Text

Proposed Tech Specs,Changing Peaking Factor Limits, Reporting,Refueling Boron Concentration,Radioactive Source Leakage Tests,Senior Reactor Operator Shift Requirements & Deleting Snubber Table
	ML20090E901
Person / Time
Site:	Prairie Island
Issue date:	07/11/1984
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20090E885	List: ML20090E882; ML20090E891; ML20090E901; ML20090E905; ML20090E917;
References
	TAC-55445, TAC-55446, TAC-55816, TAC-55817, NUDOCS 8407200093
	Download: ML20090E901 (55)
	v • d • e

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EXHIBIT B Prairie Island Nuclear Generating Plant License Amendment Request - Dated July 11, 1984 Propored changes to the Technical Specifications Appendix A of Operating Licenses DPR-42 and 60.

Pages Affected By This Amendment TS-1 TABLE TS.3.12-1 (Delete all 8 pages)

TS-ii TS.3.14-1 TS-tii TS.3.14-2 TS-iv TS.3.14-3 TS-v TS.3.14-4 TS-vi TS.3.15-1 TS-vii TABLE TS.4.10-1 (3 of 4)

TS-viii TS.4.11-2 TS-ix TS.4.12-5 TS-x TS.4.13-1 TS.1-1 TS.4.13-2 TS.2.1-2 TS.4.13-3 TS.2.1-3 TABLE TS.4.17-1 (1 of 2)

Figure TS.2~1-1 TABLE TS.4.17-2 (1 of 2)

TS.3.1-12 TABLE TS.6.1-1 TS.3.3-3 Figure TS.6.1-1 TS.3.6-2 TS.6.2-1 TS.3.6-5 TS.6.2-3 TS.3.8-1 TS.6.2-5

TS.3.8-3 TS.6.2-6 TS.3.8-4 TS.6.7-3a (Delete)

TS.3.9-1 TS.6.7-4 TS.3.9-3 TS.6.7-5 TABLE TS.3.9-1 (1 of 2) TS.6.7-6 (Delete)

TABLE TS.3.9-2 (1 of 2) TS.6.7-7 (Delete)

TS.3.10-1 TS.6.7-8 (Delete)

TS.3.10-2 TS.3.10-9 TS.3.10-10 TS.3.10-11 TS.3.10-13 Figure TS.3.10-7 TS.3.12-1 ;

8407200093 840711 PDR ADOCK 05000282 PDR p

TS-1 REV t

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 Systen Integrity TS.1-2 D. Degree of Instrume-tation 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. Qundrant Power Tilt TS.1-5 M. Rated Power TS.1-5 N. Reactor Critical TS.1-5 O. Refueling Operation TS.1-5 P. Shutdown TS.1-5 Q. Thermal 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.1-7 '

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 y,,,

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i TS-ti REV TABLE OF CONTENTS (Continued)

TS SECTION TITLE PAGE 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTING 'TS.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 TS.2.3-1 -

Instrumentation -

A. Protective Instrumentation Settings for Reactor TS.3.2-1 Trip B. Protective Instrumentation Settings for Reactor TS.2.3-3 Trip Interlocks 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. Coolant Pumps TS.3.1-1

2. Steam Generators TS.3,1-1A

3. Requiremenes for Decay Heat Removal Below TS.3.1-1A

. 350*F

4. Pressurizer , TS.3.1-2 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 E. 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 l

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TS-iii REV 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 Internal 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. Caseous 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

> Instrumentation 3.10 Control Rod and Power Distributrion Limits TS.3.10-1 A. Shutdown Reactivity TS.3.10-1 B. Power Distribution Limits TS.3.10-1 C. Quadrant Power Tilt Limits TS.3.10-4 D. Rod Insertion Limits TS.3.10-5 E. Rod Misalignment Limitations 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 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 G. Penetration Fire Barriers TS.3.14-4

.15 Event Monitoring Instrumentation TS.3.15-1 A. Process Monitors TS.3.15-1 B. Radiation Monitors TS.3.15-1

TS-iv REV TABLE OF CONTENTS TS SP."

y ION TITLE PAGE 4.O 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 TS.4.2-2 C. Records TS.4.2-3 4.3 Primary Coolant System Pressure Isolation Valves TS.4.3-1 4.4 Containment System Testr 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. Containnent 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 Tr.mperature H. Containment Shell Temperature TS.4.4-5 I. Electric Hydrogen Recombiners TS.4.4-5 4.5 Engingered Safe.ty Features TS.4.5-1 A. System Tests TS.4.5-1

1. Safety . Injection System 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 *

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 a.

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TS-v REV 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 TS.4.12-5 4.13 Snubbers TS.4.13-1 4.14 Control Room Air Treatment System Tests TS . 4.1,4-1 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. Fire Hydrant Hose Houses TS.4.16-4 G. Penetration Fire Barriers TS.4.16-4 4.17 Radiopctive 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.4.17-4 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 .

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TS-vi REV TABLE OF CONTENTS TS SECTION TITLE PAGE 5.0 DESIGN FEATURES ~TS.S.1-1 5.1 Site -

~TS.5.1-1 5.2 A. Containment Structures TS.5.2-1

1. Containment Vessel TS.S.2-1

2. Shield Building TS.S.2-2

3. Auxiliary Building Special Ventilation Zone-B. Special Ventilation Systems TS.S.2-2 C. Containment System Functional Design TS.5.2-3 5.3 Reactor TS.5.3-1 A. Reactor Core TS.5.3-1 B. Reactor Coolant System TS.5.3-1 C. Protection Systems TS.S.3-1 5.4 Engineered Safety Features TS.S.4-1 5.5 Radioactive Waste Systems TS.5.5-1 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.5.5-2

3. Solid Wastes TS.S.5-3 C. Pr9 cess and Effluent Radiological Monitoring TS.S.5-3 System 5.6 Fuel Handling. TS.S.6-1 A. Criticality Consideration TS.5.6-1 B. Spent Fuel Storage Structure TS.S.6-1 C, Fuel Handling TS.S.6-2 D. Spent Fuel Storage Capacity TS.S.6-3 j-I I

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a

e TS-vii REV TABLE OF CONTENTS (Continued)

TS SECTION TITLE PAGE 6.0 ADMINISTRATIVE CONTROLS TS.6.1-1 6.1 Organization 'TS.o.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 TS.6.2-2

4. Quorum TS.6.2-2

5. Responsibilities TS.6.2-2

6. Audit TS.6.2-3

7. Authority TS.6.2-4

8. Records 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 TS.6.2-5

3. Quorum TS.6.2-5

4. Responsibilities TS.6.2-5

5. Authority TS.6.2-6 6.eRecords TS.6.2-6

7. Procedures' TS.6.2-6 6.3 Special Inspections and Audits TS.6.3-1 6.4 Safaty 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 E. Offsite Dose Calculation Manual (0DCM) 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 Retained 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 Effluent Release TS.6.7-3 Report

6. Annual Summaries of Meteorological 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

TS-viii REV TABLE OF CONTENTS (Continued)

TS SECTION TITL2 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 TS.6.7-5 6.8 Environmental Qualification ,

TS.6.8-1 8 .

