ML20093H513

Amends 120 & 113 to Licenses DPR-42 & DPR-60,respectively, Revising TS 3.14 Re Fire Protection & Detection Sys,Limiting Conditions for Operation & TS 4.16 Re Fire Detection & Protection

ML20093H513
Person / Time
Site:	Prairie Island
Issue date:	10/06/1995
From:	Wetzel B NRC (Affiliation Not Assigned)
To:
Shared Package
ML20093H517	List: ML20093H513 ML20093H525
References
GL-86-10, GL-88-12, NUDOCS 9510200205
Download: ML20093H513 (29)

Text

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4 UNITED STATES l

j NUCLEAR REGULATORY COMMISSION WASHINGYON. D.C. 20066-0001

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NORTHERN STATES POWER COMPANY 4

DOCKET NO. 50-282 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 120 License No. DPR-42 i

1.

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

A.

The application for amendment by Northern States Power Company (the licensee) dated July 11, 1994, as supplemented April 18, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

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

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 4

D.

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

E.

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

2.

Accordingly, the following sections under paragraph 2.C of Facility Operating License No. DPR-42 are hereby amended to read as follows:

9510200205 951006 PDR ADOCK 05000282 P

PDR

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2.C.(2)

Technical Snecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.120, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

4 2.C.(4)

Fire Protection Northern States Power Company shall imp 1riment and maintain in effect all provisions of the approved fire protection program as described and referenced in the Updated Safety Analys!s Report for the Prairie Island Nuclear Generating Plant, Units 1 and 2, and as approved in Safety Evaluation Reports dated February 14, 1978, September 6, 1979, April 4, 1980, December 29, 1980, July 28, 1981, September 12, 1984, June 25, 1985, October 27, 1989, and October 6, 1995, subject to the following provision:

The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

3.

This license amendment is effective as of the date of issuance to be implemented within 30 days.

FOR THE NUCLEAR REGULATORY t.0MMISSION 0.

Beth A. Wetzel, roject Manager Project Directorate III-l Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Attachments:

Pages 3 and 4 of License No. DPR-42*

Changes to the Technical Specifications Date of Issuance:

October 6, 1995

• Pages 3 and 4 are attached, for convenience, for the composite license to reflect these changes.

ATTACHMENT TO LICENSE AMENDMENT NO. 120 FACILITY OPERATING LICENSE NO. DPR-42

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DOCKET NO. 50-282 UNIT 1 LICENSE REMOVE INSERT 1

Pages 3 Pages 3 4

4 5

TECHNICAL SPECIFICATIONS Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by i

4 amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT l

TS-iv TS-iv TS-v TS-v*

TS-vi TS-vi TS-x TS-x TS-xi TS-xi TS-xii TS-xii l

TS.1-3 TS.1-3 j

TS.3.14-1

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TS.3.14-2 TS.3.14-3 TS.3.14-4 TABLE TS.3.14-1, page 1 TABLE TS.3.14-1, page 2 TABLE TS.3.14-1, page 3 TS.4.16-1 TS.4.16-2 TS.4.16-3 TS.4.16-4 TS.4.16-5 TS.4.16-6 TS.6.1-2 TS.6.1-2 TS.6.1-3 TS.6.1-3 TS.6.2-6 TS.6.2-6 B.3.14-1 B.3.14-2 4

B.4.16-1 B.4.16-2

• Corrected page

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(2)

Pursuant to the Act and 10 CFR Part 70, to receive, possess and use at any time special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts I

required for reactor operation, as described in the Final

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Safety Analysis Report, as supplemented and amended as of Amdt.

j May 11, 1976.

No. 12 1

5-11-7' (3)

Pursuant to the Act and 10 CFR Parts 30, 40 and 70, to 4

receive, possess, and use at any time any byproduct, source i

and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation

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and radiation monitoring equipment calibration, and as l

fission detectors in amounts as required; i

(4)

Pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with j

radioactive apparatus or components; 1

(5)

Pursuant to the Act and 10 CFR Parts 30 and 70, to possess i

but not separate, such byproduct and special nuclear 1

materials as may be produced by the operation of the facility; (6)

Pursuant to the Act and 10 CFR Parts 30 and 70, to transfer Amdt.

byproduct materials from other NSP job sites for the purposes No. 86 of volume reduction and decontamination.

7-25-8 l

C.

This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations l

in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Sections i

50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is i

subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Imvel i

The licensee is authorized to operate the facility at steady i

state reactor core power levels not in excess of 1650 megawatts thermal.

l (2) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 120, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

Unit 1 i

Amendment No.120

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(3) Physical Protection The licensee shall fully implement and maintain in effect all provisions of the Commission-approved physical security, guard training and qualification, and safeguards contingency plans including amendments made pursuant to provisions of the Miscellaneous Amendments and Search Requirements revisions to 10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of Amdt.

10 CFR 50.90 and 10 CFR 50.54(p).

The plans, which contain No. 85 Safeguards Information protected under 10 CFR 73.21, are 1-5-89 entitled:

" Prairie Island Nuclear Generating Plant Physical Security Plan," with revisions submitted through November 30, 1987; " Prairie Island Nuclear Generating Plant Guard Training and Qualification Plan," with revisions submitted through February 26, 1986; and " Prairie Island Nuclear Generating Plant Safeguards Contingency Plan," with revisions submitted through August 20, 1980.

Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.

