ML20128H412

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Errata to Proposed Ts,Adding Revised Table of Contents & Making Minor Editorial Corrections
ML20128H412
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 10/03/1996
From:
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20128H409 List:
References
NUDOCS 9610090334
Download: ML20128H412 (11)
v • d • e


Text

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l 6710-96-2317 ENCLOSURE 2 l

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9610090334 961003 PDR ADOCK 05000289 P PDR

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TAllLE OF CONTENTS

, l Section E; igg 2 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2-1 2.1 Sgny_ Limits. Reactor Core 2-1 2.2 Safety Limits. Reactor System Pressurg 2-4 2.3 Limitine Safety System Settines. Protecties ,

Instrumentation 2-5 l 3 LIMITING CONDITIONS FOR OPERATION 3-1 l 3.0 Os_n.gnUction n Ref9 rements i 3-1 i 3.1 Eractor Coolant S stem 3-!a i 3.1.1 Operatiomd Coraponents 3-la 3.1.2 Pressurization, lleatup and Cooldown Limitations 3-3 3.1.3 Minimum Conditions for Criticality 3-6 3.1.4 Reactor Coolant System Activity 3-8 l 3.1.5 Chemistry (DELETED) 3-10 3.1.6 Leakage 3-12 4

3.1.7 Moderator Temperature Coefficient of Reactivity 3-16 3.1.8 Single Loop Restrictions 3-17 3.1.9 Low Power Physics Testing Restrictions 3-18 3.1.10 Control Rod Operation 3-18a 3.1.11 Reactor Internal Vent Valves (DELETED) 3-18e 3.1.12 Pressurizer Power Operated Relief Valve (PORV) and Bhick Valve 3-18d 3.1.13 Reactor Coolant System Vents (DELETED) 3-18f 3.2 Deleted 3-19 3.3 Emergency Core Cooline. Reactor Building. Emergency Cooling and Reactor Building Sprav Systems 3-21 3.4 Decay Heat Removal Capability 3-25 3.4.1 Reactor Coolant System Temperature Greater than 250 F 3 25 3.4.2 Reactor Coolant System Temperature 250 F or Less 3-26 3.5 Instrumentation Systems 3-27 3.5.1 Operational Safety Instrumentation 3-27 3.5.2 Control Rod Group and Power Distribution Limits 3-33 3.5.3 Engineered Safeguards Protection System Actuation Setpoints 3-37 3.5.4 Incore Instrumentation 3-38 3.5.5 Accident Monitoring Instrumentation 3-40a 3.5.6 Deleted 3-40f 3.6 Reactor Buildinc 3-41 3.7 1)mit Electrical Power System 3-42 3.8 Fuel Loading and Refuelin.g 3-44

3.9 Deleted 3-46 3.10 3-46 l

Miscellaneous Radioactive Materials Sources (DELETED) 3.11 Ilandline ofInadiated Fuel 3-55 3.12 Reactor Buildine Polar Crane 3-57 3.13 Secondary System Activity 3-58 3.14 Flood 3-59 3.14.1 Periodic Inspection of the Dikes Around TMI 3-59 3.14.2 Flood Condition for Placing the Unit in 110t Standby 3-60 3.15 Air Treatment Systems 3-61 3.15.1 Emergency Control Room Air Treatment System 1 61 3.15.2 Reactor Building Purge Air Treatment System 3-62a 3.15.3 Auxiliary and Fuel Handling Building Air Treatment System 3-62c 3.15.4 Fuelllandling Building ESF Air Treaunent System 3-62e ii Amendment NoJ9,72,78,97,98,J19,J 22,JM ,14I ,J67,J82,196

l TAlli E OF CONTENTS

. Section M 3.16 l

SHOCK SUPPRESSORS (SNUBBERE (DELETED) 3-63 3.17 REACTOR BUILDING AIR TEMPERATUJE 3-80 3.18 FIRE PROTECTION (DELETED) 3-86 3.19 CONTAINMENT SYSTEMS 3-95 3.20 SPECIAL TEST EXCEI'FIONS (DELETED) 3-95a 3.21 RADIOACTIVE EFFLUENT INSTRUMENTATION (DELETED) 3-96 3.21.1 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION (DELETED) 3-96 3.21.2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT 3-96 MONITORING INSTRUMENTATION (DELETED) 3.22 RADIOACTIVE EFFLUENTS (DELETED) 3-96 3.22.1 LIQUID EFFLUENTS (DELETED) 3-96 3.22.2 GASEOUS EFFLUENTS (DELETED) 3-96 3.22.3 SOLID RADIOACTIVE WASTE (DELETED) 3-96 3.22.4 TOTAL DOSE (DELETED) 3-96 l 3.23 RADIOLOGICAL ENVIRONMENTAL MONITORING (DELETED) 3-96 I 3.23.1 MONITORING PROGRAM (DELETED) 3-96  ;

