ML20135D099

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Proposed Tech Specs Incorporating Certain Requirements from Revised B&W Std TS,NUREG-1430
ML20135D099
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 12/03/1996
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20135D095 List:
References
RTR-NUREG-1430 NUDOCS 9612090243
Download: ML20135D099 (23)
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Enclosure 2 >

l Technical Specifications Revised Pages  !

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l 9612090243 961203  !

l PDR ADOCK 05000289  !

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i TABLE OF CONTENTS Secuon Eags 2 SAFETY LIMITS AND LIMITING SAFETY SYS'IEM SETTINGS 2-1 2.1 Safety Lmuts. Reactor Core 2-1 2.2 Safety Lumts. Reactor System Presswe 2-4 2.3 Lumtma Safety System Settmas. Protecnon i instrumentauon 25  ;

! 3 LIMITING CONDITIONS FOR OPERATION 3-1 >

3.0 Geneml AcnonRegarements 3-1 3.1 Reactor Coolant System 3-la 3.1.1 OperabonalCPe 3-la +

3.1.2 T.-. - *=, Heatup and Cooldown Limitahons 3-3  ;

3.1.3 Muumum Conditions for Criticality 34  ;

3.1.4 Reactor Coolant System Activity 3-8 3.1.5 Chenustry 3-10 '

3.1.6 Imkage 3-12  ;

3.1.7 Moderator Temperature Coef5cient of Reacuvity 3-16  !

3.1.8 SingleImop Resencuans 3-17 3.1.9 Imv PowerPhyncsTestingRecnchons 3-18 3.1.10 ControlRodOperabon(Delets:1) 3-18a l 3.1.11 ReactorlaternalVent Valus 318c l 3.1.12 Pressunzer Power Operssed Relief Valve (PORV)

and Block Valve 3-18d l 3.1.13 ReactorCoolant System Vents 3-18f 3.2 Qgisled 3-19 3.3 Emergenes Core Coohna. Reactor Beidma Emeraency l Coolmg and Reactor Buddag Spray Systems . 3-21  !

3-25 l 3.4 Decay Heat Removal Capstmhty l

3.4.1 Reactor Coolant System Temperature Greater than 2500F 3-25 3.4.2 Reactor Coolant System Tw e os 2500F or less 3-26 3.5 lastrumentanon Systems 3-27 3.5.1 Operanonal Safety Instrun=wahan 3-27 3.5.2 Control Rod Group and Power Distribunon Limits 3-33 3.5.3 Engmeered Safeguards Protection System Actuanon Scapomes 3-37 3.5.4 Incore Instrumesushan (Deleted) 3-38 l 3.5.5 Accident Monitoring Instrumentaban 3-40a 3.5.6 Deleted 3 40f 3.6 Reactor Buddmg 3-41 3,7 Umt ElecancalPower System 3-42 3.8 fuelI.4admaandRefucimg 3-44 3.9 Dgisted 3-46 3.10 Mi-11aa=== Radioecove Matenals Sources 3-46 3.11 Handlung.gfJIIndialgdfad 3-55 3.12 Reactor Buddmg Polar Crane 3-57 3.13 Secondan System Actmtv 3 58 3.14 Bggd 3-59 3-59  !

3.14.1 Penodic I& of the Dikes Around'IMI 3.14.2 Flood Condition for Placmg the Unit in Hot Stancby 340 AirTreatment Systems 3 61 3.15 Emergency Conuel Room Air Treatment Sysicm 3-61 3.15.1 Reactor Buddmg Purge Air Treatment System 3-62a 3.15.2 3.15.3 Auxihary and Fuelliandhng Buddmg Air Treatment System 3 62c Fuel Handhng Buddmg ESF Air Treatment Sysem 3-62c l 3.15.4

! ii A ,= t==^ No.59,72,78,97,98.II9,122,136,141,167,182.496,

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. 1 TABLE OF CONTENTS I Secuon East i

e 3.16 SHOCK SUPPRESSORS(SNUBBERS) 3-63  ;

3.17 REACIVR BUllblNG AIRTEMPERNIURE 3-80 t 3.18 FIRE PROrIECTION(DELETED) 3-86 i 3.19 CONTAINMENT SYSTEMS 3-95 j 3.20 . SPECIALTESTEXCEPTIONS(DELETED) 3-95a i 3.21 RADIOACITVE EFFLUENT INS 1RUMENTATION (DELETED) 3-%  !