(4) Fire Protection Northern States Power Company shall implement and maintain in effect all provisions of the approved fire protection program Amdt.

as described and referenced in the Updated Safety Analysis No. 12(

Report for the Prairie Island Nuclear Generating Plant, 10-6-9 Units 1 and 2, and as approved in Safety Evaluation Reports dated February 14, 1978, September 6, 1979, April 4, 1980, December 29, 1980, July 28, 1981, September 12, 1984, June 25, 1985, October 27, 1989, and October 6, 1995, subject to the following provision:

The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

2.

D.

This license iw effective as of the date of issuance and shall expire Amdt.

at midnight August 9, 2013.

No. 79 9-23-6 FOR THE ATOMIC ENERGY COMMISSION Original Signed by Roger S. Boyd A. Ciambusso, Deputy Director for Reactor Projects Directorate of Licensing

Attachment:

Change No. 3 to Appendices A and B Date of Issuance: APR 5 1974 Unit 1 Amendment No. 120

TSoiv l

l TABLE OF CONTENTS (Continued)

TS SECTION IILZ PAGE 3.10 Control Rod and Power Distribution Limits TS.3.10-1 A. Shutdown Margin TS.3.10-1 B. Power Distribution Limits TS.3.10 1 C. Quadrant Power Tilt Ratio TS.3.10 4 D. Rod Insertion Limits TS.3.10 5 i

E. Rod Misalignment Limitations TS.3.10 6 F. Inoperable Rod Position Indicator Channels TS.3.10 6 C. Control Rod operability Limitations TS.3.10 7 H. Rod Drop Time TS.3.10-7 I. Monitor Inoperability Requirements TS.3.10 8 J. DNS Parameters TS.3.10-8 l

3.11 Core Surveillance Instrumentation TS.3.11-1 3.12 Snubbers TS.3.12-1 l

3.13 control Room Air Treatment System

'TS.3.13-1 A Control Room Special Ventilation System TS.3.13-1 4

3.14 Deleted 3.15 Event Monitoring Instrumentation TS.3.15-1 A. Process Monitors TS.3.15 1 I

3. Radiation Monitors TS.3.15 1 l

C. Reactor Vessel Level Instrumentation TS.3.15-2 1

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I Prairie Island Unit 1 Amendment No. 94, 102, 120 Prairie Island Unit 2 Amendment No. E7. 95. 113 1

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TS v l

b TABLE OF CGiGr.mi (Continued)

IL.3ECTIQH M

PACE j

4.0 SURVEII.1ANCE REQUIREMENTS TS.4.0 1 4.1 Operational Safety Review 75.4.1-1 j

4.2 Inservice Inspection and Testing of Pumps and i

Valves Requirements 75.4.2 1

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A. Inspection Requirements TS.4.2 1

3. 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 Tests TS.4.4 1 i

A. Containment Leakage Tests 75.4.4-1

3. Emergency Charcoal Filter Systems TS.4.4 3 i

C. Containment Vacuum Breakers 75.4.4 4 D. Residual Heat Removal System TS.4.4-4 1

E. Containment Isolation Valves TS.4.4 5 F. Postadecident Containment Ventilation System TS.4.4-5 j

C. Containment and Shield Building Air Temperature TS.4.4 5 N. Containment Shell Temperature 75.4.4-5 j

I. Electric Hydrogen Recombiners TS.4.4 5 i

4.5 Engineered Safety Teatures 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 Vater System TS.4.5 2 i

5. Cooling Water System TS.4.5 2

3. Component Tests TS.4.5 3

1. Pumps TS.4.5 3

2. Containment Tan Motors TS.4.5-3 i

3. Valves 75.4.5-3 i

4.6 Feriodic Testing of Emergency Power System TS.4.6 1 j

A. Diesel Cenerators TS.4.6-1

3. Station Batteries TS.4.6-3

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C. Pressurizer Nester Energency Power Supply TS.4.6 3 4.7 Main Steam Isolation valves TS.4.7-1 1

4.8 steam and Power Conversion Systems TS.4.8 1 i

A. Auxiliary Feedvater System TS.4.8 1

3. Steam Cenerator Power Operated Relief Valves TS.4.8 2 C. Steam Exclusion System

.TS.4.8-2 4.9 Reactivity Anomalies TS.4.9 1 4

4.10 Radiation Environmental Monitoring Program TS.4.10 1 A. Sample Collection and Analysis TS.4.10 1

3. land Use Census TS.4.10 2 i

C. Interlaboratory Comparison Program TS.4.10 2 4.11 Radioactive Source IAakage Test TS.4.11-1 1

Prairie Island Unit 1 - Amendment No. 69, 73, 9f. JOI,103 4

Prairie Island Unit 2 - Amendment No. 63, 66. 84. 94, 96 4

Corrected page i

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TS-vi TABLE OF CONTENTS (Continued)

TS SECTION M

PACE 4.12 Steam Generator Tube Surveillance TS.4.12-1 j

A. Steam Generator Sample Selection and TS.4.12-1 Inspection B. Steam Generator Tube Sample Selection TS.4.12-1 i

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.4.14-1 l

4.15 Spent Fuel Pool Special Ventilation System TS.4.15-1 1

l 4.16 Deleted i

4.17 Radioactive Effluents Surveillance TS.4.17-1 A. Liquid Effluents TS.4.17-1 B. Caseous Effluents TS.4.17-2 i

C. Solid Radioactive Waste TS.4.17-4 D. Dose from All Uranium Fuel Cycle Sources TS.4.17-4 4

4.18 Reactor Coolant Vent System Paths TS.4.18-1 A. Vent Path Operability TS.4.18-1 B. System Flow Testing TS.4.18-1 l

4.19 Auxiliary Building Crane Lifting Devices TS.4.19-1 1

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Prairie Island Unit 1 Amendment No. 91, 99, 120 Prairie Island Unit 2 Amendment No. 84, 92, 113