3.23.2 LAND USE CENSUS (DELETED) 3-96  :

3.23.3 INTERLABORATORY COMPARISON PROGRAM (DELETED) 3-96 i 3.24 REACTOR VESSEL WATER LEVEL 3-128 4 SURVEILLANCE STANDARQS 4-1 4.1 QJtBATIONAL S AFET_Y_BEVIEW 4-1 4.2 REACTOR COOLAN T SYSTEM INSERVICE INSPECTION 4-11 4.3 TESTING FOLLQ_ WING OPENING OF SYSTE_M (DELETED) 4-13 1 4.4 REACTOR BUILDING 4-29 4.4.1 CONTAINMENT LEAKAGE TESTS 4-29 4.4.2 STRUCTURAL INTEGRITY 4-35 4.4.3 DELETED 4-37 4.4.4 HYDROGEN RECOMBINER SYSTEM 4-38 )

4.5 EMERGENCY LOADING SEOUENCE AND POWER TRANSFER, 4-39 i EMERGENCY CORE COOLING SYSTEM AND GCTOft l BUILDING COOLING SYSTEM PERIODIC TrJILN_Q 4.5.1 EMERGENCY LOADING SEQUENCE 4-39 4.5.2 EMERGENCY CORE COOLING SYSTEM 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLA'l:ON SYSTEM 4-43 4.5.4 DECAY llEAT REMOVAL SYSTEM LEAKAGE 4-45 4.6 EMERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 BEACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS 4-48 4.7.2 CONTROL ROD PROGRAM VERIFICATION 4-50

-iii-Amentiment72.8J ,JM ,J29,J37,Jf 6.J47,JH 19J .J97.498

TABl.E OF CONTENTS

! . Sections Bige I

4.8 MAIN STEAM ISOLATION VALVES 4-51 7

4.9 DECAY HEAT REMOVAL CAPABILITY - PERIODIC TESTING 4-52 4.9.1 Emergency Feedwater System Periodic Testing 4-52 (Reactor Coolant Temperature Greater Than 250 F) 4.9.2 Decay Heat Removal Capability - Periodic Testing 4-52a (Reactor Coolant Temperature 250 F or Less) i 4.10 BEACTIVITY ANOMALIES 4-53 4.11 l

l' EfACTOR COOLANT SYSTEM VENTS (DELETED) 4-54 4.12 AIR TREATMENT SYSTEMS 4-55 4.12.1 Emergency Control Room Air Treatment System 4-55 4.12.2 Reactor Building Purge Air Treatment System 4-55b

4.12.3 Auxiliary and Fuel Handling Building Air Treatment System 4-55d

! 4.13 RADIOAQVE MATERIALS SOURCES SURVEILLANCE (DELETED) 4-56

4.14 DELETED 4-56

! 4.15 MAIN STEAM SYSTEM INSERVICE INSPEGON 4-58 l 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE (DELETED) 4-59

4.17 SHOCK SUPPRESSORS (SNUBBERS)(DELETED) 4-59 l 4.18 FIRE PROTECTION SYSTEMS (DELETED) 4-72
4.19 OISGIUBE INSERVICE INSPECTION 4-77

, 4.19.1 Stetun Generator Sample Selection and Inspection Methods 4-77 j 4.19.2 Steam Generator Tube Sample Selection and Inspection 4-77 4.19.3 Inspection Frequencies 4-79 4.19.4 Acceptance Criteria 4-80 4.19.5 Reports 4-81 q 4.20 REACTOR BUILDING AIR TEMPERATURE 4-86 j 4.21 R ADIOACTIVE EFFLUENT INSTRUMENTATION (DELETED) 4-87 j 4.21.1 Radioactive Liquid Effluent Instrumentation (DELETED) 4-87

) 4.21.2 Radioactive Gaseous Process and Effluent Monitoring Instrumentation (DtiLETED) 4-90

! 4.22 EADIOACTIVE EFFLUENTS (DELETED) 4-95 4.22.1 Liquid Effluents (DELETED) 4-95 1 4.22.2 Gaseous Effluents (DELETED) 4-101

{ 4.22.3 Solid Radioactive Waste (DELETED) 4-107

! 4.22.4 Total Dose (DELETED) 4-108 l 4.23.1 Monitoring Program (DELETED) 4-117 4.23.2 Land Use Census (DELETED) 4-121 4.23.3 Interlaboratory Comparison Program (DELETED) 4-122 i

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! iv Amendment No.1I ,28,7@ .4J 47.55,72,7$ .95,97,119,J22,J29,J77,J# ,4H 2

