Text

Proposed Tech Specs for Conformance of Administrative Controls Section 6 to Guidance of STS
	ML20095J795
Person / Time
Site:	Prairie Island
Issue date:	12/14/1995
From:	; NORTHERN STATES POWER CO.
To:	;
Shared Package
ML20095J791	List: ML20095J788; ML20095J792; ML20095J795;
References
	NUDOCS 9512270401
	Download: ML20095J795 (109)
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{{#Wiki_filter:~. LICENSE AMENDMENT REQUEST DATED December 14, 1995 Conformance of Administrative Controls Section 6 -to the Guidance of Standard Technical Soecifications EXHIBIT B Appendix A, Technical Specification Pages Marked Up Pages TS-il TS-v TS-viii through TS-xiii TS.3.1-10 TS.3.1-ll Table TS.4.1-2B (Page 1) TS.4.4-4 TS.4.6-1 TS.5.1-1 TS.S.1 2 TS.6.1-1 through TS.6.1 4 Table TS 6.1-1 TS.6.2-1 through TS.6.2-7 TS.6.3-1 TS 6.4-1 TS.6.5-1 through TS.6.5-4 TS.6,0-7 (new) TS.6.0-8 (new) TS.6.6-1 TS.6.6-2 TS.6.7-1 through TS.6.7-7 TS 6.0-13 (new) B.3.1-9 B.4.4-2 9512270401 951214 ~~ PDR ADOCK 05000282 PDR ,P. ~-.

I l TS-li " E'? 91 10/27/SS 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 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 Coulant 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. Diliiind!".crirer " ::ter C :1:nt Oxyg:n, 51: ride "is O I$2 rid: Cen :ntretier TS.3.1--11 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 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

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,.m ua,Ja a. 4 %4 .A-.sa AA-.--asi, ,4,., A 4 m: TS-v "T! 1^2 12/17/92 TABLE OF CONTENTS (Continued) TS SECTION TITLE PAGE 1 4.0 SURVEILLANCE REQUIREMENTS TS.4.0 1 4.1 Operational Safety Review TS.4.1-1 4.2 Inservice Inspection and Testing of Lumps and Valves Requirements TS.4.2-1 j A. Inspection Requirements TS.4.2-1 B. Corrective Measures TS.4.2-2 i C. Records TS.4.2-3 4.3 Primary Coolant System Pressure Isolation Valves TS.4,3-1 j 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 i C. Containment Vacuum Breakers TS.4.4-4 D. gg)[ggl["rri'"r1 Mrst "r-rel Syrter TS.^ ^ ^ E. Containment Isolation Valves TS.4.4-5 F. Post Accident Containment Ventilation System TS.4.4-5 C. 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 1$.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 i

2. Containment Fan Motors TS.4.5-3

3. Valves TS.4.5-3 1

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 i C. Steam Exclusion System TS.4.8-2 4.9 Reactivity Anomalies TS.4.9-1 4.10 Radiation Environmental Monitoring Program TS.4.10-1 A. Sample Collection and Analysis TS.4.10-1 B. Land Use Census TS.4.10-2 C. Interlaboratory Comparison Program TS.4.10-2 4.11 Radioactive Source Leakage Test TS.4.ll-1 a u

TS-viii 101 .C,/ 0. g D U1.f Of TABLE OF CONTENTS (Continued) TS SECTION TITLE PAGE 6.0 ADMINISTRATIVE CONTROLS TS.6,01-1 6.1 Ruijfdsiipilipypr;--inctien TS.691-1 +~ gas.tN,;t: idea' : fi:e 2nd.^ udit TS.6.0E2-1 Dr

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.~. - -.- -.. TS-$x RE'1 9': 3/20/91 TABLE OF CONTENTS (continued) TS BASES SECTION TITLE PAGE 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit, Reactor Core B.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure B.2.2-1 2.3 Limiting Safety System Settings, Protective B.2.3-1 Instrumentetion 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability _B.3.0-1 3.1 Reactor Coolant System B.3.1-1 A. Operational Components -B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 E. De;1sted!Mau4 mum Erreter C:clent Oxy;; n. Chlerid: I As?^FIE$ rid: "eneentretier S.3.1 8 F. Isothermal Temperature coefficient (ITC) B.3.1-9 3.2 Chemical and Volume Control System B.3.2-1 3.3 Engineered Safety Features B.3.3-1 3.4 Steam and Power Conversion Systems B.3.4-1 3.5 Instrumentation System B.3.5-1 3.6 Containment System B.3.6-1 3.7 Auxiliary Electrical System B.3.7-1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Radioactive Effluents B.3.9-1 A. Liquid Effluents B.3.9-1 1 B. Gaseous Effluents B.3.9-2 L C. Solid Radioactive Waste B.3.9-4 D. Dose From All Uranium Fuel Cycle Sources B.3.9-5 E. & F. Effluent Monitoring Instrumentation B.3.9-5 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 l C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 E. Rod Misalignment Limitation B.3.10 9 F. Inoperable Rod Position Indicator Channels B.3.10-9 G. Control Rod Operability Limitations B.3.10-9 H. Rod Drop Time B.3.10-10 i I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.11-1 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13-1 l l 3.14 Deleted l 3.15 Event Monitoring Instrumentation B.3.15-1 1 1

TS-x4 Er! 99 7/9/92 TABLE OF CONTENTS (continued) TS BASES SECTION TITLE PAGE 4.0 BASES FOR SURVEILIANCE REQUIREMENTS 4.1 Operational Safety Review-B.4.1-1 4.2 Inservice Inspection and Testing of Pumps B.4.2-1 and Valves Requirements 4.3 Primary Coolant System Pressure Isolation B.4.3-1 Valves 4.4 Containment System Tests B.4.4-1 4.5 Engineered Safety Features B.4.5-1 4.6 Periodic Testing of Emergency Power Systems B.4.6-1 4.7 Main Steam Isolation Valves B.4.7-1 4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9-1 4.10 Radiation Environmental Monitoring Program B.4.10-1 A. Sample Collection and Analysis B.4.10-1 B. Land Use Census B.4.10-1 C. Interlaboratory Comparison Program B.4.10-1 4.11 Radioactive Source Leakage Test B.4.11-1 4.12 Steam Generator Tube Surveillance B.4.12-1 4.13 Snubbers B.4.13-1 4.14 Control Room Air Treatment System Tests B.4.14-1 4.15 Spent Fuel Fool Special Ventilation System B.4.15-1 4.16 Fire Detection and Protection Systems B.4.16 1 4.17 Radioactive Effluents Surveillance B.4.17-1 4.18 Reactor Coolant Vent System Paths B.4.18-1 4.19 Auxiliary Building Crane Lifting Devices B.4.19-1 l l . e

TS-x14 RT! 111 S/10/94 TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE l1 Operational Modes 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 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Gaseous Effluent Monitoring instrumentation 3.15-1 Event Monitoring instrumentation - Process & Containment 3.15-2 Event Monitoring instrumentation - Radiation 4.1-1A Reactor Trip System Instrumentation Surveillance Requirements 4.1.lB Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements 4.1-lC Miscellaneous Instrumentation 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 4.10-1 Radiation Environmental Monitoring Program (REMP) Sample Collection and Analysis 4.10-2 RFMP - Maximum Values for the Lower Limits of Detection 4.10-3 RFMP - Reporting Levels for Radioactivity Concentrations in Environmental Samples 4.12-1 Steam Generator Tube Inspection 4.13 1 Snubber Visual Inspection Interval 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation { Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Monitoring instrumentation Surveillance Requirements 4.17-3 Radioactive Liquid Waste Sampling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program l 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating l Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in f Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) 6.1-1 Minimum Shift Cree Cemperition

l TS-xii4 i RT! 109 9/3/93 APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop Operation 3.1-1 Unit.1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 DOSE EQUIVALENT I-131 Primary Coolant Specific Activity Limit Versus Percent of RATED THERMAL POWER with the Primary Coolant Specific Activity >l.0 uCi/ gram DOSE EQUIVALENT I-131 3.8-1 Spent Fuel Pool Unrestricted Region Minimum Burnup Requirements 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents 3.10 1 Required Shutdown 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 Region Minimum Burnup Requirements e i

. - ~. -. TS.3.1-10 Er! 91 10/27/99 3.1.D. MAXIMUM COOIANT 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. 4 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 hours 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 hours. b. With the specific activity of the primary coolant greater than 100/E microcurie per gram, the reactor shall be shutdown and reactor coolant system average temperature cooled to below 500'F within 6 hours. 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 EQUIVALINT 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. ^rnurl ::p:: ting :;uirrrrnt: :: idertified ir 5.' A.1.2. ^

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Tablo TS.4.1-2B (Page 1 of 2) n eu si, o n n ior. 1 TABLE TS.4.1-2B MINIMUM F1tEOUENCIES FOR SAMPLING TESTS l TEST FREOUENCY 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) (w..en at power) 4. RCS Isotopic Analysis for Iodine a) Once per 4 hours, whenever Including I-131, I-133, and I-135 the specific activity ex-caeds 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 hours fo11owin5 thermal POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour period ( above hot shutdown) 5. RCS Radiochemistry (2) Monthly i 6. RCS Tritium Activity Weekly 7. _ Din _Is.isE CS Cheri:try (Cl*g, C2) 5/'..':29 J 8. RCS Boron Concentration *(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 Concentration Monthly GHO

13. Spent Fuel Pit Boron Concentration Monthly / Weekly 4

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a TS 4.4-4 REV 115 3/9/95 1 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. l c. Halogenated hydrocarbon testing shall be performed i af ter each complete or partial replacement of a char-l coal adsorber bank or after any structural maintanance l on the system housing that could affect the charcoal l adsorber bank bypass leakage. 4 I d. Each circuit shall be operated with the heaters on at l least 10 hours every month. 5. Perform an air distribution test on the HEPA filter bank j after any maintenance or testing that could affect the air i 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 1204. i l C. Containment Vacuum Breakers I The air-operated valve in each vent line shall be tested at i quarterly intervals to demonstrate that a simulated contain-mant vacuum of 0.5 psi will open the valve and a simulated ll accident signal will close the valve. The check valves as well as the butterfly valves will be leak-tested during each refueling shutdown in accordance with the requirements of Speci-e fication 4.4.A.2. ) D. Er:# drIl ErrI P r_^re. S*.*:IO 1. Thr : ; rti n: ef the recidt:1 hrst rrrrrel cycter enternel te the 1 1stier velre: et th: :: tcinrent, ch:11 he hydre-etetically te ted fer le ' g: durin; :::F refueling chutdrer 4 i ) I 2. VI uel in:p:: tier ch:11 he rr'r f:r errrreive 1erks;: fr = i --_7------ -- ---,....__---..i_....,_...i 3 _u___ .t.. ,.t. -_-g-

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Erprire chril be red: : :: uired te reinteir 1rr' ;r rithin ^ the cer:;t-- : criteri:r in S;::ificetier ^ ^.D.? 5. If reprir: ::: n:t :: ;1 ted within ' dry:, the rereter ch:11 i he rhut d:rn r ' d:;rrrrurired until rr;rir: cre effected end the erre;*-- criterier ir 3. ther: 1: ::tiefied. 4 3

