (1) that there are no visible indications of damage or impaired operability, (2) attachments"to the supporting structure are secure, and (3) in those locations where snubber movement can be manually induced without disconnecting the snubber, that the snubber has freedom of movement and is not

' frozen up.

Snubbers which appear inoperable as a result of visual inspection may be determined Operable for the purpose of establishing the next visual inspection interval by:

Prairie Island Unit 1 - Amendment No. 14, 56, 73 Prairie Island Unit 2 - Amendment No. 8, 44, 66 s

TS.4.13-2 6

Clearly establishing the cause of the rejection for that a.

particular snubber and for others that may be generically susceptible; and b.

Functionally testing the affected snubber in the as-found condition and finding it operable per Specification 4.13.D.

However, when the fluid port of a hydraulic snubber is found to be uncovered,~the snubber shall be considered inoperable for purposes of establishing the next visual inspection interval.

All hydraulic anubbers connected to an inoperable common hydraulic fluid reservoir shall be considered as inoperable snubbers.

C.

Except as specified below, functional testing of snubbers shall be conducted at least once per 18 months during cold shutdown.

l of the total of each type snubber shall be functionally tested either Ten percent in place or in a bench test.

For each snubber that does not meet the

s functional test acceptance criteria in Specification 4.13.D below, an additional ten percent of that type of snubber shall be functionally have been tested. tested until no more failures are found or all snubbers of that typ i

x The representative sample selected for functional testing shall include the various configurations, operating environments, and the range of size and capacity of the snubbers.

shall include snubbers from the following three categories. Twenty-five perce l

The first snubber away from a reactor vessel nozzle a.

b.

Snubbers within five feet of heavy equipment (valve, pump, l

turbine, motor, etc.)

Snubbers within ten feet of the discharge of a safety / relief c.

valve Snubbers identified as "High Radiation Area" or " Difficult to Remove" s

I are exempt from functional testing provided a justifiable basis for exemption is presented for Commission review; snubber life testing is performed to qualify snubber operability for all design conditions; or snubbers of the same type, configuration, and similar service have been tested for a ten year period and no failures have occurred.

In such exempt cases, a qualitative test report shall be on file to substitute for the required functional testing.

In addition to the regular sample and specified re-sampling, snubbers which failed the previous functional test shall be retested during the next test period.

If a spare snubber has been installed in place of a failed snubber, then both the failed snubber, if it is repaired and installed in another position, and the spare snubber shall be retested Prairie Island Unit 1 - Amendment No. Id, 56, 73 Prairie Island Unit 2 - Amendment No. 8, 44,66

TS.4.13-3 C'

If any snubber selected for functional testing either fails to lockup or fails to move (i.e., frozen in place) the cause shall be evaluated and all snubbers subject to the same defect shall be functionally teste'd.

This testing is in addition to the regular sample and specified re-samples.

D.

Hydraulic snubber functional tests shall verify that:

r Activation (restraining action) is achieved within the a.

specified range of velocity or acceleration in both tension and compression.

i b.

Snubber bleed, or release rate, where required, is within the specified range in compression or tension.

For snubbers specifically required to not displace under continuous load, the ability of the snubber to withstand load without displace-ment shall be verified.

E.

An engineering evaluation shall be performed for all components supported by inoperable snubbers.

The purpose of this engineering evaluation shall be to determine if the components were adversely affected by the inoperable snubber (s) to ensure that the components remain capable of meeting the designed service.

F.

The installation and maintenance records for each snubber shall be l

reviewed at least once every 18 months to verify that the indicated service life will not be exceeded prior to the next scheduled snubber service life review.

If the indicated service life will be exceeded, the snubber service life shall be reevaluated or the snubber shall be replaced or reconditioned to extend its service life beyond the date of the next scheduled service life review.

This reevaluation, replacement, or reconditioning shall be indicated in the records.

Basis The visual inspection frequency is based upon maintaining a constant level of snubber protection to systems.

Therefore, the required inspection interval varies inversely with the observed snubber failures and is determined by the number of inoperable enubbers found during an inspection.

Inspections.

performed before that interval has elapsed may be used as a new reference point to determine the next inspection.