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. TABI E OF CONTENTS Sectiori Pagg 5 DESIGN FEATURFJ 5-1 5.1 SHE 5-1 5.2 CONTAINMENT (DELETED) 5-2 5.2.1 REACTOR BUILDING (DELETED) 5-2 5.2.2 REACTOR BUILDING ISOLAT10N SYSTEM (DELETED) 5-2 5.3 REACTOR 5-4 5.3.1 REACTOR CORE 5-4 5.3.2 REACTOR COOLANT SYSTEM 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.I NEW fTJEL STORAGE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 5.5 AIR INT AKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 6 ADMINISTRATIVE CONTR_Q.LS 6-1 6.1 RESPONSIBLU]3' 6-1 ti.2 ORGANIZATION 6-1 6.2.1 CORPORATE 6-1 6.2.2 UNIT STAFF 6-1 6.3 UNIT STAFF OUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND AUDIT 6-3 6.5.1 TECHNICAL REVIEW AND CONTROL 6-4 6.5.2 INDEPENDENT SAFETY REVIEW 6-5 6.5.3 AUDITS 6-7 6.5.4 INDEPENDENT ONSITE S AFETY REVIEW GROUP 6-8 6.6 REPORTABLE liVENT ACTION 6-10 6.7 S AFETY LIMIT VIOLATION 6-10 6.8 PROCEDURES AND PROGRAMS 6-11 6.9 REPORTINGEJiOUIREMENTS 6-12 6.9.1 ROUTINE REPORTS 6-12 6.9.2 DELETED 6-14 6.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6-17 6.9.4 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6-18 6.9.5 CORE OPER ATING LIMITS REPORT 6-19 6.10 RECORD RETENTION 6-20 -

6.11 RADIATION PROTECTION PROGRAM 6-22 6.12 HIGH RADIATION ARFA 6-22 6.13 PROCESS CONTROL PROGRAM 6-23 6.14 OFFSITE DC SE CALCULATION MANUAL (ODCM) 6-24 6.15 DELETED 6-24 6.16 POST ACCIDENT SAMPLING PROGRAMS 6-24 NL! BEG 0737 (II.B.3. II.F.I.2) 6.17 hLAJOR CHANGES TO RADIO ACTIVE W ASTE TREATMENT SYSTEMS 6-25

.v.

Amendment No.JJ J7.72.77.)29.IM , +B

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LIST OF FIGURES 4

. FIGURE TITI,E PAGE 2.1-1 Core Protection Safety Limit TMI-l 2-4a 2.1-2 DELETED 2.1-3 Core Protection Safety Bases TMI-l 2-4c 2.3-1 TMI-I Protection System Maximum Allowable Setpoints 2-11 l 2.3-2 DELETED f 3.1 1 Reactor Coolant System Heatup/Cooldown Limitations 3-Sa j (Applicable thru 10 EFPY) j 3.1-2 Reactor Coolant Inservice Leak and Hydrostatic Test 3-Sb j (Applicable thru 10 EFPY) i 1 3.1-2a Dose equivalent 1-131 Primary Coolant Specific Actu:d 3-9b Limit vs. Percent of RATED THERMAL POWER i

! 3.1-3 Limiting Pressure vs Temperature Curve for 3-18h s 100 STD cc/ Liter H30 '

t 3.5-2A DELETED thru 3.5-2M i

! 3.5-1 Incore Instrumentation Specification 3-39a l AxialImbabace Indication l

3.5-2 Incore Instrumentation Specification 3-39b Radial Flux Tilt Indication l

i j 3.5-3 Incore Instrumentation SpeciUcation 3-39c

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

. 4.17-1 Snubber Functional Test - Sample Plan 2 4-67 l i

i 5-1 Extended Plot Plan TMI N/A 3

5-2 Site Topography 5 Mile Radus N/A i

j 5-3 Gascous Effluent Release Points and Liquid Efiluent DELETED Outfall Locations (DELETED) 5-4 Relationship Between Initial Enrichment and Acceptable 5-7a j Fuel Burnup (Spent Fuel Pool A - Region II) i i vii 1

Amendment Nos.ll ,17,29,79,AJ .50.59,72,106,J09,J20,126,lM ,142,150, JM , J67, #8 1

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' 3.8.9 The reactor building purge system, including the radiation monitors which initiate purge isolation, shall be tested and verified to be operable no l more than one week prior to refueling operations. l l

3.8.10 Irradiated fuel shall not be removed from the reactor until the unit has been suberitical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Ess l 4