TS.4.6-1 REV 112 1/5/95 4.6 PERIODIC TESTING OF EMERGENCY POWER SYSTEM Aeolicability Applies to periodic testing and surveillance requirements of the emergency power system. 4 Obiective 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: Verify the fuel luvel in the day tank, a. b. Verify the fuel level in the fuel storage tank, c. DsistidVerify thet : :---1: ef dier:1 fuel fr:: the fuel eterage EEEE"I$ rithir th ::::pke'le limite specified ir T ble 1 ef ASTM D975 " "her ch:: Erd fer ficerrity, reter, 2nd : dirent. l d. Verify the fuel transfer pump can be started and transfers fuel from the storage system to the day tank. Verify the diesel generator can start and gradually accelerate. Verify l e. the generator voltage and frequency can be adjusted to 4160 i 420 volts and 60 i 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, loading and shutdown procedures where possible. i

- ~. TS.S.1-1 n__m_re n_ 1_1_,/1_1,/ o r_ 5.0 DESIGN FEATURES 5.1 SITE The Prairie Island Nuclear Generating Plant is located on property owned j by Northern States Power (NSP) Company at a site on the west bank of the j 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 nearest population center of 25,000 or more people is South Saint Paul. These site characteristics l comply with definitions in 10CFR100 (Reference 1). i 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 i .to prevent residential constraction on this land for the life of the plant (Reference 2). i These specifications use atmospheric diffusion factors based on the NRC staff evaluations. Its evaluation of accidental airborne releases is i 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 relative concentration of 1.5 x 10-s seconds per cubic meter (Reference 3). i The flood of record in 1965 produced a water surface elevation of +688 l feet MSL at the site. The calculated probable maximum flood (PMF) level is +703.6 feet mean sea level ($SL), and the estimated wave runup could 4 reach +706.7 feet MSL. (See Section 2.4.2 of this report.) Plant grade l 1evel is +695 feet MSL. Flood protection structures have been provided. The two turbine support l 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 1 flood wall, or into the various buildings (including the separate screen j 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 scre.anhouse are designed to operate up to 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. i The T :Snicel S;::ifientien 6.5 A '.r:;uire: := -- r;rney precedure thet rill 22::::itzt: ;ll t :h !f:r: f: fl::f rit:r 1:r:12 cirr: tf?2 f:: 1 Mst et t' plert cite. The :::r;:rry pr r: fur util 1::ure the a i

TS.5.1-2 -- i n,19 7,/ 0 0 --.t OT D et pre;:r er::tien f fleed prete:tien perrl: nd 22:ur: en erderly chutdrur cf the pl nt r ' pretertien f refety rel ted freilitier. This prrredure uill pr: vide fer pre;rerriz: cetier 12f:1: te precent the prrribility cf j unrefe plent :per:tien 2nd rill include requirrrent: fer peri:dic l inepretier ef firrd pretectier nrrrurer. 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. in cr r;rney precedure vill be pr:pered ir errerden:: eith Sp::ific: tier 6.5.A ' te defin: retien required fer errth;"r'er, includin; plent chutdrrn 2nd inepr: tier if en

perstienel herie errth;- '- 1: recrured et the cite.

1 l l 1 s 1 1 1 1 References 1 1. USAR, Section 2.2.1 2. USAR, Section 3.4.5 3. SER, Sections 2.3.4 and 2.3.5 l

TS.6.pt-1 E91 105 5/':/44 6.0 ADMINISTRATIVE CONTROLS 6.1 e r ; r r i e s t i r rN_sp_6u.s.i.t.ii l_it.y. A. The Dilant Nijanager shall be responsible for overall unit enfe operation and shall h:f: rentrel :fer there enrite :tifitier n rrr ery fer erf

rratier end = intenener rf the plent. 5: Plant Mene;er rhril delegate in writing the succession to this responsibility during his absence. E: Plent Frnr; r her the re:perribility fer the Fire Prete: tier Pre;rr.

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mrla.g o BT"?Ttie~ahiftr srpsrvis~oY"(SS)~^shh11 Tbs?fdsp6 ikib1'e~fsf7thiiP66Etrbl~r66in] a ~bommand function. During any 'absense"of' the ss frent che ;oomtrol room] while the > unit is,in MODR 1.,2,,3,'< or 4; ^en individual with an active

senior remotor. operator.,(SRO) 11' onse shall be designated:to assume'ths c

' oni;rol room command fuhotion, During'any' absence,of'the 88 fros'che ccone. col' room whilerthe unit-is in NODR S or^4, 'an individual vich'an 2 'aci.id SRO license' or reactor' operator' licatiseghall, tie' designsteltd s .as_ sun,e,.t_ h_e cont _ro.l. _r,oom_c; em_ arid,f_unc_ti6n.' 4 l

TS.6,01-2 R"! l'd5 5/4/93 1 .612](Orga4(attios AS Onsite and Offsite Greanizations Onsite and offsite organizations shall be established for plant operation and corporate managementMijjig}Qi,1]. The onsite and offsite organizations shall include the positions re pencible for activities affeetingplant-safetyMfith{dM1fhdijnsR@y{. l.-Linesofauthority,responsibilityandcommunicationshallbeggiasj sn;dlestablishedithyoughoug 2nd defined fer the highest management levelsy through intermediate levels) te-and 4aeludtag all operating organization positions. These relationships shall be documented and updated, as appropriate, in the fer= cf organization charts, functional descriptions of departmental responsibilities and 4 relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. These requirements 71nc10dii$ the plant' specific titles"of' those"personne1# fulfilling lthe responsibilities of the_ positions, delineated in,the,se, Technical Specifications, shall be documented in th: Oper:tirn:1 Q"clity ^ccurence Plcr er the Updated Safety Analysis Report. 2.[3.E : ch:11 be en indivital renegenent p::iti:r M 4Pglant Maianager) ir the encite ergenizetier having rerpencibility iiihill reports toi.t.h.aTch.ro.o.ii#at. WUvic. e.I6.rissw.m.s..h.w.essp sc.i.gfiWd.. Tis M3M. h responsible for overall un44-safe operation ofE6hpg.i6.I2.sr:j.;.;g.w c .e m..- .us m=... -, id Fsu ue apla,xnt and whe- .mw z: j mm www w 1 stiall have control over those onsite EspijijQM necessary ~ for safe operation and maintenance of the plant. 33.Ccorp"'6 rate 0Ecre ch:11 he en indivital enecutive preitier (V# ice Ppra ident Nuclear Cen retien) ir the effeite crganisatier hevIng ~ phyll {havpcorporate responsibility for overall plant nuclear safety and shall take any measures needed to ensure acceptable performance 4 of the staff in operating, c:aintaining and providing technical support to the plant to ensure nuclear safety.

4. The individuals who train the operating staffy 2nd th::: rhe carry out health physicsjfodstfoinii end-quality assurance functions may report to the appropriate onsite manager; however, psisjj;isdifiydiM]

they shall have sufficient organizational freedom to ensure their independence from operating pressures. BC. Plant Staff + The [ plan tZ a ta f f ' o r ga niz a ti t o d])ha11]id416dej}hsjfoJ1 ofib'gj

1. An operator to perform non;1icensed dutie's'shall'be assigned' to"escli 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 Also; if one' unit is in MODE 1, '2]

3, or 4 and the other unit is in MODE 5 or 6, as a minimum the ong site staffing shall include two senior, reactor operators (SRO) and _twd lice.n. sed reactor operators (RO)j a Each er h ty chift ch:11 be carpered f et leret the ninin= chif t creu cerperitier cheur er Table TS.6.1 -1.

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4. An individual qualified in radiation protection procedures shall be b~sWsEEEWE6ElHisif

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h. ,m- .+ ...w P_rocedures ?wh...ich.s. s - me. n_t. f.a._n L.._C_a.p _o.v_e ? _ro.,5r._am_ e 4 ~ m - ~~ - s _. J I ^11 refueling eperatien ch:11 h: directly cuperfired by : licerned S-icr Reacter Oper:ter er : Senier Reacter Oper ter 2 ne ether reneurrent rerpen: Limited te Fuc1 Mend 11ng "._.he h: s t_ e, s._. < _ ____a._.._i~..,. ._n..,__ ___._.4_

6. The cener:1 Superintendent F1:nt 06perations MpiQi@sijiijs{u(

opsdEi6ns?sishsgsdshall be fer=crl" licen :d er hold current an.4 J w SR0 license er ci=iler typ: pl:nt. ^t lecct ene :: hcr f plant renegr r-t helding : curr:nt Senier R:::ter Oper:ter licene: ch:11 b: cecigned te the pl nt oper:ti:n grcup er : 1:ng tern herie (cpprenirately tue y:cre). S ie indicidurl _t.. e _r._. _u.._._ m..__ u__ ___4___a._ __.__.._..a ____g.... 7 s he shift technica1' advisor (STA) ^shall' provide'" advisory"tschsical T 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,;3g s

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) M:rthr= Stetre P=:r c n;rny ; r r=rl :: en rutride fire prettetier j .._..3 _.,._.____-e_... ._._..__._sa_._.,._. _., e _e _s._ 2_ _e _n._.. _ _ _ - _ _... a e. t_,.. ..a _a. .n_. _s._,__.__.._ _. -. _ cerruitent rh:11 he perf: rro et intere:1: ne grerter then three yr re. [ Current Prairie Island Technical Specifications Paragraph 6.1.D was relocated to provide the substance of new Section 6.3 as follows:] Each member of the plant staff shall meet or exceed the minimum 9uali f ic a t io ns o f =mer jjp4XMW%sva.Aum' AWW<u.MP&g.va*X_ hew.wMMg-h..s%he.*<g. Aww/wwenwe.4k. gggjg* isw.:MF.w,,w.wdewa N19.1 1971 fer err;rr:ble p :itiens, except for (1) the Cencr:1 _____s_ ___._._._.....; ._.t._ .t. i _c._...__...______.._ n._.2_4_.._..__ n._.._._._._._4_._....t. ._2 7 _.. qualifientirn ef ".:;uletery cuid: 1.S, Septr 'er 1975, 2nd '@ th: Shift Manages-ps T_i3._bi_$iiiEjili.iFTi_sR.__d,._iiE.Te$, _siffffiliisW6.m. KT_sh._ifF.L.i.i_dhhid.w 1TWd.ylii.6, f iiW, lw s. 4 wm_.w m_m m who-shall have a bachelors degree or equivalent in a scientific or engineering discipline with specific training in plant design, and d response and analysis of the plant for transients and accidents, and -(-3-)- J2); the Centret Superintend =t Plant @perations jisMgwho shall meet the requirements of ANSI N18.1-1971, except that NRC license .w7 T6*H i requirements are as sPacified in Specification kT1..m.5,m Aa .1." 5: irrei".ir." pr:fr"." ch:11 be ndr the directi: :$ e der #_Q."IOd Ernber :[ Mertherr St:te: P:rer --. ;:nrnt. i 1 + 4 ,m

i TS.6.^-1 ""! 105 5/':/93 5.^ SAFE"' LIMIT '!IOLtTIOF 1 [ Deletion of this section was proposed in License Amendment Request, " Pressurizer Safety Valves and Main Steam Safety Valves Lift Setting Tolerance Change and Safety Limit curve Changes", dated May 4, 1995 )

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m_ _e. e_. _C _, D.EM.F i_ n. C -,//,/ 01_ C [6.5.B.4 is proposed to be relocated to new Specification Section 6.5.C below entitled, " Post Accident Sampling", with changes as marked up.) C. F.cintenene: 2nd T :t The fellering reintensne end t t pr::: dure: rill be der:1: ped t: enticfy

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2. "reventiv: er : rrective reintenene: ef pl:nt equiprent and cyster th t ceuld have en effect er nuclear ::fety.

