Text

Amends 141 & 132 to Licenses DPR-42 & DPR-60,respectively, Revising TS 6.0,administrative Controls & Following TS Sections Affected by Relocating,Removing & Modifying Requirements to TS 6.0
	ML20198L211
Person / Time
Site:	Prairie Island
Issue date:	12/07/1998
From:	Lyon C; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20198L214	List: ML20198L211; ML20198L221;
References
	NUDOCS 9901040103
	Download: ML20198L211 (66)
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UNITED STATES NUCLEAR REGULATORY COMMISSION wasmNGToN, D.C. 30606 0001

%,*see s NORTHERN STATES POWER COMPANY DOCKET NO. 50-282 J

PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.141 l

License No. DPR-42 1.

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

i A.

The application for amendment by Northem States Power Company (the licensee) dated December 14,1995, as supplemented on November 25,1996, April 10, September 4, and December 29,1997, January 8, March 2, June 11, August 12, and October 30,1998, 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;.

D.

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

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

2.

Accordingly, Facility Operating License No. DPR-42 is hereby amended to approve the relocation of certain Technical Specification requirements to licensee controlled documents, as described in the licensee's application dated December 14,1995, as supplemented on November 25,1996, April 10, September 4, and December 29,1997, January 8, March 2, June 11, August 12, and October 30,1998, and evaluated in the staffs safety evaluation attached to this amendment. This license is also hereby amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(2) and 2.C.(5) of Facility Operating License No. DPR-42 are hereby amended to read as follows:

9901040103 981207 I

l PDR ADOCK 05000282 a

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(2)

Technical Soecifications 1

1 The Technical Specifications contained in Appendix A, as revised through Amendment No.141, are hers y incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(5)

Additional Conditions The Additional Conditions contained in Appendix B, as revised through Amendment No.141, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Additional Conditions.

3.

This license amendment is effective as of the date of issuance, with fullimplementation of the attached technical specifications and License Condition 7 by September 1,1999.

License Condition 6 shall be implemented by the next USAR update, but no later than June 1,1999. Implementation shall also include the relocation of Technica!

l Specification requirements to the appropriate licensee-controlled documents as identified in the licensee's application dated December 14,1995, as supplemented on November 25,1996, April 10, September 4, and December 29,1997, January 8, 1

March 2, June 11, August 12, and October 30,1998, and evaluated in the staff's safety evaluation attached to this amendment.

FOR THE NUCLEAR REGULATORY COMMISSION M I.

Carl F. Lyon, Project Manager i

Project Directorate 111-1 1

Division of Reactor Projects - lil/IV Office of Nuclear Reactor Regulation Attachments: 1. Changes to the Technical Specifications

2. Appendix B, page B-2 Date of Issuance:

December 7, 1998

a-ATTACHMENT TO LICENSE AMENDMENT NO.141 FACILITY OPERATING LICENSE NO. DPR-42 DOCKET NO. 50-282 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 verticallines indicating the area of change.

REMOVE INSERT TS-ii TS-il TS-v TS-v TS-viii TS-viii TS-ix TS-ix TS-x TS-x TS-xi TS-xii TS-xiii TS.3.1-10 TS.3.1-10 TS.3.1-11 Table TS.4-1-2B (Page 1)

Table TS.4-1-28 (Page 1)

TS.4.4-3 TS.4.4-3 TS.4.6-1 TS.4.6-1 TS.S.1-1 TS.S.1-1 TS.S.1-2 TS.5.1-2 TS.6-1-1 through 6.14 Table TS.6.1-1 TS.6.2-1 through TS.6.2-7 TS.6.3-1 TS.6.5-1 through TS.6.5-8 TS.6.6-1 through TS.6.6-2 TS.6.7-1 '5 rough TS.6.7-6 Table TS.U.7-1 TS.6.0-1 through TS.6.0-17 B.3.1-8 B.3.1-8 B.3.1-9 B.4.4-3 B.4.4-3 Revise Appendix B, Additional Conditions, by removing the pages identified below and inserting the attached pages.

B-2 to 4

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TS-ii TABLP OF CONTENTS (Continued)

TS SECTION TITLE PAGE

3. LIMITING CONDITIONS FOR OPERATION 3.0 Applicability TS.3.0 1 3.1 Reactor Coolant System TS.3.1-1

- A. Operational Comp :nents TS.3.1-1

1. Reactor Coolant Loops and Coolant Circulation TS.3.1-1

2. Reactor Coolant System Pressure Control TS.3.1-3 a.

Pressurizer TS.3.1-3 i

b. Pressurizer Safety, Valves TS.3.1-3 J

c. Pressurizer Power Operated Relief i

Valves TS.3.1-4

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3. Reactor Coolant Vent System TS.3.1-5 B. Pressura/ Temperature Limits TS.3.1-6

1. Reactor Coolant System TS.3.1-6

2. Pressurizer TS.3.1-6

3. Steam Generator TS.3.1-7 I

C.

Reactor Coolant System Leakage TS.3.1-8

1. Leakage Detection TS.3.1-8

2. Leakage Limitations TS.3.1-8

3. Pressure Isolation Valve Leakage TS.3.1-9 D. Maximum Coolant Activity TS.3.1 10 E. Deleted F. Isothermal Temperature Coefficient (ITC)

TS.3.1-12 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 Systems TS.3.3-1 B. Containment Cooling Systems TS.3.3-4 C. Component Cooling Water System TS.3.3-5 D. Cooling Water System TS.3.3-7 3.4 Steam and Power Conversion System TS.3.4-1 A. Steam Generator Safety and Power Operated Relief Valves TS.3.4-1 B. Auxiliary Peedwater System TS.3.4-1 C. Steam Exclusion System TS.3.4-3 D. Radiochemistry TS.3.4-3 3.5 Instrumentation System TS.3.5-1 1

i Prairie Island Unit 1 Amendment No. 80, 91, 141 Prairie Island Unit 2 Amendment No. 73, 84, 132

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i TS-v TABLE OF CONTENTS (Continued)

TS SECTION TITLE PAGE 4.0 SURVEILLANCE 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 B.-Corrective Measures TS.4.2-2 C. Records TS.4.2-3 4.3 Primary Coolant System Pressure Isolation % ='..es TS.4.3-1 4.4 Containment System Tests TS.4.4-1 A. Containment Leakage Tests TS.4.4-1 B. Emergency Charcoal Filter Systems TS.4.4-3 C. Containment Vacuum Breakers TS.4.4-4 l

l D. Deleted E. Containment Isolation Valves TS.4.4-5 F. Post Accident Containment Ventilation System TS.4.4-5 G. Containment and Shield Building Air Temperature TS.4.4-5 H. 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 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.4.5-3

1. Pumps TS.4.5 3

2. Containment Fan Motors TS.4.5-3

3. Valves TS.4.5-3 l

4.6 Periodic Testing of Emergency Power System TS.4.6-1 A. Diesel Generators TS.4.6-1 B. Station Batteries TS.4.6-3 C. Pressurizer Heater Emergency Power Supply TS.4.6-3 4.7 Main Steam Isolation Valves TS.4.7 1 4.8 Steam and Power Conversion Systems TS.4.8 1 l

A. Auxiliary Feedwater System TS.4.8-1 B. Steam Generator Power Operated Relief Valves TS.4.8-2 C. Steam Exclusion System TS.4.8-2 4.9 Reactivity Anomalies TS.4.9-1 4.10 Deleted 4.11 Deleted i

l Prairie Island Unit 1 Amendment No. 103, 122, 141 Prairie Island Unit 2 Amendment No. 96, 115, 132

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TART.R OF CGim. dis (Continued)

TS SECTION IIILE PAGE 6.0 ADMINISTRATIVE CONTROLS TS.6.0-1

!6.1

' Responsibility TS.6.0-1 6.2-Organization TS.6,0-2 1

A.'Onsite and Offsite Organizations TS.6.0-2 i

B. Plant Staff TS.6.0-2 6.3 Plant Staff Qualifications TS.6,0-4 i

6.4 Procedures TS.6,0-5 6.5

-Programs and Manuals TS.6.0-6 l

A. Offsite Dose Calculation Manual TS.6.0-6' l

B. Primary. Coolant. Sources Outside Containment TS.6,0-6' C. Post Accident Sampling TS.6,0-7

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D. Radioactive Effluent Controls Program TS.6.0-7 i

E. Component Cyclic or Transient Limit TS.6.0-8 F.

(Reserved)

TS.6.0-8 I

G.

(Reserved)

TS.6.0-9

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H.

(Reserved)

TS.6,0-9 I.

(Reserved)

TS.6.0-9 J. Explosive Gas and Storage Tank Radioactivity TS.6.C-9 Monitoring Program l

K. Diesel Puel Oil Testing Program TS.6,0-9 l

L. Technical Specification Bases Control Program TS.6.0-10 l

M. Containment Leakage Rate Testing Program TS.6-.0-10 i

6.6 Reporting Requirements TS.6.0-12 A. Occupational' Exposure Report TS.6,0-12 B. Annual Radiological Environmental

' Monitoring Report TS.6.0-12 l

C. Radioactive Effluent Report TS.6.0-13 D. Monthly Operating Reports TS.6.0-13 E. Core Operating Limits Report (COLR)

TS.6.0-13 F. Pressure and Temperature Limit Report TS.6.0-14 6.7 High Radiation Area TS.6.0-16 i

i Prairie Island Unit 1 Amendment No. 113, 116, 141

Prairie _ Island Unit 1 Amendment No. 116, 115, 132 t

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TS ix l

TECHNICAL SPECIFICATIONS i

LIST OF TABLES I

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

3.5 2A Reactor Trip System Instrumentation 3.5-2B Engineered Safety Feature Actuation System Instrumentation

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3.15-1 Event Monitoring Instrumentation 4.1-1A Reactor Trip System Instrumentation Surveillance Requirements 4.1 1B Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements 4.1-IC Miscellaneous Instrument Surveillance Requirements 4.1-2A Minimum Frequencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements 4.12-1 Steam Generator Tube Inspection 4.13 1 Snubber Visual Inspection Interval Prairie Island Unit 1 Amendment No. [I/, IE/, 141 Prairie Island Unit 2 Amendment No. 114, 115 132

'a TS-x l

APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIqpRgi r

TS FIGURE TITLE 2.1-1 Reactor Core Safety Limits 3.1 3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant j

Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 3.8-1 Spent Fuel Pool Unrestricted Region Burnup and Decay Time Require-ments - OFA Fuel 3.8-2 Spent Fuel Pool Unrestricted Region Burnup and Decay Time Require-ments STD Fuel 3.10 1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In-Leakage Rate 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 5.6-2 Spent Fuel Pool Checkerboard Interface Requirements 5.6-3 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel. No GAD

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5.6 4 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel. No GAD 5.6 5 Spent Fuel Pool checkerboard Region Burnup and Decay Time 4

i Requirements

.0FA Fuel. 4 GAD 5.6-6 Spent Fuel Pool Checkerboard Region Barnup and Decay Time Requirements - STD Fuel. 4 GAD 5.6-7 Spent Fuel Fool Checkerboard Region Burnup and Decay Time Requirements 0FA Fuel. 8 GAD 5.6-8 Spent Fuel Fool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel. 8 GAD 5.6-9 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel. 12 GAD 5.6-10 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements STD Fuel. 12 GAD 2

5.6-11 Spent Fuel Fool Checkerboard Region Burnup and Decay Time Aequirements - 0FA Fuel. 16 or More GAD 5.6-12 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements STD Fuel. 16 or More GAD 4

Prairie Island Unit 1 Amendment No. 129, 135, 141 Prairie Island Unit 2 Amendment No. 121, 127, 132

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4 TS.3.1-10 3.1.D.