Detailed written procedures will be available for use by refueling personnel. These procedures, the above specifications, and the design of the fuel handling equipment as described in Section 9.7 of the FSAR incorporating built-in interlocks and safety features, provide assurance that no incident could occur during the refueling operations that would result in a hazard to public health and safety. If no change is being made in core geometry, one flux monitor is sufficient. This permits maintenance on the instrumentation. Continuous monitoring of the neutron flux provides immediate indication of an unsafe condition. The decay heat removal pump is used to maintain a uniform boron concentration. The shutdown margin indicated in Specification 3.8.4 will keep the core subcritical, even with all control rods withdrawn from the core (Reference 1). The boron concentration will be sufficient to maintain the core kca s 0.99 if all the control rods were removcd from the core, however, only a few control rods will be removed at any one time during fuel shuffling and replacement. The k,n with all rods in the core and with refueling boron concentration is approximately 0.9.

The specification requiring testing Reactor Building purge termination is to verify that these components will function as required should a fuel handling accident occur which resulted in the release of significant fission products. )

l Specification 3.8.10 is required as the safety analysis for the fuel handling accident was based on l the assumption that the reactor had been shutdown for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (Reference 2). l REFERENCES (1) UFS AR, Section 14.2.2.1 " Fuel Handling Accident" (2) UFSAR, Section 14.2.2.l(2) " FHA Inside Containment" 3-45 Amendment No.157 , W8

5.0 DESIGN FEATURES 5.1 S'IT.E Applicability 4

Applies to the location and extent of the exclusion boundary and low population zone.

Obiective To define the above by location and distance description.

Specification 5.1.1 The Three Mile is'and Nuclear Station Unit 1 is located in an area of low population density about ten miles southeast of Harrisburg, PA. It is in Londonderry Township of Dauphin County, Pennsylvania, about two and one-half miles north of the southern tip of Dauphin County, where Dauphin is coterminal with York and Lancaster Counties. The station is located on an island approximately three miles in length situated in the Susquehanna River upstream from York Haven Dam. Figure 5.1 is an extended plot plan of the site showing the plant orientation and immediate surroundings. The Exclusion Area as defined in 10 CFR 100.3, is a 2,000 ft. radius, including portions of Three Mile Island, the river surface around it, and a portion of Shelley Island, which is owned by Met Ed. The minimum distance of 2,000 ft, occurs on the shore of the mainland in a due easterly direction from the plant as shown on Figure 5.1 for the Exclusion Area The minimum distance to the outer boundary of the low population zone is two miles as shown on T.S. Figure 5-2, which also depicts the site topography for a radius of five miles.

5-1 Amendment No. 72,437,449

Photocell detec' tors - Operation: Normally open gate circuit closes when ultraviolet radiation from a flame,or are is present within the detector cone of vision. Performance: The detector

-arrangement with regard to sensitivity, Spectral response, and cone of vision shall be compatible with the application,in that, the response times will be ,afficiently rapid to permit suppression and or inerting of the deflagration, as applicable. Explosive Pressure Detectors - Operation: Nonnally l open contact closes when the static pressure setting is exceeded. Range: 0-5 psi; field adjustable.

Temperature detectors: Operation - The detectors operate when their temperature reaches or exceeds 190 F 10 F; and whenever the surrounding air temperature reaches a temperature above which a fire condition is considered to exist under all conditions of rates of rise as specified by Underwriters Laboratories Tests. Lpor detectors: Operation - Ionization type which functions when surrounding air reaches the alarm setpoint.

I Smoke Detectors: Ionization type which will respond to an alarm condition when products of 1 combustion are generated by a deflagration in the tunnel.

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5-9 (Page 5-10 deleted)

Amendment No.

t Enclosure 3 Certificate of Service of Errata for TMI-l Technical Specification Change Request No. 257

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METROPOLITAN EDISON COMPANY 1

JERSEY CENTRAL POWER AND LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY j GPU NUCLEAR, Inc.

Three Mile Island Nuclear Station, Unit 1

Operating License No. DPR-50

] Docket No. 50-289 l Technical Specification Change Request No. 257 - Errata 4

COMMONWEALTH OF PENNSYLVANIA )

. ) SS:

COUNTY OF DAUPHIN )

1

) Errata to Technical Specification Change Request 257 are submitted in support of Licensee's l request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear

! Station, Unit 1. As part of this request, proposed replacement pages for Appendix A are also

. included. All statements contained in this submittal have been reviewed, and all such stztements made and matters set forth therein are true and correct to the best of my knowledge.

BY- ~

ce Presider.t and Director, TMI

Swom and subscribed before me this ay of M 996.

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/ N8tary Public Nctarial Seal 4

Suzanne C. MNosik. Notary Public Londonderry Twp.. Dauphin County

My Commics.on Expires Nov. 22,1999 i Member, Pennsylvania Association of Notass 1

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