-..a_.._....._4_.,._ _4_...___-___ _r 4__-._._.._-.._._4__. ._t.._._ -_..,a-_- i. c_ s_ e t_._.._._ _ _ 7 .g _r _ _ _ ._t__ . _. _. _, _ -. _ _ _ _r -._,.. _ r._u___,__._. -_s___,.-. _C. _e,___s,-. .___-._s__. _ _r -,._a_____._ _r _. _. _, - _ _ - _ >._u___- _-._. _ _.__-_a...__._. a__-,__ _t_____. M. D.-____- P___.._1_ D_..,-o___. /, D. P_ D. \\ [ Deletion of paragraph 6.5.D, " Process Control Program (PCP)", was proposed in License Amendment Request, " Radiological Effluent Technical Specifications Conformance to Standard Technical Specifications and Ceneric Letter 89-01", dated July 17, 1995.)

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[ Revision of paragraph 6.5.E, "Offsite Dose Calculation Manual (ODCM)", was proposed in License Amendment Request, " Radiological Effluent Technical Specifications Conformance to Standard Technical Specifications and Generic Letter 89-01", dated July 17, 1995. This submittal proposes to relocate those proposed requirements to a new Specification Section 6.5.A entitled, "Offsite Dose Calculation Manual (ODCM)"] gpfmisWC6ET^mM35G~fEEFotIEsi^dsTC6KED^mst ssif f6GY hell 5: M 1-- nted to M..a srsu!::2-e 3vTEl..~i%A-program pis.vi^dihy.mw.W,f~sys~tems outside containment that reuld er ~~~ w ~ leak ~ age from th,uuuu,a n53r.Lpo.. rt.nm >n.wmnnm w - io aro contain highly radioactive fluids during a serious transient or accident to Tsihisisjas 1ow as praceiea1 level:. TEsYufsEisiYiE61TidsTp6fii1~6Hi?ufMiiWisl sif NNN N D IfE5YMM M dMEDA M EAk$$kih khsNhEN$$kh yj Q N N proj include the following: Q.Frevicien: cerrblishing pJreventive maintenance and periodic visual inspection requirements, and gbvIntegrated leak test requirements for each system at : fre ;uency net te creced refueling cycle intervalsT3_EE_N_ss.. I ? pr:gr = 2rr:pt +1e te the cr._-! !:r re: d rrihed in letter: fren L.O. Meyer, MSP, tr Directer ef Muricer Rer ter R:;uletten, 2-ted Derr9 er 31, 19'? "Le ren: Errrned inpirrrmtetier" :no March 13, 1990, "1/1/20 L:: en: Errrned Inglerentatier Additien 1 Inferr tier" j C4. P6Et"KEEidsEFSmliFi T_H._UA programW.f.6f.,f.,~ds.sT,_66_E._jif.611."_W._sf., th:11 he inplenented which wi14 ensure x. ~ ase the capability to obtain and analyze reactor coolant, radioactive ,g._.,._g., 4e e nes and particulates in plant gaseous effluentsy and containment atmosphere samples under accident conditions. The program shall include the following: [4.Trainingofpersonnelf 2b. Procedures for sampling and analysis E Ehd r }e. Provisions for maintenance of sampling and analysis equipment. DMREd f6'iiid EIVF tff1H&HFC6nEF61T?Pr6Efss u [A new paragraph 6.5.H. " Radioactive Effluent Controls Program", was proposed in License Amendment Request, " Radiological Effluent Technical Specifications Conformance to Standard Technical Specifications and Generic Letter 89-01", dated July 17, 1995. This submittal proposes to relocate those proposed requirements to a new Specification Section 6.5.D entitled, " Radioactive Effluent Controls Program"] Q CoissEnsE E* CWsifE*BFTrsssishB1 isitt TK. GT.+ Ys 6..'gfMjip=f6V,idssr ^6,Ktf61s7E6Tt.f.xe xnen~SARE.,5&E.,ti c iskYthiTU 76fElIETh nc n +. n, .=un n .s aw .uama. u u 0CCurrencasitOfensure3hatfcomponentspre[Mai,ntalDedj ~ = te .- esi a le t d..s..m.~I,n.m.i<m,,%.s o

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pre fi = 2: rind :;::d, rino dir::ti:n, 2nd :tnr ;f.::ri: ility et th: r:;u::t :f .t. c____,__e__ 5.' ^. 5.gCore Operating Limits Report T N M Y [a.Coreoperatinglimitsshallbeestablished[Hugto]and d::rrnt:d in the CORE OP"L^.TINC LI".!TS RE?.ET'he#::: each reload cycle MOMiMany remainin ~itid67past.ofareloadcycle] gigh{(orRjQMffgjgpli"^f6F"Ehe following: RPT p.HeatFluxHotChannelFactorLimit(Fo ), Nuclear Enthalpy RTP Rise Hot Channel Factor Limit (FAs ), PFDH, K(Z) and V(Z) (Specifications 3.10.B.1, 3.10.B.2 and 3.10.B.3) lia. Axial Flux Difference Limits and Target Band (Specifications 3.10.B.4 through 3.10 B.9) E3. Shutdown and Control Bank Insertion Limits (Specification 3.10.D) i$4. Reactor Coolant System Flow Limit (Specification 3.10.J) 2b.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) 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

v..e. c.,.

e_ n _eu. s_ n. c .c,u.,m.., 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 Phase II", May, 1981 WCAP-13677, "10 CFR 50.46 Evaluation Model Report: W-COBRA / TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRL0ni Cladding Options", April 1993 (approved by NRC SE dated November 26, 1993). i NSPNAD-93003-A, " Transient Power Distribution Methodology", (latest approved version) J'e.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 analysis lim ts) of the safety anal.y..w,M.w,w.1.iii.u.G.v., shutdown mar in,--end transientFM.v.w.l ji 1.i and accident iili v w 49 w ysias are met. 44.The CORE-OPE"^.T!"C -IMMS-REPORT, including any mid-cycle r to the NRC " q qspeupplied upon shall be p revisions or supplements therete, issuancer for each reload cycle uith ::pire t the Regional.^drinistr ter end Recident Inspecter. n_. D..P D. n.D..P.A..n t e_ U1.FUM.T. O ..,___......_t..__,_,_ t_ __.___n..__..c _ _ o..e_ nn.o.m..i.n, e c.u. r..M..m..e.

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::uctiel Written repert; fer the felleuing iten ch:11 b: cubritted to the apprepeiet: F"C Regi:n:1 ^.' inistrat:r:

2. Envir:n= nt:1 event: th:t indi :t: :: ::uld re: ult ir : significent envir:n :nt:1 irg :t ::: :lly rel: :d t: pl:nt :p:::tien The fe44 ewing ::: ex-ple :

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1. B.4.4-2 _ " " I _8,'S,'PS ~ 4.4 CONTAINNENT SYSTEN TESTS i Amasa continued Several penetrations of the containment vessel and the shield building j 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 i (Reference 5). Such leakage is estimated not to exceed.0254 per day. A special zone-of the auxiliary building has minimum-leakage construc-1 tion and controlled access, and is designated as a special ventilation zone where such leakage would be collected by either of two redundant 3 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-coolant accident was initially evaluated by I the AEC staff (Reference 3) assuming primary containment leak rate of 0.5% per day at the peak accident pressure. Another conservative assumption in j the calculation is that primary containment leakage directly to the ABSVZ i is 0.1% per day and leakage directly to the environs is 0.014 per day. The resulting two-hour doses at the nearest SITE BOUNDARY and 30-day doses at the low population zone radius of 14 miles are less than guidelines } presented in 10CFR100. Initial leakage testing of the shield building and the ABSV resulted j in a greater inleakage than the design basis. The staff has reevaluated j doses for these higher inleakage rates and found that for a primary containment leak rate of 0.254 per day at peak accident pres-j sure, the offsite doses are about the same as those initially calculated j for higher primary containment leakage and lower secondary containment in-leakage (Reference 6). e_ u_._ n. _ f 2.. _ 3_ u..._ n-__.., c....-_-

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. - -. _ =. 4 LICENSE AMENDMENT REQUEST DATED December 14, 1995 Conformance of Administrative Controls Section 6 to the Cuidance of Standard Technical Soecificatiens i I EXHIBIT C i j Appendix A, Technic'al Specification Pages Revised Pages i i TS-ii TS-v TS-viii through TS-xii TS.3.1-10 i Table TS.4.1-2B (Page 1) TS.4.4-4 1 TS.4.6-1 i TS.5.1-1 TS.5.1-2 TS.6.0-1 i TS.6,0-2 TS.6,0-3 i TS.6,0-4 TS.6.0-5 4 TS.6.0-6 TS.6.0-7 TS.6.0-8 TS.6,0-9 TS.6.0 10 TS.6,0-11 TS.6.0-12 TS.6,0-13 TS.6,0-14 TS.6,0-15 B.4.4-2 1 l l \\

-. _ _ _._._ _ m. __. __..._. __ h TS-li TABLE OF CONTENTS (Continued) l l TS SECTION TITLE PACE 3. LIMITINC 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 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 Cenerator 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.3.1-9 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 1 3.4 Steam and Power Conversion System TS.3.4-1 A. Steam Cenerator Safety and Power Operated Relief Valves TS.3.4-1 i B. Auxiliary Feedwater System TS.3.4-1 C. Steam Exclusion System TS.3.4-3 ) D. Radiochemistry TS.3.4-3 l 3.5 Instrumentation System TS.3.5 1 1

TS-v TABLE OF CONTENTS (Continued) 1 l TS SECTION TITLE PAGE l 4.0 SURVEILIANCE REQUIREMENTS 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 f 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 D. Deleted E. Containment Isolation Valves TS.4.4-5 F. Post Accident Containment Ventilation System TS.4.4-5 C. 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 l A. System Tests TS.4.5-1 l

1. Safety Injection System TS.4.5-1 l

2. Containment Spray System TS.4.5-1 l

3. Containment Fan Coolers TS.4.5-2 i

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 i

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 Radiation Environmental Monitoring Program TS.4.10-1 A. Sample Collection and Analysis TS.4.10-1 B. Land Use Census TS.4.10-2

c. Interlaboratory Comparison Program TS.4.10-2 4.11 Radioactive Source Leakage Test TS.4.ll-1 i

l

~. - -.. - -.... TS-viii TABLE OF CONTENTS (Continued) TS SECTION TITLE PACE 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.6,0-5 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.6,0-8 F. (Rese rved) TS.6,0-8 G. (Reserved) TS.6.0-8 H. (Rese rved) 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 6.6 Reporting Requirements TS.6,0-ll A. Occupational Exposure Report TS.6,0-ll B. Annual Radiological Environmental Monitoring ReportTS.6.0-ll C. Radioactive Effluent Report TS.6,0-11 D. Monthly Operating Report TS.6,0-12 l E. Core Operating Limits Report (COLR) TS.6,0-12 6.7 High Radiation Area TS.6.0-14 l 1