MAXIMUM COOLANT ACTIVITY

1. The specific activity of the primary coolant (except as specified in 3.1.D.2 and 3 below) shall be limited to:

a. Less than or equal to 1.0 microcuries per gram DOSE EQUIVALENT I 131. and

b. Less than or equal to 100/E microcuries per gram of gross radioactivity.

2. If a reactor is critical or the reactor coolant system average temperature is greater than or equal to 500*F:

a. With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure TS.3.1-3. the reactor shall be shutdown and reactor coolant system average temperature cooled to below 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the specific activity of the primary coolant greater than 200/E microcurie per gram, the reactor shall be shutdown and reactor coolant system average temperature cooled to below 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

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3.

If a reactor is at or above COLD SHUTDOWN. with the specific activity of the primary coolant greater than 1.0 microcurie per l

gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.1-2B until the specific activity of the primary coolant is restored to within its limits.

Next pages are Figure TS.3.1-3 and TS.3.1 12.

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4 Prairie Island Unit 1 Amendment No. 91, 141 Prairie Island Unit 2 Amendment No. 84, 132

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u Table TS.4.1 2B (Page 1 of 2)

TABLE TS.4.1-2R

- MINIMUM FREQUENCIES FOR SAMPLING TESTS IKEI FREQUENCY 1.

RCS Gross 5/ week

' Activity Determination 2.

RCS Isotopic Analysis for DOSE 1/14 days (when at power)

EQ9IVALENT I-131 Concentration 3.

RCS Rsdiochemistry E determination 1/6 months (1) (when at power) 4.

RCS Isotopic Analysis for Iodine a) Onct per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever Including I-131. I-133 and I-135 the specific activity ex-coeds 1.0 uCi/ gram DOSE EQUIVALENT I-131 or 100/E uCi/ gram (at or above cold shutdown). and b) One sampla between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following THERMAL POWER-change exceeding 15 percent of the RATED THERMAL POWER within a one hour period (above hot shutdown) 5.

RCS Radiochemistry (2)

Monthly 6.

RCS Tritium Activity Weekly 7.

Deleted 8.

RCS Boror Concentration *(3) 2/ Week (4) 9.

R'4ST Boron Concentration Weekly

10. Boric Acid Tanks Boron Concentration 2/ Week

11. Caustic Standpipe NaOH Concentration Monthly

12. Accumulator Boron Concentration Monthly 13.

Spent Fuel Pit Boron Concentration Weekly

Required at all times.

I Prairie Island Unit 1 Amendment No. ffi, Iff, 141 Prairie Island Unit 2 Amendment No. 104, 121, 132 o

TS.4.4-3 b.

Cold DOP testing shall be performed after each complete or i

partial replacement of a HEPA filter bank or after any l

structural maintenance on the system housing that could affect l

the HEPA bank bypass leakage.

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Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a charcoal adsorber bank or after any structural maintenance on the system housing that i

could affect the charcoal adsorber bank bypass leakage.

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d.

Each circuit shall be operated with the heaters on at least 10 l

hours every month.

5.

Perform an air distribution test on the HEPA filter bank after any maintenance or testing that could affect the air distribution within the systems. The test shall be performed at rated flow l

rate ( 10%). The results of the test shall show the air distribution is uniform within 20%.

C.

Containment Vacuum Breakers The air-operated valve in each_ vent line shall be tested at quarterly intervals to demonstrate that a simulated containmant vacuum of 0.5 psi will open the valve and a simulated accident signal will close the valve.

I The check valves as well as the butterfly valves will be leak-tested in accordance with the requirements of Specification 4.4.A.3.

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l Prairie Island Unit 1 Amendment No. 115, 126, 141 l

Prairie Island Unit 2 Amendment No. 108, 118, 132 l'

TS.4.6-1 I

i 4.6 PERIODIC TESTING OF EMERGENCY POWER SYSTEM i

i Aeoliembility I

Applies to periodic testing and surveillance requirements of the emergency power system.

a Qhdretive l'

To verify that the emergency power sources and equipment are OPERABLE.

i Semelfication

'The following tests and surveillance shall be performed:

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i A. Diesel Generators

1. At least once each month, for each diesel generator:

Verify the fuel level in the day tank.

a.

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Verify the fuel level in the fuel storage tank.

c.

Deleted d.

Verify the fuel transfer pump can be started and transfers fuel from the storage system to the day tank.

t Verify the diesel generator can start and gradually accelerate.

e.

Verify the generator voltage and frequency can be adjusted to 4160 t j

420 volts and 60 1.2 Hz. Subsequently, manually sychronize the i

generator. gradually load to at least 1650 kW (Unit 2: 5100 kW to 5300 KW). and operate for at least 60 minutes. This test should be t

conducted in consideration of the manufacturer's recommendations regarding engine prelube, warm-up. loading and shutdown procedures i

where possible.

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Prairie Island Unit 1 Amendment No. 103, 113, 141 Prairie Island Unit 2 Amendment No. 96, 10ti, 132

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TS.5.1-1 i

i 5.0 DESIGN FEATURES 2

l 5.1 SITE The Prairie Island Nuclear Generating Plant is located on property owned i

by Northern States Power (NSP) Company at a site on the west bank of the 1

Mississippi River, approximately 6 miles northwest of the city of Red Wing Minnesota. The minimum distance from the center line of either reactor to the site exclusion boundary is 715 meters, and the low population zone distance is 1-1/2 miles. The neacest population center of 25.000 or more people is South Saint Paul. These *,ite characteristics j

comply with definitions in 10CFR100 (Reference 1).

i The U.S. Army Corp of Engineers controls the land within the exclusion i

area that is not owned by NSP. The Corps has made an agreement with NSP f

to prevent residential construction on this land for the life of the j

plar.t (Reference 2).

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These specifications use atmospheric diffusion factors based on the NRC staff evaluations. Its evaluation of accidental airborne releases is based on a relative concentration of 9.8 x 10" seconds per cubic meter l

at the site boundary. Its evaluation of routine releases is based on a relative concentration of 1.5 x 10 seconds per cubic meter (Reference j'

3).

4 The flood of record in 1965 produced a water surface elevation of +688 feet MSL at the site. The calculated probable maximum flood (PMF) level 4

is +703.6 feet mean sea level (MSL) and the estimated wave runup could reach +706.7 feet MSL. (See Section 2.4.2 of this report.)

Plant grade j

level is +695 feet MSL.

Flood protection structures have been provided. The two turbine support facilities, the common auxiliary building, and the two shield buildings I

have been physically connected by a concrete flood wall, most of the length of which constitutes the concrete foundation walls for the various buildings. The top of this wall supports the metal siding for the j

buildings at about elevation +705 feet MSL. Fourteen doors through the flood wall, or into the various buildings (including the separate screen j

+

house), are provided with receivers for the erection of flood protection i

panels to prevent flood water from reaching safety related facilities.

The cooling water pumps in the screenhouse are designed to operate up to i

a flood level of +695 feet MSL without flood protection measures, and up to a level of +707 feet MSL with the erection of flood protection panels.

The main transformer foundation is at +695 feet MSL. The transformer will function to a flood level of +698 feet MSL.

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l Frairie Island Unit 1 Amendment No. 80, 141 Prairie Island Unit 2 Amendment No. 73, 132

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-s T5.5.1-2 5.1 SITE (continued)

The plant is designed for a design basis earthquake having a horizontal ground acceleration of 0.12g and an operational basis earthquake having a horizontal ground acceleration of 0.06g.

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i References 1.

USAR. Section 2.2.1 2.

USAR. Section 3.4.5 3.

SER Sections 2.3.4 and 2.3.5 L

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Prairie Island Unit 1 Amendment No. 80, 91, 141 Prairie Island Unit 2 Amendment No. 73, 84, 132 l

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TS.6.0-1 l

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6.0 ADMINISTRATIVE CONTROLS 6.1 Responsibility A. The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility during his absence.

The plant manager or his designee shall approve, prior to implementation, each proposed test. experiment or modification to systems or equipment that affect nuclear safety.

B. The shift supervisor (SS) shall be responsible for the control room command function. During any absence of the SS from the control room while the unit is in MODE 1. 2. 3. or 4. an individual with an active senior reactor operator (SRO) license shall be designated to assume the control room command function. During any absence of the SS from the control room while the unit is in MODE 5 or 6. an individual with an active SRO license or reactor operator license shall be designated to assume the control room command function.

Prairta Island Unit 1 Amendment No. ff, fdi, 141 Prairie Island Unit 2 Amendment No. 55, 98, 132

TS.6.0 2 l

6.2 Organization

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A. Onsite and offsite organizations Onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for ac:ivities affecting safety of the nuclear powcr plant.

1. Lines of authority, responsibility and communication shall be defined and established throughout highest management levels, intermediate levels, end all operating organization positions.

These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the plant specific titles of tLsse personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the Updated Safety Analysis Report.

2. The plant manager shall report to the corporate officer specified in 6.2.A.3. shall be responsible for overall safe operation of the plant. and shall have control over those onsite activities necessary for safe operation and maintenance of the plant.

3. A corporate officer shall have corporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating, maintaining and providing technical support to the plant to ensure nuclear safety.

4. The individuals who train the operating staff, carry out health physics, or perform quality assurance functions may report to the appropriate onsite manager: however, these individuals shall have sufficient organizational freedom to ensure their independence from operating pressures.

B. Plant Staff The plant staff organization shall include the following:

1.

An operator to perform non licensed duties shall be assigned to each reactor containing fuel and one additional operator to l

perform non-licensed duties shall be assigned when either or both reactors are operating in MODES 1. 2. 3. or 4.

2.

At least one licensed operator shall be present in the control room for each reactor containing fuel. In addition, while either unit dw in MODE 1. 2. 3. or 4. at least one licensed senior reactor operator shall be present in the control room.

Prairie Island Unit 1 Amendment No. 105, 120, 141 Prairie Island Unit 2 Amendment No. 98, 173, 132

t TS.6.0-3 l

B.

Plant Staff (continued) 3.

Shift crew composition may be less than the minimum requirement of 10CFR50.54(m)(2)(i.) and 6.2.B.1 and 6.2.B.7 for a period of

' time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

j 4.

An individual qualified in radiation protection procedures shall be on site when fuel is in a reactor. The position may be vacant for not more than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence. provided immediate action is taken to fill the required position.

5.

Administrative procedures shall be developed and implemented to limit the working hours of personnel who perform safety related functions (e.g., licensed SRos. licensed Ros, health physicists.

auxiliary operators, and key maintenance personnel).

The procedures shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime.

Any deviations from the working hour guidelines shall be authorized in advance by the Plant Manager or designee in accordance with approved administrative procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or designee, to ensure that excessive hours have not been assigned. Routine deviation from the working hour guidelines shall not be authorized.