TS x TABLE OF CONTENTS (continued)

TS BASES SECTION IIILg PACE 2.0 BASES POR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit, Reactor Core B.2.1-1 l

2.2 Safety Limit, Reactor Coolant System Pressura B.2.2 1

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2.3 Limiting Safety System Settings, Protective 5.2.3 1 Instrumentation

^l 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 i

3.1 Reactor Coolant System B.3.1 1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1 4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 E. Maximum Reactor Coolant Oxygen, Chloride B.3.1-8 and Fluoride Concentration F. Isothermal Temperature coefficient (ITC) 5.3.1-9 3.2 Chemical and Volume Control System B.3.2 1 3.3 Engineered Safety Features B.3.3 1 3.4 Steam and Power Conversion Systems B.3.4 1 3.5 Instrumentation System B.3.5-1 l

3.6 Containment System B.3.6 1 3.7 Auxiliary Electrical System B.3.7-1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Radioactive Effluents B.3.9-1 A. Liquid Effluents B.3.9-1 B. Caseous Effluents B.3.9-2 C. Solid Radioactive Waste B.3.9-4 D. Dose From All Uranium Fuel Cycle Sources B.3.9-5

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E. & F. Effluent Monitoring Instrumentation B.3.9 5 3.10 Control Rod and Power Distribution Limits B.3.10 1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10 9 F. Inoperable Red Position Indicator Channels B.3.10 9 4

C. Control Rod Operability Limitations B.3.10-9 H. Rod Drop Time B.3.10 10 I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.11-1 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13 1 3.14 Deleted 3.15 Event Monitoring Instrumentation B.3.-15-1 Prairie Island Unit 1 Amendment No. 91, 94, 120 Prairie Island Unit 2 Amendment No. 84, 87, 113

i TS-xi

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TAELE OF CONTENTS (continued) 1 TS BASES SECTION g

PAGE 4.0 BASES FOR SURVEILLANCE REQUIREMENTS 4.1 Operational Safety Review B.4.1-1 4.2 Inservice Inspection and Testing of Pumps B.4.2-1 and Valves Requirements 4

4.3 Primary Coolant System Pressure Isolation B.4.3-1 Valves 4.4 Containment System Tests B.4.4-1 4.5 Engineered Safety Features B.4.5-1 1

4.6 Periodic Testing of Emergency Power Systems B.4.6-1 I

4.7 Main Steam Isolation Valves B.4.7-1 1

4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9-1 l

4.10 Radiation Environmental Monitoring Program B.4.10-1 i

A. Sample Collection and Analysis B.4.10-1 i

B. Land Use Census 5.4.10-1 C. Interlaboratory Comparison Program B.4.10 1 4.11 Radioactive Source Leakage Test B.4.11-1 i

4.12 Steam Generator Tube Surveillance B.4.12-1 4.13 Snubbers B.4.13-1 4.14 Control Room Air Treatment System Tests B.4.14-1 4.15 Spent Fuel Pool Special Ventilation System B.4.15-1 4.16 Deleted 4.17 Radioactive Effluents Surveillance B.4.17-1 4.18 Reactor Coolant Vent System Paths B.4.18-1

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4.19 Auxiliary Building Crane Lifting Devices B.4.19-1 d

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J Prairie Island Unit 1 Amendment No. 91. 99, 120 Prairie Island Unit 2 Amendment No. 84, 92, 113

l TS xii i

TECHNICAL SPECIFICATIONS i

LitT OF Tam re

.i TS TABLE IZEZ

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i 1-1 operational Modes J

i 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5 2A Reactor Trip Systes Instroentation i.

3.5 2B Engineered Safety Feature actuation System Instrumentation 3.9 1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Caseous Effluent Monitoring instrumentation i.

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3.15-1 Event Monitoring Instrumentation - Process & Containment 3.15 2 Event Monitoring Instrumentation - Radiation 4'.1 1A Reactor Trip System Instrumentation Surveillance Requirements

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4.1 1B Engineered Safety Feature Actuation System Instrumentation j

Surveillance Requirements j

4.1 IC Miscellaneous Instrumentation Surveillance Requirements i

4.1 2A Minimum Frequencies for Equipment Tests i

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4.1 23 Minimum Frequencies for Sampling Tests 4

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4.2 1 Special Inservice Inspection Requirements f

j 4.10 1 Radiation Environmental Monitoring Program (RIMP)

Sample collection and Analysis i

4.10 2 RmP - Maximum Values for the Lower Limits of Detection 4.10 3 j

RmP - Reporting Levels for Radioactivity Concentrations in Environmental Samples i

4.12 1 Steam Cenerator Tube Inspection j

4.13 1 Snubber Visual Inspection Interval l

4.17 1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirementa 3

4 4.17-2 Radioactive caseous Effluent Monitoring Instrumentation i

surveillance Requirements 4.17 3 Radioactive Liquid Waste Sampling and Analysis Program 4

i 4.17-4 Radioactive Caseous Vaste Sampling and Analysis Program 3

5.5 1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Cenerating a

Plant (Per Unit) 5.5 2 Anticipated Annual Release of Radioactive Nuclides in j

Caseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1 1 Minimum shift crew Composition Prairie Island Unit 1 Amendment No. 98,197. fif,120

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Prairie Island Unit 2 Amendment No. )J.Ipp. 104, 113

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TS.1-3 i

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i DOSE EOUIVAT FNT I-131 DOSE EQUIVA1.ENT I-131 is that concentration of I-131 (uci/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1 132, I.133,1-134, and I-135 i

actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844,

" Calculation of Distance Factors for Power and Test Reactor Sites".