- ~. _ _.... l TS-ix TABLE OF CONTENTS (continued) TS BASES SECTION TITLE PAGE 2.0 BASES FOR SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 Safety Limit, Reactor Core B.2.1-1 2.2 Safety Limit, Reactor Coolant System Pressure B.2.2-1 2.3 Limiting Safety System Settings, Protective B.2.3-1 Instrumentation 3.0 BASES FOR LIMITING CONDITIONS FOR OPERATION 3.0 Applicability B.3.0-1 3.1 Reactor Coolant System B.3.1-1 A. Operational Components B.3.1-1 B. Pressure / Temperature Limits B.3.1-4 C. Reactor Coolant System Leakage B.3.1-6 D. Maximum Coolant Activity B.3.1-7 E. Deleted F. Isothermal Temperature Coefficient (ITC) B.3.1-9 3.2 Chemical and Volume Control System B.3.2-1 3.3 Engineered Safety Features B.3.3-1 3.4 Steam and Power Conversion Systems B.3.4-1 3.5 Instrumentation System B.3.5-1 3.6 Containment System B.3.6-1 3.7 Auxiliary Electrical System B.3.7-1 3.8 Refueling and Fuel Handling B.3.8-1 3.9 Radioactive Effluents B.3.9-1 A. Liquid Effluents B.3.9-1 B. Caseous Effluents B.3.9-2 C. Solid Radioactive Waste B.3.9-4 i D. Dose From All Uranium Fuel Cycle Sources B.3.9-5 E. & F. Effluent Monitoring Instrumentation B.3.9-5 3.10 Control Rod and Power Distribution Limits B.3.10-1 A. Shutdown Margin B.3.10-1 B. Power Distribution Control B.3.10-1 C. Quadrant Power Tilt Ratio B.3.10-6 D. Rod Insertion Limits B.3.10-8 i E. Rod Misalignment Limitation B.3.10-9 ] F. Inoperable Rod Position Indicator Channels B.3.10-9 G. Control Rod Operability Limitations B.3.10-9 H. Rod Drop Time B.3.10-10 I. Monitor Inoperability Requirements B.3.10-10 J. DNB Parameters B.3.10-10 3.11 Core Surveillance Instrumentation B.3.11-1 3.12 Snubbers B.3.12-1 3.13 Control Room Air Treatment System B.3.13-1 3.14 Deleted 3.15 Event Monitoring Instrumentation B.3.15-1 t S

TS-x l j l l TABLF OF CONTENTS (continued) TS BASES SECTION TITLE PAGE 4.0 BASES FOR SURVEILIANCE REQUIREMENTS 4.1 Operational Safety Review B.4.1-1 4.2 Inservice Inspection and Testing of Pumps B.4.2-1 and Valves Requirements 4.3 Primary Coolant System Pressure Isolation B.4.3-1 Valves 4.4 Containment System Tests B.4.4-1 4.5 Engineered Safety Features B.4.5-1 4.6 Periodic Testing of Emergency Power Systems B.4.6-1 4.7 Main Steam Isolation Valves B.4.7-1 4.8 Steam and Power Conversion Systems B.4.8-1 4.9 Reactivity Anomalies B.4.9-1 4.10 Radiation Environmental Monitoring Program B.4.10-1 A. Sample Collection and Analysis B.4.10-1 B. Land Use Census B.4.10-1 C. Interlaboratory Comparison Program B.4.10-1 4.11 Radioactive Source Leakage Test B.4.ll-1 4.12 Steam Generator Tube Surveillance B.4.12-1 4.13 Snubbers B.4.13-1 4.14 Control Room Air Treatment System Tests B.4.14-1 4.15 Spent Fuel Pool Special Ventilation System B.4.15-1 4.16 Fire Detection and Protection Systems B.4.16-1 4.17 Radioactive Effluents Surveillance B.4.17-1 4.18 Reactor Coolant Vent System Paths B.4.18-1 4.19 Auxiliary Building Crane Lifting Devices B.4.19-1 0 1

l TS-xi TECHNICAL SPECIFICATIONS LIST OF TABLES TS TABLE TITLE l-1 Operational Modes 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 3.9-1 Radioactive Liquid Effluent Monitoring Instrumentation 3.9-2 Radioactive Caseous Effluent Monitoring instrumentation 3.15-1 Event Monitoring instrumentation - Process & Containment 3.15-2 Event Monitoring instrumentation - Radiation 4.1-1A Reactor Trip System Instrumentation Surveillance Requirements 4.1-1B Engineered Safety Feature Actuation System Instrumentation Surveillance Requirements 4.1-lc Miscellaneous Instrumentation 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 4.10-1 Radiation Environmental Monitoring Program (REMP) Sample Collection and Analysis 4.10 2 RFMP - Maximum Values for the Lower Limits of Detection 4.10-3 RFMP - Reporting Levels for Radioactivity Concentrations in l Environmental Samples 4.12-1 Steam Generator Tube Inspection 1 4.13-1 Snubber Visual Inspection Interval 4.17-1 Radioactive Liquid Effluent Monitoring Instrumentation Surveillance Requirements 4.17-2 Radioactive Gaseous Effluent Monitoring instrumentation Surveillance Requirements 4.17-3 Radioactive Liquid Waste Sampling and Analysis Program 4.17-4 Radioactive Gaseous Waste Sampling and Analysis Program 5.5-1 Anticipated Annual Release of Radioactive Material in Liquid Effluents From Prairie Island Nuclear Generating Plant (Per Unit) 5.5-2 Anticipated Annual Release of Radioactive Nuclides in Gaseous Effluent From Prairie Island Nuclear Generating Plant (Per Unit) ]

... -. ~ S TS-xii l APPENDIX A TECHNICAL SPECIFICATIONS LIST OF FIGURES J TS FIGURE TITLE 2.1-1 Safety Limits, Reactor Core, Thermal and Hydraulic Two Loop 4 Operation 3.1 1 Unit 1 and Unit 2 Reactor Coolant System Heatup Limitations 3.1-2 Unit 1 and Unit 2 Reactor Coolant System Cooldown Limitations 3.1-3 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 Minimum Burnup Requirements 3.9-1 Prairie Island Nuclear Generating Plant Site Boundary for Liquid Effluents 3.9-2 Prairie Island Nuclear Generating Plant Site Boundary for Gaseous Effluents 3.10-1 Required Shutdown Margin Vs Reactor Boron Concentration 4.4-1 Shield Building Design In-Leakage Rate i 5.6-1 Spent Fuel Pool Burned / Fresh Checkerboard Cell Layout 5.6-2 Spent Fuel' Pool Checkerboard Region Minimum Burnup Requirements t i 1

0 1 TS.3.1-10 3.1.D.MAXINUM C001 ANT 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: 1 With the specific activity of the primary coolant greater than a. 1.0 microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours 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 hours, b. With the specific activity of the primary coolant greater than 100/E microcurie per gram, the reactor shall be shutdown and reactor coolant system average temperature cooled to below 500*F within 6 hours. 3. If a reactor is at or above COLD SHUTDOWN, with the specific activity of the primary coolant greater than 1 2 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.

i Tc.ble TS.4.1-28 (Page 1 of 2) l l TABLE TS.4.1-25 MINIMUM FREOUENCIES FOR SAMPLING TESTS ) TEST FREOUENCY 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 hours, 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 betwe,en 2 and 6 hours following thermal POWER change exceeding 15 percent of the RATED THERMAL POWER within a one hour l period ( above hot shutdown) l ( 5. RCS Radiochemistry (2) Monthly 6. RCS Tritium Activity Weekly 7. Deleted l 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 ms)

13. Spent Fuet fit Boron Concentration Monthly / Weekly i

s TS 4.4-4 d 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 i could affect the HEPA bank bypass leakage. Halogenated hydrocarbon testing shall be performed c. after each complete or partial replacement of a char-coal 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 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 rate (1104). The results of the test shall i show th6 air distribution is uniform within 1204. C. Containment Vacuum Breakers 1 j The air operated valve in each vent line shall be tested at quarterly intervals to demonstrate that a simulated contain-mant 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 during each refueling shutdown in accordance with the requirements of Speci-fication 4.4.A.2. ) i i

TS.4.6-1 4.6 PERIODIC TESTING OF EMERGENCY POWER SYSTEM Apollcability Applies to periodic testing and surveillance requirements of the emergency power system. Obiective 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: 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 the day tank. Verify the diesel generator can start and gradually accelerate. Verify e. 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 KU), and operate for at least 60 minutes. This test should be conducted in consideration of the manufacturer's recommendations regarding engine prelube, warm-up, loading and shutdown procedures where possible. i 4

~ i I 4 i TS.5.1-1 ~ i ' 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 1 population zone distance is 1-1/2 miles. The nearest population center of 25,000 or more people is South Saint Paul. These site characteristics comply with definitions in 10CFR100 (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 boundary. Its evaluation of routine releases is based on a relative concentration of 1.5 x 10-5 seconds per cubic meter (Reference 3), s 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 +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, and the two shield buildings have been physically connected by a concrete flood wall, most of the 4 length of which consti*utes 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 separate screen house), are provided with receivers for the erection of flood protection a panels to prevent flood water from reaching safety related facilities. d 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 i 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.

TS.S.1-2 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. 1 } l l l References 1. USAR, Section 2.2.1 2. USAR, Section 3.4.5 3. SER, Sections 2.3.4 and 2.3.5

TS.6.0-1 l 6.0 ADMINISTRATIVE CONTROLS 6.1 Responsibility A. The plant manager shall be responsible for overall unit operation ari shall delegate in writing the succession to this responsibility during his absence. The plant manager or his designes 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.

TS.6,0-2 6.2 Organization A. Onsite and Offsite Ornanizations 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 vice president 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 vice 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 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. Also, if one unit is in MODE 1, 2, 3, or 4 and the other unit is in MODE 5 or 6, as a minimum the on-site staffing shall include two senior reactor operators (SRO) and two licensed reactor operators (RO).

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.

TS.6,0-3

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 hours in order to accommodate unexpected absence of on-duty shif t 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 i

on site when fuel is in a reactor. The position may be vacant for not more than 2 hours, in order to provide for unexpected absence, provided immediate action is taken to fill the required position. j

5. The amount of overtime worked by plant staff members performing safety related functions shall be limited and controlled by procedures which implement an NRC approved program.

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

7. The shift technical advisor (STA) shall p.* -r ' % advisory technical support to the shift supervisor in the areas of thermal hydraulics, reactor engineerin5. 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.

s t 1 i

TS.6,0-4 6.3 Plant Staff Qualifications 1 Each member of the plant staff shall meet or exceed the minimum qualifications of Regulatory Guide 1.8, Rerision 1, September 1975 except for (1) personnel who perform the function of shift technical advisor shall have a bachelors degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and 4 (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. 4 l l I

TS 6,0-5 6.4 Procedures Written procedur.s shall be established, implemented, and maintained covering the following activities: J 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 l requirements of NUREG-0737, Supplement 1, as stated in Generic ' etter 82-33; 1 C. Quality control for effluent and environmental monitoring; D. Fire protection program implementation; and E. All programs specified in Specification 6.5. 4 I s

TS.6,0 6 6.5 Programs and Manuals The following programs shall be established, implemented and maintained. A. Offsite Dose Calculation Manual (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 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:

sufficient information to support the change (s) together with a. y 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 10 CFR 20.1302, 40 CFR190, 10 CFR 50.36a, and 10 CFR 50, 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 i

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 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., 1 month and year) the change was implemented shall be indicated. j B. Primary Coolant Sources Outside Containment This program provides controls to minimize leakage from those portions of systems outside containment that could contain highly i 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: i

1. Preventive maintenance and periodic visual inspection i

requirements, and

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

TS.6.0 7 C. Post Accident Sampline This program provides controls that ensure the capability to obtain and analyze reactor coolant, radioactive gases, and particulates 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, waste 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 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 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;

3. Monitoring, sampling, and analysis of radioactive liquid and 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 and projected dose contributions from radioactive effluents for the current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least monthly;

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

-~ i i TS.6,0-8 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 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 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 B 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 uranien fuel cycle sources, conforming to 40CFR190, 1

E. Component Cyclic or Transient Limit This program provides controls to track the USAR, Section 4.1.4 i cyclic and transient occurrences to ensure that components are maintained within the design limits. s F. (Reserved) G. (Reserved) H. (Reserved) i I. (Reserved) 4 1 a

TS.6,0-9 J. Exolosive Gas and Storare Tank Radioactivity Monitorina Program 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 4

system and a surveillance program to ensure the limits are maintained. Such limits shall be appropriate to the system's design criteria; 1 f

2. A surveillance program to ensure that the quantity of radioactivity contained in each gas storage tank is less than or l

equal to 78,000 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 d

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. Diesel Fuel Oil Testing Program I 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 4 1 Table 1 of ASTM D975-77 when checked for viscosity, water, and sediment. L. Technical Specifications Bases Control Program i 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 1

license; or

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

i r

TS.6,0-10

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 Specification 6.5.L.2 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, l.

