ML20237B506

Proposed Tech Specs Re Conformance of Administrative Controls Section 6 to Guidance of Std Tech Specs

ML20237B506
Person / Time
Site:	Prairie Island
Issue date:	08/12/1998
From:	NORTHERN STATES POWER CO.
To:
Shared Package
ML20237B497	List: ML20237B494 ML20237B500 ML20237B506
References
NUDOCS 9808180348
Download: ML20237B506 (39)

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I ATTACHMENT 1 SUPPLEMENT 8 to LICENSE AMENDMENT REQUEST DATED December 14,1995 Conformance of Administrative Controls Section 6 to the Guidance of Standard Technical Specifications Appendix A, Technical Specification Pages Marked Up Pages as proposed in this Supplement (shaded material to be added, strike-through material to be removed)

TS.6.0-2 (from December 29,1997 supplement)

TS.6.0-10 (from Supplement 4)

TS.6.0-13 (from Supplement 6) l 9808180348 980012 PDR ADOCK 05000282 P

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TS.6.0-2 6.2 Organization A. Qasite 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 activities affecting safety of the nuclear power plant.

1. Lines of authority, responsibility and communication shall be defined and established 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 bffis^e'rv4ee peesident 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 Eff5Eirlvice president 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 l

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 unit is in MODE 1, 2.

3, or 4, at least one licensed senior reactor operator shall be present in the control room.

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TS.6.0-10 L. Technical Specifications Bases Control Prqgram This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Bases or 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 license: or

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(aiaEdij above shall be reviewed and approveg by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with USAR updates.

M. Containment Leakage Rate Testing Pr2 gram 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 Program," 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 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..

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 t-_____________

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TS.6.0-13 o

C. Radioactive Effluent ReppII 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 10CFR50. Appendix I,Section IV.B.1.

D. Monthly Operatine Reports 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

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submitted on a monthly basis no later than the 15th of each month j

following the calendar month covered by the report.

l E. Core Operating Limits Report (COLR)

1. Core operating limits shall be established prior to each reload 4

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

l a Heat Flux Hot Channel Factor Limit (Fa"l), Nuclear Enthalpy Rise Hot Channel Factor Limit (Fo/" ). PFDH K(Z) and V(Z) l (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 Evaluation Methods for Application to PI Units"(latest approved version)

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s, ATTACHMENT 2 SUPPLEMENT 8 to LICENSE AMENDMENT REQUEST DATED December 14,1995 Conformance of Administrative Controls Section 6 to the Guidance of Standard Technical Specifications Appendix A, Technical Specification Pages Revised Pages as proposed in this Supplement (Changes from current Technical Specifications sidelined)

TS.6.0-2 TS.6,0-10 TS.6,0-13 l

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TS.6.0-2 l

6.2 Organization j

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

1. Lines of authority, responsibility and communication shall be defined and established 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 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 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 is in MODE 1,

2. 3, or 4, at least one licensed senior reactor operator shall be present in the control room.

v.

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TS.6,0-10 g

L. Ischnical Specifications Baggs Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Bases or 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 change in the Technical Specifications incorporated in a.

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the license: or l

b. a change to the USAR or Bases that involves an unreviewed l

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 j

6.5.L.2.a and b above shall be reviewed and approved by the i

NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with USAR updates.

M. Containment Leakage Rate Testing Ptsgram 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 Program." 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 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..

l For pipes connected to systems that are exterior to both the shield building and the auxiliary building special ventilation i

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.,

I Prior to unit startup, following testing in accordance with the program. the combined leakage rate acceptance criteria

u.

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TS.6,0-13 I

C. Radioactive Effluent Report I

The Radioactive Effluent Report covering the operation of the plant during the previous calendar year shall be submitted by May 15 of I

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 10CFR50, Appendix I, Section IV.B.1.