3.21.1 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION (DELETED) 3-%

3.21.2' RADIOACTIVE GASEOUS PROCESS AND ElTLUENT 3 96 l MONITORING INSTRUMENTATION (DELETED) t 3.22 RADIOACTIVE EFFLUENTS (DELETED) 3-%

3.22.1 LIQUID EFFLUENTS (DELEIED) 3-%

j 3.22.2 GASEOUS EFFLUENTS (DELETED) 3-%

3.22.3 SOLID RADIOACITVE WASTE (DELETED) 3-%

3.22.4 'IUTAL DOSE (DELETED) 3-%

1 3.23 RADIOLOGICAL ENVIRONMENTAL MONITORING (DELEIED) 3-%

4 3.23.1 MONITORING PROGILAM(DELETED) 3-%

3.23.2 LAND USE CENSUS (DELETED) 3-%

! 3.23.3 IN1ERLABORATORY COMPARISON PROGRAM (DELEIED) 3-%

3.24 BEACIOR VESSEL WATER LEVEL 3 128 b 4 SURVEILLANCE STANDARDS 41 4

4.1 OPERATIONAL SAFETY REVIEW 41 l 4.2 REACIOR COOLANT SYSTEM INSERVICE INSPECTION - 4-11 4.3 'IESTING FOLLOWING OPENING OF SYSTEM (DELETED) 4 13 4.4 REACIOR BUILDING 4-29 4.4.1 CONTAINMENTLEAKAGETESTS 4-29

! 4.4.2 STRUCTURALINTEGRITY 4-35 4.4.3 DELETED 4-37 l' 4.4.4 HYDROGEN RECOMBINER SYSTEM 4-38 4.5 EMERGENCYIDADING SEOUENCE AND POWER'IRANSFEll 4-39

EMERGENCY CORE COOLING SYSTEM AND REACiOR j BUILDING COOLING SYSTEM PERIODIC TESTING
4.5.1 EMERGENCYIDADING SEQUENCE 4 39 l

' 441  !

4.5.2 EMERGENCY CORE COOLING SYSTEM 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4-43 4.5.4 DECAY HEATREMOVAL SYSTEMlEAKAGE 4-45 1 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 REACIOR CONTROL ROD SYSTEM TESTS 4-48 l

4.7.1 CONIROL ROD DRIVE SYSTEM FUNCTIONAL'IESTS 4-48

, 4.7.2 CONIROL ROD PROGRAM VERIFICATION (DELETED) 4-50l i

-iii-Amendment 72 ,81,198 ,129 .137 ,144 ,147 ,I ll8 ,19I ,197 , 498;

UST OF FIGURES FIGURE TIT 1.E 1%g.E 2.1 1 Core Protection Safety Umit 'D41-1 2-4a 2.1 2 DELETED i 2.1-3 Core Protection Safety Bases TMI-1 2-4c 2.3-1 TMI-! Protection System Maximum Allowable Setpoints 2 11 .

2.3-2 DELETED 1

3.11 Reactor Coolant System Heatup/Cooldown Umitations 3-Sa (Applicable thru 10 EFPY) 3.1-2 Reactor Coolant Inservice Imak and Hydrostatic Test 3-5b (Applicable thru 10 EFPY) 3.1-2a Dose equivalent I-131 Primary Coolant Specific Actual 3 9b Umit vs. Percent of RATED THERMAL POWER 3.1-3 DELETED 3.5-2A DELETED thru ,

3.5-2M 3.5-1 lacore Instrumentation Specification 3-39a Axial Irnhalance Indication 3.5-2 Incore Instrunwetation Specification 3 39b j Radial Flux Tilt Indication 3.5-3 Incore Instmmentation Specification 3-39c i

3.11-1 Transfer Path to and from Cask Loading Pit 3-56b  !

4,17-1 Snubber Functional Test - Sample Plan 2 4-67 5-1 Extended Plot Plan TMI N/A j 5-2 Site Topography 5 Mile Radius N/A 5-3 Gaseous Effluent Release Points and Uquid Effluent N/A Outfall Locations 5-4 Minimum Burnup Requirements for Fuel in Region II of 5-7a the Pool A Storage Racks 5-5 Minimum Burnup Requirements for Fuel in the Pool "B" 5-7b Storage Racks vii Amendment Nos. :, !?, 2^,3^,05,'^,72, ^5, ^^,:2^,125,:34,

e2,:5^,152,157,ISS,IM,

3.1.10 CONTROL ROD OPERATION l

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i (Page 3-18b deleted) 3-1Sa Aneximent No.

3.3 EMERGENCY CORE COOLING. REACTOR BUILDING EMERGENCY COOLING AND REACTOR BUILDING SPRAY SYSTEMS i

Aeolicability Applies to the operating status of the emergency core cooling, reactor building emergency cooling, and reactor building spray systems.

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Ob_iective To define the conditions necessary to assure immediate a,allability of the emergency core cooling, reactor building emergency cooling and reactor building spray systems.