I-AVERAGE DISINTEGRATION ENERGY I shall be the average (weighted in proportion to the concentration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than iodines, with half lives greater than 15 minutes, making up at least 95%

of the total non-iodine activity in the coolant.

1 CASEOUS RADWASTE TREATMENT SYSTEM l

The CASEQUS RADWASTE TREATMENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

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Prairie Island Unit 1 Amendment No. 91, Ill, 120 Prairie Island Unit 2 Amendment No. 84, 104, 113 a

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3.

At least two licensed operators shall he present in the control room during a reactor startup, a scheduled reactor 4

shutdown, and during recovery from a reactor trip. These 1

i operators are in addition to those required for the other j

reactor.

4.

An individual qualified in radiation protection procedures shall be on site when fuel is in a reactor, j

5.

All refueling operations shall be directly supervised by a i

licensed Senior Reactor Operator or a Senior Reactor Operator Limited to Fuel Handling who has no other concurrent respons.

ibilities during this operation.

6.

The General Superintendent Plant Operations shall be formerly 2

licensed or hold a current license on a similar type plant.

4 7.

At least one member of plant management holding a current Senior Reactor Operator license shall be assigned to the plant operations group on a long tern basis (approximately two years). This i

individual shall not be assigned to a rotating shift.

)

D.

Each member of the plant staff shall meet or exceed the minimum qualifications of ANSI N18.1-1971 for comparable positions, except I

for (1) the General Superintendent Radiation Protection who shall meet or exceed the qualifications of Regulatory Guide 1.8, September 1975, i,

and (2) the Shift Manager who shall have a bachelors degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for 4

transients and accidents, and (3) the General Superintendent Plant Operations who shall meet the requirements of ANSI N18.1-1971, except j

that NRC license requirements are as specified in Specification 6.1.C.7.

The training program shall be under the direction of a designated member j

of Northern States Power management.

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Prairie Island Unit 1 Amendment No. 82, 105, 120 Prairie Island Unit 2 Amendment No. 75, 98, 113 4

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TS.6.1 3 t

E.

Administrative procedures shall be developed and implemented to limit the working hours of unit staff who perform safety-related functions; e.g.,

senior reactor operators, reactor operators, health physicists, auxiliary operators, and key maintenance personnel.

Procedures shall include the j

following provisions:

l 1.

Adequate shift coverage shall be maintained without routine heavy j

use of overtime. The objective shall be to have operating personnel l

work a nominal 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week while the plant is operating. However, in i

the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

a. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight excluding shift turnover time.

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b. Overtime should be limited for all nuclear plant staff personnel so

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that total work time does not exceed 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, i

nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48 hour5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> period, nor more than 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> in any seven day period, all excluding shift turnover time.

Individuals 1

should not be required to work more than 15 consecutive days without two consecutive days off.

A break of at least eight hours including shift turnover time should c.

I be allowed between work periods.

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d.

Except during extended shutdown periods, the use of overtime should 2

be considered on an individual basis and not for the entire staff on I

a shift, Shift Emergency Coordinator (SEC) on-site rest time periods shall e.

not be considered as hours worked when determining the total work time for which the above limitations apply.

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i Prairie Island Unit 1 Amendment No. 64, 105. 120 Prairie Island Unit 2 Amendment No. 58, 98, 113

l TS.6.2-6 l

f.

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

l g.

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

h.

All procedures required by these Technical Specifications, including i

implementing procedures of the Emergency Plan, and the Security Plan (except as exempted in Section 6.5.F), shall be reviewed initially and periodically with a frequency commensurate with their safety i

significance but at an interval of not more than two years.

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Maintenance work requests and their associated procedures shall be 1

reviewed per the requirements of Section 6.2.C.

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Special reviews and investigations, as requested by the Safety Audit I

i Committee, j.

Review of investigative reports of unplanned releases of radioactive t

material to the environs.

k.

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

j 1.

The review of safety evaluations, when safety evaluations are required by 10 CFR Part 50, Section 50.59, for procedures or procedure changes to verify that such actions do not constitute an unreviewed safety question.

Fire Protection Program and implementing procedures and the submittal m.

1 of recommended changes to the Safety Audit Commmittee.

j 5.

Authority The OC shall be advisory to the Plant Manager.

In the event of a disagreement between the recommendations of the OC and the Plant i

Manager, the course determined by the Plant Manager to be the more i

conservative will be followed. A written summary of the disagreement j

will be sent to the Vice President Nuclear Generation and the Chairman

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of the SAC for review.

6.

Records Minutes shall be recorded for all meetings of the DC and shall identify all documentary material reviewed. The minutes shall be distributed to each member of the OC, the Chairman and each member of d

the Safety Audit Committee, the Vice President Nuclear Generation and others designated by the OC Chairman.

7.

Procedures 4

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

f.

a. Responsibility and authority of the group Prairie Island Unit 1 Amendment No. 96, 105, 120 Prairie Island Unit 2 Amendment No. 89, 98, 113 i

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k UNITED STATES 3

j NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. - aaat

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NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 j

PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.113 License No. DPR-60 1.

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

A.

The application for amendment by Northern States Power Company (the licensee) dated July 11, 1994, as supplemented April 18, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and i

regulations set forth in 10 CFR Chapter I; 5

B.

The facility will operate in conformity with the application, the provisions of the A:t, and the rules and regulations of the Comission; 4

1 C.

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

safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations, t

D.