D. Konthly Operating Renorts

. Routine reportsJof operating statistics and shutdown experience, including documentation of all 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 Operating Limits Report (COLR)

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

I

, a ATTACHMENT 3 SUPPLEMENT '8 I

LICENSE AMENDMENT REQUEST DATED December 14. 1995 Conformance of Administrative Controls Section 6 to the Guidance of Standard Technical Specification Final License Amendment Pages Note: All of the current Technical Specification Chapter 6 pages were deleted.

including Table 6.1 1. by the original submittal and replaced with pages-TS.6,0-1 through TS.6.0-15. Pages TS.6.0-16 and TS.6,0-17 have been added by subsequent supplements.

1

- Page Number Source Document Page Number Source' Document

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TS.6.0-3 Supplement 3 TS-ii Original submittal TS v Supplement (11/25/96)

TS.6,0-4 Supplement (11/25/96)

TS-viii Supplement 4 TS 6,0-5 Supplement 2 TS-ix Supplement 4 TS.6,0-6 Original submittal TS x Supplement 7 TS.6,0-7 Supplement (11/25/96)

TS xi' deleted - Supplement 3 TS,6.0-8 Supplement (11/25/96)

TS-xii deleted.- Supplement 3 TS.6.0-9 Supplement'5 TS-xiii deleted - Original TS.6,0-10 Supplement 8 submittal TS.6.0-11 Supplement 4 TS.3.1-10 Supplement 7 TS.6.0-12 Supplement 4 TS.3.1-11 deleted - Original submittal TS.6.0-13 Supplement 8 Table

' Supplement 7 TS.6.0-14 Supplement 7 TS.4.1-2B (Page 1 of TS.6.0-15 Supplement 7 2)

TS.6.0-16 Supplement 7 l

TS.4.4-3 Supplement 7 I

TS.6,0 17 Supplement 8 l

TS.4.6-1 Original submittal B.3.1-8 Sapplement 7 TS.5.1-1 Original submittal B.3.1-9 deleted - Original TS.5.1-2 Original submittal submittal TS.' 6. 0 - 1 Original submittal B.4.4-3 Supplement 4 TS.6,0-2 Supplement 8

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TS-ii TABLE 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 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.

Pressurizer TS.3.1-3

b. Pressurizer Safety Valves TS.3.1-3 c.-Pressurizer Power Operated. Relief Valves TS.3.1-4 3'. Reactor Coolant Vent Sys':em TS.3.1-5 w.

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 TS.3.1-8

2. Leakage Limitations TS.3.1-8

3. Pressure Isolation Valve Leakage TS.3.l 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 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 I

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TS-v TABLE OF CONTENTS-(Continued)

TS SECTION.

TITLE

. PAGE 4'. 0 SURVEILLANCE REQUIREMENTS TS.4.0-1 4.1

' Operational SaFoty neview 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 L

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, j

I. Electric Hydrogen Recombiners TS.4.4-5 4.5 Engineered Safety Features TS.4.5-1 l

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'A.

System Tests TS.4.5-1

1. Safety. Injection System TS.4.5-1 l

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 l

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 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 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 j

L 4.11 Deleted i

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TS-viii TABLE OF CONTENTS (Continued)

TS SECTIOl{

TITLE PAGE 6.0 ADMINISTRATIVE CONTROLS TS.6,0-1 61 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.6,0-5 6.5 Programs and Manuals TS.6.0-6 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 D. Radioactive Effluent Controls Program TS.6,0-7

.E.

Component Cyclic or Transient Limit TS.S 0-8 F.

(Reserved).

TS.6,0-8 G.

(Reserved)

TS.6,0 8 H.

(Reserved)

TS.6,0-8 I.