6.

The operations manager or assistant operations manager.shall hold an SRO license.

7.

The shift technical advisor (STA) shall provide advisory technical support to the shift supervisor in the areas of i

thermal hydraulics, reactor engineering and plant analysis with regard to the safe operation of the unit. Personnel performing the function of the STA shall be assigned to the shift crews' a unit is in MODE 1. 2

3. or 4.

Prairie Island Unit 1 Amendment No. 103, 120, 141 Prairie Island Unit 2 Amendment No. ?8, 113, 132

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6.3 Plant Staff Qualifications Each member of the plant staff shall meet or exceed the minimum qualifications of Regulatory Guide 1.8. Revision 1. September 1975 except for (1) personnel who perform the function of shift technical advisor shall hold an SRO license and 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 (2) the operations manager who shall meet the requirements of ANSI N18.1-1971, except that NRC license requirements are as specified in Specification 6.2.B.6.

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Prairie Island Unit 1 Amendment No. 165, 120, 141 Prairie Island Unit 2 Amendment No. 98, 113, 132 e

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l TS.6.0-5 l

6.4 Procedures Written procedures shall be established, implemented, and maintained covering the following activities:

A. The applicable procedures recommended in Regulatory Guide 1.33.

Revision 2. Appendix A. February 1978; B. The emergency operating procedures required to implement the requirements of NUREG-0737 and to NUREG-0737 Supplement 1. as stated in Generic Letter 82-33:

C. Quality control for effluent and environmental monitoring:

D. Fire protection program implementerion: and E. All programs specified in Specification 6.5.

L i

Prairie Island Unit 1 Amendment No. 61, 80, 141 Prairie Island Unit 2 Amendment No. 55, 73, 132 l

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i 1

TS.6,0 6 l

4 6.5 Programs and Manuals The following programs shall be established, implemented and maintained.

2 A. Offsite Dese calculation Manumi (ODCM)

The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct 4

of the radiological environmental monitoring program. The ODCM shall also contain the radioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Monitoring. and Radioactive Effluent Reports required by Specification 6.6.B and Specification 6.6.C.

Changes to the ODCM:

1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

I a.

sufficient information to support the change (s) together with the appropriate analyses or evaluations justifying the change (s).

9 b.

a determination that the change (s) maintain the levels of radioactive effluent control required by 10CFR20.1302, 40CFR190 10CFR50.36a. and 10CFR50. Appendix I and not adversely impact the accursey or reliability of effluent, dose or setpoint calculations:

2. Shall become effective after approval by a member of plant management designated by the Plant Manager.

t

3. Shall be submitted to the NRC in the form of a complete legible copy of the entire ODCM as a part of or concurrent with the Radioactive Effluent Report for the period of the report in which any change in the ODCM was mads. Each change shall be identified by markings in the margin of the affected 1

pages, clearly indicating the area of the page that was changed. The date (i.e., month and year) the change was implemented shall be indicated.

B. Primary Coolant Sources outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practical. The systems include portions of Residual Heat Removal. Safety Injection, and Containment Spray Systems. The program shall include the following:

Prairie Island Unit 1 Amendment No. 114, 122, 141 Prairie Island Unit 2 Amendment No. 107, 115, 132

=.

TS.6.0-7 l

B. Primary Coelant Sources Outside Containment (continued)

1. Preventive maintenance and periodic visual inspection i

requirements, and

2. Integrated leak test requirements for each system st refueling cycle intervals or less.

C. Post Accident Sameline This program provides controls that ensure the capability to obtain and analyza reactor coolant, radioactive gases, and particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following:

j

1. Training of personnel;

2. Procedures for sampling and analysis; and

3. Provisions for maintenance of sampling and analysis equipment.

D. Radionetive Effluent Centrols Progrga This program conforms to 10CFR50.36a for the control of radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably achievable.

4 This pro,c em shall allocate releases equally to each unit. The liquid t.?deste t&,r.tment system. waste gas treatment system.

containutnt purge release vent, and spent fuel pool vent are 3

shared by both units. Experience has also shown that contributions from both units are released from each auxiliary building vent. Therefore, all releases will be allocated equally in determining conformance to the design objectives of 10CFR50 Appendix I.

The program shall be contained in the ODCM. shall be implemented by procedures, and shall include remedial actions to be taken 3

whenever the program limits are exceeded. The program shall include the following elements:

1. Limitations en the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance i

tests and setpoint determination in accordance with the methodology in the ODCM;

2. Limitation on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to Appendix B to 10CFR20.1 - 20.601. Table II. Column 2 i

Prairie Island Unit 1 Amendment No. 114, 122, 141 Prairie Island Unit 2 Amendment No. 107, 115, 132

_ _. _. ~ _ _.. -. _ - _.

,_-___m__m_._

h.

TS.6,0 8 l

'D. Radiometive Effluent Controls PrenIAE (continued)

3. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10CFR20.1302 and with the j

methodology and parameters in the ODCM:

4. Limitations on the annual and quarterly doses or dose I

commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10CFR50. Appendix I:

5. Determination of cumulative dose contributions from radioactive effluents for the current calendar quarter and current calendar

' year in accordance with the methodology and parameters in the l

ODCM. Determination of projected dose contributions for radioactive effluents in accordance with the methodology in the 4

ODCM at least monthly:

)

1

6. Limitations on the functional capability and use of the liquid and gaseous effluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of one month from the liquid effluent releases vould exceed 0.12 mrem to the total body or 0.4 mrem to any organs or from the gaseous affluent releases would exceed 0.4 mrad for gamma air dose. 0.8 mrad for beta air dose, or 0.6 mtem organ dose; I

1.

7. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with Appendix B to 10CFR20.1 -

4 20.601. Table II. Column 1 i

8. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas

+

j beyond the site boundary, conforming to 10CFR50. Appendix I:

9. Limitations on the annual and quarterly doses to a member of the public from iodine-131, iodine-133, tritium, and all radionuclides in particulate form with half lives greater than eight days in gaseous effluents released from each unit to areas beyond the site

[

boundary, conforming to 10CFR50. Appendix Is and l

10. Limitation on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40CFR190.

j I

E. Comeonent Cvelle or Transient L4=4t This program provides controls to track the USAR. Section 4.1.4 cyclic and transient occurrences to ensure that components are maintained within the design limits.

j' F.

(Reserved) i Prairie Island Unit 1 Amendment No. 80, 122, 141 Prairie' Island Unit 2 knendment No. 73, 115, 132 1

s r--

.______m i

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-r-

.y.

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TS.6.0-9 l

a j

G.

(Reserved)

H.

(Reserved) 1 I.

(Reserved)

J. Exnlosive Gas and Storana Tank Radioactivity Monitorine Pronram

)

This ptogram provides controls for potentially explosive gas mixtures 1

contained in the waste gas holdup-system the quantity of radioactivity contained in gas storage tanks, and the quantity of j

radioactivity contained in unprotected outdoor liquid storage tanks.

1 j

The program shall include:

1

1. The limits for concentration of oxygen in the waste gas holdup i

system aad a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria:

2. A surveills"se program to ensure that the quantity of radioactivity contained.in each gas storage tank is less than or equal to 78.800 curies of noble gases (considered as dose equivalent Xe-133); and 3

3. A surveillance program to ensure-that the quantity of radioactivity contained in each of the following tanks shall be limited to 10 curies, excluding tritium and dissolved or entrained 4

i noble gases:

i Condensate storage tanks Outside temporary tanks

4. The provisions of TS 4.0 are applicable to the Explosive Gas and i

Storage Tank Radioactivity Monitoring Program surveillance frequencies.

K. Diesel Fuel 011 Testina Pronram i

A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The l

program shall include sampling and testing requirements, and l

acceptance criteria, all in accordance with the limits specified in Table 1 of ASTM D975-77 when checked for viscosity, water, and I

sediment. Acceptability of new fuel oil shall be determined prior to l

addition to the safeguards storage tanks. Testing of diesel fuel oil stored in safeguards storage tanks shall be performed at least every 31 days. The provisions of TS 4.0 are applicable to the Diesel Fuel oil Testing Program surveillance frequencies.

i Prairie Island Unit 1 Amendment No. 103, 122, 141 Prairie Island Unit 2 Amendment No. 96, 115, 132

]

e TS.6.0 10 l

L. Technical Seecifications Bases Centrol Prenram This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Beses of the Technical Specifications shall be made under appropriate administrative controls and reviews.

2. Licensees may make changes to Bases without prior NRC approval j

provided the changes do not involve either of the following:

a. a change in the Technical Specifications incorporated in the licenses or

b. a change to the USAR or Bases that involves an unreviewed safety question as defined in 10CFR50.59.

3. The Bases Centrol Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.

4. Proposed changes that meet the criteria of Specifications j

6.5.L.2.a and b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented j

without prior NRC approval shall be provided to the NRC on a j

frequency consistent with USAR updates.

l M. Centainment Lenkane Rate Testine Prenram i

A program shall be established to implement the leakage rate j

testing of the containment as required by 10 CFR 50.54(o) and 10 i

CFR 50. Appendix J. Option B. as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163. " Performance-Based Containment Leak-Test Program." dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident is less than the containment i

internal design pressure.

P.,

of 46 psig.

The maximum allowable primary containment leakage rate. L..

at P..

shall be 0.25% of primary containment air weight per day. For pipes connected to systems that are in the auxiliary building special ventilation zone, the total leakage shall be less than 0.1% of primary containment air weight per day at pressure P..

For pipes connected to systems that are exterior to both the shield building and the auxiliary building special ventilation zone. the total leakage past isolation valves shall be less than 0.01% of primary containment air weight per day at pressure P..

j Leakage Rate acceptance criteria are:

a. Primary containment leakage rate acceptance criterion is s 1.0 L.,

Prior to unit startup, following testing in accordance with the program, the combined leakage rate acceptance criteria Prairie Island Unit 1 Amendment No. 126, 141

{

Prairie Island Unit 2 Amendment No. 118, 132 I

._... - _. ~ _ -. - -. - - _

TS.6.0-11 l

M. Containment Lankana Rate Testinn Prenram (continued) are s 0.60 L. for all components subject to Type B and Type C tests and so.75 L. for Type A tests;

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is s 0.05 L. when testad at '

246 psig

2) For each door intergasket test. leakage rate is s 0.01 L.

when pressurized to 210 psig.

l The provisions of 4.0.A do not apply to the test frequencies T

specified in the Containment Leakage Rate Testing Program. The Containment Leakage Rate Testing Program stipulates acceptable extension of test intervals.

l The provisions of 4.0.B (except that the allowed surveillance intervals-are defined by the Containment Leakage Rate Testing Program) are applicable to the Containment Leakage Rate Testing Progotm.

l Ii-I i

l I

l l

t

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1 4

Prairie Island Unit 1 Amendment No. 17(, 141 Prairie Island Unit 2 Amendment No. 118, 132 1

.. s' l

TS.6,0-12 i

l

)

6.6 Reporting Requirements The following reports shall be submitted in accordance with 10CFR50.4 A. Occuestional Ercosure Renort l

A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 arem/yr and their associated man-rem exposure according to work and job functions, e.g..

reactor operations and surveillance.

inservice inspection. routine maintenance, special maintenance (describe maintenance). waste processing. and refueling. This 1

tabulation supplements the requirements of 10CFR20.2206. The dose assignments to various duty functions may be estimated based on pocket dosimeter. TLD. or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions. This report shall be submitted by April 30 of

{

each year.