4 1 i 4 d 4 a l 4

TS.6.0-ll l 6.6 Reporting Requirements The following reports shall be submitted in accordance with 10CFR50.4 A. Occupational Exposure Report 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 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 Report 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 Cuide 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 ahall also include the following: a summary description of the radiciogical environmental monitoring program; a map of all sampling locations keyed to a table giving distances and directions from one ret.ctor; and the results of licensees participation in the Interlaborat ory Comparison Program defined in the ODCM. C. Radioactive Effluent Report 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 10CER50.36a and 10CFR50, Appendix I, Section IV.B.l. I l

l TS.6,0-12 D. Monthly Operating Reports Routine reports of operating statistics and shutdown experience 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 Operatine Limits Reoort (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:

un

a. Heat Flux Hot Channel Factor Limit (Fn

), Nuclear Enthalpy RTP Rise Hot Channel Factor Limit (FAs ), 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 Evaluation Methods for Application to PI Units" (latest approved version) 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 XN NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II", May, 1981 WCAP-13677, "10 CFR 50.46 Evaluation Model Report: W-COBRA / TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRL0m Cladding Options", April 1993 (approved by NRC SE dated November 26, 1993).

. -. _ -... _ -.. - _. ~..- TS.6,0-13 l l NSPNAD-93003-A, " Transient Power Distribution Methodology", (latest approved version) 1

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 j

shutdown margin, transient analysis limits and accident analysis limits) of the safety analysis are met. l

4. The COLR, including any midcycle revisions or supplements, shall i

be provided upon issuance for each reload cycle to the NRC. 4 1 j 1 i l l i 4 i i i 5 j l i 4 e i i l --.m..,

TS.6.0-14 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 mrem /hr but less than 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 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 or equal to 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 precedures to provide positive exposure control over the activities being performed within the area.

~. _- -. l TS.6.0-15 i 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 j for purposes 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 warning device. i i l T 1 i i l 1 1 i 1 4 1 8 d

. ~. i B.4.4 2-4.4 CONTAINMENT SYSTEN 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 4 leakage being conveyed across the annulua by the penetrations themselves, thus bypassing the function of the Shield Building ventilation System (Reference 5). Such leakage is estimated not to exceed.0254 per day. A special zone of the auxiliary building has minimum-leakage construc-tion and controlled access, and is designated as a special ventilation zone where such leakage would be collected by either of two redundant traina 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 i i and charcoal filters which exhaust to the shield building exhaust stack. k 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 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 f at the low population zone radius of 14 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 0.254 per day at peak accident pres-sure, the offsite doses are about the same as those initially calculated 4 for higher primary containment leakage and lower secondary containment in-leakage (Reference 6). i l 1 i 1 1 l i l l 1 4 e .m.

LICENSE AMENDMENT REQUEST DATED December 14, 1995 Conformance of Administrative Controls Section 6 To the Guidance of Standard Technical Soecifications r P EXHIBIT D Standard Technical Specification Pages Marked Up Pages 5.0-1 5.0-2 Insert 1 5.0-3 Insert 2 5.0 4 Insert 3 5.0-5 Insert 4 5.0-6 5.0-7 5.0-8 5.0-9 5.0-10 5.0-11 5.0-12 5.0-13 5.0-14 5.0-15 5.0-16 5.0-17 5.0-18 5.0-19 5.0-20 Inserts 5 and 6 5.0-21 5.0-22 5.0-23 5.0 24 ..L..

LICENSE AMENDMENT REQUEST DATED November 2, 1995 Conformance of Administrative Controls Section 6 l To the Guidance of Standard Technical Snecifications EXHIBIT D Standard Technical Specification Pages Marked Up Pages 5.0-1 5.0-2 Insert 1 5.0-3 Insert 2 5.0-4 Insert 3 5.0-5 Insert 4 5.0-6 5.0-7 5.0-8 5.0 9 5.0-10 5.0-11 5.0-12 5.0-13 5.0-14 5.0 15 5.0-16 5.0-17 5.0-18 5.0-19 5.0-20 Inserts 5 and 6 5.0-21 5.0-22 5.0-23 5.0-24

Respons@444y T5,C.0-(" G4.0 ADMINISTRATIVE CONTROLS 6 4.1 Responsibility N V V V g% V \\/ M N 6.) -5.1.1 AThef.PlantSu"inucndent]shallberesponsibleforoverallunit operation andMall delegate in writing the succession to this responsibility during his absence. ro cx w a y < The Plant Superinsendent] or his designee shall approve, prior to implementation, each proposed test experiment or-modification to sy3tems or equipment that affect nuclear safety. @, The-f)consnand function.fhiftfupervisor (SS)}-shall be responsible S.I.2 raom During any absence of the-ESSF from the control room while the unit is in MODE 1. . 3. or 4. an individual with an active Senior Aeactor. erator (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/eactorgperatorlicenseshallbedesignatedt'oassumethe control room command function. n n n n. n n n v v v 9-9- y v w w g

w. n T6. C,0 -2 t0 ADMINISTRATIVE CONTROLS 6-G.2 Organization s _ m m n v ~ v w' ~ - - V N/ v: y 4 5.2.1A Onsite and Offsite Oraanizations 4 Onsite and offsite organizations shall be established for me-operation and corporate management, respectively. The onsite and offsite organizations shall include the positions for activities affecting safety of the nuclear power plant.

l. +.

Lines of authority, res?onsibility, and communication shall be defined and establis1ed throughout highest management levels intermediate levels, and all oaerating organization positions. These relationships shall Je documented and updated, as appr~opriate, in organization charts, functional descriptions of departmental responsibilities and relationships, and job descriptions for key personnel positions, or in equivalent forms of documentation. d cJ $ b"I M These i requirementsVs#1 hall be documented in the [FSAn]: 14 d y fa 5' rtrod h, tb. co<prve thi avh

2.. b.

The 4/lant Superin@Ed"eM<shallibe responsible for overall, cif,Dy* safe operation of the plant and 'shall have control over Ife i those onsite activities nec,essary for safe operation and 6, z. /),3> i maintenance of the plant-;.. proNc+p 4g 1 t)ste 9 corporateresponsibilityforov%4veposition]shallhave Iba &-spcci fice corporate exes

3. s.

erall 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 +-s e, Lh 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 the.ir independence from operating pressures. Plth -5 2-G 8 0944 Staff Pad f The unit staff organization shall include the following: MM -A non-licensed operatorAshall be assigned to each reactor ),-a-containing fuel and e additionalVnon-licensed cperator dM. bn Ofc VM op4 h pcenn 0" h peaF - (ccntinued) I n m n _m A ^ _~_ m, ~ vf 'rv$ bib b.O-2 SeV 1. O b J l ,J

Insert 1, STS page 5.0-2 including the plant specific titles of those personnel fulfilling the responsibilities of the positions delineated in these Technical Specification, 1 Exhibit D Standard Technical Specifications Marked Up Pages

1 0"gan' cat i on-3.2 T 5. C. b 3 i (o + 2 Organization m n n A o n em y, n y v v v _x v f s v v y -5.2.2 Unit Staff (ccntinuedh W W E I N v cry bob OLf t-4+ operating in MODES I,9 control oos4 rem-which-a-reactor 5 shall be assigned for 1 2. 3. or 4. 33wp 7 Y uiTit sttes-w oth units shutdown require a total of t3te erators for the M4 @ t ack 04% M i 3 2, 4. Atleastonelicensed,h:typ$eratorGOtshallbepresent in the control room M er fu ic ia the reacto In g addition, while unit is in MODE 1, 2. 3. or 4. at least onelicensedJenior eactor #perator +SR@ shall be present in the control room. g b g 3 -c,- Shift crew composition may be less than the minimum / requirement of 10 CFR 50.54(m)(2)(1) and 4-A-2-a and G-2/pg for a period of time not to exceed 2 hours in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew y g-A[ composition to within the minimum requirements. .] y pQ 4 Lf th A [J G ith fhysic /cchnigi;n] shall be on site when fuel is M. reactor. The position may be vacant for not more _ A} \\ A q./ than 2 hours, in order to provide for unexpected absence. provided immediate action is taken to fill the required I w h d) m, position. oN Administrative procedures shall be developed and implemented 7 (limit the working hours of unit staff who perform safet) rentted functions (e.g., licensed SR0s. licensed R03 < health physicts . auxiliary operators, and key maintepante personnel). / i Adequate shift cov r e shall be maint-e~1ned without routine eobjectdesiallbetohave heavy use of overtime. l operating personnel work 40hourweekwhiletheun'p4r12]hourday, no n 1Ssoperating. However, in the event that unforeseen oblems requ(re substantial amounts eling, major maintenanc Q. periods of ed, or during extended of overtime to be r major plant shutdown for r modi ficati , on a temporary basis the following guidelines shall followed: 4 An individual should not be permitted td work more n 16 hours straight, excluding shift turnover time: (ccntinucd) n n n n n n__ _ _ - y V / [ e g

.. - =. i e Insert 2, STS page 5.0-3 i .J Also, if one unit is in MODE 1, 2, 3, or 4 and the other unit is 4 in MODE 5 or 6, as a minimum the on-site staffing shall include two senior reactor operators (SRO) and two licensed reactor operators (RO). ) i Exhibit D Standard Technical Specifications Marked Up Pages i i

or-ganizatjog-u. c. (,4-2 Organization n ^ n v v v v v v 1 ~ 5:2 Unit StaM-feont4nuedt i i 2. An individual should not be permitted to work more thaj / 16 hours in any 24 hour period, nor more than 24 houps' in any 48 hour period, nor more than 72 hours in nf 1 day period, all excluding shift turnover ti 4 i 3. A brea of at least 8 hours shoulo be all d between work pe ' ds, including shift turnover 1me: 4. Except during tended shutdown riods, the use of overtime should considered an individual basis and not for the ent~ e sta on a shift. g'u uperin(delinesshallbeauthorized Any deviation from the abov in advance by the [Pla t dent] or his designee, in accordance with app d administra ve procedures. or by higher levels of agement, in accor aqce with established i procedures and th documentation of theT sis for granting the deviati l l Contr shall be included in the procedures such at in idual overtime shall be reviewed monthly by the Plant 4 i perintendent] or his designee to ensure that excessive i hours have not been assigned. Routine deviation from the above guidelines is not authorized. l E-3 S. The amount of overtime worked by unit staff members [ performing safety related functions shall be limited and ( 1 ccordance with-the-NRC-Pol 4cy-Statement-on _) controlled i p(Generic !. h er 02-12). h w. -) working hour ^tt by pro u y t.5 whn p bew N I The-fherations )(anager or /ssistant fperations Manager} gE 3 6. -f. shall 1old an SRO license. p eg mw. i 7,-9. The ghift technical /dvisor (STA) shall provide advisory technical support to theJhift fupervisor4S91n the areas of thermal hydraulics. reactor engineering, and plant analysis with regard to the safe operation of the unit. #- addition. the STA-shall mcct the quaMf4 cst-tens-specvfied-by the Ccrission Policy Statcm:nt on-Engineering-Experti-sev SM-ft, gw s w J-3 n p n n . - ~ i y y v u w y j j WCC 3I3 M* be v' 1. 04/07/N