(Reserved)

TS.6,0-8

.J. Explosive Gas and Storage Tank Radioactivity TS.6,0-9 Monitoring Program K. Diesel-Fuel Oil Testing Program TS.6,0-9 L. Technical Specification Bases Control Program TS.6,0-9 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 l

B. Annual Radiological Environmental Monitoring Report.

TS.6.0-12 C. Radioactive Effluent Report TS.6,0-12 D. Monthly Operating Report TS.6,0-13 E. Core Operating Limits Report (COLR)

TS.6,0-13 6.7 High Radiation Area TS.6,0-15 I

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TS-ix TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE

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1-1 Operational Modes 3.5-1 Engineered Safety Features Initiation Instrument Limiting Set Foints 3.5-2A Reactor Trip System Instrumentation 3.5-2B Engineered Safety Feature Actuation System Instrumentation 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 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 e

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TS-x APPENDIX A TEJHNICAL SPECIFICATIONS LIST OF FIGURES 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 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 5.6-2 Spent Fuel Pool Checkerboard Interface Requirements 5.6-3 Spent Fuel Fool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel, No GAD 5.6-4 Spent Fuel Fool checkerboard Region Burnup and Decay Time Requirements - STD Fuel, No GAD 5.6-5 Spent Fuel Fool Checkerboard Region Burnup and Decay Time Requirements - 0FA Fuel, 4 GAD 5.6-6 Spent Fuel Fool Checkerboard Region Burnup 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 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 Qecay Time Requirements - STD Fuel 12 GAD 5.6-11 Spent Fuel Fool 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 I

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TS.3.1-10 3.1.D.

MAXIEUM 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:

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

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:

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

1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

b. With the specific activity of the primary coolant greater than 100/2 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

3.

If a reactor is at or above COLD SHUTDOWN. with the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram. perform the sampling and analysis requirements of item 4a of Table 4.1-2B until the specific activity of the primary coolant is restored to within its limits.

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

'r Table TS.4.1-2B (Page 1 of 2)

TABLE TS.4.1-2B MINIMUM FREQUENCIES FOR SAMPLING TESTS IESI FREQUENGX 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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, whenever Including I-131 I-133, and I-135 the specific activity ex-ceeds 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> following THERMAL POWER change exceeding' is percent of the RATED THERMAL POWER within a one hour period (above hot shutdown) 5.

RCS Radiochemistry (2)

Mont'hly 6.

RCS Tritium Activity Weekly 7.

Deleted 8.

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

RWST 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.

'WQ.

18 TS.4.4-3 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.

Halogenated hydrocarbon testing shall be performed afte,r each c.

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 <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 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 q'arterly u

intervals to demonstrate that a simulated containment vacuum of 0.5 psi will open the valve and a simulated accident signal 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.

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TS.4.6 1 4.6 PERIODIC TESTING OF EMERGENCY POWER SYSTEM Applicability Applies'to periodic testing and surveillance requirements of the emergency power system.

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

Specification The following tests and surveillance shall be performed:

A. Diesel Generators

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

a.

Verify the fuel level in the day. tank.

j.

b.

Verify the fuel level in the fuel storage tank.

I j

c.. Deleted d.

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

e.

Verify the diesel generator can start and gradually accelerate.

Verify'the generator voltage and frequency can be adjusted to 4160 420 volts and 60 1.2 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. loadi ; and shutdown procedures where possible.

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TS.S.1-1 i

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'5.0 DESIGN FEATUREE 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 l

Mississippi River, approximately 6 miles northwest of the city of Red i

Wing, Minnesota. The minimum distance from the center line of either j

reactor to the site exclusion boundary is 715 meters, and the low l

population zone distance is 1-1/2 miles. The nearest population center of l

25,000 or more people is South Saint Paul. These site characteristics comply with definitions in 10CFR100 (Reference 1).

'The ti.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).

l 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 boundary. Its evaluation of routine releases is based on a 4

relative concentration of 1.5 x 10 seconds per cubic meter (Refer ~ence 3).

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 is +70?.6 feet mean cea 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, and 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 doors through the flood wall, or into the various buildings (including the separcte screen house), are provided with receivers for the erection of flood protection panels to prevent flood water from reaching safety related faci 11tles.