However, the results of such early inspections performed before the original required time interval has elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval.

will override the previous schedule.Any inspection whose results require a shorter i Prairie Island Unit 1 - Amendment No. Id, gg 73 Prairie Island Unit 2 - Amendment No. 8, 44,b6

Tcble TS.4.17-1 (Page 1 of 2)

TABLE TS.4.17 RADI0 ACTIVE LIQUID EFFLUENT MONITORING INST SURVEILLANCE REQUIREMENTS Channel Check Source Functional Frequency Check Test Calibration Instrument (4)

Frequency Frequency Frequency Liquid Radwaste Effluent Daily during Prior to Quarterly (1) At least once every Line Gross I,adioactivity releases each Monitor 18 months (3) release Liquid Radwaste Effluent Daily during Line Flow Instrument releases At least once every l 18 months Steam Generator Blowdown Daily during Monthly Quarterly (l) At least once every Gross Radioactivity releases Monitors 18 months (3)

Steam Generator Blowdowd Daily during Flow releases At least once every l 18 months Turbine Building Sump Daily during Continuous Composite releases At least once every Samplers (Includes 18 months sample

^

volume check)

Discharge Canal Daily during Monthly Quarterly (2) At least once every Monitor releases 18 months (3)

Discharge Canal Daily during Flow. Instruments releases At least once every l 18 months Condensate Storage Daily Tank Level Monitors Quarterly At least once every-18 months Level Monitors for Daily when Quarterly At least once every Temporary Outdoor in use Tanks Holding when in 18 months when in Radioactive Liquid use use Prairie Island Unit 1 - Amendment No. 59, 73 Prairie Island Unit 2 - Amendment No. 53, 66

Table TS.4.17-2 (Page 1 of 2)

TABLE TS.4.17 RADIOACTIVE CASEOUS EFFLUENT MONITO SURVEILLANCE REQUIREMEh7S Channel Check Source Functional Frequency Check Test Calibration Instrument (4)

Frequency Frequency Frequency Waste Gas Holdup System Daily during Explosive Gas (Oxygen)

Monthly ( )

Quarterly (5) system Manitors operation Effluent Release Points (Unit No. 1 Reactor Bldg. Unit No. 1 Aux Bldg, Unit No. 2 Reactor Bldg, Unit No. 2 Aux Bldg, Spent Fuel Pool, Radwaste Bldg)

Noble Gas Activity Daily during Monthly

Quarterly (1) At least once every Monitor (4) releases (Except Radwaste 18 months (3)

Building)

Noble Gas Activity Daily during Monthly Quarterly (2) At least once every Monitor Radwaste releases Building (4) 18 months (3)

Iodine and Weekly Particulate Samplers Sampler Flow Rate Weekly Monitor At least once every l 18 months Air Ejector Noble Gas Daily during Monthly Quarterly (2) At least once every Monitors (Each Unit) releases 18 months (3)

A source check of the applicable noble gas monitor sha21 Se conducted prior to waste gas decay tinc ot cont.ainmeat 1. urge release.

Prairie Island Unit 1 - Amendment No. 59, 73 Prairie Island Unit 2 - Amendment No. 55, 66

4*.

TABLE TS.6.1-1 gg i

YY MINIMUM SHIFT CREW COMPOSITION (Note 1 and 3) os :-

ow I l CATECORY Ii BOTH UNITS IN COLD SMUTDOWN ONE UNIT IN COLD SHUTDOWN BOTH UNITS ABOVE COLD a$

OR REFUELING SIRITDOWN OR REFUELING SHUTDOWN AND SHUTDOWN ONE UNIT ABOVE COLD SHUTDOWN aa 4

No. Licensed Senior 2 (Note 2)

@,,E Operators (LS0) 2 (Notes 2, 4) 2 (Note 4) l 5

3M Total No. Licensed 4

i 4

Operators (LSO & LO) 5 Total No.' Licensed &

6

$y Unlicensed Operators 8

7 I

w Shif t Technical Advisor 0

1 1

j NOTES:

1.

Shif t crew composition may be one less then the minimum requirements for a period of time not to exceed two hours in order to accommodate an unexpected abserice of one duty shif t crew member provided immediate action is taken to restore the shif t crew composition to l

within the minimum requiremente specified.

2.