Insert 3, STS page 5.0-4 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. Exhibit D Gtandard Technical Specifications Marked Up Pages i l

Unit-Statf-Oua4444 cat 4ons-5.0 'T5. G. 0-V 5.0 ADMINISTRATIVE-CONTR0tt Plant-(o4;3 p Staff Qualifications g g (, h ru ote: Minimum qualifications for members of the unitAff 3 hat 1 be specified n overall ualification stat ntw efe'rencing an ANSI qualifications. Generall me ing individal position Standard acceptable to the or ferable: however the second method e to those unit staffs requi -special p lon statements because of unique organizationa structurD N _ 1, 5 e f[evnbet M 5 5.3.1 Each member of the unit staff shall meet or exceed the minimum qualifications of { Regulatory Guide 1.8. Revision .1907. Or more recent re'/isionsror ^NSI Standar4-aseeptable-to-t-RC-st+ff3-The-staff-nct covered by-fRegtriatory-Guide-h8} shaWmeet or exceed-the-minimum-qua hf4 ca t40ns-o f-[Regul a ti onsr Regul atory Guidesvor--ANSI-Standards-acceptable-t&NRC-stafft p c.f p R n m n n n m y y y 3 v v v v v -WOC ST 5.0-5 -Rev-1. 04/0 N95-

Insert 4, 3TS page 5.0-5 except for (1) personnel who perform the function of shift technical advisor shall have a bachelors degree or equivalent in a scientific or engineering discipline with specific training in plant design, and response and analysis of the plant for transients and accidents, and (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. Exhibit D Standard Technical Specifications Marked Up Pages

Sra>c+durs-3.4-T5. G. o - y 0.0 ADMINISTRATIVE CONTR3tS M.4 Procedures p. Q 'v' (g 's.- L. \\_ / 'g,' 5.4.1 Written procedures shall be established, implemented. and maintained covering the following activities: l A +. The applicable procedures recommended in Regulatory Guide 1.33. Revision 2. Appendix A. February 1978: B 4. Theemergencyoperatingproceduresreguiredtoimplementthe requirements of NUREG-0737 and to NURuG-0707. Supplement 1. as stated in -EGeneric Letter 82-33}; control c -e. Quality a:gtrance for effluent and environmental monitoring: O4 Firefrotection/rogramimplementation: and G E +. All programs specified in Specification 6.5. n n n n n Aj v / \\/ ~g m-- P WOG STS-5.0 5 Rev 1. 04' W E

= t Srogr m: cad-Menea+s-5.5 'TS. O O-G 5.0 AD"INISTRATIVE CONTROLS C-5. 5 Programs and Manuals n n n m o n n n v ,n v v v v v v v v v v v v v v v w The following programs shall be established implemented. and maintained. 4-4 [} Offsite Dose Calculation Manual (ODCM) a 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; and b. 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 Operating, and Radioactive Effluent Release Reports required by Specification [5.6.2] and Specification [5.6.3]. Licensee initiated changes to the ODCM: a. Shall be documented and records of reviews performed shall be retained. This documentation shall contain: This sedion is nd

- [pr v1ou y P "

1. sufficient information to support the change (s) together with the appropriate analyses or evaluations submitted for review by the NRC justifying the change (s), and on July 17,1995 2. a determination that the change (s) maintain the levels of radioactive effluent control required by 10 CFR 20.1302. 40 CFR 190. 10 CFR 50.36a. and 10 CFR 50. Appendix I. and not adversely impact the accuracy or reliability of effluent, dose, or setpoint calculations: b. Shall become effective after the approval of the [ Plant Superintendent]: and c. 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 Release Report for the period of the report in which any change in the ODCM was made. Each change shall be identified by markings in the margin of the affected pages. clearly indicating the area of the -(continued)- n y, WOG-STS-3 0-7 Rev 1. 04/07/95 l .I

Progr r: and Manuc h T S. CO ~} 5.5 Programs and Mantteh 7__m _n_ ~ 5.5.1 Off+ite 00se Calculet4en-Hentt&H00EM) (contintred)- page that was changed, and shall indicate the date (i.e.. month and year) the change was implemented. 4 podtm5 01 0.5.2 @. Primary Coolant Sources Outside Containment R.s.sibl O Re mo vo I 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 tog j levels as low as practicable. The systems include [Rccircul ion g -Speay Safety Injection. Chcmical and Volume Control. gas-str4pper,-and-Mydrogen-RpineG. The program shall include the following: awdCM M cd 5grqSyrem5, j +. Preventive maintenance and periodic visual inspection requirements: and 2 b. Integrated leak test requirements for each system at refueling cycle intervals or less. i 5.5.3 C. Post Accident Samolino i This program provides controls that ensure the capability to obtain and analyze reactor coolant. radioactive gases, and i particulates in plant gaseous effluents and containment atmosphere samples under accident conditions. The program shall include the following: 1 i +, Training of personnel; 4

2. 4h Procedures for sampling and analysis: and z, &

Provisions for maintenance of sampling and analysis equipment. 5-5-4 C. Radioactive Effluent Controls Procram This section is not This program conforms to 10 CFR 50.36a for the control of ma*ed up since it radioactive effluents and for maintaining the doses to members of the public from radioactive effluents as low as reasonably was previously submitted for achievable. The program shall be contained in the ODCM Shall be review by the NRC implemented by procedures, and shall include remedial actions to on July 17,1995 i (continued) m nn _r ~ -~. m ~ l 6 5.0-8 Rev 1. 04/07/95 4

.~_ accgran and Maak-T g,C,,0- 8 5-5-Programs-and44anuals ^ ^^ ^ - A n-n m' , y-y m 5.0.4 Redicactive Cf&ent-GentmH-Procram-(continued) be taken whenever the program limits are exceeded. The program shall include the following elements: a. Limitations on the functional capability of radioactive liquid and gaseous monitoring instrumentation including sur veillance tests and setpoint determination in accordance i with the methodology in the ODCM: 1 b. Limitations on the concentrations of radioactive material released in liquid effluents to unrestricted areas. conforming to 10 CFR 20. Appendix B. Table 2 Column 2: ~ c. Monitoring, sampling, and analysis of radioactive liquid and gaseous effluents in accordance with 10 CFR 20.1302 and with the methodology and parameters in the ODCM: 4 This section is not d. Limitations on the annual and quarterly doses or dose marked up since it commitment to a member of the public from radioactive was previously materials in liquid effluents released from each unit to submitted for unrestricted areas, conforming to 10 CFR 50. Appendix I: review by the NRC on July 17.1995 i e. Determination of cumulative and arojected dose contributions from radioactive effluents for t1e current calendar quarter and current calendar year in accordance with the methodology and parameters in the ODCM at least every 31 days: f. 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 i period of 31 days would exceed 2% of the guidelines for the annual dose or dose commitment, conforming to 10 CFR 50. i Appendix I: g. Limitations on the dose rate resulting from radioactive material released in gaseous effluents to areas beycrd the site boundary conforming to the dose associated with 10 CFR 20. Appendix B. Table 2. Column 1: h. 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 10 CFR 50. Appendix I: (continued) m n -n nnn - pnn ,_n WOG STS-G.0 Rev 1. 04/07/05-I i r

eegeamMnd Manuair 4+

i . 5-5-ProgramundJianuals ,~ m - ,~ se v y s, 5.5.4 Radioactive Effluent Controls Pr umam Montimed) 1. Limitations on the annual and cuarterly doses to a member of the public from iodine 131, iocine-133. tritium, and all radionuclides in particulate form with half lives > 8 days in gaseous effluents released from each unit to areas beyond the site boundary, conforming to 10 CFR 50. Appendix I: and j. Limitations on the annual dose or dose commitment to any i member of the public due to releases of radioactivity and to radiation from uranium fuel cycle sources. conforming to i 40 CFR 190. E. Comoonent Cyclic or Transient Limit 4 s.s.s i u

4. l. 4-l This program provides controls to track thefSAR. Section [

]. Cyclic and transient occurrences to ensure that components are maintained within the design limits. ~ Pre-Stressed Concrete Containment Tendon Surveillance Proaram / gram provides controls for monitoring any ten ~ deg i in re-stressed concrete contJa'n.mants, including effectiveness s corrosion protectierrmedium, to ensure containment struc tegri yMhe program shall include 1 baseline measurements pr' initial operations. The Tendon Surveillance Pro

m. inspection requencies, and acceptance criteria sh e in accordance with [Regu atory Guide 1.35, Revie~ -

. 1989]. The provisions of SR 3.0.2 and SR 3.0.3 are applicable to t Tendon Surveillance Program inspection frequencies. E. (@tSt.YVtb 5:5J R ea c tor-Cool a nt-Pumo-F4vwhee M ns cec t i on-Proc r a m-- X_- l N j This program sh. T1 provide.ftthe-inspeftion of each reactor 1 a coolant pump fl heepenhe recommendat4cns~oLRegulatory. P of Regulatory Guide 1.14, Pevision 17 August-1925 (g, ( kSW Vt i \\ A. _A Os .Ng v v v v v g g., ) MSTS 0-10-- - Rev-1--04/07/95

P cgram; aM-MontnHs- -6 + r 5 Programs and Manuals (continued) m nn m nm m - m -m .v y y w v y v v y m v m y f 5.5.8 Inservice Testina Proaram This. program provides controls for inservice testing f ASME Code lass 1, 2. and 3 components including applicable s ports. The p gram shall include the following: a. esting frequencies specified in Section I of the ASME Bonier and Pressure Vessel Code and app 1 cable Addenda as fod ws: ASME Bo er and Pressure Vessel Co and applicable denda terminology r Required Frequencies inservice test for performing inservice activities testina activit-ies 3 Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or ever 3 months At least once per 92 days Semiannually o every 6 mo hs At least once per 184 days Every 9 mo s least once per 276 days Yearly or annually At least once per 366 days Biennia y or every 2y rs At le t once per 731 days b. T provisions of SR 3.0.2 are applicabl to the above equired Frequencies for performing inserv' e testing activities: The provisions of SR 3.0.3 are applicable to in rvice testing activities: and d. Nothing in the ASME Boiler and Pressure Vessel Code s 11 be construed to supersede the requirements of any TS. lj. l b5t ( V4 ) ) ton 45G)-Tube-Surve4WneeJrocram b.5-g Steam Genera Reviewer's N~ote-TheJ.icensee's curreJntcensing' basis steam generator tube surveilliiid - icemFnts shall be relocated from the LC0 and ' uded re. An appropFTate-admin' trative controls ormat should be used. i 7_Jrogra n nf n n q m_ m. + .m.-- WCG STS 5.0-11 Rcv 1. 04/07/ %- r.. s