B. Annual Radiolonical Environmental Monitorine Renort t

The Annual Radiological Environmental Monitoring Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting i

period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM), and in 10CFR50. Appendix I. Sections IV.B.2. IV.B.3, and IV.C.

4 The Annual Radiation Environmental Monitoring Reports shall include summarized and tabulated resulta in the format of Regulatory Guide 4.8. December 1975 of all radiological environmental samples taken during the report period. In the event that some results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results.

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

f The reports shall also include the following: a summary description of the tsdiological environmental monitoring program a map of sampling locations keyed to a table giving distances and directions from the reactor site; and the results of licensees participation in the Interlaboratory Comparison Program defined in the ODCM.

4 l

Prairie Island Unit 1 Amendment No. 91, 122, 141 Prairie Island Unit 2 Amendment No. 84, 115, 132

r TS.6.0-13 l

C. Radioactive Effluent Reeort The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10CFR50.36a and 10CFR$0. Appendix I.Section IV.B.1.

D. Monthlv Oeeratina Reoorts Routine reports of operating statistics and shutdown experience.

including documentation of all challenges to the pressurizer power operated relief valves or pressurizer safety valves, shall be 4

submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

i E. Core Ooeratina Lim 4ts Reeert (COLR)

J

1. Core operating limits shall be established prior to each reload

~

cycle, or prior to any remaining portion of a reload cycle. and shall be documented in the COLR for the following:

a. Heat Flux Hot Channel Factor Limit (Fa""). Nuclear Enthalpy Rise Hot Channel Factor Limit (Fo/" ). PFDH. K(Z) and V(Z)

(Specifications 3.10.B.L. 3.10.B.2 and 3.10.B.3)

b. Axial Flux Difference Limits and Target Band (Specifications 3.10.B.4 through 3.10.B.9)

c. Shutdown and Control Bank Insertion Limits (Specification 3.10.D)

d. Reactor Coolant System Flow Limit (Specification 3.10.J)

2. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC.

specifically those described in the following documents:

NSPNAD-8101 A.

" Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version)

NSPNAD-8102 A.

" Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to T2 Units"(latest approved version)

Prairie Island Unit 1 Amendment No. 91, 122, 141 Prairie Island Unit 2 Amendment No. 84, 115, 132

TS.6.0-14 E. Core Deeratinn Limits Reeert (COLR)

(continued)

WCAP-9272-P-A. " Westinghouse Reload Safety Evaluation Methodology".

July. 1985 WCAP 10054 P-A. " Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code". August, 1985 1

WCAP 10924-P A " Westinghouse Large-Break LOCA Best-Estimate Methodology". December. 1988

{

WCAP-10924-P-A. Volume 1. Addendum 4. " Westinghouse Large Break LOCA Best Estimate Methodology". August. 1990 XN-NF-77-57 (A)

XN-NP-77-57 Supplement I (A). " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors PhaseII". May, 1981 WCAP-13677 "10 CFR 50.46 Evaluation Model Report: H-COBRA / TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOa Cladding Options". April 1993 (approved by NRC SE dated November

26. 1993)

NSPNAD 93003-A. " Transient Power Distribution Methodology". (latest approved version)

3. The core operating limits shall be determined such that all applicable limits (e.g.. fuel thermal mechanical limits. core thermal-hydraulic limits. ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.

4. The COLR. including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

F. Pressure and Temeerature Limit Reeert

1. RCS pressure and temperature limits for heatup, cooldown. low temperature operation, criticality. and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following Technical Specification sections:

3.1.A.1.c(4). 3.1.A.2.c(2), 3.1.A.2.c(3). 3.1.B.1.a.

3.3.A.3 3.3.A.4. 3.3.A.5, and Table 4.1-1C.

2. The analytical methods used to determine the RCS pressure and temperature limits and Cold overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC. specifically those described in the following document:WCAP-14040-NP A. Revision 2. " Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Includes any exemption granted by NRC to ASME Code Case N 514)

Prairie Island Unit 1 Amendment No. 122, 135, 141

-Prairie Island Unit 2 Amendment No. 115, 127, 132

. _ =.. -

i Y

TS,6.0-15 l

F. Pressure and Temnerature Lim 4t Renert (continued)

3. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto. Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline material properties will be submitted to the NRC prior to issuance of an updated PTLR.

1 l

1 1

Prairie Island Unit 1 Amendment No. 122. 135, 141 Prairie Island Unit 2 Amendment No. 115, 127, 132

.e 7.,

l' l

TS.6.0-16

~

i

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6.7 High Radiation Area i

I A. Pursuant to 10CFR20. paragraph 20.1601(c) in lieu of the

'i requirements of 10CFR20.1601. each high radiation area, as defined in j

10CFR20.'in which the intensity of radiation is greater than 100 i-arem/hr but less than or equal to 1000 ares /hr. shall be barricaded I

j and conspicuously posted as a high radiation area and entrance j

thereto shall be controlled by requiring issuance of a Radiation Work

)

Permit (RWP). Individuals qualified in radiation protection j_

procedures (e.g., health physics technicians) or personnel i

continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates'less than or equal to q

1000 mrea/hr. provided they are otherwise following plant radiation j

protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas t

shall be provided with or accompanied by one or more of the j'

following:

1. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

,4 i

2. A radiation monitoring device that continuously integrates the l

radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this j

monitoring device may be made after the dose rate levels in~the area have been established and personnel are aware of them.

3._An individual qualified in radiation protection procedures with a i

radiation dose rate monitoring device, who is responsible for e

providing positive control over the activities within the area and i

shall perform periodic radiation surveillance at the frequency specified by the radiation protection manager.

B. In addition.to the requirements of Specification 6.7.A above areas with radiation levels greater than 1000 mrem /hr shall be provided vith locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of the stay time specification of the RWP. direct or remote (such as closed circuit TV or transmitting radiation monitoring device) continuous surveillance may be made by. personnel qualified in radiation protection procedures 4

to provide positive exposure control over the activities being.

performed within the area.

Prairie Island Unit 1 Amendment No. 70, 122, 141 Prairie Island Unit 2 Amendment No. 64, 115, 132

e'

~..

TS.6,0-17 6.7~

High Radiation Area (continued)

C. For individual high radiation areas with radiation levels of greater than 1000 arem/hr, accessible to personnel, that are located within large areas such as reactor containment, where no enclosure exists for puposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barriended and conspicuously posted, and a flashing light shall be activated as a warning device.

1 4

l l

F 1

j Prairie Island Unit 1 Amendment No. 70, 122, 141 Prairie Island Unit 2 Amendment No. 64, 115, 132 i

e."

. t.

i.,

B.3.1-8 3.1 REACTOR COOLANT SYSTEM Bases continued D.

Maximum Coolant Activity The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the SITE BOUNDARY will not exceed an t

appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values j

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

Specification 3.1.D.2. permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure TS.3.1-3. accommodates possible iodine i

spiking phenomenon which may occur following changes in THERMAL POWER.

j operation with specific activity levels exceeding 1.0 microcuries/ gram DOSE EQUIVALENT i-131 but within the limits shown on Figure TS.3.1 3 l

should be minimized since the activity levels allowed by Figure TS.3.1 3 increase the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose at the SITE BOUNDARY by a factor of up to 20 following a postulated steam generator tube rupture.

j i

Reducing RCS temperature to less than 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the 1

I atmospheric steam relief valves. The surveillance requirements in Table TS.4.1-2B provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

Next page is B.3.1-10.

l Prairie Island Unit 1 Amendment No. 91, 106, 141 Prairie Island Unit 2 Amendment No. 84, 99, 132 i

l

m,., -.

7 g

B.4.4-3 4.4 CONTAINMENT SYSTEM TESTS Bases continued Several penetrations of the containment vessel and the shield building could, in the event of leakage past their isolation valves, result in leakage being' conveyed across the annulus by the penetrations themselves, thus bypassing the function of the Shield Building Ventilation System (Reference 5). Such leakage is estimated not to exceed.025% per day. A special zone of the auxiliary building has minimum-leakage construction.and controlled access, and is designated as a special ventilation zone where such leakage would be collected by either of two redundant trains of the Auxiliary Building Special Ventilation System. This system, when activated, will supplant the normal ventilation and draw a vacuum throughout the zone such that all outleakage will be through particulate and charcoal filters which exhaust to the shield building exhaust stack.

The design basis less-of-coolant accident was initially evaluated by the AEC staff (Reference 3) assuming primary containment leak rate of 0.5%

per day at the peak accident pressure. Another conservative assumption in the calculation is that primary containment leakage directly to the ABSVZ is 0.1% per day and leakage directly to the environs is 0.01% per day.

The resulting two-hour doses at the nearest SITE BOUNDARY and 30 day doses at the low population zone radius of 1% miles are less than guidelines presented in 10CFR100.

Initial leakage testing of the shield building and the ABSV resulted in a greater inleakage than the design basis. The staff has reevaluated doses for these higher inleakage rates and found that for a primary containment leak rate of 0.25% per day at peak accident pressure, the offsite doses are about the same as those initially calculated for higher primary containment leakage and lower secondary containment in-leakage (Reference 6).

Prairie Island Unit 1 Amendment No. 115, 126, 141 Prairie Island Unit 2 Amendment No. 108, 118, 132

[...

5 t

APPENDIX B e

ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. DPR-42 Amendment implementation Number Additional Condition Qalg 141

6. Relocate current Technical By the next USAR I

Specification 3.1.E, Maximum update, but no later I

i Reactor Oxygen, Chloride and than June 1,1999 i

Fluoride Concentration, I

Technical Specification 5.1 i

flood shutdown requirements to I

the USAR.

I 141

7. Relocate current Technical By September 1, i

Specification 4.6.A.1.c, Diesel 1999 i

Fuel OilTesting, requirements I

to the Diesel Fuel Oil Testing I

Program.

I i

j B-2 Amendment No.141

.' pa curg,

[

\\

UNITED STATES g

j NUCLEAR REGULATORY COMMISSION o

WASHlWGTON, D.C. 3060H001 NORTHERN STATES POWER COMPANY DOCKET NO. 50-306 PRAIRIE ISLAND NUCLEAR GENERATING PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 132 License No. DPR-60 1.

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

A.

The application for amendment by Nr.,rthern States Power Company (the licensee) dated December 14,1995, as supplemented on November 25,1996, April 10, September 4, and December 29,1997, January 8, March 2, June 11, August 12, and October 30,1998, complies with the standards and requiremer'.ts 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 hesith and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

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

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

2.

Accordingly, Facility Operating License No. DPR-60 is hereby amended to approve the relocation of certain Technical Specification requirements to licensee-controlled documents, as described in the licensee's application dated December 14,1995, as supplemented on November 25,1996, April 10, September 4, and December 29,1997, January 8, March 2, June 11, August 12, and October 30,1998, and evaluated in the staff's safety evaluation attached to this amendment. This license is also amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraphs 2.C.(2) and 2.C.(5) of Facility Operating License No. DPR-60 are hereby amended to read as follows:

. (2)

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.132, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

(5)

Additional Conditions The Additional Conditions contained in Appendix B, as revised through

(

Amendment No.132, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Additional Conditions.