Does not include the IIcensed Senior Reactor Operator, or Senior Reactor Operator Limited to Fuel Handling, supervising refueling operatons.

3.

Each LSO and LO shall be licensed on each unit.

4.

One LSO shall be in the control room at all times when a reactor is above cold shutdown l

)

g n

H 8

SENIOR VIE PRESIDENT h

POWER SLPPLf S '.

a

.a.

I I

l N

VICE PRESIDENT fC PLANT ENCIEERING y

VICE PKSIDENT g

.o AW CONSTRUCTION OUALITy MANAGER SYSTEM PROD vlCE PRESIDENT ASSURANCE FLEL OP & MAINTENANCE NUCLEAR SLFPLY

.I E M RATION e V

w

.?

NW l

EMRAI. MANAGER NUCLEAR PLANTS E

g EA00UARTERS NUCLEAR CROUP n

h3 PLANT MANAGER MANAGERS MANAGER I

PR000 Cit 0N e4 (PRAIRIE ISLAND MJCLEAR AR MANAGER l

U TRAINING

& HONTICELLO)

ANALYSIS TECHNICAL NUCLEAR SUPPORT SERVICES f

SERVICES d

O l

ON-SITE n

TRAINING (M-SITE TEcellCAL 8

y l

SERVICES GROLPS g

l STRATI (De y

I i

I l

I I

w g

g g g g-

?

gg --

SAFETY AuurT E

o, Cn coggg,77g,34c, g

o liAS THE RESPONSIBILITY FOR THE FIRE PROTECTION PROGRAM g

FIGtIRE TS.6.1-1 NSP CORPORATION ORGANIZATION RELATIONSHIP TO ON-SITE OPERATING OpGAN ZATIONS L

lCAO ORT 5611 l I

~

TS.6.2-1 6.2 Review and Audit Organizational units for the review and audit of facility operations shall be constituted and have the responsibilities and authorities outlined below:

A.

Safety Audit Committee (SAC)

The Safety Audit Committee provides the independent review of plant operations from a nuclear safety standpoint. Audits of plant operation are conducted under the cognizance of the SAC.

,/

/

1.

Membership The SAC shall consist of at least five (5) persons.

a.

b.

The SAC chairman shall be an NSP representative, not having line responsibility for plant operation, appointed by the Vice President Nuclear Generation. Other SAC members shall

(

be appointed by the Vice President Nuclear Generation or by g

such other person as he mai designate. The Chairman shall appoint a Vice Chairman from the SAC membership to act in his absence.

No more than two members of the SAC shall be from groups c.

holding line responsibility for operation of the plant.

d.

A SAC member may appoint an alternate to serve in his absence, with concurrence of the Chairman. No more than one alternate shall serve on the SAC at any one time. The alternate member shall have voting rights.

2.

Qualifications The SAC members should collectively have the capability a.

required to review activities in the following areas:

nuclear power plant operations, nuclear engineering, chemistry and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering, quality assurance practices, and other appro-priate fields associated with the unique characteristics of the nuclear power plant.

Prairie Island Unit 1 Amendment No. 13, 47,[/[,73-Prairie Island Unit 2 Amendment No. 7, 41, ///,66 i

i i

e

,._._,--,,a-

TS.6.2-3 f.

Investigation of all Reportable Events and events requiring Special Reports to the Commission.

Revisions to the Facility Emergency Plan, Facility Security Plan, and the g.

Fire Protection Program.

h.

Operations Committee minutes to determine if matters considered by that Committee involve unreviewed or unresolved safety questions.

i.

Other nuclear safety matters referred to the SAC by the Operations Committee, plant management or company management.

i

j. All recognized indications of an unanticipated deficiency in some aspect of design or operation of safety-related structures systems, or components.

k.

Reports of special inspections and audits conducted in accordance with specification 6.3.

1.

Changes to the Offsite Dose Calculation Manual (ODCM).

a Review of investigative reports of unplanned releases of radioactive m.

material to the environs.

6.

Audit - The operation of the nuclear power plant shall be audited formally under the cognizance of the SAC to assure safe facility operation.

3

a.

