ML19260A147

Tech Spec Change Request 62 Supporting Licensee Request to Change App a to License DPR-50 Re Definition of Limiting Conditions for Operation Relating to Temps in Containment Per NRC 770516 Request.Certificate of Svc Encl

ML19260A147
Person / Time
Site:	Three Mile Island
Issue date:	07/22/1977
From:	Arnold R METROPOLITAN EDISON CO.
To:
Shared Package
ML19260A143	List: ML19260A142 ML19260A145 ML19260A147
References
NUDOCS 7910290761
Download: ML19260A147 (7)

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' METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Scecification Change Recuest No. 62 This Techaical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.

METROPOLITAN EDISOI COMPANY

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By /

e Piesident Sworn and subscribed to me this M

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day of , 1977.

O V T ,

Notary Public' NO- 'y r.;_,_

%d , ( t. , , , 3 t?s.iouso g 1480 214

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UNITED STATES OF MERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 IE GOPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 62 to Appendix A of the Operating License for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with the U. S. Nuclear Regulatory Co::=ission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States r. ail, addressed as follows:

Mr. Weldon B. Arehart Mr. Harry B. Reese, Jr.

Board of Supervisors of Board of County Comissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Court House Middletown, Pennsylvania 17057 Harrisburg, Pennsylvania 17120 METROPOLITAN EDISON COMPANY

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By V ce President N Dated: July 22, 1977 1480 215

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Three Mile Island nuclear Station, Unit 1 Operating License No. DPR-50 Docket No. 50-289

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Technical Scecification Chance Recuest No. 62 The licensee requests that tha attached revised pages be added to the existing Technical Specifications, Appendix A.

Reasons for Change Reauest The reasons for this addition to the Technical Specifications is to conply with the NRC's letter, dated May 16, 1977, requesting a submittal defining the limiting conditions for operation relating to temperatures in the containment.

Safety Analysis Justifyine Change Based on ectual operating experience, an analysis has been performed using a containment temperature of 130 F for elevations above 320'

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and 120 F for elevations below 320'. This analysis was performed in accordance with Working Stress Design provisions of ACI318-63 Acceptance criteria was based on the allovable stresses specified in Section III, Division 2, issued January 1, 1975 of the ASME Boiler and Pressure Vessel Code. The stresses resulting from the increased operating temperature were found not to be detrimental to the performance of the structure. In addition, the most critical section under the loading conditions specified above was checked using the Ultimate Strength Design provisions of ACI318-63. A safety factor of 1.63 vas found using this design method and confirmed the adequacy of the structure.

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I fe1# wor 761 1480 216

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TABLE OF CO:iTE:iTS Section Page 3.1 3 MINIMUM CONDITIONS FOR CRITICALITY 3-6 3.1.h REACTOR COOLAIIT SYSTEM ACTIVITY 3-8 3.1.5 CHE4ISTRY 3-10 3.1.6 LEAKAGE 3-12 3.17 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY 3-16 3.1.8 SINGLE LCOP RESTRICTIONS 3-17 3.1 9 LOW POWER PHYSICJ TESTING RESTRICTIONS 3-18 3.1.10 CONTROL ROD OPERATION 3-18a 3.2 MAKEUP A!iD PURIFICATION AND CHDfICAL ADDITION SYSTEMS 3-19 3.3 DERGriCY CORE COOLING, REACTOR BUILDING DERGE:!CY COOLING, AND REACTOR BUILDING SPRAY SYSTDtS 3-21 3.h TURBINE CYCLE 3-25 35 INSTRUME:iTATION SYSTE4S 3-27 3 5.1 OPERATIONAL SAFETY INSTRUMBITATION 3-27 3 5.2 CO:iTROL R0D GROUP AND POWER DISTRIBUTION LIMITS 3-33 353 EIGINEERED SAFEGUARDS PROTECTION SYSTEM ACTUATION SETPOINTS 3-37 3 5.h INCORE INSTRUMENTATION 3-38 3.6 REACTOR BUILDING 3 h1 3.7 UNIT ELECTRICAL POWER SYSTE! 3 h2 3.3 FUEL L0 APING AND RE W ELING 3-kh

, 39 RADIOACTIVE MATERIALS 3-h6 3.10 MISCELLANECUS RADIOACTr/E MATERIALS SOURCES 3-h6 3.11 *WiDLING OF IRRADIATED FUEL 3-55 3.12 , REACTOR DUILDING POLAR CRANE 3-57 3.13 SECONDARY SYSTE4 ACTIVITY 3-58 3 1h FLOOD 3-59 3.1h.1 PERIODIC INSPECTION OF THE DI?ES AROUND TMI 3-59 3 1h.2 FLOOD CONDITION FOR PLACING THE UNIT IN HOT STriDBY 3-60 3.15 (RESERVED) 3.16 SHOCK SUPPRESSORS (SNUBBERS) 3-63 3 17 REACTOR BUILDI::G AIR TD4PERATURE 3-80 h SURVEILIXiCE STEIDARDS h_1

• h.1 OPERATIONAL SAFETY REVIEW h-1 4.2 REACTOR COOLANT SYSTDI INSERVICE INSPECTION h-11 h.3 TESTING FOLLOWIi!G OPENING OF SYSTDI k-28 h.h REACTOR BUILDI:!G h-29 h.h.1 CONTAIN:.EUT LEAK AGE TESTS h-29 h.h.2 STRUCTURAL INTEGRITY h-35 h.h.3 HYDROGEN PURGE SYSTE! h-37 h.5 EERGENCY LOADING SEQUENCE AIID POWEP WliSFER, EEFOE:!CY CORE COOLING SYSTDI ANO REACTOR EUILDING COOLING SYSTE! PERIODIC TESTING h-39 h.5 1 EERGE;CY LOADING SEQUENCE h-39 h.5 2 EERGE' ICY CORE C00LI:iG SYSTri h_hl L.5.3 REACTOR BUILDING COOLING AND ISOLATICN SYSTE4 h-h3 h.5.h DECAY HEAT RE:0 VAL SYSTDi LEAEAGE h-h5 11 1480 217