.Progr4r45-and-Haauals- ) 5.5-s s?5 Programs and Manuals (continued) f\\g '/ sg n ; ,% y vq; ,c 5.5.10 Secondary Water Chemistrv Proaram Th1 rogram provides controls for monitoring seconday water chemistry to inhibit SG tube degradation and low p,retsure turbine 4 disc str corrosion cracking. The program sh r include: a. Identifi tion of a sampling schedule or the critical l variables a ntrol points for t e variables: b. Identification ofthe procedu s used to measure the values of the critical varlables: c. Identification of pro ssM ampling points, which shall 4 include monitoring e discita ge of the condensate pumps for evidence of conde,wr in leak (ag / d. Procedures o'r the recording and mah ement of data: ^ e. Proce res defining corrective actions fop 4 off control 11 and po' chemistry conditions: f. A procedure identifying the authority responsib1 r the interpretation of the data and the sequence and tim of administrative events, which is required to initiate corrective action. T. 65d d \\ 5'.s E kni,ilation FilteMeWeooram "!FM A program shall be established to implement the following require'd tetting of Engineered Safety Feature (ESF) filter ventilation systems at the frequencies specified in [ Regulatory Guide ]. and in acco'rdance with [ Regulatory Guide 1.52. Revisi '2. ASME N510-1989 and AG-1]. N / a. Demonstrate forgach of the ESF, systems that an inplace test of the high efficiency par,t1'ttnate air (HEPA) filters shows a penetration and sy' step bypass < [0.05]% when tested in accordance with [ Tst'ory Guide 1.52. Revision 2. and ASME N510-1989] at ystem f1'owrate specified below [ 10%]. S Ventilation System Flowrate / Q % - g -- r' ^- r ~_ WOG STS-5.0-12 {cv 1. 04/07/05

or-ogram: and-Hanu d r l a.a \\ 5.5 Programs and Manuals ~ ,A^f'y ^ y m m m y m ~ y m 5'.%.11 Ventilation Filter Testino Proaram (VFTP) (continued) b. Demonstrate for each of the ESF systems that an inpla test of the charcoal adsorber shows_a penetration and sy m bypass < [0.05]% when tested in accordance with [ ulatory Guide 1.52. Revision 2 and ASME N510-1989] at t system flowrate specified below [ 10%]. ESF Ventilation System F wrate / c. Demonstrate (oreachoftheESF stems that a laboratory test of a sample of the charco adsorber, when obtained as aescribec in CRhgulettry buidt .52. Revision 2]. Ms. Fvw methyl iodide penstration lpss than the value specified below when tested ihsaccorfance with [ ASTM D3803-1989] at a temperature of s [30'Cs] and greater than or equal to the relative humidity spec' (ed below. ESF Ventila on Syst'em Penetration RH N Reviewer's N . Allowable penetration - [100% - methyl iodide efficiency r charcoal credited in staff safe'ty evaluation]/ (safety f or). K 4 Safety actor - [5] for systems with heaters. \\ = [7] for systems without heaters. 3 Demonstrate for each of the ESF systems that the pressure drop across the combined HEPA filters. the prefilters.'and the charcoal adsorbers is less than the value specified \\ below when tested in accordance with [ Regulatory Guide 1.52, 4 1 (continued) c. n ne m pfp _n. WOG STS 5.0-10 - Rev 1. 04/07/95-

} Progr= ard "anueh-T5.G.o-5.5 Programs-and-Manuals ^^ ~~ -N ^ ~ ~ ^ s v v v % ~ v v v

5. 5. fl Ventilation Filter Testina Proaram (VFTP)

(continued) n nd ASME N510-1989] at the system flowr te specified NESFVentilationSystem Det a P Flowrate e. Demonstrate that t heat I for each of the ESF systems dissipate the v lue specifiedtelow [ 10%) when tested in accordanc i+h[ASMEN510-1989( "$1-Ven[Plation Systs h % N Wattage 4 N N The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the FTP T test frequencies. x G.5.12 3, Exolosive Gas and Storace Tank Radioactivity Monitorina Procram 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 or fed into the offgas treatment system, and the quantity of radioactivity contained in unprotected outdoor liquid storage tanks]. The gaseous radioactivity quantities shall be determined following the This section is not methodology in [ Branch Technical Position (BTP) ETSB 11-5. mirked up since it " Postulated Radioactive Release due to Waste Gas System Leak or Failure"]. The liquid radwaste quantities shall be determined in w:s previously submitted for accordance with [ Standard Review Plan. Section 15.7.3. " Postulated review by the NRC Radioactive Release due to Tank Failures"]. on July 17,1995 The program shall include: a. The limits for concentrations of hydrogen and oxygen in the [ Waste Gas Holdup System] and a surveillance program to ensure the limits are maintained. Such limits shall be (continuedF nm ^ & n ^fv~ ^ ^ ^v m WOG STS 5.0-14 I<ev 1. 04/07/05

.~ .=. l Progrc = and Mangc19 5.5 i'iograms and Manuals } [w m - K mp yn O.S.12 E x Di os i vedies-ard-Storace-Ta nk-Ra dioac t4 vity-Moni t ori no-P roc ram j 4 cent +nuedh appropriate to the system's design criteria (i.e., whether i or not the system is designed to withstand a hydrogen explosion): b. A surveillance program to ensure that the quantity of radioactivity contained in [each gas storage tank and fed 4 into the offgas treatment system] is less than the amount that would result in a whole body exposure of = 0.5 rem to any individual in an unrestricted area. in the event of [an j uncontrolled release of the tanks' contents]; and i c. A surveillance program to ensure that the quantity of ThD section is not h martee p s:nce r radioactivity c9ntaWf in all wtdoor liquid redwaste tarL4@ the.t are not surroun6d by liners, dixes, or sails, d,paole was previously submitte for of holding the tanks' contents and that do not have tank miew by the NRC overflows and surrounding area drains connected to the on July 17,1995 [ Liquid Radwaste Treatment System] is less than the amount i that would result in concentrations less than the limits of 10 CFR 20. Appendix B. Table 2. Column 2. at the nearest potable water supply and the nearest surface water supply in j an unrestricted area in the event of an uncontrolled release of the tanks' contents, i The provisions of SR 3.0.2 and SR 3.0.3 are applicable to the Explosive Gas and Storage Tank Radioactivity Monitoring Program surveillance frequencies. l l 5.5.10 K, Diesel Fuel Oil Testina Proaram i j A diesel fuel oil testing program to imalement required testing of i both new fuel oil and stored fuel oil s1all be established. The programshallincludesamplingandtestingre3uirements.and acceptance criteria, all in accordance with Nplicable ASTM 005 7 7 The purpose of-the progranris-tofestabMsh-the-bandard5. -dli} gl foHowing: ?cceptability of new fuel-eil for use-prior-to pu Iid tb h mb S 4 eddition-to 1 c. storage tanks-by detemintrq-that-the-fuel-oi4-hast 1 en m! gravity or an-absolute-spec 4f4e-gr-av4trwithin WLo h cb dd b hmits,-- Vis Loin, Lvchv> c& S thvkh y (-- RD- -D -^- D- ^ ff%f P-WOC STS 5.0-15 Rev 1. 04/07/95 ~

r0grr and Maaueb- 'T.S.G.6 m 5.5 Progr = =d Manuals K Qf f ^f f p K-. F543 Diesel Fuel Oil Testino Proaram (continued) / / 2. a flash point and kinematic viscosity with 'Timits for ASTM 20 fuel 011. and 3. a clea d bright appeara - ith proper color: b. Other properties for ? D fuel oil are within limits within 31 days f wing sampktas and addition to storage tanks; and N c. particulate concentration of the ciLis s 10 mg/l when tested every 31 days in accordance with ASTM'0-2276. Method A-2 or A-3. N 5.5.14-f,. Technical Soeci fications ffS+- Bases Control Procram This program provides a means for processing changes to the Bases of these Technical Specifications. wkwcd5p%[scchia I.,s-Changes to the Bases of the--7lir shall be made under appropriate administrative controls and reviews.

1. 4.

Licensees may make changes to Bases without prior NRC approval provided the changes do not involve either of the following: LAnia)S Nf'u. M. P r

n. -1.

a change in the W incorporated in the license: or A Le a change to the upd;ted sSAR or Bases that involves an unreviewed safety question as defined in 10 CFR 50.59. 3'-s. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the fSAR.

' posed changes that meet the criteria of Specification Pro Q-d-

.s.14b above shall be reviewed and approved by the NRC rior to implementation. Changes to the Bases implemented withoutpriorNRCapprovalshallbegrovidedtot1eNRCona g* g* g frequencyconsistentwith10CQ^50.,1fet-L4.50)$ WP) h - n n mA n m n ~ ~ s v-q y v -v WOC STS -5.0-16 Rev 1. 04/07/95

l f*rograms end 44&ntrals-95-5.( P{ogeiiis end Fianuels (contirgedt m v v v v v v yvv f v vyvv v 5\\E.15 Safety Function Determination Proaram (SFDP) NThis program ensures loss of safe nd appropriate actions taken. Upon entry into LC0 3.0.6. n evaluation shall be made to determine if loss of safet function xists. Additionally, other appropriate actions may taken as a 4'equit of the support system inoperability and corr sponding r excbption to entering supported system Condition a Required s Actions. This program implements the requirement of LCO 3.0.6. The SFD shall contain the following: a. ProviQonsforcrosstraincheckstoensrealossofthe capabilt y to perform the safety functiAn assumed in the accident alysis does not go undetected; b, Provisions for ensuring the plant maintained in a safe condition if a loss of function c dition exists: c. Provisions to ens e that an i perable supported system's Completion Time is ot inappr riately extended as a result ofmultiplesupportNystemi.operabilities: and d. Other appropriate limit t' ns and remedial or compensatory actions. A loss of safety function e ists when, assuming no concurrent single failure, a safety f6nction ssumed in the accident analysis cannot be performed. For the purpoqe of this program a loss of safety function may ex'st when a support system is inoperable, and: a. A required sy em redundant to the system (s) supported by the inoperab e support system is also inoperable; or system redundant to the sy\\(s) in turn b. A require stem supporte bytheinoperablesupportedsfstemisalso inoper le: or \\\\ c. A re ired system redundant to the support system (s) for the sup orted systems (a) and (b) above is also inoperable. The S P identifies where a loss of safety function ekists. If a loss of safety function is determined to exist by this ' program, th appropriate Conditions and Required Actions of the LCO in w ch the loss of safety function exists are required to b,e. ntered. n m n n n n n m n n n n n Am v v v v x-v s s j s WOG STS -G.0-17 Rev 1. 04/07/95

Reporting Rcquirement+ T5. C. oA 5.0 ADMINISTRATI"E CONTROLS 6-+.6 Reporting Requirements f m-T_ ^ m m y m m m m y The following reports shall be submitted in accordance with 10 CFR 50.4.