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

E.

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

satisfied.

2.

Accordingly, the following sections under paragraph 2.C of Facility Operating License No. DPR-60 are hereby amended to read as follows:

J l

2.C.(2)

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.113, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

2.C.(4)

Fire Protection Northern States Power Company shall implement and maintain in effect all provisions of the approved fire protection program as described and referenced in the Updated Safety Analysis Repcrt for the Prairie Island Nuclear Generating Plant, Units 1 and 2, and as approved in 1

Safety Evaluation Reports dated February 14, 1978, September 6, 1979, April 4, 1980, December 29, 1980, July 28, 1981, September 12, 1984, June 25, 1985, October 27, 1989, and October 6, 1995, subject to the following provision:

The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

3.

This license amendment is effective as of the date of issuance to be implemented within 30 days.

7 FOR THE NUCLEAR REGULATORY COMMISSION 1

0-Beth A. Wetzel, P oject Manager Project Directorate III-1 i

Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation i

Attachments:

1 Pages 3 and 4 of License No. DPR-60*

l Changes to the Technical Specifications Date of Issuance: October 6, 1995 i

l

• Pages 3 and 4 are attached, for convenience, for the composite license to reflect these changes.

1

t 0

ATTACHMENT TO LICENSE AMENDMENT NO. 113 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 UNIT 2 LICENSE

^

REMOVE INSERT Pages 3 Pages 3 4

4 TECHNICAL SPECIFICATIONS 1

Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.

REMOVE INSERT TS-iv TS-iv TS-v TS-v*

TS-vi TS-vi TS-x TS-x TS-xi TS-xi TS-xii TS-xii TS.1-3 TS.1-3 TS.3.14-1 TS.3.14-2 TS.3.14-3 TS.3.14-4 TABLE TS.3.14-1, page 1 l

TABLE TS.3.14-1, page 2 TABLE TS.3.14-1, page 3 TS.4.16-1 l

TS.4.16-2 TS.4.16-3 TS.4.16-4 TS.4.16-5 TS.4.16-6 TS.6.1-2 TS.6.1-2 TS.6.1-3 TS.6.1-3 TS.6.2-6 TS.6.2-6 1

B.3.14-1 B.3.14-2 B.4.16-1 B.4.16-2 j

• Corrected page i

i

e i

3 (4)- Pursuant'to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess and use in amounts as required any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis or instrument and equipment calibration or associated with

Amdt, radioactive apparatus or components; No. 6 i

5-11-76 (5) Pursuant to the Act and 10 CFR Parts 30 and 70, to possess but not separate, such byproduct and special nuclear materials as may be produced by the operation of the i

facility; 4

(6) Pursuant to the Act and 10 CFR Parts 30 and 70, to transfer

^* E I

byproduct materials from other NSP job sites for the purposes i

j of volume reduction and decontamination.

gy !

7-25-89 C.

This amended license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Sections 4

50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power Lavel The licensee is authorized to operate the facility at steady state reactor core power levels not in excess of 1650 megawatts thermal.

^1 (2) Technical Specifications

)

The Technical Specifications contained in Appendix A, as revised through Amendment No. 113, are hereby incorporated in the license. The licensee shall operate the facility in l

accordance with the Technical Specifications.

(3) Physical Protection 4

i The licensee shall fully implement and maintain in effect all provisions of the Commission approved physical security, j

guard training and qualification, and safeguards contingency Amdt j

plans including amendments made pursuant to provisions of the No. 78 Miscellaneous Amendments and Search Requirements revisions t 1-5-89 j

10 CFR 73.55 (51 FR 27817 and 27822) and to the authority of 10 CFR 50.90 and 10 CFR 50.54(p). The plans, which contain i

j Safeguards Information protected under 10 CFR 73.21, are j

entitled:

" Prairie Island Nuclear Cenerating Plant Physical j

Security Plan," with revisions submitted through November 30, I

i Unit 2 Amendment No. 113

4 (3)

Physical Protection--continued 1987; " Prairie Island Nuclear Generating Plant Guard Training Amdt.

and Qualification Plan," with revisions submitted through No. 7E February 26, 1986; and " Prairie Island Nuclear Generating 1-5-85 Plant Safeguards Contingency Plan," with revisions submitted through August 20, 1980. Changes made in accordance with 10 CFR 73.55 shall be implemented in accordance with the schedule set forth therein.

i (4) Fire Protection Northern States Power Company shall implement and maintain in Amdt effect all provisions of the approved fire protection program b

No as described and referenced in the Updated Safety Analysis gQ Report for the Prairie Island Nuclear Generating Plant, Units 1 and 2, and as approved in Safety Evaluation Reports dated February 14, 1978, September 6, 1979, April 4, 1980, December 29, 1980, July 28, 1981, September 12, 1984, June 25,1985, October 27, 1989, and October 6, 1995, subject to the following provision:

s t

The licensee may make changes to the approved Fire Protection Program without prior approval of the Commission only if those changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.

2.

D.

This license is effective as of the date of issuance and shall expire Amdt.

at midnight October 29, 2014.