Soecification ,

l 3.3.1 The reactor shall not be made critical unless the following conditions are met:

3.3.1.1 Iniection Systems j 1

a. The borated water storage tank shall contain a minimum of 350,000 gallons of water )

i having a minimum concentration of 2,500 ppm boron at a temperature not less than 40*F. If the boron concentration or water temperature is not within limits, restore the BWST to OPERABLE within 8 hrs. If the BWST volume is not within limits, restore the BWST to OPERABLE within one hour. Specification 3.0.1 applies.

b. Two makeup pumps are operable in the engineered safeguards mode powered from independent essential buses. Specification 3.0.1 applies.

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c. Two decay heat removal pumps are operable. Specification 3.0.1 applies.
d. Two decay heat removal coolers and their cooling water supplies are operable. (See Specification 3.3.1.4) Specification 3.0.1 applies.

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e. Two BWST level instrument channels are operable. j i f. The two reactor building sump isolation valves (DH-V6A, DH-V6B) shall be either manually or remote-manually operable. Specification 3.0.1 applies. ,

l 3.3.1.2 Core Flooding System

a. Two core flooding tanks each containing 1040 i 30 ft' of borated water at 600125 psig shall be available. Specification 3.0.1 applies.
b. Core flooding tank boron concentration shall not be less than 2,270 ppm boron.
c. The electrically operated discharge valves from the core flood tank will be assured open by administrative control and position indication lamps on the engineered safeguards status panel. Respective breakers for these valves shall be open and conspicuously marked. A one hour time clock is provided to open the valve and remove power to the valve. Specification 3.0.1 applies.

Amendment No. M,98, MB, 3-21

d. One core flood tank pressure instrumentation channel and one core flood tank level instrumentation channel per tank shall be operable,
e. . Core flood tank (CFT) vent valves CF-V3A and CF-V3B shall be closed and the breakers to the CFT vent valve motor operators shall be tagged open, except when adjusting core flood tank level and/or pressure.. Specification 3.0.1 applies.

3.3.1.3 Ranctor Buildine Sorav Svstam and Panctor Buildine Fmeramney Co3]ing Svstam The following components must be OPERABLE:

a. Two reactor building spray pumps and their associated spray nozzles headers and two reactor building emergency cooling fans and associated cooling units (one in each train).

Specification 3.0.1 applies,

b. The sodium hydroxide (NaOH) tank shall be maintained at 8 ft.16 inches lower than the BWST level as measured by the BWST/NaOH tank differential pressure indicator.

'Ihe NaOH tank concentration shall be 10.01.5 weight percent (%). If the NaOH concentration is not within limits, restore to OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. If the BWST/NaOH tank level differential is not within limits, restore to OPERABLE within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

c. All manual valves in the discharge lines of the sodium hydroxide tank shall be locked open.

3.3.1.4 Cooline Water Svstems - Specification 3.0.1 applies.

a. Two nuclear service closed cycle cooling water pumps must be OPERABLE,
b. Two nuclear service river water pumps must be OPERABLE.
c. Two decay heat closed cycle cooling water pumps must be OPERABLE.
d. Two decay heat river water pumps must be OPERABLE,
e. Two reactor building emergency cooling river water pumps must be OPERABLE.

3.3.1.5 - Engineered Safeguards Valves and Interlocks Associated with the Systems in Specifications 3.3.1.1, 3.3.1.2, 3.3.1.3, 3.3.1.4 are OPERABLE. Specification 3.0.1 applies.

3.3.2 Maintenance or testing shall be allowed during reactor operation on any component (s) in the makeup and purification, decay heat, RB emergency cooling water, RB spray, CFT pressure instrumentation, CFT level instrumentation, BWST level instrumentation, or cooling water systems which will not remove more than one train of each system from service.

Components shall not be removed from service so that the affected system train is inoperable for more than 72 consecutive hours. if the system is not restored to meet the requirements of Specification 3.3.1 within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the reactor shall be placed in a HOT SHUTDOWN condition within six hours.

3-22 Amendment No.33 ,80 ,98 ,137 , W4, 499r 1

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3.4 D"ICAY HEAT REMOVAL CAPABILITY i Aeolicability Applies to the operating status of systems and components that function to remove decay heat when one or mere fuel bundles are located in the reactor vessel.

Obiective To define the conditions necessary to assure continuous capability of decay heat removal.*

Specification j 3.4.1 Reactor Coolant System temperature greater than 250*F.

3.4.1.1 With the Reactor Coolant System temperature greater than 250*F, three independent EFW l pumps and associated flow paths shall be OPERABLE ** with:

a. Two EFW pumps, each capable of being powered from an OPERABLE emergency bus, and one EFW pump capable of being powered from an OPERABLE steam supply system.