Audits of selected aspects of plant operation, as delineated in Paragraph 4.4 of ANSI N18.7-1972, shall be performed with a frequency commensurate with their nuclear safety significance and in a manner to assure that an audit of all nuclear safety-related activities is completed within a period of two years. The audits shall be performed in accordance with appropriate written instructions and procedures.

b.

Audits of aspects of plant radioactive effluent treatment and radio-logical environmental mcnitoring shall be performed as follows:

1.

Implementation of the Offsite Dose Calculation Manual at least once every two years.

2.

Implementation of the Process Control Program for solidification of radioactive wastes at least once every two years.

3.

The Radiological Environmental Monitoring Program and the results thereof, including quality controls, at least once every year, Periodic review of the audit program should be performed by the SAC at c.

least twice a year to assure its adequacy.

d.

Written reports of the audits shall be reviewed by the Vice President l

Nuclear Generation, by the SAC at a scheduled meeting, and by members of management having responsibility in the areas audited.

Prairie Island Unit 1 - Amendment No. 49, 59, 61, 73 Prairie Island Unit 2 - Amendment No. 43, 53, 55, 66

q TS.6.2-5 B.

Operations Committee (OC) 1.

Membership The Operations Committee shall consist of at least six (6) members drawn from the key supervisors of the onsite staff.

The Plant Manager shall serve as Chairman of the OC and shall appoint a Vice Chairman from the OC membership to act in his absence.

2.

Meeting Frequency The Operations Committee will meet on call by the Chairman or as requested by individual members and at least monthly.

3.

Quorum A majority of the permanent members, including the Chairman or Vice Chairman 4.

Responsibilities - The following subjects shall be reviewed by the Operations Committee:

Proposed tests and experiments and their results, a.

b.

Modifications to plant systems or equipment as described in the Updated Safety Analysis Report and having nuclear safety significance or which involve an unreviewed safety question as defined in Paragraph 50.59 (c), Part 50, Title 10, Code of Federal Regulations.

Proposals which would effect permanent changes to normal c.

and emergency operating procedures and any other proposed changes or procedures that will affect nuclear safety as determined by the Plant Manager.

d.

Proposed changes to the Technical Specifications or operating licenses.

All reported or suspected violations of Technical Specifica-e.

tions, operating license requirements, administrative procedures, operating procedures.

Results of investigations, includinz evaluation and recommendations to prevent recurrence will be reported in writing to the Vice President Nuclear

~

l Generation and to the Chairman of the Safety Audit Committee.

Prairie Island Unit 1 Amendment No. 48,[,73 Prairie Island Unit 2 Amendment No. 43e[/P/,66

i TS.6.2-6 f.

Investigations of all Reportable Events and events requiring Special Reports to the Commission.

g.

Drills on emergency procedures (including plant evacuation) and adequacy of communication with offsite support groups.

h.

All procedures required by these Technical Specifications, including implementing procedures of the Emergency Plan, and the Security Plan, shall be reviewed initially and periodically with a frequency commensurate with their safety significance but at an interval of not more than two

years, i.

Special reviews and investigations, as requested by the Safety Audit Committee.

j. Review of investigative reports of unplanned releases of radioactive material to the gnvirons.

k.

All changes to the Process Control Program (PCP) and the Of fsite Dose Calculation Manual (ODCM).

5.

Authority

-6 The OC shall be advisory to the Plant Manager.

In the event of a disagree-ment between the r2 commendations of the OC and th,e Plant Manager, the course determined by the Plant Manager to be the more conservative will be followed.

A written summary of the disagreement will be sent to the General Manager Nuclear Plants and the Chairman of the SAC for review.

6.

Records Minutes shall be recorded for all meetings of the OC and shall identify

. all documentary material reviewed.

The minutes shall be distributed to each member of the OC, the Chairman and each member of the Safety Audit Committee, the General Manager Nuclear Plants and others designated by the OC Chairman or Vice Chairman.

7.

Procedures A written charter for the OC shall be prepared that contains:

Responsibility and authority of the group a.

b.

Content and method of subnission of presentations to the Operations Committee Mechanism for scheduling meetings c.

d.

Provision for meeting agenda Unit 1 - Amendment No. 49,/y4/,73 Unit 2 - Amendment No. 43,g/g/,66

TS.6.7-4 7.