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TS-ix REV TECHNICAL SPECIFICATIONS LIST 0F TABLES TS TA"LE 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 g 3.14-1 Safety Related Fire Detection Instruments 3.15-1 Event Monitoring Instrumentation - Process 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 Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements 4.4-1 Unit I and Unit 2 Penetration Designation for Leakage Tests 4.10-1 Radiation Environmental Monitoring Program (REMPO)

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 tbnitoring Instrumentation Surveillance Rquirements 4.17-3 Radloactive Liquid Waste Sampling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Ef fluents 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 Plant (Per Unit) -

6.1-1 Minimum Shif t Crew Composition

r- -- - -

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TS-x REV 1,

s 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 Ccolant System Heatup Limitations 3.1-2 Unit I and Unit 2 Reactor Coolant System Cooldown Limitations

-3.1-3 Effect of Fluence and Copper Content on Shift of RT for Reactor Vessel Steels Exposed to 550* Temperature NDT 3.1-4 Fast Neutron Fluence (E >l MeV) as a Function of Full Power Service Life 3.1-5 DOSE EQUIVALENTS t-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uC1/ 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 Nornalized Operating Envelope 3.10-6 Deviation from Target Flux Difference as a Function of Thermal Power 3.10-7 F /F Function l 3.10-8 VkZ)#NsaFunctionofCoreHeight 4.4-1 Shield Building Design In-Leakage Rate 4.10-1 Prairie Island Nuclear Generating Plant Radiation Ehvironmental Monitoring Program (Sample Location Map) 4.10-2 Prairie Island Nuclear Generating Plant Radiation Environmental Monitoring Program (Sample Location Map) 6.1-1 NSP Corporate Organizational Relationship to On-Site Operating Orga'11zation 6.1-2 Prairie Island Nuclear Generating Plant Functional Organization for On-site Operating Group v

TS.1-1 REV 1.0 DEFINITIONS The -succeeding frequently 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 to the Commission.

B. (Deleted) 9 4 e

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a TS.2.1-2 REV The solid curves of Figure TS 2.1-1 represent the loci of points of tnermal power, ccalant pressure, and coolant average temperature for which either the coolant enthalpy at the core exit is limiting or the DNB ratio is limiting. For the 1685 psig and 1985 psig curves, the coola~nt average enthalpy at the core exit is equal to saturated water enthalpy below power levels of 81% and 65% respectively. For the 2235 psig and 2385 psig curves, l the coolant average temperature at the core exit is equal to 650*F below power levels of 58% and 66% respectively. For all four curves, the DNBR l is equal to 1.3 at higher power levels. The area of safe operation is below these curves.

The plant conditions required to violate the limits in the lower power range are precluded by the salt'-actuated safety valves on the steam generators.

The highest nominal setting of the steam generator safety valves is 1129 psig (saturation temperature 560*F). At zero power the difference between primary coolant and secondary coolant is zero and at full power it is 50*F. The reactor conditions at which steam generator safety valves open is shown as a dashed line on Figure TS.2.1-1.

Except'for special tests, power operation with only one loop or with natural circulation is not allowed. Safety limits for such special tests will be determined as a part of the test procedure. <

The curves are based on the following nuclear hot channel factors ( ):

F H = 1.65 [1 + 0.2(1-P)]; and F = 2.32 This combination of hot channel factors is higher than that calculated at full power for the range from all control rods fully withdrawn to maximum allowable control rod insertion. The control rod insertion limits. are covered by Specification 3.10. Adveric power distribution factors could occur at lower power levels because additional control rods are in the core.

However, the control rod insertion limits specified by Figure TS.3.10-1 assure that the DNB ratio is always greater at part power thar; at full power.

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

s e TS.2.1-3 REV The Reactor Control and Protective System is designed to prevent any anticipated combination of tryg ent conditions that would result in a DNB ratio of less than 1.30 -

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References:

L (1) -FSAR, Section 3.2.2 ,

_ (2) FSAR, Section 3.2.1 (3) FSAR, Section 14.1 I

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Figure TS.2.1-1 REV Safety limits, Reactor Core Thermal and Hydraulic Two-loop Operation 880 .

650 . ...... .. ... ... .. ... . . . . . . . . . . ... ... . ..... .. . .... ... . . . . . . . . . . . . . . . . . . .... .. . . .

. c 0 . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . .. ... .. . . .. ...... ....

630- ;--

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820- ~ ~ - - .i - - . - - - ---

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2305 psig 570 ..... ... ...

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1.ocus of Points et which Steam -

Generator Safety Velves Open :

. . 2235 E. s'* E 560- .' - - ~ ~ : - : - -

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. 1985 Psi 8 .

1 . .

l 550- - - - - - - - -

0 .............. .

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i. : 1685 psig .

530- ............ ...... .. . . . . . . . . . . .. . ..... . . . . . . . . . . . . . . . . . . . . .

520 '

0 $0 4o s'0 s'O 150 150 1do 160 Rated Core Power (%)

Figure TS.2.1-3

TS.3.1-12 REV

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 nicrocuries per gram, perform the sampling and analysis require-ments of item 4a of Table 4.1-2B until the specific act i vity 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.

. BASES The limitations on the specific activity of the primary coolant ensure that tha 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 I on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative ln 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.E.2, permitting power operation to continue for limited l time periods with the prir.ary 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 l" which may occur following changes in thermal power. Operation with specific activity levels exetding 1.0 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 levele allowed by i

L

TS.3.3-3 REV

c. The spray additive tank contains not less than 2590 gal..ons of solution with a sodium hydroxide concentration of 9% to 11% by weight inclusive.

d. Manual valves in the above sys ems 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 operation, changes in valve position will be under direct administrative control.

s

e. Automatic valves, interlocks, ducts, dampers, controls and piping associated with the above components and required for accident condi-tions are operable,

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 motor control center supply breakers open.