No. 72 9-23-E FOR THE ATOMIC ENERGY COMMISSION Original Signed by A. Giambusso A. Ciambusso, Deputy Director for Reactor Projects Directorate of Licensing Date of Issuance:

OCT 29 1974 Unit 2 j

Amendment No. 113 1

1

TS iv i

TABLE OF CONTENTS (Continued)

TS SECTION M

PACE __

3.10 Centrol Rod and Fower Distribution Limits T8.3.10 1 A. Shutdown Margin TS.3.10 1

5. Fower Distribution Limits TS.3.10 1 C. Quadrant Power Tilt Ratio 75.3.10 4 B. 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 C. Control Rod Operability Limitations TS.3.10 7 N. Rod Drop Time TS.3.10-7 j

I. Monitor Inoperability Requirements TS.3.10-8 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 Roos Air Treatment System TS.3.13-1 A Control Room Special Ventilation System TS.3.13-1

{

3.14 Deleted i

3.15 Event Monitoring Instrumentation TS.3.15-1 A. Process Monitors TS.3.15-1 B. Radiation Monitors TS.3.15-1 C. Reactor Vessel Level Instrumentation TS.3.15-2 j

4 1

1 4

i Prairie Island Unit 1 Amendment No. 94. 102, 12 Prairie Island Unit 2 Amendment No. 27, 95, 11

l l.-

l TS v l

TABLE OF CGWiiCi fContinuadi TS SECTION M

PACE j

4.0 SURVEII.1 ALICE REQUIREMENTS TS.4.0 1 i

4.1 operational Safety Review TS.4.1-1 4.2 Inservice Inspection and Testing of Pumps and Valves Requirements TS.4.2-1 A. Inspection Requirements TS.4.2-1

3. Corrective Measures TS.4.2-2 i

C. Records TS.4.2-3 4.3 Frimary Coolant Systes Pressure Isolation Valves T8.4.3-1 4.4 Containment System Tests 2

TS.4.4 1 j

A. Containment Leakage Tests TS.4.4-1 i

3. Emergency Charcoal Filter Systems TS.4.4-3 C. Containment Vacuus Breakers TS.4.4 4 j

D. Residual Heat Removal System TS.4.4 4 i

E. Containment Isolation Valves TS.4.4-5 l

F. Post adecident containment Ventilation System TS.4.4-5 i

C. Containment and Shield Building Air j

Temperature T5.4.4 5 l

N. Containment Shell Temperature TS.4.4 5 I. Electric Hydrogen Recombiners TS.4.4-5 4.5 Engineered Safety Features TS.4.5 1 A. System Tests T5.4.5 1 1

1. Safety Injection Syctem TS.4.5 1

2. Containment spray System T5.4.5-1 j

3. Containment Fan Coolers TS.4.5 2 j

4. Component Cooling Vater System TS.4.5 2

5. Cooling Water System T5.4.5 2

8. Component Tests 75.4.5 3

1. Pumps TS.4.5 3 I

2. Containment Fan Motors TS.4.5 3

3. Valves TS.4.5 3 4.6 Feriodic Testing of Emergency Power System TS.4.6 1 A. Diesel Ceneratora TS,4.6 1 t

3. Station Batteries TS.4.6-3 C. Pressurizer Heater Emergency Power Supply TS.4.6 3 4.7 Main Stena Isolation valves TS.4.7-1 4.8 Steam and Power Conversion Systems TS.4.8-1

{

A. Auxiliary Feedwater System

)

5. Steam Generator Power Operated Relief Valves TS.4.8 2 TS.4.8-1 C. Steam Exclusion System 4

75.4.8 2

)

4.9 Reactivity Anosalies TS.4.9 1 4.10 Radiation Environmental Monitoring Program TS.4.10 1 1

A. Sample Collection and Analysis TS.4.10 1 1

8.14nd Use Census l

TS.4.10 2 C. Interlaboratory Comparison Program TS.4.10 2 1

4.11 Radioactive Source 14akage Test TS.4.11-1 Prairie Island Unit 1 - Amendment No. 69, 73, 91 JOI,103 Prairie Island Unit 2 - Amendment No. 63, 66, 84, 94, %

i Corrected page

TSovi

\\

a TABLE OF CONTENTS (Continued)

TS SECTION M

PAGE 4.12 Steam Generator Tube Surveillance TS.4.12 1 A. Steam Cenerator 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.4.14-1 4.15 Spent Fuel Fool Special Ventilation System TS.4.15-1 4.16 Deleted 4

4.17 Radioactive Effluents Surveillance TS.4.17-1 A. Liquid Effluents TS.4.17-1 B. Caseous 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 1

TS.4.18-1 A. Vent Path Operability TS.4.18-1 B. System Flow Testing TS.4.18-1 4.19 Auxiliary Building Crane Lifting Devices TS.4.19-1 4

)

1 Prairie Island Unit 1 l

Amendment No. 91, 99,120 Prairie Island Unit 2 Amendment No. 84. 92,113 I 1

4 i

TS x

~

TAB 12 OF CONTENis (continued)

TS BASES SECTION IIILE PACE 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS j

j 2.1 Safety Limit, Reactor Core B.2.1 1 2.2 Safety Limit, Reactor Coolant System Pressure B.2.2-1 2.3 Limiting safety System Settings, Protective B.2.3-1 4

Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 4

j 3.0 Applicability B.3.0 1 l

4 3.1 Reactor Coolant System B.3.1 1 j

A. Operational Components B.3.1-1 i

B. Pressure / Temperature Limits B.3.1 4 1

C. Reactor Coolant System Laskage B.3.1-6 4

D. Maximum Coolant Activity B.3.1-7 E. Maximum Reactor Coolant Oxygen, Chloride B.3.1-8

)

and Fluoride Concentration F. Isothermal Temperature coefficient (ITC) 8.3.1-9 1

1 3.2 Chemical and Volume Control System B.3.2 1 3.3 Engineered Safety Features B.3.3 1 3.4 Steam and Power Conversion Systems B.3.4 1