The cooling water pumps in the screenhouse are designed to operate up to a flood level of +695 feet MSL without flood protection measures, and up to a level ot' +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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• TS.S.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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References 1.

USAR. Section42.2.1 f

2.

USAR. Section 3.4.5 3.

SER. Sections 2.3.4 and 2,.3.5 l

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• 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 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 i

the control room while the unit is in MODE 5 or 6 an individual with J

an active SRO license or reactor operator license shall be designated to assume the control room command function.

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6.2 Organization A. Onsite and Offsite Organizatiana 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.

1. Lines of authority, responsibility and communication shall be defined and established 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 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 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 is in MODE 1, 2, 3, or 4, at least one licensed senior l

reactor operator shall be present in the control room.

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_____________1_.__.___._____._._____.______.

h.

TS.6.0-3 l

i-

.B.

Elant 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 <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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 l

for not more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,'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 SR0s, licensed Ros, health physicists, L

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 j

accordance with approved administrative procedures and with l

documentation of the basis for granting the deviation.

(-

l 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.

1

a.

TS.6.0-4 l

l 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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TS.6.0-5 l

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

A. The applicable proceduren 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. Suppl.ement 1. as stated $n Generic Letter 82-33:

C. Quality control for effluent and environmental monitoring:

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

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TS.6,0-6 l

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

A. Offsite Dose Calculation Manual (ORCM1 1

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 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:

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 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 made. Each change l

l shall be identified by markings in the margin of the affected l

pages, clearly indicating the area of the page that was l

changed. The date (i.e., month and year) the change was i

l implemented shall be indicated.

B. Primary Coolant Sources Outside Containment i

This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly radioactive fluido 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:

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b-TS.6.0-7 l

f B.

Primary Coolant Sources Outside Containment (continued)

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 Samoling This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulate in plant gaseous effluents and containment 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. Radioactive Effluent Controls Program 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.

This program shall allocate releases equally to each unit. The 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 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 prcgram 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 l

include the following elements:

1. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including surveillance l

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

J

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:

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La TS.6,0-8 c

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D. Radioactive Effluent Controls Program (continued)-

3. Monitoring, sampling, and analysis of radioactive liquid and 1

gaseous effluents in accordance with 10CFR20.1302 and with the methodology and parameters in the ODCM:

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:

5. Determination of cumulative dose contributions from radioactive effluents for the curre.t calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM. Determination of' projected dose contributions for radioactive effluents in accordance with the methodology in the ODCM at least monthly:

6. Limitations on the functional capability and use of the liquid and-L gaseous effluent treatment systems to ensure that appropriate i

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 would exceed 0.12 nrem 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 mrad for beta air dose, or 0.6 mrem 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 P to 10CFR20.1 -

20,601. Table II, Column 1:

8. Limitations on the annual and quarterly air doses resulting from noble gases released in gaseous effluents from each unit to areas 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 I: 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. Camponent Cvelic or Transient Limit l

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.

F.

(Reserved)

A,-

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TS.6.0-9 l

G.

(Reserved) i I

'H.

(Reserved)

I, (Reserved)

J. Explosive Gas and Storage Tank Radioactivity Monitoring Program This program provides controls for potentially explosive gps 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 i

equivalent Xe-133); and

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:

l Condensate storage tanks l

Outside temporary tanks l

l 4'.

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

K. Didsel Fuel Oil Testing Program 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 i

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.

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TS.6.0-10 l

L. Technical Specifications Bases Control Pro &IAM This program provides a means for processing changes to the Bases of these Technical Specifications.

1. Changes to the Bases or 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 change in the Technical Specifications incorporated in a.

the license: or

b. a change to the USAR or Bases that involves an unrevi&wed 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.a and b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with USAR updates.

M. Contain.mmLthkage Rate Testing Procram 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 Program," 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-

)

pipes connected to systems that are in the auxiliary building i

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..