3.

This license amendment is effective as of the date ofissuance, with fullimplementation of the attached technical specifications and License Condition 7 by September 1,1999.

License Condition 6 shall be implemented by the next USAR update, but no later than June 1,1990. Implementation shall also include the relocation of Technical Specification requirements to the appropriate licensee-controlled documents as identified in the licensee's application dated December 14,1995, as supplemented on November 25,1996, April 10, September 4, and December 29,1997, January 8, March 2, June 11, August 12, and October 30,1998, and evaluated in the staff's safety evaluation attached to this amendment.

FOR THE NUCLEAR REGULATORY COMMISSION CAEy Carl F. Lyon, Project Manager Project Directorate ill-1 Division of Reactor Projects -lil/IV Office of Nuclear Reactor Regulation Attachttients: 1. Changes to the Technical Specifications

2. Appendix B page B-2 I

Date of issuance: December 7, 1998

ATTACHMENT TO LICENSE AMENDMENT NO.132 FACILITY OPERATING LICENSE NO. DPR-60 DOCKET NO. 50-306 l

i Revice 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 verticallines indicating the area of enange.

l REMOVE INSERT TS-li TS-il

~

TS-v TS-v TS-viii TS-viii TS-ix TS-ix TS-x TS x TS-xi TS-xii TS-xiii TS.3.1-10 TS.3.1-10 l

TS.3.1-11 Table TS.4-1-2B (Page 1)

Ta' le TS.4-1-2B (Page 1) o TS.4.4-3 TS.4.4-3 TS.4.6-1 TS.4.6-1 TS.5.1-1 TS.S.1-1 TS.5.1-2 TS.S.1-2 TS.6-1-1 through 6.1-4 1

Table TS.6.1-1 TS.6.2-1 through YS.6.2-7 TS.6.3-1 TS.6.5-1 through TS.6.5-8 TS.6.6-1 through TS.6.6-2 TS.6.7-1 Srough TS.6.7-6 Table TS.6.7-1 TS.6.0-1 througn TS.6.0-17 1

B.3.1-8 B.3.1-8 B.3.1-9 B.4.4-3 B.4.4-3 Revise Appendix B, Additional Conditions, by removing the pages identified below and inserting the attached pages.

B-2

.. _.. ~ -. ~ _.. -.--

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4 7S 11 TABLR OF CONT PTS (Continued) 4 TS SECTIDH TITLE

... PAGE 4 -

1

3. LIMITING CONDITIONS FOR OPERATION 3.0 App,licability TS.3.0-1 3.1 Reactor Coolant System TS.3.1 1 A. Operational Components TS.3.1-1

1. Reactor Coolant Loops and Coolant Circulation TS.3.1-1

2. Reactor Coolant System Pressure Control TS.3.1-3

a. Pressuriser TS.3.1-3

b. Pressuriser Safety Valves TS.3.1 3

c. Pressurizer Power Operated Relief Valves T5.3.1-4

3. Reactor Coolant Vent System TS.3.1-5 B. Pressure / Temperature Limits TS.3.1-6

1. Reactor Coolant System TS.3.1-6

2. Pressurizer TS.3.1 6

3. Steam Generator TS.3.1-7 C.

Reactor Coolant System Leakage TS.3.1-8

1. Leakage Detection T5.3.1-8

2. Leakage Limitations TS.3.1-8

3. Pressure Isolation Valve Leakage TS.3.1-9 D. Maximum Coolant Activity TS.3.1-10 E. Deleted F. Isothermal Temperature Coefficient (ITC)

TS.3.1-12 3.2 Chemical s:.d Volume Control System TS.3.2 1 3.3 Engineere6 Safety Peatures TS.3.3 1 A. Safety Injection and Residual Heat Removal Systsas TS.3.3-1 B. Containment Cooling Systems TS.3.3-4 C. Component Cooling Water System TS.3.3 5 D. Cooling Water System TS.3.3 7 3.4 Steam and Power Conversion System TS.3.4 1 A. Steam Generator Safety and Power Operated Relief Valves TS.3.4-1 B. Auxiliary Feedwater System TS.3.4 1 C. Steam Exclusion System TS.3.4-3 D. Radiochemistry TS.3.4-3 3.5 Instrumentation System TS.3.5 1 Prairie Island Unit 1 Amendment No. 80, 91, 141 Prairie Island Unit 2 Amendment No. 73, 84, 132

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TS-v e

TABLR OF CONTENTS (Continued)

TS SECTION IIILE PAGE 4.0 SURVEILLANCE EEQUIREMENTS TS.4.0-1 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 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 Tests TS.4.4-1 A. Containment Leakage Tests TS.4.4-1 B. Emergency Charcoal Filter Systems TS.4.4-3 C. Containment Vacuum Breakers TS.4.4-4 l

D. Deleted E. Containment Isolation Valves TS.4.4-5 F. Post Accident Containment Ventilation System TS.4.4-5 G. Containment and Shield Building Air Temperature TS.4.4-5 H. 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 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

2. Component Tests TS.4.5 3

1. Pumps TS.4.5-3

2. Containment Fan Motors 75.4.5-3

3. Valves TS.4.5-3 4.6 Periodic Testing of Emergency Power System TS.4.6-1 i

A. Diesel Generators T5.4.6 1 B. Station Batteries TS.4.6 3 C. Pressurizar Heater Emergency Power Supply TS.4.6-3 1

I 4.7 Main Steam Isolation Valves TS.4.7 1 4.8 Steam and Power Conversion Systems 75.4.8 1 A. Auxiliary Feedwater System TS.4.8-1

3. Steam Generator Power Operated Relief Valves TS.4.8-2 C. Steam Exelusion System TS.4.8 2 4.9 Reactivity Anomalias TS.4.9-1 4.10 Deleted 4.11 De,1stad j

l Prairie Island Unit 1 Amendmer.t No. 103, 122, 141 Prairie Island Unit 2 Amendment No. 96, 115, 132

-e TS-viii TABLE OF CONTENTE (Continued)

TS SECTION TITLE PAGE 6.0 ADMINISTRATIVE CONTROLS TS 6,0-1 6.1 Responsibility TS.6.0-1 6.2 Organization TS.6.0-2 A. Onsite and Offsite Organizations TS.6.0-2 B. Plant Staff-TS.6.0-2 6.3 Plant Staff Qualifications TS.6.0-4 6.4 Procedures TS.6,0-3 6.5 Programs and Manuals TS.3.0-6 A. Offsite Dose Calculation Manual TS.6.0-6 B. Primary Coolant Sources outside Containment TS.6.0-6 C. Post Accident Sampling TS.6,0-7 I

D. Radioactive Effluent Controls Program TS.6.0-7 E. Component Cyclic or Transient Limit TS.6.0-8 F.

(Reserved)

TS.6.0-8 G. '(Reserved)

TS.6,0-9 H.

(Reserved)

TS.6.0-9 I.

(Reserved)

TS.6.0-9 l

J. Explosive Gas and Storage Tank Radioactivity TS.6.0 Monitoring Program K. Diesel Fuel Oil Testing Program TS.6.0-9 4

L. Technical Specification Bases Control Program TS.6,0-10 M. Containment Leakage Rate Testing Program TS.6.0-10 6.6 Reporting Requirements TS.6.0-12 A. Occupational Exposure Report' TS.6.0-12 B.' Annual Radiological Environmental Monitoring Report TS.6,0-12 C. Radioactive Effluent Report TS.6.0-13 D. Monthly Operating Reports TS 6,0-13 E. Core Operating Limits Report (COLR)

TS.6.0-13 F. Pressure and Temperature Limit Report TS.6.0-14 6.7 High Radiation Area TS.6.0-16 Prairie Island Unit 1 Amendment No. 123, 126, 141 Prairie Island Unit 1 Amendment No. 116, 118, 132

i TS-ix l

4 TECHNICAL SPECIFICATIONS LIST OF TABLES TS TART.R IIILE 1-1 Operational Modes i

3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Points 3.5-2A Reactor Trip System Instrumentation 3.5-2B Engineered Safety Feature Actuation System Instrumentation J

3.15-1 Event Monitoring Instrumentation 4.1-1A Reactor Trip System Instrumentation Surveillance Requirements 4.1-1B Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements 4.1-1C Miscellaneous Instrument Surveillance Requirements 4.1-2A Minimum Fregoencies for Equipment Tests 4.1-2B Minimum Frequencies for Sampling Tests 4.2-1 Special Inservice Inspection Requirements 4.12-1 Steam Generator Tube Inspection 4.13-1 Snubber Visual Inspection Interval Prairie Island Unit 1 Amendment No. (([ (([, 141 Prairte Island Unit 2 Amendment No. 114. 115, 132 w

TS-x l

APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES 1

)

TS FIGURE TITLE l

2.1-1 Reactor Core Safety Limits 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 uCi/ gram DOSE EQUIVALENT I-131 3.8 1 Spent Fuel Pool Unrestricted Region Burnup and Decay Time Require-ments - 0FA Fuel 3.8-2 Spent Fuel Pool Unrestricted Region Burnup and Decay Time Require-ments - STD Fuel 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4 1 Shield Building Design In Leakage Rate 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 4

5.6-2 Spent Fuel Pool Checkerboard Interface Requirements 5.6-3 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements 0FA Fuel. No GAD 5.6 4 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel. No GAD 5.6-5 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel. 4 GAD 5.6-6 Spent Fuel' Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel. 4 GAD 5.6-7 Spent Fuel Pool Checkerboard Region Burnup and Decay Time 1

Requirements 0FA Fuel. 8 GAD 5.6 8 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements STD Fuel. 8 GAD 5.6-9 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel. 12 GAD 5.6-10 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel. 12 GAD 5.6-11 Spent Fuel Pool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel, 16 or More GAD 5.6 12 Spent Fuel Fool Checkerboard Region Burnup and Decay Time Requirements - STD Fuel, 16 or More GAD l

i i

Prairie Island Unit 1 Amendment No. 129. 135, 141 Prairie Island Unit 2 Amendment No. 121, 127, 132

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TS.3.1-10 3.1.D.

MAXIMUM COOLANT ACTIVITY

1. The specific activity of the primary coolant (except as specified in 3.1.D.2 and 3 below) shall be limited to:

a. Less than or equal to 1.0 microcuries per gram DOSE EQUIVALENT I-131. and

b. Less than or equal to 100/E microcuries per gram of gross radioactivity.

i

2. If a reactor is critical or the reactor coolant system average temperature is greater than or equal to 500*F:

a. With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months during one continuous time interval or exceeding the limit line shown on Figure TS.3.1 3 the reactor shall be shutdown and reactor coolant system average temperature cooled to below 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

b. With the specific activity of the primary coolant greater than a

100/E microcurie per gram, the reactor shall be shutdown and 4

reactor coolant system average temperature cooled to below 500*F within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

3.

If a reactor is at or above COLD SHUTDOWN. with the specific activity cf the primary coolant greater than 1.0 micro:urie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per i

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

1 i

Next pages are Figure TS.3.1 3 and TS.3.1-12.