(2) For Unit 2 operation, containment spray system motor-operated valves MV32106 and MV32109 shall be closed and shall have the motor control center supply breakers open.

2. During startup operation or power operation, any one of the following conditions'of inoperability may exist for.1.ekch 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 .

a. One fan coolur unit or one duct for a fan cooler unit 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 . 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 . JThe remaining containment spray pump and the four fan units shall be demonstrated to la operable before initiating repairs and once every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereaf ter.

i l

L

TS.3.6-2

. REV

4. Positive reactivity changes shall not be made by boron dilution when containment system integrity is not intact unless the reactor is initially suberitical by at least 10% AK/K.

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

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

TS.3.6-5 REV This specification also prevents positive insertion of reactivity whenever containment integrity is not maintained if such addition would violate the respective shutdown margins. Effectively, the baron concentration must be maintainted at a predicted concentration which shall have the reactor initially suberitical by at least 10% AK/K if the containment system is to be disabled with the vessel open.

The 2 psig limit on internal pressure provides adequate margin between themaximuminternalpressureof46psigandthepeak29ecident pressure resulting from the postulated Design Basis Accident.

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.

Specifying a minimum temperature of 30*F will provide adequate margin above NDTT during power operation when containment is required.

Theconsyryaygyecalculationofoff-sitedosesforthelossofcoolant accident is based on an initial shield building annulus 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.

The specified 44 k6femperature difference is consistent with the LOCA accident analysis 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 conservative calculation of doses for these conditions indicated that changing allowable containment leak .5%/ day to 0.25%/ day -

would offset the increased leakage. 5f te fr High efficiency particulate absolute (HEPA) filters are installed before the charcoal adsorbers to prevent clogging of the fodine adsorbers for all emergency air treatuent 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% throut;h DOP testing and a charcoal adsorber leakage of less than 1% through halogenated

- hydrocarbon testing. The laboratory carbon sample test results 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

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

Obj ectives To ensure that no incident could occur during fuel handling and refueling operations that would affect public health and safety.

Specification A. Durin'g 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 shall be operable or locked closed in each line which penetrates the contain-ment and provides a direct path from containment 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 continuously monitored by at least two neutron monitors, each with continuous visual indication in the control room and one with audible indication in the containnent, which are in service whenever core geometry is being changed. When core geometry is not being changed, at least one neutron flux monitor shall be in service. ,

4. During reactor vessel head removal and while loading and unloading fuel from the reactor, the reactor shall be suberitical by at least 10% AK/K.

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

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. The required water level shall be verified prior to moving fuel assemblies or control rods and at least once every day while the 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 l 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, both residual heat removal loops shall be operable.

8. If Specification 3.8.A.6 or 3.8.A.7 cannot be satisfied, all fuel handling l 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.

TS.3.8-3 REV BASES 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 1fPerations that would result in a hazard to public health and safety.' Whenever changes are not being made in core geometry, one flux monitor 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.

The shutdown margin indicated in A.4 above will keep the core subcritical, l even if all control rods were withdrawn from the core. During refueling, the reactor refueling cavity is filled with approximately 275,000 gallons of borated water. The boron concentration of this water is sufficient to maintain the reactor suberitical by approximately 10% LK/K in the cold condition with all rods inserted, and will also maintain the cort 2fub-critical even if no control rods were inserted into the reactor Periodic checks of refueling water -boron concentration insure that proper shutdown margin is maintained. A.9 allows the control room operator to l inform the manipulator operator of any impending unsafe condition detected from the main control board indicators during fuel movement.

No movement of fuel in the reacter 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 fu The delay time is consistent with the fuel handl'ing accidentanalysis.gg)

The spent fuel assemblies will be loaded into the spent fuel cask for shipment

. to a reprocessing plant af ter sufficient decay of fission prbducts. In loading t

i I

i l

TS.3.8-4 REV the cask into a carrier, there is a potential drop of 66 feet (5) . 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 ficsion products from all fuel assemblies in the cask would esult 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 Spe'nt Fuel Pool Special Ventilation System ( is a safeguards system which 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 exhaust 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 each SFPSVS filter train. The charcoal adsorbers are installed to reduce the potential release of radioiodine 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 radioactive 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 acciaent 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. The water level may also be lowered below 20 feet for upper internals removal / replacement. 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, (2) during latching / unlatching and upper 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. '

I References (1) FSAR Section 9.3.2 (2) FSAR Table 3.2.1-1 (3) FSAR Section 14.2.1 (4) FSAR Section 9.6 (5) FSAR Page 9.5-20a w

TS.3.9-1 REV 3.9 RADIOACTIVE EFFLUENTS Applicability Applies at all times to the liquid and gaseous' radioactive efffuents from the plant and the solidification and packaging for shipment of colid radioactive waste.

Objective To implement tha requirements of 10CFR20,10CFR71,10CFR50 Section 50.36a, Appendix A and Appendix I to 10CFR50, 40CFR141, and 40CFR190 pertaining to radioactive effluents.

Specifications A. Liquid Effluents

1. Concentration

a. Tre concentration of liquid radioactive material released at

.any time from the site (Figure 3.9-1) shall be limited to the concentrations.specified in 10 CFR Part 20, Appendix B, Table II, Column 2, for radionucludes other thsn dissolved or entrained noble gases. For dissolved or entrained noble gases, the concentration shall be limited to 2 x 10 uci/ml 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 and provide a special report to the Commission within 30 days. l

2. Dose

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

1. During any calendar quarter to 13.0 mrem to the total body and to 110 mrem to any organ, and

2. During any calendar year to 16.0 mrem to the total body and to 120 mrem to any organ.

i

TS.3.9-3 REV

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

a. The dose rate at any time due to radioactive materials released in gaseous effluents from the site (Figure 3.9-2) shall be limited to the following values:

1. The dose rate limit for noble gases shall be <500 mrem / year to the total body and 13000 mrem / 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 11500 mrem / 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 and provide a special report to the Commission within 30 days.

2. Dose from Noble base;

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 110 mrad for gamma radiation and 120 mrad for beta radiation, and

2. During any calendar year, to 120 mrad for game-a radiation and 140 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 identi-fies the cause(s) for exceeding the limit (s) and defines the corrective action (s) taken to reduce the releases and the pro-posed corrective actions to be taken to assure the' subsequent releases will be within the above limits.