(1) With one pump or flow path inoperable, restore the inoperable pump or flow path to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in COLD SHUTDOWN within '

the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. j (2) With more than one EFW pump or flow path inoperable, restore the inoperabie pumps or flow paths to OPERABLE status within one hour or be in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the j following 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

NOTE: When EF-P-1 and EF-P-2A or EF-P-2B become inoperable due to TS surveillance, entry into this LCO may be delayed for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. I

b. Four of six turbine bypass valves OPERABLE. With more than two turbine bypass valves inoperable, restore operability of at least four turbine bypass valves within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
c. The condensate storage tanks (CS'ij O'ERABLE with a minimum of 150,000 gallons of condensate available in each CST.

(1) With a CST inoperable, restore the CST to operability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and COLD SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

(2) With more than one CST inoperable, restore the inoperable CST to OPERABLE status or be subcritical within I hour, in at least HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in COLD SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

  • These requirements supplement the requirements of Sections 3.1.1.1.c, 3.1.1.2,3.3.1 and 3.8.3.
    • HSPS operability is specified in Section 3.5.1.

3-25 Amendment No.4 ,78 ,98 ,119 ,124 ,162 ,-19%

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I j 3.5.2 CONTROL ROD GROUP AND POWER DISTRIBUTION LIMITS I j Annlicability I His specification applies to porver distribution and operation of control ro6s during power operation.  ;

Obiective l l To assure an acceptable core power distribution during power operation, to set a limit on potential l reactivity insertion from a hypothetical control rod ejection, and to assure core subcriticality after a  !

i_ reactor trip. l Specification

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} 3.5.2.1 He available shutdown margin shall not be less than one percent AK/K with the highest l worth control rod fully withdrawn. <

3.5.2.2 Operation with inoperable rods:

a. Operation with more than one inoperable rod as defined in Specification 4.7.1 and 4.7.2.3 in the safety or regulating rod banks shall not be permitted Verify SDM 2:l %

, Ak/k or initiate boration to restore within limits within I hour. The reactor shall be' brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

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b. If a control rod in the regulating and/or safety rod banks is declared inoperable in the withdrawn position as defined in Specification Paragraph 4.7.1.1 and 4.7.1.3, an evaluation shall be initiated immediately to verify die existence of one percent Ak/k hot shutdown margin. Boration may be initiated to merease the available rod worth either to compensate for the worth of the inoperable rod or until the regulating banks are fully withdrawn, whichever occurs first. Simultaneously a program of exercising the remaining regulating and safety rods shall be initiated to verify operability,
c. If within one hour of determination of an inoperable rod as defined in Specificati on 4.7.1, and once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter, it is not determined that a one percent ak/k hot shutdown margin exists combining the worth of the inoperable rod with each of the other rods, the reactor shall be brought to the HOT SHUTDOWN condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> until this margin is established.
d. Following the determination of an inoperable rod as defined in Specification 4.7.1, all rods shall be exercised within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and exercised weekly until the rod problem is solved,
e. If a control rod in the regulating or safety rod groups is declared inoperable per 4.7.1.2 and cannot be aligned per 3.5.2.2.f, power shall be reduced to diO% of the thermal power allowable for the reactor coolant pump combination within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and the overpower trip setpoint shall be reduced to :f/0% of the thermal power allowable within 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Verify the potential ejected rod worth (ERW) is within the assumptions of the ERW analysis and verify peaking factor (Fq(Z) and ph) limits per the COLR have not been exceeded within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. l l

l v 3-33 Amendment No. M, WA (5-18-76) l I

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f. If a control rod in the regulating or axial power shaping groups is declared inoperable i per Specification 4.7.1.2, operation may continue provided that within I hour the rods l l in the group are positioned such that the rod that was declared inoperable is maintained within allowable group average position limits of Specification 4.7.1.2.
g. If the inoperable rod in Paragraph "e" above is in groups 5,6,7, or 8, the other rods in the group may be trimmed to the same position. Normal operation of 100 percent of the thermal power allowable for the reactor coolant pump combination may then continue provided that within I hour the rod that was declared inoperaMe is maintained within l j allowable group average position limits in 3.5.2.5.

3.5.2.3 The worth of single inserted control rods during criticality is limited by the restriction of l Specification 3.1.3.5 and the Control Rod Position Limits defined in Specification 3.5.2.5.

3.5.2.4 Quadrant Tilt:

a. Except for physics tests, the quadrant tilt, as determined using the full incore system (FIS), shall not exceed the values in the CORE OPERATING LIMITS REPORT.