6 Prairie Island Unit 1 Amendment No. 91, 141 Prairie Island Unit 2 Amendment No. $4, 132

't.

Tablo TS.4.1-2B (Page 1 of 2)

TABLE TS.4.1-2B MINIMUM FRT.0UENCIES FOR SAMPLING TESTS e

g FREQUENCY 1.

RCS Gross 5/ week Activity Determination 2.

RCS Isotopic Analysis for DOSE 1/14 days (when at power)

EQUIVALENT I-131 Concentration 3.

RCS Radiochemistry E determination 1/6 months (1) (when at power) 4.

RCS Isotopic Analysis for Iodine a) Once per 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , whenever Including I-131. I-133. and I 135 the specific activity ex-coeds 1.0 uCi/ gram DOSE EQUIVALENT I-131 or 100/E uCi/ gram (at or above cold shutdown), and b) One sample between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months following THERMAL POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period (above hot shutdown) 5.

RCS Radiochemistry (2)

Monthly 6.

RCS Tritium Activity Weekly 7.

Deleted 8.

RCS Boron Conctntration*(3) 2/ Week (4) 9.

RWST Boron Concentration Weekly

10. Boric Acid Tanks Boron Concentration 2/ Week 4

11. Caustic Standpipe NaOH Concentration Monthly

12. Accumulator Boron Conesntratio Monthly i

13.

Spent Puel Pit Boron Concentration Weekly

' Required at all times.

Prairie Island Unit 1 Amendment No. III, Iff, 141 Prairie Island Unit 2 Amendment No. 104, 121, 132 ee e w wes m eoa

. e

e TS.4.4-3 r

b.

Cold DOP testing shall be performed after each complete or partial replacement of a HEPA filter bank or after any structural maintenance on the system housing that could affect the HEPA bank bypass leakage, c.

Halogenated hydrocarbon testing shall be performed after each complete or partial replacement of a charcoal adsorber bank or after any structural maintenance on the system housing that could affect the charcoal adsorber bank bypass leakage.

d.

Each circuit shall be operated with the heaters on at least 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months every month.

5.

Perform an air distribution test on the HEPA filter bank after any maintenance or testing that could affect the air distribution within the systems. The test shall be performed at rated flow rate ( 10%). The results of the test shall show the air distribution is uniform within 120%.

C.

Centainment vacuum Ernakers The air-operated valve in each vent line shall be tested at quarterly intkrvals to demonstrate that a simulated containmant vacuum of 0.5 psi vill open the valve and a simulated accident sigaal will close the valve.

The check valves as well as the butterfly valves will be leak-tested in accordance with the requirements of Specification 4.4.A.3.

Prairie Island Unit 1 Amendment No. 115, 126, 141 Prairie Island Unit 2 Amendment No. 108. 118, 132 l

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TS.4.6-1 i

4.6 PERIODIC TESTING OF EMERGENCY POWER SYSTEM

{

Annlicability l

Applies to periodic testing and surveillance requirements of the emergency

)

power system.

Obinetive 1

i To verify that the emergency power sources and equipment are OPERABLE.

Snecifientio1 The following tests and surveillance shall be performed:

A. Diesel Generatern

1. At least.once each month, for each diesel generators Verify the fuel level in the day tank.

a.

b.

Verify the fuel level in the fuel storage tank.

c.

Deleted d.

Verify the fuel transfer pump can be started and transfers fuel from the storage system to tiis day tank.

i a.

Verify the diesel generator can start and gradually accelerate.

Verify the generator voltage and frequency can be adjusted to 4160 !

420 volts and 60 1 1.2 Hz. Subsequently, manually sychronize the generator gradually load to at least 1650 kW (Unit 2: 5100 kW to 5300 KW) and operate for at least 60 minutes. This test should be conducted in consideration of the manufacturer's recommendations regarding engine prelube warm-up. Ioading and shutdown procedures l

where possible.

Prairie Island Unit 1 Amendment No. 103, 113, 141 Prairie Island Unit 2 Amendment No. 96, 106, 132

TS.5.1 1 5.0 DESIGN FEATURES 5.1 SITE The Prairie Island Nuclear Generating Plant is located on property owned by Northern States Power (NSP) Company at a site on the west bank of the Mississippi River, approximately 6 miles northwest of the city of Red Wing. Minnesota. The minimum distance from the center line of either reactor to the site exclusion boundary is 715 meters, and the low population zone dictance is 1-1/2 miles. The nearest population center of 25.000 or more people is South Saint Paul. These site characteristics j

comply with definitions in 20CFR100 (Reference 1).

The U.S. Army Corp of Engineers controls the land within the exclusion

)

area that is not owned by NSP. The Corps has made an agreement with NSP to prevent residential construction on this land for the life of the plant (Reference 2).

These specifications use atmospheric diffusion factors based on the NRC staff evaluations. Its evaluation of accidental airborne releases is based on a relative concentration of 9.8 x 10" seconds per cubic meter at the site boundcry. Its evaluation of routine releases is based on a 4

relative concentration of 1.5 x 10 seconds per cubic meter (Reference 3).

The flood of record in 1965 produced a water surface elevation o* +688 feet MSL at the site. The calculated probable maximum flood (PMF) level is +703.6 feet mean sea level (MSL), and the estimated wave runup could reach +706.7 feet MSL. (See Section 2.4.2 of this report.) Plant grade level is +695 feet MSL.

Flood protection structures have been provided. The two turbine support facilities, the common auxiliary building. mad the two shield buildings have been physically connected by a concrete flood wall, most of the length of which constitutes the concrete foundation walls for the various buildings. The top of this wall supports the metal siding for the buildings at about elevation +705 feet MSL. Fourteen do6rs through the flood wall, or into the various buildings (including the separate screen house), are provided with receivers for the erection of flood protection panels to prevent flood water from reaching safety related facilities.

The cooling water pumps in the screenhouse are designed to operate up to a flood level of +695 feet MSL without flood protaction measures, and up j

to a level of +707 feet MSL with the erection of flood protection panels.

The main transformer foundation is at +695 feet MSL. The transformer will function to a flood level of +698 feet MSL.

)

1 Prairie Island Unit 1 Amendment No. 80, 141 Preirie Island Unit 2 Amendment No. 73, 132

4

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s TS.5.1-2 5.1 SITE (continued)

The plant is designed for a design basis earthquake having a horizontal ground acceleration of 0.12g and an operational basis earthquake having a 4

horizontal ground acceleration of 0.06g.

s 1

a 1

References 1.

USAR. Section 2.2.1 2.

USAR. Section 3.4.5 3.

SER. Sections 2.3.4 and 2.3.5 Prairie Island Unit 1 Amendment No. 80, 91, 141 Prairie Island Unit 2 Amendment No. 73, 54, 132

5 TS.6.0 1 l

6.0 ADMINISTRATIVE CONTROLS 6.1 Responsibility A. The plant manager shall be responsible for overall unit operation and shall delegate in writing the succession to this responsibility l

during his absence.

l The plant manager or his designee shall approve, prior to implementation, each proposed test. experiment or modification to systems or equipment that affect nuclear safety.

B. The shift supervisor (SS) shall be responsible for the control room command function. During any absence of the SS from the control room while the unit is in MODE 1. 2. 3. or 4. an individual with an active senior reactor operator (SRO) license shall be designated to assums i

the control room command function. During any absence of the SS from the control room while the unit is in MODE 5 or 6. an individual with an active SRO license or reactor operator license shall be designated to assume the control room command function, l

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Prairie Island Unit 1 Amendment No. II, fdi, 141 Prairie Island Unit 2 Amendment No. 55, 98, 132 1

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TS.6.0 2 l

6.2 Organization A. Onsite and offsite ornanfrations onsite and offsite organizations shall be established for plant operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

4

1. Lines of, authority, responsibility and communication shall be defined and establishsd throughout highest management levels.

intermediate levels, and all operating organization positions.

These relationships shall be documented and updated, as appropriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements, including the plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specifications, shall be documented in the Updated Safety Analysis Report.

2. The plant manager shall report to the corporate officer specified

)

i in 5.2.A.3. shall be responsible for overall safe operation of the plant. and shall have control over those onsite activities l

necessary for safe operation and maintenance of the plant.

3. A corporate officer shall have corporate responsibility for j

overall plant nuclear safety and shall take any measures needed to ensure acceptable performance of the staff in operating.

maintaining and providing technical support to the plant to ensure 1

nuclear safety.

4. The individuals who train the operating staff, carry out health i

physics, or perform quality assurance functions may report to the appropriate onsite manager; however, these individuals shall have i

sufficient organizational freedom to ensure their independence from operating pressures.

B. Plant Staff j

The plant staff organization shall include the following:

1.

An operator to perform non licensed duties shall be assigned to j

each reactor containing fuel and one additional operator to perform non-licensed duties shall be assigned when either or both reactors are operating in MODES 1, 2, 3. or 4.

2.

At least one licensed operator shall be present in the control room for each reactor containing fuel. In addition, while either c

unit is in MODE 1. 2. 3. or 4. at least one licensed senior reactor operator shall be present in the control room.

Prairie Island Unit 1 Amendment No. 105, 120, 141 Prairie Island Unit 2 Amendment No. 98, 113, 132

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R.

e TS.6.0 3 l

B.

flant Staff (continued) 3.

Shift crew composition may be less than the minimum requirement of 10CFR50.54(m)(2)(i) and 6.2.B.1 and 6.2.B.7 for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

4.

An individual qualified in radiation protection procedures shall be on site when fuel is in a reactor. The position may be vacant for not more than'2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , in order to provide for unexpected absence, provided immediate action is taken to fill the required position.

5.

Administrative procedures shs11 be developed and implemented to limit the working hours of personnel who perform safety related functions (e.g.

licensed SRos licensed Ros. health physicists, auxiliary operators, and key m41ntenance personnel).

The procedures shall include guidelines on working hours that ensure adequate shift coverage shall be maintained without routine heavy use of overtime.

Any deviations from the working hour guidelines shall be authorized in advance by the Plant Manager or designee in accordance with approved administrative procedures and with documentation of the basis for granting the deviation.

Controls shall be included in the procedures such that individual overtime shall be reviewed monthly by the Plant Manager or designee to ensure that excessive hours have not been assigned. Routine deviation from the working hour guidelines shall not be authorized.

6 The operations manager or assistant operations manager shall hold an SRO license.

7.

The shift technical advisor (STA) shall provide advisory technical support to the shift supervisor in the areas of thermal hydraulics reactor engineering, and plant analysis with regard to the safe operation of the unit. Personnel performing the function of the STA shall be assigned to the shift crew when a unit is in MODE 1. 2. 3. or 4.

Prairie Island Unit 1 Amendment No. 105, 120. 141 Prairie Island Unit 2 Amendment No. 93, 113, 132

TS.6.0-4 l

1 6.3 Plant Staff Qualifications Each member of the plant staff shall meet or exceed the minimum qualifications of Regulatory Guide 1.8. Revision 1. September 1975 except for (1) personnel who perform the function of shift technical advisor shall hold an SRO license and 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 accidentr. and (2) the operations manager who shall meet i

the requirements of ANSI N18.1-1971. except that NRC license requirements are as specified in Specification 6.2.B.6.

i 4

1 a

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1 Prairie Island Unit 1 Amendment No. 105, 120, lol Prairie Island Unit 2 Amendment No. 98, 113, 132 o

TS.6.0-5 l

6.4 Procedures Written procedures shall be established, implemented and maintained covering the following activities:

)

'I A. The applicable procedures recommended in Regulatory Guide 1.33.