--~,-w ~v - - -- - - < -

4 .

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

REV TABLE TS.3.9-1 RADIOACTIVE LIQUID EFFLUENT MONITORING INSTRUMENTATION MINIMUM ~

CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. Gross Radioactivity Monitors Providing Automatic Termination of Release

a. Liquid Radwaste Effluent Line i During releases 1

b. Steam Generator Blowdown 1/ Unit During releases 2 Effluent Line

2. Flow Rate Measurement Devices

.a. Liquid Radwaste Effluent Line 1 During releases 4 requiring throt-tling of flow

b. Steam Generator Blowdown Flow 1/ Gen During releases 4 1

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

a. Condensate Storage Tanks 1/ Unit When tanks ~

5 are in use

b. Temporary Outdoor Tanks 1/ Tank When tanks 5 Holding Radioactive Liquid are in use

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

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

REV TABLE TS.3.9-2 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION MINIMUM

~

CHANNELS INSTRUMENT OPERABLE APPLICABILITY ACTION

1. Waste Gas Holdup System 2 During system 2 Explosive Gas operation (Oxygen) Monitors

2. Effluent Release Points (Unit No.1 Reactor Bldg, Unit No. 1 Aux Bldg, Unit No. 2 Reactor Bldg, Unit No. 2 Aux Eldg.

Spent Fuel Pool, Radwaste Bldg)

a. Ibble Gas Activity 1 During releases 4,5,7 Monitor *

b. Iodine Sampler 1 During releases 3 Cartridge ,

c. Particulate Sampler 1 During releases 3 Filter

~

d. Sampler Flow 1 During releases 1 Integrator l

3. Air Ejector Noble Gas 1 During power 6 Monitors (Each Unit) operation ,

L

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

,m _ __ _ _ _ _ , - - _ _ _ _ _ ,

TS.3.10-1 REV 3.10 C0FIROL R0D AND POWER DISTRIBUTION LIMITS a

Applicability Applies to the limits on core fission power distribution and to ths limits on control rod operations.

- Objective To assure 1) core suberiticality after reactor trip, 2) acceptable core power distributions during power operation, and 3) Limited potential reactivity in-sertions caused by hypothetical control rod ejection.

Specification A. Shutdown Reactivity The shutdown margin with allowance for a stuck control rod assembly shall exceed the applicable value shown in Figure TS.3.10-1 under all steady-state operating conditions, except for physics tests, from zero to full power, i including effects of axial power distribution. The shutdown margin as used here is defined as the amount by which the reactor core would be suberitical at het shutdown conditions if, all control rod assemblies were tripped, assuming that the highest worth control rod assembly remained fully withdrawn, and assuming no changes in xpnon.or boron concentration.

B.- Power Distribution Limits

1. At all times, except dgring igw power physics testing, measured hot channel factors, F and F shall meet the followiOg limikH s: , as defined in the bases, Fq x 1.03 x 1.05 1[Fq(F4H }

F * +

H 1 Q

~

where the following definitions apply:

l (a) K(Z) is the axial dependence function shown in Figure TS.3.10-5.

i l

(b) Z is the core height location.

(c) P is the fraction of full power at which the core is operating.

N In the F limit determination when P 10.50, set P = 0.50.

(d) F 0 (Ff H) is the enthalpy rise dependent heat flux factor shown in Figure TS.3.10-7. 2.21/P shall be used for Westinghouse assemblies.

l (e) F ikHFig(F 0reTS.3.10-7.

) is the heat flux1.55 dependent shall be enthalpy used for rise factor shown Westinghouse assemblies.

t

' " ' ' 9

TS.3.10-2 REV N N N N (f) FO r F^H s de m ed as the measured F O r F. , respectively, with the smallest margin or greatest excess o limit.

(g) 1.03 is the engingering hot channel factor, F applied to the measured F toaccountformanufacturibg, tolerances.

(h) 1.05 is applied to the measured F to account for measure-ment uncertainty.

N (1) 1.04 is applied to the measured F AH t account for measure-ment uncertainty.

N N

2. Hot channel factors, F and F q shall be measured and the targetfluxdifferencedetermk!e,d,atequilibriumconditions according to the following conditions, whichever occurs first:

(a) At least once per 31 effective full-power days in conjunction with the target flux difference determination, or (b) Upon reaching equilibrium conditions after exceeding the reactor power at which target flux difference was last determined, by 10%

or more .of rated power.

F

' O (equil) shall meet the following limit for the middle axial 80%

of the core:

F q (equil) x V(Z) x 1.03 x 1.05 1[Fq (F )/P] x K(Z) l vbere V(Z) is defined in Figure 3.10-8 and other terms are defined in 3.10.B.1 above.

3. (a) If either measured hot channel factor exceeds its limit specified in

.3.10.B.1,reducereactor.powerandthehigh'neutroafluxtripsetpoint by 1% for each percent that the measured F or 5% for'each percent that 9 Then follow themeasuredFhexceedsthe3.10.B.111mit.

3.10.B.3.(c).

(b) If the measured Fg (equil) exceeds the 3.10.B.2 limits but not the 3.10.B.1 limit, take one of the following actions:

. 1. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months place the reactor in an equilibrium configuration for which Specification 3.10.B.2 is satisfied, or

. 2. Reduce reactor power and the high neutron Qux trip setpoint by 1% for each percent that the measured Ff' (equil) x 1.03 x 1.05 x V(Z) exceeds the limit.

I i

l

TS.3.10-9 REV mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded. .

Dqring ogeration, the plant staff compares the measured hot channei factors, F' and F LOCA anah)s,es(described

. The terms onlater) to the side the right limitsofdetermined in the the equations in transient and section 3.10.B.1 represent the analytical limits. Those terms on the left side represent the measured hot channel factors corrected for engineering, calculational, and measurement uncertainties.

Meight Dependent Heat Flux Hot Channel Factor is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and.

rods. flux hot channel factor is [F q(FTp)maximumvalueoftheheightdependentheat y /P] X K(Z) for Exxon fuel and (2.21/P) X K(Z) for Westinghouse fuel.

. s The F0 (F ) and F1 (F )n functions shown in Figure TS.3.10-;7 are based on two LOCA ana yse 1)" Ft P = 1.55 and F = 2.32 and 2) at F1 H 1.65 and F = 2.28.