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TABLE OF CONTENTS Page Section k.6 DIERGE*!CY POWER SYSTE*4 PERIODIC TESTS h-h6 k.7 REACTOR CONTROL ROD SYSTEM TESTS h-48 h.7 1 CONTROL ROD DRIVE SYSTEM FUNCTIONAL TESTS h-h8 h.7.2 CONTROL R0D PROGRAM VERIFICATICN h-50 h.8 MAIN STEAM ISOLATION VALVES h-51 49 DIERGENCY FEEDWATER PUMPS PERIODIC TESTING h-52 h.9 1 TEST h-52 h.9 2 ACCEPTANCE CRITERIA '

h-52 h.10 REACTIVITY ANCMALIES h-53 h.11 SITE E'iVIRONMENTAL RADI0 ACTIVITY SURVEY h-Sh h.12 CONTROL ROCM FILTERING SYSTEM h-55 4.12.1 OPERATING TESTS h-55 h.12.2 FILTER TESTS h-55 h.13 RADI0 ACTIVE MATERIALS SOURCES SURVEILLANCE h-56 h.14 REACTOR 3UILDING FURGE EXHAUST SYSTEM k-57 h.15 MAIN STEAM SYSTDI INSERVICE INSPECTION h-58 h.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE h-59 h.17 SHOCK SUPPRESSORS (SNUB 3ERc-)' h-60 h.19 REACTOR BUILDING AIR TEMPERATURE h-72 l 5 DESIGN FEATURES 5_1 6

51 SITE 5-1 5.2 CONTAI!REIT 5-2 5.2.1 REACTOR BUILDING 5-2 5 2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 53 REACTOR 5h 5 3.1 REACTOR CORE 5h 5.3.2 REACTOR COOLANT SYSTEM 5h 5,h UEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.h.1 NEW FUEL STORAGE 5-6 5.h.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTAKE TU? RIEL FIRE FROTECTION SYSTDIS 5-8

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6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-2 6.2.1 0FFSITE 6-2 6.2.2 FACILITY STAFF 6-2 6.3 STATION STAFF QUALIFICATIONS 6-3 6.L TRAINING 6-3 6.5 REVIEW AND AUDIT 6-3 6.5 1 PLANT OPERATICUS REVIEW CO:0!IT"EE (PORC) 6-3 6.5 2.A MET-ED CORPORATE TECENICAL SUPPCRT STAFF 6-5 6.5 2.3 GENERAL OFFICE REVIEW BOARD (GORB) 6-7 6.6 REPORTABLE OCCURRENCE ACTICN 6-10 6.7 OCCURRENCES INVOLVIN3 A SAFETY LIMIT VIDIATION 6-10a 6.8 PROCEDURES 6-11

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h.19 REACTOR BUILDI:!G AIR TEMPEFATL"RE Acolicability This specification applies to the average air temperature of the primary containment during power operations.

Obiective To assure that the temperatures used in the safety analysis of the reactor building are not exceeded.

Specification h.18.1 When the reactor is critical, the reactor building temperature vill be checked once each twenty-fcur (2h) hours. If any detector exceeds 1300F (120oF belov elevation 320) the arithmetic average vill be ec=puted to assure compliance with Specification 3.17.1.

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h-72 1

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3.17 REACTOR BUILDING AIR TEMPERATURE Apelicability Tnis specification applies to the average air temperature of the primary containment during power operations.

Objective To assure that the temperatures used in the safety analysis of the Reactor Building are not exceeded.

Srecification 3 17 1. Primary cor'sinment average air te=perature above Elev. 320 shall not excee. .. 3 F and average air temperature belov Elev. 320 shall not exceed i G0F.

3.17.2 If, while the reactor is critical, the above stated te=perature limits are exceeded the average temperature shall be reduced to the above limits within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or be in at least HOT STANDBY vithin the next six (6) hours and in COLD SHUTDOWU within the following thirty (30) hours.

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3 17 3 The primary containment average air temperature shall be calculated  !

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a) The average temperature above elevation 320 vill be calculated i by taking the arithmetic average of the temperature from the ,

available detectors above elevation 320. (No = ore than four  :

(h) of the nineteen (19) detectors may be out of service) l

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b) The average temperatures belov elevation 320 vill be calculated

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by taking the arithmetic average of the temperature from the  !

available detectors belov elevation 320. (No more than two (2) ',

of the five (5) detectors may be out of service)

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PT Location PT Location i

1 SE Dome Elev h54 13 UE Wall Elev 326 2 NW Doce Elev h5h 14 S Wall Elev 326 3 N Wall Elev h5h 15 NW Wall Elev 326 i h E Wall Elev h54 16 E Sec Shield Elev 326 j 5 S Wall Elev h5h 17 S 12 Wall Elev 321 6 W Wall Elev h5h 18 NE Wall Elev 29h*  ;

7 NE Wall Elev 39h 19 S Wall Elev 29h* ,

8 S Wall Elev 39h 20 tiW Wall Elev 29h*

9 NW Wall Elev 39h 21 E Sec Shield Elev 29h* ,

10 NE Wall Elev 36L 22 :iW See Shield Elev 29h*

11 S Wall Elev 36h 23 NE See Shield Elev 36L 12 NW Wall Elev 36h 2h N Sec Shield Elev 36h ,

• detectors located belov elevation 320 3-80 143) 22h