The FIS is OPERABLE for monitoring quadrant tilt provided the number of valid symmetric string individual SPND signals in any one quadrant is not less than the limit in the CORE OPERATING LIMITS REPORT.

b. When the full incore system is not OPERABLE and except for physics tests quadrant tilt .

as determined using the power range channels for each quadrant (out of core detector j system)(OCD), shall not exceed the values in CORE OPERATING LIMITS REPORT.

c. Whe'n neither detector system above is OPERABLE and, except for physics tests, quadrant tilt as determined using the minimum incore system (MIS), shall not exceed <

I the values in the CORE OPERATING LIMITS REPORT.

d. Except for physics tests if quadrant tilt exceeds the tilt limit, allowable power shall be l reduced 2 percent for each I percent tilt in excess of the tilt limit. For less than four pump operation, thermal power shall be reduced 2 percent below the thermal power i

allowable for the reactor coolant pump combination for each I percent tilt in excess of the tilt limit.

e. If quadrant power tilt exceeds the tilt limit then within a period of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, the quadrant power tilt shall be reduced to less than the tilt limit except for physics tests, or the following verifications and/or adjustments in setpoints and limits shall be made:
1. Verify Fn(Z) and ph are within limits of the COLR once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and restore QPT to :ssteady state limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or perform steps 2,3, &4 below.

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l 3-34

, Amendment No. H, 29, 39, 40, 60, 90, +26, 442, +50; i

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2. He prda*=1 system reactor powerfunhalance mvelope trip setpoints shall be reduced l 2 percent in power for each I percent tilt, in excess of the tilt limit, or when thermal power is equal to or less than 50% full power with four reactor coolant pumps runmng,

! set the nuclear overpower trip seapomt equal to or less than 60% full power i

3. He control rod group withdrawal limits in the CORE OPERATING LIMITS l REPORT shall be reduced 2 percent in power for each I percent tilt in excess of the tilt limit.

l 4. The operatxmal imbalance limits in the CORE OPERATING LIMITS REPORT shall }

be reduced 2 percent in power for each I percent tilt in excess of the tilt limit.

f. Except for physics or dia,.c dc testmg, if quadrant tilt is in excess of the maximum tilt limit l defined in the CORE OPERATING LIMITS REPORT and using the applicable detector i systan defined in 3.5.2.4.a, b, and c above, reduce thermal power to $15% FP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. l Diagnostic testmg dunng power vi.Gon with a quadrant tilt is p..iattad provideC Uat the l thermal power allowable is twW as stated in 3.5.2.4.d above f

j g. Quadrant tilt shall be momtored on a nunimum frequency ofonce every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> whm the l

QPT alarm is inoperabic and every 7 days when the alarm is operable durug power operatum above 15 pera:nt of rated power When QPT has been restored to ssteady state limit, verify hourly for 12 consecutive hours, or until venfied acceptable at h95% FP.

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3-34a Air.cz4reet No. 39, 38, 39; 40, 46, 60, 430, E!6, 442, MG, #2;

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! 3.5.2.5 Control Rod Positions l l

a. Operating rod group overlap shall not exceed 25 percent .t5 percent, between two  :

sequential groups except for physics tests. f

b. Position limits are specified for regulating control rods. Except for physics tests or ,

exercising control rods, the regulating control rod insertion / withdrawal limits are l' specified in the CORE OPERATING LIMITS REPORT.

1. If regulating rods are inserted in the restricted operating region, corrective measures shall be taken immediately to achieve an acceptable control rod position.

Acceptable control rod positions shall be attained within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and Fn(Z) and p,shall be verified within limits once every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, or power shall be reduced to ,

spower allowed by insertion limits. l

2. If regulating rods are inserted in the unvc=Lable operating region, initiate l boration within 15 minutes to restore SDM to h1 MK/K, and restore regulating l rods to within restricted region within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce power to spower allowed by rod insertion limits.
c. Safety rod limits are given in 3.1.3.5.

3.5.2.6 'Ihe control rod drive patch panels shall be locked at all times with limited access to be authorized by the Director, Operations and Maintenance, TMI.

l 3.5.2.7 Axial Power imbalance:

a. Except for physics tests the axial power imbalance, as determined using the full incore system (FIS), shall not exceed the envelope defined in the CORE OPERATING LIMITS REPORT.