Revision 2. Appendix A. February 1978:

B. The emergency operating procedures required to implems.nt the requirements of NUREG-0737 and to NUREG 0737. Supplement 1. as stated in Generic Letter 82-33:

C. Quality control for effluent and environmental monitoring:

D. Fire protection program implementations and E. All programs specified in Specification 6.5.

j Prairie Island Unit 1 Amendinent No. 61, 80, 141 Prairie Island Unit 2 Amendment No. 55, 73, 132 u,

e,e TS.6.0-6 l

6.5 Programs and Manuals The following programs shall be established, implemented and maintained.

A. offsite Desa calculation Manum 1 foncM)

The ODCM shall contain the methodology and parameters used in the calculation of offsite doses resulting from radioactive gaseous and liquid effluents in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the radiological environmental monitoring program. The CDCM shall also contain the rcdioactive effluent controls and radiological environmental monitoring activities and descriptions of the information that should be included in the Annual Radiological Environmental Monitoring and Radioactive Effluent Reports required by Specification 6.6.B and Specification 6.6.C.

Changes to the ODCH:

1. Shall be documented and records of reviews performed shall be retained. This documentation shall contain:

a.

sufficient information to support the change (s) together with the appropriate analyses or evaluations justifying the change (s).

b.

a determination that the change (s) maintain the levels of radioactive effluent control required by 10CFR20.1302, 40CFR190 10CFR50.36a. and 10CFR50. Appendix I, and not adversely impact the accuracy or reliability of effluent.

dose or setpoint calculations:

2. Shall become effective after approval by a member of plant management designated by the Plant Manager.

3. Shall be submitted to the NRC in the form of a complete legible copy of the entire CDCM as a part 9f or concurrent with the Radioactive Effluent Report for the period of the 3

report in which any change in the ODCM was mad.e. Each change shall be identified by markings in the margin of the affected pages clearly indicating the area of the page that was changed. The date (i.e.. month and year) the change was implemented shall be indicated.

3. Pr4 mary Coelant Seurens outside Centningang This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluids during a serious transient or accident to levels as low as practical. The systems include portions of Residual Heat Removal. Safety Injection, and Containment Spray Systems. The program shall include the following:

Prairie Island Unit 1 Amendment No. 114, 122. 141 Prairie Island Unit 2 Amendment No. 107, 115, 132 I

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4 TS.6.0-7 l

I B. Pr4 mary Coelant sources outside containment (continued) j'

1. Preventive maintenance and periodic visual inspection requirements, and

}

2. Integrated leak test requirements for each system at refueling cycle intervals or less.

C. Post Accident Same14ns This program provides controls that ensure'the capability to obtain and analyze reactor coolant. radioactive gases, and j

particulates in plant gaseous effluents and containment i

atmosphere samples under accident conditions. The program shall

' include the following:

1. Training of personnel:

2. Procedures for sampling and analysis: and

3. Provisions for maintenance of sampling and analysis equipment.

D. Radiometive Effluent Centrols Premram 1

This program conforms to 10CFR50.36a for the control of radioactive effluente and for maintaining the doses to members of the public from radioactive affluents as low as reasonably I

achievable.

1 This program shall allocate releases equally to each unit. The i-liquid radwaste treatment system, vaste gas treatment system.

containment purge release vent. and spent fuel pool vent are shared by both units. Experience has also shown that j

contributions from both units are released from each auxiliary l

building vent. Therefore, all releases will be allocated equally in determining conformance to the design objectives of 10CFR50.

Appendix I.

The program shall be contained in the ODCM, shall be implemented by procedures, and shall include remedial actions to be taken whenever the program limits are exceeded. The program shall include the following elements:

1. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance tests and se point determination in accordance with the methodology in the ODCM:

h

?

2. Limitation on the concentrations of radioactive material released in liquid effluents to unrestricted areas, conforming to Appendix B :o 10CFR20.1 - 20.601. Table II. Column 2:

1i Prairie Island Unit 1 Amendment No, 114, 122, 141 Prairie Island Unit 2 Amendment No. 107, 115, 132 4

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4 3

n.*

+

TS.6,0 8 l

D. Radioactive Effluent Controls Pronram (continued) l-

3. Monitoring, sampling, and analysis of radioactive liquid and i

gaseous effluents in accordance with 10CFR20.1302 and with the j

methodology and parameters in the ODCM.

l 4 Limitations on the annual and quarterly doses or dose commitment to a member of the public from radioactive materials in liquid effluents released from each unit to unrestricted areas, conforming to 10CFR50, Appendix I:

l

5. Determination of cumulative dose contributions from radioactive affluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM. Determination of projected dose contributions for

)

radioactive affluents in accordance with the methodology in the ODCM at least monthly; i

}

6. Limitations on the functional capability and use of the liquid and gaseous affluent treatment systems to ensure that appropriate portions of these systems are used to reduce releases of radioactivity when the projected doses in a period of one month 1

from the liquid effluent releases would exceed 0.12 arem to the total body or 0.4 mrem to any organ: or from the gaseous effluent releases would exceed 0.4 mrad for gamma air dose. 0.8 stad for beta air dose, or 0.6 arem organ dose

7. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beyond the site boundary conforming to the dose associated with Appendix B to 10CFR20.1 -

20.601, Table II. Column 1

8. Limitations on the annual and quarterly adr doses resulting from noble gases released in gaseous effluents from each unit to areas beyond the site boundary, conforming to 10CFR50, Appendix I

9. Limitatiens on the annual and quarterly doses to a member of the public from iodine-131, iodine 133, tritium, and all radionuclides in particulate form with half lives greater than eight days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10CFR50, Appendix Is and

10. Limitation on the annual dose or dose commitment to any member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources, conforming to 40CFR190.

E. gggnonent Cvelle er Translant Limit This program provides controls to track the USAR, Section 4.1.4 cyclic and transient occurrences to ensure that components are unintained within the design limits.

F.

(Reserved)

Prairie Island Unit 1 Amendment No. 80, 122, 141 Prairie Island Unit 2 Amendment No. 73, 115, 132

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TS.6,0-9 l

G.

(Reserved)

H.

(Reserved)

I. (Reserved) l J. Erniesive Gas and Storana Tank Radiometivity Monitorine Premram This program provides controls for potentially explosive gas mixtures contained in the waste gas holdup system, the quantity of radioactivity contained in gas storage tanks, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks.

The program shall include:

1. The limits for concentration of oxygen in the waste gas holdup system and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria:

2. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than or equal to 78.800 curies of noble gases (considered as dose equivalent Xe-133); and I

3. A surveillance program to ensure that the quantity of radioactivity contained in each of the following tanks shall be limited to 10 curies, excluding tritium and dissolved or entrained noble gases:

Condensate storage tanks outside temporary tanks

4. The provisions of TS 4.0 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies.

K. Diasal Fuel Oil Testino Prenegg A diesel fuel oil testing program to implement required testing of both new fuel oil and stored fuel oil shall be established. The program shall include sampling and testing requirements, and acceptance criteria, all in accordance with the limits specified in Table 1 of ASTM D975-77 when checked for viscosity, water, and sediment. Acceptability of new fuel oil shall be determined prior to addition to the safeguards storage tanks. Testing of diesel fuel oil stored in safeguards storage tanks shall be performed at least every 31 days. The provisions of TS 4.0 are applicable to the Diesel Fuel 011 Testing Program surveillance frequencies.

1 Prairie Island Unit 1 Amendment No. 103, 122, 141 Prairie Island Unit 2 Amendment No. 96, 115, 132

TS.6.0-10 l

L. Technical Sencifications Bases Centrol Prenram This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Bases of the Technical Specifications shall be made under appropriate administrative controls and reviews.

~

2. Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following:

a. a change in the Technical Specifications incorporated in the licenses or f

b. a change to the USAR or Bases that involves an unreviewed safety question as defined in 10CFR50.59.

3. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the USAR.

4. Proposed changes that meet the criteria of Specifications 6.5.L.2.

and b above shall be reviewed and approved by the NRC prj zo implementation. Changes to the Bases implemented without

ior NRC approval shall be provided to the NRC on a freque-

consistent with USAR updates.

M. Centainment Leakane Rate Testinn Prenram A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50. Appendix J. Option B. as modified by approved exemptions.

This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163. " Performance-Based Containment Leak-Test Prosram." dated September 1995.

The peak calculated containment internal pressure for the design basis loss of coolant accident is less than the containment internal design pressure. P.,

of 46 psig.

The maximum allowable primary containment leakage rate.

L.,

at P.,

shall be 0.25% of primary containment air weight per day. For i

pipes connected to systems that are in the auxiliary building special ventilation zone, the total leakage shall be less than 0.1% of primary containment air weight per day at pressure P..

For pipes connected to systems that are exterior to both the chield building and the auxiliary building special ventilation zone. the total leakage past isolation valves shall be less than 0.01% of primary containment air weight per day at pressure P..

Leakage Rate acceptance criteria are:

a. Primary containment leakage rate acceptance criterion is s 1.0 L..

Prior to unit startup, following testing in accordance with the program, the combined 3eakage rate acceptance criteria Prairie Island Unit 1 Amendment No. 126, 141 Preirie Island Unit 2 Amendment No. 118, 132 m

TS.6,0-11 l

M. Containment Leakane Rate Testine Pronram (continued) are s 0.60 L. for all components subject to Type 3 and Type C tests and s0.75 L. for Type A tests;

b. Air lock testing acceptance criteria are:

1) Overall air lock leakage rate is s 0.05 L. when tested at 246 psig

2) For each door intergasket test. leakage rate is s 0.01 L.

when pressurized to 210 psig.

The provisions of 4.0.A do not apply to the test frequencies specified in the Containment Leakage Rate Testing Program. The Containment Leakage Rate Testing Program stipulates acceptable extension of test intervals.

The provisions of 4.0.B (except that the allowed surveillance intervals are defined by the Containment Leakage Rate Testing Program) are applicable to the Containment Leakage Rate Testing Program.

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i Prairie Island Unit 1 Amendment No. ff(, 141 Prairie Island Unit 2 Amendment No. 118, 132

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TS.6.0 12 l

6.6 Reporting Requirements The following reports shall be submitted in accordance with 10CFR50.4 A. Occuestional Ernemure Reeert A tabulation on an annual basis of the number of station, utility and other personnel (including contractors) receiving exposures greater than 100 mrem /yr and their associated man-rem exposure according to work and job functions, e.g..

reactor operations and surveillance, inservice inspection, routine maintenance, special maintenance (describe maintenance), vaste processing, and refueling. This tabulation supplements the requirements of 10CFR20.2206. The dose assignments to various duty functions may be estimated based on pocket dosimeter. TLD. or film badge measurements. Small exposures totalling less than 20% of the individual total dose need not be accounted for. In the aggregate, at least 80% of the total whole body dose received from external sources should be assigned to specific major work functions. This report shall be submitted by April 30 of each year.

i B. Annual Radielenical Environmental Monitorine Reeert The Annual Radiological Environmental Monitoring Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include summaries, interpretations, and analyses of trends of the results of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with the objectives outlined in the Offsite Dose Calculation Manual (ODCM). and in 10CFR50. Appendix I. Sections IV.B.2. IV.B.3. and IV.C.