Each pin's F'q;and bouads of FikUre TS.3.h, ith uncertain 10-7 (as modified by K(Z)91es) must be shown gnd P appropriately). to operate These w functionsarefuelassemblylimitsandtheF,FIH q s need not be de same for every assembly.

These LOCA analyses covered peak pellet burnups from 0 to 55,000 MWD /MTU.

The K(Z) function shown in Figure TS.3.10-5 is a normalized function having limits for three regions. The K(Z) specified for the lowest six feet of the core is arbitrarily flat since the lower part of the core is generally not limiting. Above that region, the K(Z) value is based on large and small break LOCA analyses. The K(Z) in the uppermost region is limited to reduce the PCT expected during a small break LOCA since this region of the core is expected to uncover temporarily for some small break LOCAs. Figure TS.3.10-5 shows the operating envelope for F = 2.32 and F'l = 1.55. The operating envelope for

~

F = 2.28 and F'? = 1. 9 bounding for alI Fcombinations . F}q51sshown lessonlimiti!g, therefore Figure TS.3.

Figure TS.3.10-7 F is the measured Nuclear Hot Channel Factor, defined as the maximum local hSatfluxinthecoredividedbytheaverageheat flux in the core. Heat fluxes are derived from measured neutron fluxes and fuel enrichment.

V(Z) is an g'xially dependent function applied to the equilibrium measured F to bound F s that could be measured at non-equilibrium conditions. This function 10 based on power distribution control analyses that evaluated the effect of burnable poisons, rod position, axial effects and xenon worth.

F Engineering Heat Flux Hot Channel Factor is defined as the allowance on h9a,tfluxrequiredformanufacturingtolerances. The engineering factor allows for local variations in enrichment, pellet density and diameter, surface area of the fuel rod and eccentricity of the gap between pellet and clad. Combined statistically the net effect is a factor of 1.03 to be applied to fuel rod surface heat flux.

)

TS.3.10-10 REV The 1.05 multiplier accounts for uncertainties associated with measure-ment of the power distribution with the moveable incore detectors and the use of those measurements to establish the ' assembly local power distribution.

F (equil) is the measured limiting Fq obtained at equilibrium conditions dQring target flux determination.

Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio

- of the integral of linear pcwer along the rod with the highest integrated power to the average rod powgr. The maximum value of the Nuclear Enthalpy Rise Hot Channel Factor is F AH ( Q

+* (~}

When a measurement of F AH is taken, experimental error must be allowed for and 4 percent is the appropriate allowance for a' full core map taken with the movable incore detector flux capping system.

Measurements of the hot channel factors are required as part of startup physics bests, at least once each effective full power month of operation, and whenever abnormal power distribution conditions require a reduction

- of core power to.a level based on measured hot channel factors. The incore map taken following initial loading provides confirmation of the basic nuclear design bases including proper fuel loading patterns. The periodic monthly incore mapping provides additional assurance that the nuclear design bases remain inviolate and identify operational anomalies which would otherwise affect these bases.

For normal operation, it is not necessary to measure these quantities. ,

Instead it has been determined that, provided certain conditions are observed, the hot channel factor limits will be net; these conditions are as follows:

1. Control rods in a single bank move together with no individual rod insertion differing by more than 15 J

e b

TS.3.10-11 REV inches from the bank demand position. An accidental misalignment limit of 13 steps precludes a rod misalign-ment greater than 15 inches with consideration of maximum instrumentation error.

2. Control rod banks are sequenced with overlapping banks as described in Technical Specification 3.10. ,

3. The control bank insertion limits are not violated.

4. Axial power distribution control procedures, which are given in terms of flux difference control and control bank insertion limits are observed. Flux difference refers to the difference in signals between the top and bottom halves of two-section excore neutron detectors. The flux difference is a measure of the axial offset which is defined as the difference in normalized power between the top and bottom halves of the core.

The permitted relaxation in F"H and F allows for radial power shapechangeswithrodinsertkontotheinsertionlimits. It has been determined that provided the above conditions 1 through 4 are obsegved, these hot channel factor limits are met. In specification 3.10, F arbitrarily limited for P 10.5 (except forlowpowerphysicstestsh.is The proceduras for axial power distribution control referred to above are designed to minimize the effects of xenon redistribution on the axial power distributicn during load-follow maneuvers. Basically control of flux difference is required to limit the difference between the current value of Flux Difference ( AI) and a reference value which corresponds to the full power equilibrium value of Axial Of fset (Axial Of fset = AI/ fractional power). The reference value of flux difference varies with power level and burnup but expressed as axial "

offset it varies only with burnup.

The technical specifications on power distribution control assure that the Height Dependent Heat Flux Hot Channel Factor upper bound envelope l of Figures TS.3.10-5 times TS.3.10-7 is not exceeded and lenon distri-butions are not developed which at a later time, would cause greater local power peaking even though the flun dif ference is then within the limits specified by the procedure.

The target (or reference) value of flux difference is determined as follows: At any time that equilibrium xenon conditions have been established, the indicated flux difference is noted with the full length rod control rod bank more then 190 steps withdrawn This value, divided by the fraction of full power at which the core was operating is the full power value of the target flux difference.

Values for all other core power levels are obtained by multiplying the full power value by the fractional power. Since the indicated equilibrium was noted, no allowances for excore detector error are necessary and indicated deviation of !5 percent I are permitted from the indicated reference value. Figure TS.3.10-6 shows the allowed deviation from the target flux difference as the function of thermal power.

e TS.3.10-13 REV resulting from operation within the target band. The consequences of being outside the +5% target band but within the Figure TS.3.10-6 limit for power levels between 50% and 90% has been evaluated and determined to result in acceptable peaking factor values. - Therefore, while the l deviation exists the power level is limited to 90 percent or lower depending on the indicated axial flux dif ference. In all cases the

+5 percent target band is the Limiting Conditica for Operation. Only when the target band 'a violated da the limits under Figure TS.3.10-6 apply.

If, for any reason, the indicated axial flux dif ference is not controlled within *.he +5 percent band for as long a period as one hour, then xenon distributions may be significantly changed and operation at 50 percent is required to protect against potentially more severe consequences of some accidents.

As discussed above, the essence of the procedure ic to maintain the xenon distribucion in the core as close to the equilibrium full power condition as possible. This is accomplished by using the boron system to position the full length control rods to produce the required indicated flux difference. .

For Condition II events the core is protected from overpower and a minimum DNBR of 1.30 by an automatic protection system. Compliance with operating procedures is assumed as a precondition for Condition II transients, however, operator error and equipment malfunctions are separately assumed to lead to the cause of the transients considered.