5. C.1 h.

Occucational Radiation Exoosure Reoort h .................---N0TE---------------------- ~ A single suhiiItt'aPmay4eJDade for a mWe station. The submittal should combine s to all units at the station. A tabulation on an annual basis of the number of station utility, ey du De *and other personnel (including contractors) receiving exposures -100 mrem /yr and their associated man rem exposure according to work and job functionsf e.g., reactor operations and surveillance, f inservice inspection, routine maintenance, special maintenance (4 describe maintenance-})the requirements of 10 CFR 20.2206. waste processing, and refueling +. This tabulation supplements The dose assignments to various duty functions may be estimated based on pocket dosimeter, thermoluminescent dosimeterM. or film _g b badge measurements. Smallexposurestotallingf20%ofthe 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. The report shall be submitted by April 30 of each year. [-The-initia, report cha bc submitted by,^pr44-30-of-4he-year following-the-4 nit 4abetsit4cahtyd 5--6-2 g, Annual Radioloaical Environmental Coeratina Re00rt -: c;C_ .....................N0TE ------------ M 1on. The ade for a multjplemnT A single submit ahn submittal should combine all units at the station. n The Annual Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall This section is not be submitted by May 15 of each year. The report shall include inam up since n summaries interpretations, and analyses of trends of the results ttj fo of the radiological environmental monitoring program for the reporting period. The material provided shall be consistent with review by the NRC the objectives outlined in the Offsite Dose Calculation Manual on July 17,1995 ,A n p ^ m n ^ n _m "00 STS 5.0-10 Ra 1. 04/07/05

Reper-ting Requirements-5.S T.s, c.0 - i1 t 4A Danne, t.i nn D.a.n,i.n i. r. a.ma. n.t. c..- m y v m v s 5.0,2 -Annud= Redid oeicM=Envierc.entd=0oera tino=Recort=4conM nued ) (00CM), and in 10 CFR 50. Appendix I. Sections IV.B 2. IV.B.3, and IV.C. The Annual Radiological Environmental Operating Report shall j include the results of analyses of all radiological environmental samples and of all environmental radiation measurements taken during the period pursuant to the locations specified in the table This section ts nd and figures in the 00CM. as well as summarized and tabulated merked up since it results of these analyses and measurements [in the format of the was previously table in the Radiological Assessment Branch Technical Position. submined for Revision 1 November 1979]. [The report shall identify the TLD review by the NRC results that represent collocated dosimeters in relation to the i '" J"'Y 17' '"' NRC TLD program and the exposure period associated with each i result.] In the event that some individual results are not available for inclusion with the report, the report shall be J submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary j report as soon as possible. 5.0.0 C, Radioactive Effluent "c' nc ReDort ........................N0TE------------------------

E I

A single su may be made for a multiple unit-statili The submittal should com ions comm raTT units at the station; however, for units w ate radwaste systems, the submittal shall s e e releases o adica ive material from each unit j The Radioactive Effluent Rclerc Report covering the operation of the unit shall be submitted in accordance with 10 CFR 50.36a. This section is nd marked up since it The report shall include a summary of the quantities of was previously radioactive liquid and gaseous effluents and solid waste released submitted for from the unit. The material provided shall be consistent with the roview by the NRC objectives outlined in the 00CM and Process Control Program and in on July 17,1995 conformance with 10 CFR 50.36a and 10 CFR 50, Appendix I. Section IV.B.1. 4 ,P y V ~ ~ fcontinuedP W9G-STS-S.0-19 Rev 1. 04/07/95

~. _. i RCporting-Rewi+emente 6 ~f5. G, 0 5.0 Reporting Requirements (coat-inued)= ^, ff fy - ^ ^ v\\ v v q m y v v v y +* 0, Month 1v Ooeratina Reoorts l Routine reports of operating statistics and shutdown experienceE-i including documentation of a challenges to the pressur-Mer--imer cperated relicf valves -Or pressurizcr safety-vehes.]= shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report. are opeJTacy.im'ks Pzpak L 4-6-5E.- CORE OPEPJTINC LIti!TS REPORF (COLR) ), -e-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:,g 33tg5 7he TrTatvidual-specificarians that addr-ess-cccc uperating _ limits must be ceferenced here. q, e-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: pg(, 3destTfy-the Io!Lical Report (s) by number. tit;1e JatennF NRC staff approvaT-~dctamentu 4 the staff Safety a on r_t fe-a ptsfit' spec 1 ethodolog

3. +.

The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal-mechanical limits. N b# - core thermal-hydraulic limits. Emcrgcncy Corc CO^' 3 Systems-4ECCS) limits nuclear limits such as , transient N9W analysis limits. and accident analysis limits) the safety analysis are met. q +. The COLR including any midcycle revisions or supplements. shall be provided upon issuance for each reload cycle to the NRC. 5.' " Reactor Coolant System (RCSP PRE 55URE-AND-TEMPERATURETMITS7 PORT (PTLR) ~ N ~U RCS press temperatur s for heat up, cooldown, a. low temperature on criticality, and hydrostatic _~4 [7 ,nhK^_nf EJ~ ~ ~ M06-STS-5.0-20 4ev 1. 04/07/95 l

'l Insert 5, STS page 5.0-20

a. Heat Flux Hot Channel Factor Limit (F[ ), Nuclear Enthalpy Rise Hot Channel Factor Limit (F.

), 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) a e

Insert 6, STS page 5.0-20 NSPNAD-8101-A, " Qualification of Reactor Physics Methods for 2 Application to PI Units" (latest approved version) i l NSPNAD-8102-A, " Prairie Island Nuclear Power Plant Reload Safety Evaluation Methods for Application to PI Units" (latest approved version) f WCAP-9272-P-A, " Westinghouse Reload Safety Evaluation Methodology", July, 1985 1 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 I Methodology", December, 1988 ) WCAP-10924-P-A, Volume 1, Addendum 4, " Westinghouse Large Break LOCA Best Estimate Methodology", August, 1990 l XN-NF-77-57 (A), XN-NF-77-57, Supplement 1 (A), " Exxon Nuclear Power Distribution Control for Pressurized Water Reactors Phase II", May, 1981 WCAP-13677, "10 CFR 50.46 Evaluation Model Report: E-COBRA / TRAC 2-Loop Upper Plenum Injection Model Update to Support ZIRLOm Cladding Options", April 1993 (approved by NRC SE dated November 26, 1993). NSPNAD-93003-A, " Transient Power Distribution Methodology", (latest approved version) Exhibit D Standard Technical Specifications Marked Up Pages

Reporting Requiremers-- l W j J i 5.6 Reporting Requirements pp-yp a 5.6. Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS RE )0RT (PTLR) (cont 1nued) testing as well as-heatup and cooldown rates shall b established and documented in the PTLR for the foil wing: [The individual specifications that address RCS pr ssure and temperature limits must be referenced here.] b. The analytical methods used to determine the S pressure dtemperaturelimitsshallbethoseprevipdslyreviewed a approved by the NRC. specifically thos described in the foripwing documents: [ Identify the NRC aff approval docu nt by date.] c. The PTL shall be provided to the NR upon issuance for each reactor v sel fluence period and f r any revision or supplement thereto. \\ Reviewers' Notes: Tqemethodology or the calculation of the P-T limits for NRC approval-should i ude the following provisions: 1. The methodology sh y de ibe how the neutron fluence is calculated (referenc Regulatory Guide when issued). 2. The Reactor Vessel ial Surveillance Program shall comply with Appendi Hto\\10CFR50. The reactor vessel material irradiat n survettlance saecimen removal schedule shall be provid , along with,how t1e specimen examinations shall be used update the PT(R curves. N 3. Low Tempera re Overaressure Protection (LTOP) System lift setting li its for t1e Power Operated Relief Valves (PORVs), develope using NRC-approved methoddlogies may be included in the LR. 4. The djusted reference temperature (ART) or each reactor be line material shall be calculated accgunting for r diation embrittlement, in accordance with egulatory Guide .99 Revision 2. 5 The limiting ART shall be incorporated into the alculation of the 3ressure and temperature limit curves in abqordance with NUREG-0800 Standard Review Plan 5.3.2. Pressur - Temperature Limits. (continucy n_n n y ,_ ~ p n WOErSTS = - 5. 0 -Rev 1. 04/07/9P

Reporting Rcquircment+- 5 5.6 Reporting Requirements ,K-np e n ~/, 5. 6 Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR) (continued) 6. The minimum temperature requirements of Appendix G to 10 CFR f Part 50 shall be incorporated into the pressure and temperature limit curves. 7. LicenseeswhohaveremovedtwoormorecapsuJesshould compare for each surveillance material the Jneasured increase reference temperature (RT r) to the predicted increase in e where the redicted increase in RT' is based on the R 1: shift in Rfer plus the two standard deviation value mea (2a,) pecified in Regulatory Guide 139. Revision 2. If the value exceeds the predicte#'value (increase RT[1 +lice measur 20 ). th demonstra how the results affect the approved methodology. 5.6.7 EDG Failure Reoort If an individual emerge'ncy diesel generator (EDG) experiences four or more valid failures ih t e'last 25 demands, these failures and any nonvalid failures expe enced by that EDG in that time period shall be reported within )0 ays. Reports on EDG failures shall include the informationp ecommqnded in Regulatory Guide 1.9. Revision 3. Regulatory / Position \\C.S. or existing Regulatory _ Guide 1.108 reportin requirement \\ N 5.6.8 PAM Reoort When a report'is required by Condition B'or G of LCO 3.3.[3]. " Post Accide'nt Monitoring (PAM) Instrumentation." a report shall be submit sd within the following 14 days. 'The report shall outline he preplanned alternate method of monitoring. the cause of thex noperability and the plans and scheduTA for restoring the instr entation channels of the Function to OPERABLE status. / \\\\ \\ \\ fn _ e_ p n qn n_ p n \\ %.VI I L i l II.4G bl / --WOG STS 5.0-22 --Rev 1. 04/07/95

~. - Repcrting 90guircments-5 ReportingRequirements (continued) [. O. M-M. A. e O. A O.. A. s 5.6.9 Tendon Surveillance Reoort An bnormal degradation of the containment struct e detected duri he tests required by the Pre-stressed Concrete Containment Tendon veillance Program shall be reporteddo the NRC within 30 days. The report shall include a desc 'p' tion of the tendon condition, tfie condition of the concre especially at tendon s anchorages),theQnspectionprocedur , the tolerances on _ cracking, and the tqrrective acti taken. 5.6.10 Steam Generator Tube Insoector Reoort rts requ\\ (by the Licensee's current Reviewer's Note-ire licensing basis'ye arding steam gerte ator tube surveillance requirements shall be included here. appropriate administratWe controls format should b used. Revijver's Note: These reports may be require scovering inspection. test, and maintenance activities. These reports are f d d ,x etermined on an in ivi ual basis for each unit and'tbeir preparation and submittal are designated in the Technical _Speci fications. n n n ,/ V V 'V' b-Q' 'G 'G V V V WOC STS -;.0 Rev 1. 04/07/95-

[High % d utio cAce d M 5-0-ADMINISTRATIVE-CONTROES-Q-{&. 7 HighRad1ationArea}7 4 n ,v l -5,L 4 Pursuant to 10 CFR 20. paragraph 20.1601(c) in lieu of the requirements of 10 CFR 20.1601, each high radiation area, as i defined in 10 CFR 20, in which the intensity of radiation is > 100 mrem /hr but < 1000 mrem /hr, shall be barricaded and I 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 s 1000 mrem /hr. provided they are otherwise following plant radiation protection [h procedures for entry into such high radiation areas. j

MNof Any individual or group of individuals permitted to enter such rOview by the NRC areas shall be provided with or accompanied by one or more of the l

on July 17,1995 following: a. A radiation monitoring device that continuously indicates the radiation dose rate in the area. / b. A radiation monitoring device that continuously integrates ) the radiation dose rate in tne 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. c. 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] in the RWP. 5.7.2 In addition to the requirements of Specification 5.7.1. areas with radiation levels a 1000 mrem /hr shall be provided with locked or continuously guarded doors to prevent unauthorized entry and the 1 keys shall be maintained under the administrative control of the Shift Foreman 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 l (continue s n_n ^ nqn n o m h h$h .0-r\\eV /I -}}

ML20095J795

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