{

3.5 Instrumentation System B.3.5 1 q

3.6 Containment System B.3.6-1 j

3.7 Auxiliary Electrical System B.3.7-1 1

3.8 Refueling and Fuel Handling B.3.8-1 3.9 Radioactive Effluents B.3.9 1 A. Liquid Effluents B.3.9-1 i

B. Caseous Effluents B.3.9-2 C. Solid Radioactive Waste B.3.9-4 D. Dose From All Uranium Fuel Cycle Sources B.3.9 5 i

E. & F. Effluent Monitoring Instrumentation B.3.9 5 l

3.10 Control Rod and Power Distribution Limits B.3.10 1 l

A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 i

C. Quadrant Power Tilt Ratio B.3.10-6 l

D. Rod Insertion Limits B. 3.10 '

E. Rod Misalignment Limitation B.3.10 9 F. Inoperable Rod Position Indicator Channels B.3.10 9 C. Control Rod Operability Limitations B.3.10-9

)

H. Rod Drop Time B.3.10-10 l

I. Monitor Inoperability Requirements B.3.10-10 I

J. DNB Parameters B.3.10-10

]

3.11 Core Surveillance Instrumentation B.3.11 1 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13 1 l

3.14 Deleted i

3.15 Event Monitoring Instrumentation B.3.15 1 3

Prairie Island Unit 1 Amendment No. 91, 94, 120 Prairie Island Unit 2 Amendment No. 84, 87, 113 m

~. -_

),*

TS xi 1

4 TABLE OF CONTENTS (continued)

TS BASES SECTION g

FACE 4.0 BASES POR.SURVEILIANCE REQUIREMENTS 4.1 Operational Safety Review B.4.1-1 4.2 Inservice Inspection and Testing of Pumps B.4.2-1 4

and Valves Requirements 4.3 Primary Coolant Systes Pressure Isolation B.4.3-1

]

Valves 4.4 containment Systes Tests B.4.4-1 4.5 Engineered Safety Features B.4.5 1 4.6 Periodic Testing of Emergency Power Systems B.4.6 1 l

4.7 Main Steam Isolation Valves B.4.7-1 j

4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9-1 4.10 Radiation Environmental Monitoring Program B.4.10 1 A. Sample Collection and Analysis B.4.10-1 B. Land Use Census B.4.10-1 l

C. Interlaboratory Comparison Program B.4.10-1 4.11 Radioactive Source Leakage Test 5.4.11-1 4.12 Steam Generator Tube Surveillance B.4.12-1 4.13 Snubbers B.4.13-1 4.14 Control Room Air Treatment System Tests B.4.14-1 i

4.15 Spent Fuel Fool Special Ventilation System B.4.15 1 4.16 Deleted 4.17 Radioactive Effluents Surveillance B.4.17-1 4.18 Reactor Coolant Vent System Paths B.4.18 1 4.19 Auxiliary Building Crane Lifting Devices 5.4.19-1 i

Prairie Island Unit 1 Amendment No. 91, 99, 120 Prairie Island Unit 2 Amendment No. 84, 92, 113

Tsozii

(

TECHNICAL SPECIFICATIONS i

i LIST OF Tam et i

ILIMM IIILE

]

1-1 Operational Modes 3.5 1 Engineered Safety Features Initiation Instrument Limiting Set Points i

3.5 2A Reactor Trip System Instroentation 1

3.5 28 Engineered safety Testure actuation System Instrumentation 3.9 1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9 2 Radioactive Caseous Effluent Monitoring instrumentation 4

h 3.15 1 Event Monitoring Instrumentation - Process & Containment 3.15 2 Event Monitoring Instrumentation - Radiation j

4*.1 1A Reactor Trip System Instrumentation Surveillance Requirements 4.1 18 Engineered Safety Teature Actuation System Instrumentation Surveillance Requirements i

4.1 1C Miscellaneous Instrumentation Surveillance Requirements l

4.1 2A Minimum Trequencies for Equipment Tests 4.1 25 Minimum Trequencies for Sampling Tests j

4.2 1 Special Inservice Inspection Requirements i

4.10 1 4

Radiation Environmental Monitoring Frogram (REMP)

)

Sample collection and Analysis 4.10 2 RIHp - Maximum Values for the Lower Limits of Detection 1

4.10-3 RTMP - Reporting Levels for Radioactivity Concentrations in Environmental samplas t

i 4.12 1 Steam Generator Tube Inspection I

j 4.13 1 Snubber Visual Inspection Interval i

4.17 1 i

Radioactive Liquid Effluent Monitoring Instrumentation surveillance Requirements 4.17 2 Radioactive Caseous Effluent Monitoring Instrumentation i

surveillance Requirements j

4.17-3 Radioactive Liquid Waste $ampling and Analysis Program j

4.17-4 Radioactive Caseous Vaste sampling and Analysis Program 5.5 1 Anticipated Annual Release of Radioactive Material in 1

Liquid Effluents From Frairie Island Nuclear Cenerating Plant (For Unit) i 5.5 2 Anticipated Annual Release of Radioactive Nuclides in l

i Caseous Effluent From Frairie Island Nuclear Generating Flant (For Unit) 4 5.1 1 Minimum shift crew Composition Prairie Island Unit 1 Mendment No. 78,157. fif, 120 Prairie Island Unit 2 h endment No. )J.Zpp, 104, 113

l TS.1 3 l

l i

DOSE EOUITALENT I-131 DOSE EQUITALENT I-131 is that concentration of I-131 (uci/ gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I 131, I-132, I-133, I-134, and I-135 actually present.

The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID 14844,

}

" Calculation of Distance Factors for Power and Test Reactor Sites".