Quadrant Power Tilt Limits Quadrant power tilt limits are based on the following considerations. Fre-quent power tilts are not anticipated during normal operation since this phenomenon is caused 'uy some asymmetric perturbation, e.g. rod misalignment, x-y xenon transient, or inlet temperature mismatch. A dropped or misaligned rod will easily be detected by the Rod Position Indication System or core instrumeatation per Specification 3.10.F. and core limits protected per Specification 3.10.E. A quadrant tilt by some other means (x-y xenon tran-sient, etc.) would not appear instantaneously, but would build up over several hours and the quadrant tilt limits are set to protect against this situation. They also serve as a backup protection against the dropped or misaligned rod.

Operational experience shows that normal power tilts are less than 1.01.

Thus, sufficient time is available to recognize the presence of a tilt and correct the cause before a severe tilt could build up. During start-up and power escalation, however, a large tilt could be initiated.

Therefore, the Technical Specification has been written so as to prevent escalation above 50 percent power if a large tilt is present.

O ,

FIGURE TS.3.10-7 REV F

q VERSUSLHF f FMION 2.33 ,

t i i

I i

l 2.32 ,

~\ x- l i

l i

i i

x f N

\ .e 2.31 i

\ N.

A i

.hx F

9

'xx -

2.30 \ x f i !

+ 1 I \

' i ~-~~

i 2,29 -

x x

\ x x

< ' N i 2.28 i i i

1,55 1.60 1.65 F

AH

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

Obj ective To ' define those conditions of snubber operability necessary to assure safe reactor operation. -

l 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 Specifica-tion 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.

I BASES

~

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.

e r -

e 7 TS.3.14-1 REV 3.14 FIRE DETECTION AND PROTECTION SYSTDIS Applicability Applies to instrumustation and plant systems used for fire detection and protection of the nuclear safety-related structures, systems, and components of the plant.

Obj ective To insure that the structures, systems, and components of the plant important to nuclear safety are protected from fire damage.

Specification A. Fire Detection Instrumentation

1. Except as specified below, the minimum fire detection instrumentation for each fire detection zone shown in Table 3.14-1 shall be operable 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. IfSpecification3.14.I.1cannotbemet:

a. Within one hour,' establish a fire watch patrol to inspect the zone with the inoperable instrumentsat least once per hour.

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

~

b. Restore the inoperable instruments to operable status within 14 days or submit a special report to the Commission within 30 days outlining the cause of the malfunction and the plans for restoring the instruments to operable status.

B. Fire Suppression Water System

1. Except as specified in 3.14.B.2 or 3.14.B.3 below, the system shall be operable at all times with:

a. The following pumps, including automatic initiation logic, operable and capable of delivering at icast 2000 gpm at a discharge pressure of 108 psig,

l. Diesel-driven fire pump

2. Motor-driven fire pump

3. Screen wash pump

a e TS.3.14-2 REV

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.

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 l outlining the plans and procedures to be used to provide for the loss of redundancy in the Fire Suppression Water system. With an inoperable pump, perform the surveillance required by Specification 4.16.B.2.

3. With the fire suppression water system otherwise inoperable.

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

b. Provide a special report to the Commission within 30 days outlining l the actions taken and the plans and schedule for restoring the system to operable status,

c. If Specification 3.14.B.3.a cannot be met, the reactors shall 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 within 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . .

C. Spray and Sprinkler Systems

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

a. Auxiliary Feed Pump Room WP-10

b. Diesel Generator Areas PA-1

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

d. Unit No. 1 Electrical Penetration Area PA-4

e. Unit No. 2 Electrical Penetration Area PA-6

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

g. Screenhouse PA-9

2. If Specificaiton 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..

o .

TS.3.14-3 REV D. Carbon Dioxide System

1. Except as specified in 3.14.D.3 below, the CO system protecting 2

the relay and calbe spreading room area shall be operable with a minimun level of 60% in the C0 storage tank.

3

2. During those periods when the relay and cable spreading room area is normally occupied, automatic initiation of the CO system may 2

be bypassed. During those periods when the area is normally unoccupied, the CO system shall be capable of automatic initiation 2

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.

E. 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

c. Safety Related Areas of Screenhouse

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

9 C TS.3.14-4 REV 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:

a. Unit No. 1 Reactor Buildir.g

b. Unit No. 2 Reactor Building

c. Turbine Building

d. Auxiliary Building

e. Screenhouse

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 l outlining the cause of the inoperability and the plans and schedule for restoring the houses to Operabl* status.

_ G. Penetration Fire Barriers

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 Op~erable 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 barriers to Operable status.

p. .

TS.3.15-1 REV 3.15 E7ENT 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.

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

Specification A. Process Monitors l

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 Shutdown within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

B. Radiation Monitors l

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

3. With the number of Operable event monitoring insttumentation 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.

I

TABLE TS.4,10-1' (Pegt 3'of 4)

PRAIRIE ISLAND NUCLEAR GENERATING PLAlfr RADIATION ENVIRONMEffrAL MONITORING PROGRAM SAMPLE COLLECTION AND ANALYSIS Number of Samples Exposure Pathway and . Sampling and Type and Frequency and/or Sample Sample Locations ** Collection Frequency of Analysis 3.. WATERBORNE (continued).

[ d. Sediment from One sample upstream of Semiannually Gairaa isotopic analysis shoreline plant, one sample down- of each sample

stream of plant, and one i

from shoreline of recreational area. .

4. INGESTION 4 ,

s. Milk One sample from dairy Monthly or hiweekly Gamma isotopic and I-131 farm having highest D/Q, if animals are on analysis of each sample one sample from each of pasture three dairy farms cal-culated to have doses f rom I-131 > 1 mrem /yr, and one sample from 10-20 miles ,

b. Fish and One sample of one game Semiannually Gamma isotopic analysis Invertebrates specie of fish located on each sample (edible N upstream and downst ream portion only on fish) a '

l of the plant site

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One sample of Invertebrates 5 upstream and downstream of , ,L the plant site m E

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

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TS.4.11-2 REV D. Tests resulting in 0.005 microcuries or more of removable ,

contamination on the test sample shall be reported to the Conaission 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 fellowing records shs11 he retained for i years:

a. Test results in microcuries, for tests performed pursuant to TS 4.11.