The FIS is operable for monitoring axial power imbalance provided the number of valid self powered neutron detector (SPND) signals in any one quadrant is not less than the limit in the CORE OPERATING LIMITS REPORT.

b. When the full incore detector system is not OPERABLE and except for physics tests axial power imbalance, as determined using the power range channels (out of core detector system)(OCD), shall not exceed the envelope defined in the CORE OPERATING LIMITS REPORT.
c. When neither detector system above is OPERABLE and, except for physics tests axial power imbalance, as determined using the minimum incore system (MIS), shall not exceed the envelope defined in the CORE OPERATING LIMITS REPORT.
d. Except for physics tests if axial power imbalance exceeds the envelope, corrective measures (reduction of imbalance by APSR movements and/or reduction in reactor I

power) shall be taken to maintain operation within the envelope. Verify Fn(Z) and p",

are within limits of the COLR once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> when not within imbalance limits.

l 3-35  ;

l i Amendment No. M, M, 99, 38, 39, 50, He, M6, M9, MG, H9; i

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e. If an acceptabl] axial power imbalance is not achieved eithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, reactor pow:r shall be reduced to :c40% FP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. j i
f. Axial power imbalance shall be monitored on a minimum frequency of once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when axial power imbalance alarm is OPERABLE, and every I hour when imbalance alarm is inoperable during power operation above 40 percent of rated power.

3.5.2.8 A power map shall be taken at intervals not to exceed 31 effective full power days using the l incore instrumentation detection system to verify the power distribution is within the limits shown in the CORE OPERATING LIMITS REPORT.

l Bases The axial power imbalance, quadrant power tilt, and control rod position limits are based on LOCA analyses which have defined the maximum linear heat rate. These limits are developed in a manner that ensures the initial condition LOCA maximum linear heat rate will not cause the maximum clad temperature to exceed 10 CFR 50 Appendix K. Operation outside of any one limit alone does not necessarily constitute a situation that would cause the Appendix K Criteria to be exceeded should a LOCA occur. Each limit represents the boundary of operation that will preserve the Acceptance Criteria even if all three limits are at their maximum allowable values simultaneously. The effects of the APSRs are included in the limit development. Additional conservatism included in the limit development is introduced by application of:

a. Nuclear uncertainty factors
b. Thermal calibration uncertainty
c. Fuel densification effects
d. Hot rod manufacturing tolerance factors 1
e. Postulated fuel rod bow effects
f. Peaking limits based on initial condition for Loss of Coolant Flow transients.

The incore instrumentation system uncertainties used to develop the axial power imbalance and quadrant tilt limits accounted for various combinations of invalid Self Powered Neutron Detector (SPND) signals. If the number of valid SPND signals falls below that used in the uncertainty analysis, then another system shall be used for monitoring axial power imbalance and/or quadrant tilt.

For axial power imbalance and quadrant power tilt measurements using the incore detector system, the mimmum incore detector system consists of OPERABLE detectors configured as follows:

1 Axial Power imbalance

a. Three detectors in each of three strings shall lie in the same axial plane with one plane in each axial core half,
b. The axial planes in each core half shall be symmetrical about the core mid-planes.
c. The detectors shall not have radial symmetry.

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! Quadrant Power Tilt

a. Two sets of four detectors shall lie in each core half. Each set of four shall lie in the same axial plane. The two sets in the same core half may lie in the same axial plane.
b. Detectors in the same plane shall have quarter core radial symmetry.

3-35a l

Amendment No. H, 99, 38, 39, 50, MO, M6, 449, MO, M7, MB, 1

A system of 52 incore fhtx detector assemblies with seven hars per assembly has been prowaed primarily for fuel management purposes 'Ihe system includes data display and record functions and is also used for out of-core nuclear instrumentaten calibration and for core power distribution w i';cetice.

l a. 'Ihe outef-core instrumentaten calibration includes:

l 1. Calibrations of the split h s at initial reactor startup, durug the power mlation

! Program, and ps-M thereafter.

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2. A m..peisce check with the incore mstrumer*=tian in the event one of the four.

out ofcore power range detector assemblics gives abnormal readmgs dunng operation

3. Cu 0... 4cm that the out ofcore axial power splits are as avp~*M
b. Core power distributen venfication includes:

b 1. Measurc.aent at low power initial reactor startup to check that pour distributon is consistent with <=WW l

2. Subsequent checks dunng operaton to ensure that power distributen is consistent with calculations
3. Indmation of power distributen in the event that abnormal situatons occur dunng nector operaten.
c. 'Ihe muumum requinment for 23 individual moore detectors is based on the followirg:

1

1. An W* axial imbalance irmiic4ies can be obtamed with nine individual detectors Figure 3.5-1 shows a typical set of three detector stnngs with three &tectors per string that will indsate an axial imhalance The three deterrnr stnngs are the center one, one from the inner ring of sy ....si; cal stnngs and one from the outer ring of sy. .wical strmgs
2. Figure 3.5-2 shows a typical detecten scheme whch will indsate the radial power distributen with 16 individual detectors. The rendungs from two detectors in a radial quadrant at either plane can be - .p M with readmgs from the other quadrants to measure a radial flux tilt.  ;

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3. Figure 3.5-3 combines Figures 3.5-1 and 3.5-2 to illustrate a typical se of 23 individual detectors that can be specified as a muumum for axial imbalance i

drh...;..eion and radial tilt indication, as well as for the hn . ides ofgross core power distributons l