E AVERACE DISINTEGRATION ENERCY j

E shall be the average (weighted in proportion to the concentration of i

I each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (in MeV) for isotopes, other than todines, with half lives greater than 15 minutes, making up at least 954 of the total non iodine activity in the coolant.

CASEOUS RADVASTE TREATMENT SYSTEM i

The CASEOUS RADWASTE TREATHENT SYSTEM shall be any system designed and installed to reduce radioactive gaseous effluents by collecting primary coolant system offgases from the primary system and providing for delay or holdup for the purpose of reducing the total radioactivity prior to release to the environment.

4 e

t

)

Prairie Island Unit 1 Amendment No. 9I, III. 120 j

Prairie Island Unit 2 Amendment No. 84, 104. 113 1

5 l

TS.6.102

,1

)'

3.

At least two licensed operators shall he present in the control room during a r. actor startup, a scheduled reactor i

3 shutdown, and during recovery from a reactor trip. These operators are in addition to those required for the other reactor.

4.

An individual qualified in radiation protection procedures shall.be on site when fuel is in a reactor.

i 5.

All refueling operations shall be directly supervised by a licensed Senior Reactor Operator or a Senior Reactor Operator Limited to Puol Handling who has no other concurrent respons-j ibilities during this operation.

4 6.

The General Superintendent Plant Operations shall be formerly licensed or hold a current license on a similar type plant.

]

7.

At least one member of plant management holding a current Senior Reactor operator license shall be assigned to the plant operations group on a long term basis (approximately two years). This individual shall not be assigned to a rotating shift.

i D.

Each member of the plant staff shall meer or exceed the minimum qualifications of ANSI N16.1-1971 for comparable positions, except for (1) the General Superintendent Radiation Protection who shall meet j

or exceed the qualifications of Regulatory Cuide 1.8, September 1975, and (2) the Shift Mana5er who shall have a bachelors degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (3) the General Superintendent Plant i

operations who shall meet the requirements of ANSI N18.1-1971, except t

that NRC license requirements are as specified in Specification 6.1.C.7.

The training program shall be under the direction of a designated aember of Northern States Power management.

i i

i l

Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 82, 105, 120 Amendment No. 75, 98, 113 I

w 4.

~, - -,

i TS.6.1 3 E.

Administrative procedures shall be developed and implemented to limit the I

working hours of unit staff who perform safety related functions; e.g.,

senior reactor operators, reactor operators, health physicists, auxiliary i

operators, and key maintenance personnel.

Procedures shall include the following provisions:

1.

Adequate shift coverage shall be maintained without reatine t.sr. -

use of overtime. The objective shall be to have operating person el work a nominal 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> week while the plant is operr. ting. However, in the event that unforeseen problems require substantial amounts of overtime to be used, or during extended periods of shutdown for refueling, major maintenance or major plant modifications, on a temporary basis, the following guidelines shall be followed:

1

s. An individual should not be permitted to work more than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> straight excluding shift turnover time.

~

b. Overtime should be limited for all nuclear plant staff personnel so that total work time does not exceed 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in any 24-hour period, nor more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in any 48-hour period, nor more than 84 hours9.722222e-4 days <br />0.0233 hours <br />1.388889e-4 weeks <br />3.1962e-5 months <br /> in <

any seven day period, all excluding shift turnover time.

Individuals I

should not be required to work more than 15 consecutive days without two consecutive days off, A break of at least eight hours including shift turnover time should c.

be allowed between work periods.

d.

Except during extended shutdown periods, the use of overtime should be considered on an individual basis and not for the entire staff on a shift, Shift Emergency Coordinator (SEC) on site rest time periods shall s.

not be considered as hours worked when determining the total work time for which the above limitations apply.

l c

Prairie Island Unit 1 Prairie Island Unit 2 Amendment No. 64, 105, 120 Amendment No. 58, 98, 113 w

-u m.-

-m

I t

TS.6.2-6

s f.

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

4 3

Drills on energency procedures (including plant evacuation) and

]

adequacy of communication with offsite support groups, h.

All procedures required by these Technical Specifications, including implementing procedures of the Emergency Plan, and the Security Plan l

(except as exempted in Section 6.5.F), shall be reviewed initially

{

and periodically with a frequency commensurate with their safety 1

significance but at an interval of not more than two years.

Maintenance work requests and their associated procedures shall be reviewed per the requirements of Section 6.2.C.

l 1.

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

Committee.

)

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

J j

k.

All changes to the Process Control Program (PCP) and the Offsite Dese calculation Manual (ODCM).

t 1.

The review of safety evaluations, when safety evaluations are required by 10 CFR Part 50, Section 50.59, for procedures or l

procedure changes to verify that such actions do not constitute an j

unreviewed safety question.

i 4

m.

Fire Protection Program and implementing procedures and the submittal i

of recommended changes to the Safety Audit Commmittee.

l 5.

Authority The OC shall be advisory to the Plant Manager.

In the event of a i

disagreement between the recommendations of the OC and the Plant Manager, the course determined by the Plant Manager to be the more i

conservative will be follove1. A written summary of the disa5reement l

will be sent to the Vice P;e11 dent Juclear Ceneration and the Chairman i

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 distributec to each member of the 00, the Chairman and each member of the Safety AuJf t Committee, the Vice President Nuclear Ceneration and i

others designatsi by the OC Chairman.

7.

Procedures l

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

a. Responsibility and authority of the group l

Prairie Island Unit 1 Amendment No. 96, 105, 120 t

Prairie Island Unit 2 Amendment No. 89.

98, 113 I

.-