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

Bases Licensee's program, facilities, 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 prcgram to detect anl control contamination of areas in the plant by such radioactive materials. Small quant'*.ies 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 would

{

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 alphe emitters) are exempt from leakage tests in this specification. The acceptance critaria of less than 0.005 microcuries on the test sample is also based on 10 CFR 70.39(c).

= .

TS.4.12-5 REV

. 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.

E. Reports

1. Following each in-service inspection of steam generator tubes, if there are any.tutes requiring plugging, the number of tubes plugged in each steam generator shall be submitted in a

. special report to the Commission within 15 days.

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 provide a description of investigations conducted to determine cause of the tube degradation and corrective measures taken to prevent recurrence.

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9 6

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i., . <.m = - - - - r - =-'e- ~ * * - - ---

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TS.4.13-1 REV 4.13 SNUBBERS Applicability Applies to periodic testing and surveillance requirements of safety related hydraulic snubbers.

0bjective To verify the integrity and operability of hydraulic snubbers.

Specification -

The following surveillance requirements apply to all safety related snabbers. These requirements augment the inspections required by Section XI of the ASME Code.

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 12 months ! 25%

2. . 6 months t 25%

3,4 124 days ! 25%

. 5,6,7 . 62 days t 25%

8 or more 31 days t 25%

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

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 shall verify (1) that therc are no visiblew indications of damage or impaired operability, (2) attachments to the supporting structure are secure, and (3) in those locations where snabber 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:

4 a

9

TS. 4.13-2 REV

a. Clearly establishing the cause of the rejection for that 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 suubber is found to be uncovered, the snubber shall be considered inoperable for purposes of establishing the next visual inspection interval. All hydraulic snubbers 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 l conducted at least once per 18 months during cold shutdown. Ten percent of the total of each type snubber shall be functionally tested either _ in place or in a bench test. For each snubber that does not meet the functional test acceptance criteria in Specifica-tion 4.13.D below, an additional ten percent of that type of snubber shall be functionally tested until no more failures are found or all snubbers of that type have been tested.

The representative sample selected for functional testing shall include the various configurations, operating environments, and the range of size and capacity of the snubbers. Twenty-five l percent of the sample shall include snubbers from the following three categories. ,

a. The first snubber away from a reactor vessel nozzle

b. Snubbers within five feet of heavy equipment (valve, pump, turbine, motor, etc.)

c. Snubbers within ten feet of the discharge of a safety / relief valve Snubbers identified as "High Radiation Area" or " Difficult to l Remove" 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 re-tested during the next test period. If a spare snubber has.been installed in place of a failcd snubber, then both the failed snubber, if it is repaired and installed in another position, and the spare snubber shall be retested.

TS.4.13-3 REV 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 tested. This testing is in addition to the regular sampic and specified re-samples.

D. Hydraulic snubber functional tests shall verify that: ,

a. Activation (restraining action) is achieved within the specified range of velocity or acceleration in both tension and compression.

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 disp. lace-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 l be reviewed at least once every 18 mcaths 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 re-evaluated or the snubber shall be replaced or reconditioned to extend its service life beyond the date of the next scheduled service life review. This re-evaluation, replacement, or recondi-tioning shall be indicated in the records. ,

BASES 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 ty.the number of inoperable snubbers found during an inspection.

Inspections performed before that interval has elapsed may be used as a new reference point to determine the next inspection. However, the results of such early inspecitons performel before the original required time interval hat- elapsed (nominal time less 25%) may not be used to lengthen the required inspection interval. Any inspection whose results require a shorter inspection interval will override the previous schedule.

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

o' .

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

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

Liquid Radwaste Effluent Daily during Prior to Quarterly (1) At least once every Lins Gross Radioactivity releases Each Release 18 months (3)

M:nitor Liquid Radwaste Effluent Daily during - -

At least once every Lina Flow Instrument releases 18 months Stm.m Generator Blowdown Daily during Monthly Quarterly (1) At least once every Grors Radioactivity releases 18 months (3)

Monitors Stitm Generator Blowdown Daily during - -

At least once every Flow , Releases 18 months Turbine Building Sump Daily dpring - -- -

C:ntinuous Composite releases Sirplers (Includes sampl o volume check)

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

Dircharge Canal Daily during -- --

At least once every Flow Instruments releases 18 months Condensate Storage Daily --

Quarterly At least once every Trek Level Monitors 18 months L1 val Monttors for Daily when -

Quarterly At least once every T:cporary Outdoor in use when in use 18 months when in Trnks Holding use Radioactire Liquid

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

REV TABLE TS.4.17 RADIOACTIVE GASEOUS EFFLUENT MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMESTS Source -

Channel Check Check Functional Test Instrument Frequency Frequency Frequency Calibration Frequency W2cta Gas Holdup System Daily During ---

Monthly (2) Quarterly (5)

Explosive Qps System (Oxygen) Monitors Operation Effluent Release Points (Unit No. 1 Reactor Bldg, Unit No. 1 Aux Bldg, Unit No. 2 Rsactor Bldg, Unit No. 2 Aux Bldg.

Spent Fuel Pool, Radwaste Bldg)

Noble Gas Activity Da%1y During Monthly

Quarterly (1) At least once every Monitor (4) Releases 18 months (3)

(Except Radwaste ,

Building)

Noble Gas Activity Daily During Monthly Quarterly (2) At least once every Monitor Radwaste Releases 18 months (3)

Building (4)

Iodine and Weekly --- --- ---

Particulate Samplers Sampler Flow Weekly --- ---

At least once every Integrator 18 months Air Ejector Noble Gas Daily During Monthly Quarterly (2) At least once every Monitors (Each Unit) Releases 18 months (3) o A source check of the' applicable noble gas monitor shall be conducted prior to each wasta gas decay tank or containment purge release.

e TABLE TS.6.1-1 MINIMlE SilIFT CREW COMPOSITION (Notes 1 and 3)

CATEGORY BOTil UNITS IN COLD Sl!UTDOWN ONE UNIT IN COLD SilUTDOWN BOTil UNITS ABOVE COLD OR REFilELING SHUTDOWN OR REFUELING SilUTDOWN AND SilUTDOWN ONE UNIT ABOVE COLD SilUTDOWN No. Licensed Senior 2 (note 2) 2 (notes 2, 4) 2 (note 4) l i Operators (LS0)

Total No. Licensed 4 . 4 5 Operators (LSO & LO)

Total No. Licensed & 6 7 8 Unlicensed Operators Shift Technical Advisor 0 1 1 NOTES:

1. Shift crew composition may be one less than the minimum requirements for a period of time not to exceed two hours in order to accommodate an unexpectrd absence of one duty shift ]

crew member provided immediate action is taken to restore the shif t crew composition to within the minimum requirements specified.