Leakage Rate acceptance criteria are:

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l 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 A

.A.

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TS.6.0-11 l

M. Containment Leakage Rate Testing Program (continued) are s 0.60 L for all components subject to Type B 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' l

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

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The provisions of 4.0.B (except that the allowed survei,11ance l

intervals are defined by the Containment Leakage Rate. Testing Program) are applicable to the Containment Leakage Rate Testing Program.

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6.6 Reporting Requirements I

The following reports shall be submitted in accordance with 10CFR50.4

)

I A. Occupational Exeggure Reoort

]

l 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), waste processing, and refueling. This tabulation supplements the requirements of 10CFR20.2206. The dose l

l assignments to various duty functions may be estimated based on 1

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 j

major work functions. This report shall be submitted by April 30 of j

each year.

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B. Annual Radiological Environmental Monitoring Reoort 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 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 j

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.

.s

• TS 6.0 13 l

C. Radioactive Effluent Reoort 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 i

from the plant. The material provided shall be consistent with the objectives outlined in the ODCM and in conformance with 10CFR50.36a and 10CFR50, Appendix I, Section IV.B.1.

D. Monthlv-Ocerating Reports l

Routine. reports of operating statistics and shutdown experience.

l including documentation of all 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.

l E. Core coeratinn Limits Report (COLR)

I

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

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"'l), Nuclear Enthalpy Rise Hot Channel Factcr Limit (Fa ""' ). PFDH, K (Z) and V(Z) n (Specifications 3.10.B.1, 3.10.B.2 and 3.10.B.3)

b. Axial Flux Difference Limits and Target Band l

(Specifications 3.10.B.4 through 3.10.B.9)

]

c.

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

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 PI Units"(latest approved version) 1

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lo-t TS.6.0-14 l

E. Ljre Operating Limits Reoort (COLR)

(continued) l WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology",

July, 1985 i

WCAP-10054-P-A, " Westinghouse Small Break ECCS Evaluation Model l

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 XN-NF-77-57 (A), XN-NF-77-57 Supplement 1 (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 ZIRL0nc Cladding Options". April 1993 (approved by NRC SE dated November l

26, 1993)

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

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 j

limits) of the safety analysis are met.

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

l-F. Pressure and Temperature Limit Report

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 l

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.

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3.3.A.3, 3.3.A.4.

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

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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) l

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TS.6.0-15 l

F. Ergagure and Temperature Limit Report (continued)

}

l 3 The PTLR shall be provided to the NRC upon issuance for each

(

reactor vessel fluence period and for any revision or supplement i

thereto.

Changes to the curves, setpoints, or parameters in the 3

f PTLR resulting from new or additional analysis of beltline material properties will be submitted to the NRC prior to issuance of an updated PTLR.

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TS.6.0-16 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 j

10CFR20, in which the intensity of radiation is greater than 100

)

mrem /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 mrem /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 1

area have been established and personnel are aware of them.

3. An individual qualified in radiation protection procedures with a radiation dose rate monitoring device, who is responsible for providing positive control over the activities within the area and 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 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.

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TS.6.0-17 6.7 High Radiation Area (continued) j C. For individual high radiation areas with radiation levels of greater than 1000 mrem /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 j

individual area, that individual area shall be barricaded and conspicuously posted, and a flashing light shall be activated as a warning device.

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m B.3.1-8 3.1 REACTOR COOLANT SYSTEM Haann continued D.

Maximum Coolant Activity The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the SITE BOUNDARY will not exceed an appropriately small fraction of Part 100 limits following a steam

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

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

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

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.

f Next page is B.3.1-10.

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t 4-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.

l The design basis loss-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 l

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

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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 l

containment leak rate of 0.25% per day at peak accident pressure, the l

offsite doses are about the same as those initially calculated for i

higher primary containment leakage and lower secondary containment in-leakage (Reference 6).

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