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! 3-35b l Amendment No. 450,45h i

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1 1he 25 i S% overlap between successive control rod groups is allowed since the worth of a rod is lower at the upper and lower part of the stroke Control rods are arranged in groups or banks de6ned as follows:

i j Group Function J

l Safety 4 2 Safety 3 Safety l

4 Safety 5 R eil=6ag 2

6 F Tit =6ag 7 F eitadag l

8 APSR(axialp::wcrshaping rod bank) i Control red groups are withdrawn in sequence 1,ey.wg with group 1. Groups 5,6 and 7 are ossiopped 25

%. The normal position at power is for group 7 to be partially inserted.

1 The rod position limits are based on the most hnuting of the followmg three criteda: ECCS power peakmg, shutdown magpn, and potatial ejected rod worth. As di--A above, cu...yliss with the ECCS power pealong critenon is ensured by the rod position limits. & muumum available rod worth, consistent with the rod position limits, provides for actueving hot shutdown by reactor trip at any time, assunung the lughest worth control rod that is withdrawn remains in the full out position (Reference 1). The rod position limits also ensure that inserted rod groups will not contam single rod worths greater than: 0.65% delta k/k at rated power.

These values have been shown to be safe by the safety analysis of the h>W1 rod ejection accident (Reference 2). A maximum single inserted control rod worth of 1.0% delta k/k is allowed by the rod position lirrits at hot zero power. A single inserted contiel rod worth 1.0% delta k/k at t r =.g of hfe, hot, zero power would result in a lower transient peak thermal power and, therefore, less severe mvisu e=1 ena-v=- than 0.65% delta k/k ejected rod worth at rated power The plant computer will scan for tilt and imh 1m anal will satisfy the techincal spacinannn requirements if the computer is out of service, then manual abitanna for tilt above 15 percent power and imbalance above 40 percent power must be performed as speci6ed until the computer is returned to service.- l l

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3-36 Airedrst No. 4-7,29,39,40, M,424,442,4M,4Mr ,

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! Reduction of the nuclear overpower trip setpoint to 60% full power whim thermal power is equal to or less l than 50% full power maintains both core protection and an operability magpn at reduced power similar to that l l at full power.

Dunng the physics testing program, the high flux trip setpomts are narJni+.dvely set as follows to assure an  :

WW safety marginis provided i Test Power Test Setooint l

O 4% .

s80 90 %  !

>80 105.1 %

)

I REFERENCES (1) UFSAR, Secten 3.2.2.1.2 " Reactivity Control Distribution" l

(2) UFSAR, Secton 14.2.2.2 " Rod Ejection Accident" i i

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3-36a Amendment No. 39,4M,442,460,4M; 1

., .~su_,.... . . ~ . - ~ . - - ..~,.s s.-. ..-n-- - - ~ ~ . . = . - . .--+..u. .-a.+ .~n,...~.a- .a u - s ~--- a . > . . .- -

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3.5.4 INCOREINSTRUMENTATION i

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(Page 3-39 deleted) i 3-38 Amenenent No. 4M;

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Es pageintentionally bbek s

3-40 Anadred No.45h

TABLE 4,12 MINIMUM EQUIPMENTTEST FREQUENCY ltsm Its Freauency

1. Control Rods Rod drop times ofall Each Refuelmg shutdown fulllength rods ,
2. Control Rod Movernent ofeach rod Every 92 days,when l Movement reactoris critical
3. Pressuruer Setpomt In whc with the Safety Valves Inservice Testmg Program
4. Main Steam Setpomt In a &c withthe Safety Valws Insernce Testmg Program
5. Refuehng System Functonal Start ofeach Interlocks refueling penod
6. Main Steam (See Sectum 4.8)

Isoladon Valves

7. Reactor Coolant Evaluate Daily,when reactor System Immage coolant system Lw.iare is greater than 525'F
8. (oeieted) - -
9. Spent Fuel Functional Each refueling penod Cooling System prior to fuelhandhng
10. Intake Pump (a) Silt Accumulabon- Not to exceed 24 months House Floor Visualine (Elevatum ofIntake Pump 262 R. 6 in.) House Floor (b) Silt AcamaW Quarterly Measurement of Pump House Flow
11. Pressurner Block Funcuonal* Quarterly Valve (RC-V2)
  • Funcuan shall be d=--%.ied by w. tug the valve through one complete cycle of full travel.