2. Does not include the licensed Senior Reactor Operator, or Senior Reactor Operator limited to Fuel llandling, supervising refueling operations. g

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3. Each LSO and to shall be licensed on each unit. $

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4. One LSO shall be in the control room at all times when a reactor is above cold shutdown. F cn T

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SENIOR VICE PF.ISIDENT POWER SUPPLY I

VICE PRESIDENT v!CE PRESIDENT I l PC LY f MANAGER VICE PRESIDENT PLANT ENGINEER 1rdi OUALITY SYS*2H PROD 0 FUEL NUCLEAR AND CONSTRUCTION ASSURANCE OP & MAINTENANCE GENERATION

l SUPPLY I I

CEEh -R GENERAL MANAGER NUCLEAR PLANTS S NUCLEAR GROUP I I

_1 MANAGERS

-R PRODUCTION NUCLEAR

~=a TECtmICAL

-R NUCLEAR SUPPORT i

(PRAIRIE ISLAND TRAINING ANALYSIS SERVICES SERVICES

& MONT! CELL 0s g

OH-SITE ON-SITE TECtetICAL I g

I TRAINING SERVICES CROUPS

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t_______gpggg, ,_________________-________ cg',';,%'c, _

g a HA5 INE RESPONSIBILITY FOR THE AY .

FIRE PROTECTION PROGRI.H FIGURE T9.6.1-1 NSP CDRPORATION ORGANIZATION

i KLATIONSHIP TO ON-SITE ~

CI'ERATING ORGANIZATIONS 4 !

l CAD ORTSill l

TS.6.2-1 REV 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

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

b. The SAC chairman shall be an NSP representative, not having line responsibility #cr plant operation, appointed by the Vice President Nuclea. Generation. Other SAC members shall be appointed by the Vice President Nuclear Generation or by such other person as he may designate. The Chairman shall appoint a Vice Chairman from the SAC membership to act in his absence.

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

a. The SAC members should collectively have the capability required to review act1vities in the following are'as: nuclear power plant operations, nuclear engineering, chemistry and radiochemistry, metallurgy, instrumentation and control, radiological safety, mechanical and electrical engineering, quality assurance practices, and other appropriate fields associated with the unique characteristics of the nuclear power plants.

TS.6.2-3 REV

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

g. Revisions to the Facility Emergency Plan, Facility Security Plan, and the Fire Protection Program,

h. Operations Committee minutes to determine if =atters considered by

~

that Committee involve unreviewed or unresolved safety questions.

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

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 Of fsite Dose Calculation Manual (ODCM) .

m. Review of investigative reports of unplanned releases of radioactive material to the environs.

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

a. Audits of selected aspects of plant operation, as delinected 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 instruct' ions and procedures.

b. Audits of aspects of plant radioactive effluent treatment and radio-logical environmental monitoring shall be perforced 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.

c. Periodic review of the audit program should be performed by the SAC at icast 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 b..ving responsibility in the areas audited.

TS.6.2-5 REV 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.

I

3. Quorum A majority of the permanent members, includ.ing the Chairman or Vice Chairman

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

a. Proposed tests and experiments and their results,

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. ,

2. Proposals which would effect p9rmanent changes to normal and emergency operating protzdures and any other proposed changes or procedures that .1111 af fect nuclear saf_ety as determined by the Plant Manat;er,

d. Froposed changes to the Technical Specifications or operating licenses.

e. All reported or suspected violations of Technical Specifica-tions, operating license requirements, administrative procedures, operating procedures. Results of investigations, including 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.

i

=

TS.6.2-6 REV

f. Investigation 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 Specificaitonis, 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 byt at an interval of not more than two years. -

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

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

k. All changes to the Process Control Program (PCP) and the Offsite Dose Calculation Manual (0DCM).

S. Authority The OC shall be advisory to the Plant >bnager. In the event of a disagreement.between the' recommendations of the OC and the 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.

i

6. Records Minutes shall be recorded for all meetings of the OC and shall identify all documentary material reviewed. The minutes shall be distributed by the OC Chairman or Vice Chairman.

~

7. Procedures ,

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

a. Responsibility and authority of the group

b. Content and method of submission of presentations to the Operations Committee

c. Mechanism for scheduling meetings

d. Provision for meeting agenda

TS.6.7-4 REV 4

7. Report of Safety and Relief Valve Failures and Challenges. An annual report of pressurizer safety and reliev 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:

a. The Commission shall be notified by a report submitted pursuant to the requirements of Section 30.73 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 e .

(a) Annual Radiation Environmental Monitoring Reports covaring 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 assess-ment of the observed impacts of the plant operation on the environment. The reports shall also include the risults of b 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 environmeatal 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.

TS.6.7-5

- REV (d) The reports shall also include the following: a summary description of the radiological environmental monitoring program; a map of all sampling locations keyed to a tabla giving distances and directions from one reactor; add the results of licensee participation in the Interlaboratory Comparison Program, required by Specification 4.10.C.l.

2. Environmental Special Reports l (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. 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 Environmental Special Report as soon as l practicable.

3. Other Environmental Reports (non-radiological, non-acuatic)

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

~

a. Environmental events that indicate or could result in a signi-ficant environmental impact causually related to plant opera-tion. The following are examples: 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 organisms or conditions. This report shall be submitted within 30 days of the event and shall (a) describe, analyze, and evaluate the event, including extent and nagni-tude of the impact and plant operating characteristics, (b) describe the probable cause of the event, (c) indicate the action taken to correct the reported 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) indicate the agencies notified and their preliminary responses,

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 an evaluation of the environmental impact of the pro-posed activity and shall be submitted 30 days prior to imple-menting the proposed change, test or experiment.

D. Special Reports Unless otherwise indicated, special reports required by the Technical Specification shall be submitted to the appropriate NRC Regional Administration within the time period specified for each report.

.

v t e Letter Sequence Other
TAC:55445, (Open) TAC:55446, (Open) TAC:55816, (Open) TAC:55817, (Open)
Initiation Request, Request, Request Acceptance... Administration Questions Answers Results Approval... Other: ML20090E901, ML20090E905, ML20090E917, ML20101A242
MONTHYEAR1984-07-11 [Table View]

ML20090E901

Text

References:

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