Report of Safety and Relief Valve Failures and Challenges.

An annual l

report of pressurizer safety and relief valve failures and challenges shall be submitted prior to March 1 of each year.

B.

Reportable Events The following actions shall be taken for Reportable Events:

The Commission shall be notified by a report submitted pursuant to a.

the requirements of Section 50.73 to 10 CFR Part 50, and b.

Each Reportable Event shall be reviewed by the Operations Committee and the results of this review shall be submitted to the Safety Audit Committee and the Vice President Nuclear Generation.

C.

Environmental Reports The reports listed below shall be submitted to the Administrator of the appropriate Regional NRC Office or his designate:

1.

Annual Radiation Environmental Monitoring Report (a)

Annual Radiation Environmental Monitoring Reports covering the operation of the program during the previous calendar year shall be submitted prior to May 1 of each year.

(b)

The Annual Radiation Environmental Monitoring Reports shall include summaries, interpretations, and an analysis of trends of the

.results of the radiological environmental surveillance activities for the report period, including a comparison with preoperational studies, operational controls (as appropriate), and previous environmental surveillance reports and an assessment of the observed impacts of the plant operation on the environment.

The reports shall also include the results of land use censuses required by Specification 4.10.B.l.

If harmful effects or evidence of irreversible damage are detected by the monitoring, the report shall provide an analysis of the problem and a planned course of action to alleviate the problem.

(c)

The Annual Radiation Environmental Monitoring Reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8, December 1975 of all radiological environmental samples taken during the report period.

In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

The missing data shall be submitted as soon as possible in a supplementary report.

(d)

The reports shall also include the following:

a summary descrip-tion of the radiological environmental monitoring program; a map of all sampling locations keyed to a table giving distances and Prairie Island Unit 1 - Amendment No. 54, 59, 73 Prairie Island Unit 2 - Amendment No. 48, 53, 66

TS.6.7-5 C

directions from one reactor; and the results of licensee participa tion in the Interlaboratory Comparison Program, required by Speci fication 4.10.C.1.

2.

Environmental Special Reports I

(a)

When radioactivity levels in samples exceed limits specified in Table 4.10-3, an Environmental Special Report shall be submitted within 30 days from the end of the affected calendar quarter.

l For certain cases involving long analysis time, determination of quarterly averages may extend beyond the 30 day period.

In these cases the potential for exceeding the quarterly limits will be reported within the 30 day period to be followed by the Environ-mental Epecial Report as soon as practicable.

i I

3.

Other Environmental Reports (non-radiological, non-aquatic)

Written reports for the following items shall be submitted to the appropriate NRC Rggional Administrator:

Environmental events that indicate or could result in a significant a.

environmental impact causually related to plant operation.

following are examples:

The excessive bird impaction; onsite plant or animal disease outbreaks; unusual mortality of any species protected by the Endangered Species Act of 1973; or increase in nuisance

'i organisms or conditions.

. ~

This report shall be submitted within 3.0 days of the event and shall (a) describe, analyze, and evaluate the including extent and magnitude of the impact and plant

event, operating characteristics, (b) describe the probable cause of the (c) indicate the action taken to correct the reported event

event, (d) indicate the corrective action taken to preclude repetition of the event and to prevent similar occurrences involving similar components or systems, and (e their preliminary responses. ) indicate the agencies notified and b.

Proposed changes, test or experiments which may result in a significant increase in any adverse environmental impact which was not previously reviewed or evaluated in the Final Environ-mental Statement or supplements thereto.

This report shall include

and shall be submitted 30 days prior to implementing change, test or experiment.

D.

Special Reports Unless otherwise indicated, special reports required by the Technical Speci-fication shall be submitted to the appropriate NRC Regional Administrator within the time period specified for each report.

Prairie Island Unit 1 - Amendment No. 54, 59, 73 Prairie Island Unit 2 - Amendment No. 48, 53, 66

\\

l i

._

v t e Letter Sequence Other
TAC:55445, (Open) TAC:55446, (Open)
Initiation Request, Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML20090E901, ML20090E905, ML20090E917, ML20101A242, ML20129E520
MONTHYEAR1984-05-31 [Table View]

ML20129E520

Text

Attachment:

Attachment:

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