4-8 Amendment No.5568 ,78,149,I75, M8;

O TABLE 4.1-3 E

@ MINIMUM SAMPLING FREQUENCY

% FREOUENCY

" ITEM CHECK E

F 1. Reactor Coolant a. Specific Actisity Detemun-aticato compare tothe Atleast once each 7 days dunng POWER OPERATION, HOT STANDBY,

'l y 100/CpCi/gmlimit STARTUP,and HOT SHUTDOWN.

b. Isotopic Analysis for DOSE i) I per 14 days dunng poweroperations w EQUIVALENTl-131 Concentra- ii) One Sample between 2 and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

~

tion following a1NERMAL POWER change E exceeding 15%ofthe RATED THERMAL t

  • POWERwithin aone hour penod dunng power operation, start-up, and hot staruby.

iii) # Once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />,whenew:r the f* specific activity exceeds 1.0 pCi/ 3

  • gram DOSE EQUIVALENTI-131 or 100/EpCi/gramdunng allmodes but refueling
c. Radinen; cal for E I per 6 months
  • dunns power ILo-Jus -

operation ,

d. Chemistry (C1, F and 02 ) 5 tunes / week whenT,is greater

. than 200*F.

e. Baron concentration 2 times / week
f. Tritium Radioactisity Monthly
2. Borated Water Storage Boron conwnuat:an Weeklyand after each makeup when f TankWater Sample reactor coolant system pressureis greater than 300 psig or T,is greaterthan200 F. ,
3. Core FloodmgTank Boron concentration Monthly and after each makeup whai RCS Water Sample pressure is greater than 700 psig. .

'g , s. o 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS J

Aeolicability i

Applies to the surveillance of the control rod system.

[

Obiective i

To assure operability of the control rod system.

Specification 4.7.1.1 De control rod trip insertion time shall be measured for er;.;h control rod at either full flow or no flow conditions following each refueling outage prior to

return to power. He maximum control rod trip insertion time for an operable control rod drive mechanism, except for the axial power shaping rods (APSRs), l from the fully withdrawn position to 3/4 insertion (104 inches travel) shall not -

exceed 1.66 seconds at hot reactor coolant full flow conditions or 1.40 seconds for the hot no flow conditions (Reference 1). For the APSRs it shall be i demonstrated that loss of power will not cause rod movement. If the trip j insertion time above is not met, the rod shall be declared inoperable.

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! 4.7.1.2 If a control rod is misaligned with its group average by more than an indicated

nine inches, the rod shall be declared inoperable and the limits of Specification 3.5.2.2 shall apply. ne rod with the greatest misalignment shall be evaluated 3

first. The position of a rod declared inoperable due to misalignment shall not be ,

included in computing the average position of the group for determining the l operability of rods with lesser misalignments.

} 4.7.1.3 If a control rod cannot be exercised, or if it cannot be located with absolute or

relative position indications or in or out limit lights, the rod shall be declared to

[

be inoperable.

i BAHE ,

j' ne control rod trip insertion time is the total elapsed time from power interruption at the j control rod drive breakers until the control rod has actuated the 25% withdrawn refererse j swi:ch during insertion from the fully withdrawn position. He specified trip time is based

upon the safety analysis in UFSAR, Chapter 14 and the Accident Parameters as specified therein.

i Each control rod drive mechanism shall be exercised by a movement of a minimum of 3% j of travel at a minimum of every 92 days. This requirement shall apply to either a partial or fully withdrawn control rod at reactor operating conditions. Exercising the drive

.r mechanisms in this manner provides assurance of reliability of the mechanisms.

f a  !

4-48 Amendment No. +5h t

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3 4.7.2 CONTROL ROD PROGRAM VERIFICATION (Group vs. Core Positions) ,

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This page intentionally left blank 4-50 Amendment No. 46h

Enclosure 3 Certificate of Service for TMI-l Technical Specification Change Request No. 253 l

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

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION 4 IN THE MATTER OF GPU NUCLEAR, Inc. j DOCKET NO. 50-289 LICENSE NO. DPR-50 i

CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No. 253 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with executives of Londonderry Township, Dauphin County, Pennsylvania; Dauphin County, Pennsylvania; and the Pennsylvania Department of Environmental Resources, Bureau of Radiation Protection, by deposit in the ' United States mail, addressed as follows:

Mr. Darryl LeHew, Chairman Mr. Russell L. Sheaffer, Chairman

. Board Supervisors of _ Board of County Commissioners i Londonderry Township of Dauphin County 7 R.D, #1, Geyers Church Road P.O. Box 1295 -

Middletown, PA 17057 Harrisburg,PA 17108 4

Director, Bureau of Radiation Protection PA Department of Environmental Resources ,

P.O. Box 2063 i 4

Harrisburg,PA 17120 - j 1

ATTN: Mr. Stan T. Maingi l

GPU NUCLEAR CORPORATION BY: Ie -

^ Vice President and Director, TMI  !

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TE: l_ s M ,

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