The Annual Radiation Environmental Monitoring Reports shall include summarized and tabulated results in the format of Regulatory Guide 4.8. December 1975 of all radiological environmental samples taken during the report period. In the event that some results are not

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available for inclusion with the report. the report shall be submitted noting and explaining the reasons for the missing results.

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

The reports shall also include the following: a summary description of the radiological environmental monitoring program: a map of sampling locations keyed to a table giving distances and directions from the reactor site: and the results of licensees participation in the Interlaboratory Comparison Program defined in the ODCM.

Prairie Island Unit 1 Amendment No. 91, 122, 141 Prairie Island Unit 2 Amendment No. S#, 115, 132

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N TS.6,0 13 l

i C. Radienetive Effluent Raeort The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of each year. The report shall include a summary of the quantities of radioactive liquid and gaseous affluents and solid waste released from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conforaance with 10CFR50.36a and 10CFR50. Appendix I.Section IV.B.I.

D. Month 1v Ocaratino Racerts j

e Routine reports of operating statistics and shutdown experience.

I including documentation of al.; challenges to the pressurizer power operated relief valves or pressurizer safety valves. shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

E. Core Deeratine L4mits Reeert (COLR) l

1. Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

i l

a. Heat Flux Hot Channel Factor Limit (Fa""). Nuclear Enthalpy l

Rise Hot Channel Factor Limit (Fa/#' ). PFDH. K(Z) and V(Z)

(Specifications 3.10.B.1, 3.10.B.2 and 3.10.B.3)

b. Axial Flux Difference Limits and Target Band (Specifications 3.10.B.4 through 3.10.B.9)

c. Shutdown and Control Bank Insertion Limits (Specification 3.10.D)

d. Reactor Coolant System Flow Limit (Specification 3.10.J)

2. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

NSPNAD-8101 A. " Qualification of Reactor Physics Methods for Application to PI Units" (latest approved version)

NSPNAD-8102-A. " Prairie Island Nuclear Power Plant Reload Safety j

Evaluation Methods for Application to PI Units"(latest approved version)

Prairie Island Unit 1 Amendment No. 91, 122, 141 Prairie Island Unit 2 Amendment No. 84, 115, 132

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4

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TS.6.0 14 E. Core Oeeratinn 14m4ts Reeert (COLR)

(continued) i WCAP 9272-P A. " Westinghouse Reload Safety Evaluation Methodology".

July 1985 WCAP-10054-P-A " Westinghouse Small Break ECCS Evaluation Model using the NOTRUMP Code". August. 1985 WCAP-10924 P A. Westinghouse Large Break LOCA Best-Estimate Methodology". December. 1988 WCAP-10924 P-A. Volume 1. Addendum 4. " Westinghouse Large Break LOCA Best Estimate Methodology". August, 1990 KN NP-77-57 (A). XN NP 77-57. Supplement 1 (A). " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors PhaseII" May. 1981 WCAP 13677 "10 CPR 50.46 Evaluation Model Report: M COBRA / TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLow Cladding options". April 1$93 (approved by NRC SE dated November 26, 1993)

I NSPNAD-93003-A. " Transient Power Distribution Methodology".(latest approved version)

3. The core operating limits shall be determine / such that all applicable limits (e.g., fuel thermal-mechanical limits. core thermal-hydraulic limits. ECCS limits, nuclear limits such as shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met.

4. The COLR. including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

P. Pressure and Temeerature Limit Reeert

1. RCS pressure and temperature limits for heatup, cooldown. Iow temperature operation, criticality, and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following Technical Specification sections:

3.1. A.1.c (4). 3.1. A. 2. c (2 ). 3.1. A. 2. c (3 ). 3.1.B.1. a.

3.3.A.3, 3.3.A.4. 3.3.A.5, and Table 4.1-1C.

2. The analytical methods used to determine the RCS pressure and temperature limits and Cold overpressure Mitigation System setpoints shall be those previously reviewed and approved by the NRC. specifically those described in the following document:WCAP-14040 NP A. Revision 2. " Methodology Used to Develop Cold overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves" (Includes any exemption granted by NRC to ASME Code case N 514)

Prairie Island Unit 1 Amendment No. 122, 135, 141 Prairie Island Unit 2 Amendment No. 115, 127, 132

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TS.6.0-15 l

P. Pressure and Temnaratura Lim 4t Raeort (Continued)

3. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto. Changes to the curves, setpoints, or parameters in the PTLR resulting from new or additional analysis of beltline j

material properties will be submitted to the NRC prior to issuance of an updated PTLR.

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1 Prtirie Island Unit 1 Amendment No. 122, 135, 141 Prairie Island Unit 2 Amendment No. 125, 127, 132

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l TS.6.0-26 6.7 High Radiation Area A. Pursuant to 10CFR20. paragraph 20.1601(c). in lieu of the requirements of 10CFR20.1601. each high radiation area, as defined in 10CFR20. in which the intensity of radiation is greater than 100 arem/hr but less than or equal to 1000 mrem /hr. shall be barricaded and conspicuously posted as a high radiation area and entrance thereto shall be controlled by requiring issuance of a Radiation Work Permit (RWP). Individuals qualified in radiation protection procedures (e.g.. health physics technicians) or personnel continuously escorted by such individuals may be exempt from the RWP issuance requirement during the performance of their assigned duties in high radiation areas with exposure rates less than or equal to 1000 mres/hr. provided they are otherwise following plant radiation protection procedures for entry into such high radiation areas.

Any individual or group of individuals permitted to enter such areas shall be provided with or accompanied by one or more of the following:

1. A radiation monitoring device that continuously indicates the radiation dose rate in the area.

2. A radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received. Entry into such areas with this monitoring device may be made after the dose rate levels in the area have been established and personnel are aware of them.

3. An individual qualified in radiation protection procedures with a radiation dose cate monitoring device, who is responsible for providing porltive control over the activities within the area and shall perform periodic radiation surveillance at the frequency 1

specified by the radiation protection manager.

]

B. In addition to the requirements of Specification 6.7.A above, areas j

with radiation levels greater than 1000 mrem /hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the keys shall be maintained under the administrative control of the Shift Supervisor on duty or health physics supervision. Doors shall remain locked except during periods of access by personnel under an approved RWP that shall specify the dose rate levels in the immediate work areas and the maximum allowable stay times for individuals in those areas. In lieu of the stay time specification of the RWP. direct or remote (such as closed circuit TV or transmitting radiation monitoring device) continuous surveillance may be made by personnel qualified in radiation protection procedures to provide positive exposure control over the activities being performed within the area.

Prairie Island Unit 1 Amendment No. 70, 122, 141 Prairie Island Unit 2 Amendment No. 64, 115, 132

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e TS.6.0-17 6.7 High Radiation Area (continued)

C.-For individual high radiation areas with radiation levels of greater than 1000 arem/hr. accessible to personnel, that are located with!n large areas such as reactor containment, where no enclosure exists 3

for puposes of locking, or that cannot be continuously guarded, and where no enclosure can be reasonably constructed around the individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a varning device.

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t 1

1 4

3 Prairie Island Unit 1 Amendment No. 70, 122, 141 Prairie Island Unit 2 Amendment No. 64, II5, 132

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B.3.1-8 3.1 REACTOR COOLANT SYSTEM Bases continued D.

Maximum Coolant Activity The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses at the SITE BOUNDARY will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady j

state primary to secondary steam generator leakage rate of 1.0 gpm. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of typical site locations. These values are conservative in that specific site parameters of the Prairie Island site. such as SITE BOUNDARY location and meteorological conditions, were not considered in this evaluation.

1 Specification 3.1.D.2. permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microcuries/ gram DOSE EQUIVALENT I-131. but within the allowable limit shown on Figure TS.3.1-3 accommodates possible iodine l

spiking phenomenon which may occur following changes in THERMAL POWER.

Operation with specific activity levels exceeding 1.0 microcuries/ gram i

DOSE EQUI 7ALENT I-131 but within the limits shown on Figure TS.3.1-3 should be minimized since the activity levels allowed by Figure TS.3.1-3 increase the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months thyroid dose at the SITE BOUNDA2Y by a factor of up to 20 following a postulated steam generator tube rupture.

1 Reducing RCS temperature to less than 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves. The surveillance requirements in Table TS.4.1-2B provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.

Next page is B.3.1-10.

Prairie Island Unit 1 Amendment No. 91, 106, 141 Prairie Island Unit 2 Amendment No. 84, 99, 132 1

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B.4.4-3 4.4 CONTAINM W SYSTEM TEs u Bases continued SeveraA penetrations of the containment vessel and the shield building could in the event of leakage past their isolation velves, result in leakage being conveyed across the annulus by the penetrations themselves. thus bypassing the function of the Shield Building Ventilation System (Reference 5). Such leakage is estimated not to exceed.025% per day. A special zone of the aurt:tary building has minimum leakage construction and controlled access, and is designated as a special ventilation zone where such leakage would be collected by either of two redundant trains of the Auxiliary Building Special Ventilation System. This system. when activated, will supplant the normal ventilation and draw a vacuum throughout the zone such that all outleakage will be through particulate and charcoal filters which exhaust to the shield building exhaust stack.

The design basis loss-of-coelant accident was initially evaluated by the AEC staff (Reference 3) assuming primary containment leak rate of 0.5%

per day at the peak accident pressure. Another conservative assumption in the calculation is that primary containment leakage directly to the ABSVZ is 0.1% per day and leakage directly to the environs is 0.01% per day. The resulting two-hour. doses at the nearest SITE BOUNDARY and 30-day doses at the low population zone radius of 1% miles are less than guidelines presented in 10CFK100.

Initial leakage testing of the shield building and the ABSV resulted in

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a greater inleakage than the design basis. The staff has reevalu Led doses for these higher inleakage rates and found that for a primary containment leak rate of 0.25% per day at peak accident pressure. the offsite' doses are about the same as those initially calculated for higher primary containment leakage and lower secondary containment

)

in-leaktge (Reference 6).

Prairie Island Unit 1 Amendment No. 115, 126, 141 Prairie Island Unit 2 Amendment No. 108, 118, 132 L

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APPENDIX B ADDITIONAL CONDITIONS FACILITY OPERATING LICENSE NO. DPR-60 Amendment implementation Number Additional Condition Data 132

6. Relocate current Technical By the next USAR I

Specification 3.1.E, Maximum update, but no later i

Reactor Oxygen, Chloride and than June 1,1999 i

Fluoride Concentration, I

Technical specification 5.1 I

flvod shutdown requirements I

to the USAR (TRM).

I 132

7. Relocate current Technical By September 1, I

Specification 4.6.A.1.c, D!esel 1999 l

Fuel Oil Testing, requirements I

to the Diesel Fuel Oil Testing i

Program.

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1 B-2 Amendment No.132 1

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ML20198L211

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