ML20003G188

Application to Amend License DPR-50 Submitted as Tech Spec Change Request 99 to Revise App A.Certificate of Svc Encl

ML20003G188
Person / Time
Site:	Three Mile Island
Issue date:	04/24/1981
From:	Hukill H METROPOLITAN EDISON CO.
To:
Shared Package
ML20003G185	List: ML20003G184 ML20003G188 ML20003G191
References
NUDOCS 8104280445
Download: ML20003G188 (4)

Text

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O METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISIAND NUCLEAR STATION, UNIT 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 99 This Technical 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.

1 METROPOLITAN EDISON COMPANY By ,

Direitor, DC-1 H. D. Hukill Sworn and subscribed to me this,d V dayofN/JA~ , 1981.

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UNITED STATES OF AERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET No. 50-289 LICENSE NO. DPR-50 METROPOLITAN EDISON COMPANY This is to certify that a copy of Technical Specification Change Request No. 99 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 Commission and been served on the chief executives of Londonderry Township, Dauphin County, Pennsylvania and Dauphin County, Pennsylvania by deposit in the United States mail, addressed as follows:

Mr. Donald Hoover, Chairman Mr. John E. Minnich, Chairman Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R.D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 Harrisburg, PA 17120 METROPOLITAN EDISON COMPANY By ,

' Director, TMI-l H. D. Hukill Dated: April 24, 1981

Three Mlle Island Nuclear Station, Unit 1 Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No. 99 The Licensee requests that the attached pages 111, 4-9, 4-10, 4-35, 4-51, and 4-54 replace the respective existing Tech. Spec. pages.

Reason for Change Request These administrative changes are requested for the following reasons:

99.1 - Speciff .ation 4.11 " Site Environmental Radioactivity Survey" refers the tr ader to Section 6.4 of Appendix B. There is no such section in Appendix B. Therefore, this section should be deleted.

99.2 Specification 4.8.1 states that a closure time of approximately 112 sec. shall be verified. The TMI-1 FSAR Section 10.2.1.3 states < 120 sec. is the limit. The erroneous time of 112 sec.

was obtained from a test procedure, which was later corrected to 120 sec. Therefore, the closure time in the Technical Speci-fication should be revised to < 120 sec.

99.3 - Table 4.1-3 item 6 requires a twice per week boron concentration check of the Boric Acid Mix Tank or the Reclaimed Boric Acid Tank.

A footnote should be added tc this specification to allow relief from this sampling when the tanks are empty.

99.4 - Specification 4.4.2.1.1 " Containment Tendons" states that only the tendon surveillance done ac one and three years following initial structural integrity used Regulatory Guide 1.35 Rev.1. This statement was correct when the Change Request was submitted for NRC review. However, the five-year surveillance was performed prior to receipt of Amendment 59 and was therefore also done per Regulatory Guide 1.35 Rev. 1. Therefore, the abose referenced section has been rewritten to reflect what actually occurred.

Safety Analysis Justifying Change Because these changes are administrative in nature, they have no effect on nuclear safety. Therefore, no safety analysis is necessary.

License Amendment Classification (10CFR 170.22)

These changes are administrative in nature. Therefore, enclosed please find the proper remittance of $1200.00.

TABLE OF CONTENTS Section Page 4.4.2 STRUCTURAL INTEGRITY 4-35 4.4.3 HYDROGEN PURGE SYSTEM 4-37 4.5 DiERGENCY LOADING SEOUENCE AND POWER TRANSFER, EMERGENCY CORE COOLING SYSTEM AND REACTOR BUILDING COOLING SYSTEM PERIODIC TESTING 4-39 4.5.1 EMERGENCY LOADING SEQUENCE 4-39 4.5.2 EMERGENCY CORE COOLING SYSTEM 4-41 4.5.3 REACTOR BUILDING COOLING AND ISOLATION SYSTEM 4-43 4.5.4 DECAY HEAT RDiOVAL SYSTEM LEAKAGE 4-45 4.6 DfERGENCY POWER SYSTEM PERIODIC TESTS 4-46 4.7 REACTOR CONTROL ROD SYSTEM TESTS 4-48 4.7.1 CONTROL ROD DRIVE SYSTDi FUNCTIONAL TESTS 4-48 4.7.2 CONTROL ROD PROGRAM VERIFICATION 4-50 4.8 MAIN STEAM ISOLATION VALVES 4-51 4.9 EMERGENCY tTEDWATER PUMPS PERIODIC TESTING 4-52 4.9.1 TEST 4-52 4.9.2 ACCEPTANCE CRITERIA 4-52 4.10 REACTIVITY ANOMALIES 4-53 4.11 INTENTIONALLY BLANK 4-54 l 4.12 AIR TREATMENT SYSTEMS 4-55 4.12.1 DfERGENCY CONTROL ROOM AIR TREATMENT SYSTDi 4-55 4.12.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTDi 4-55b 4.12.3 AUXILIARY AND FUEL HANDLING EXHAUST AIR TREATMENT SYSTEM 4-55d 4.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 REACTOR BUILDING PURGE EXHAUST SYSTEM 4-57 4.15 MAIN STEAM SYSTDi INSERVICE INSPECTION 4-58 4.16 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 4.17 SHOCK SUPPRESSORS (SNUBBERS) 4-60 4.18 FIRE PROTECTION SYSTDiS 4-72 4.18.1 FIRE PROTECTION INSTRUMENTS 4-72 4.18.2 FIRE SUPPRESSION WATER SYSTDi 4-73 4.18.3 DELUGE / SPRINKLER SYSTDi 4-74 4.18.4 CO2 SYSTDi 4-74 4.18.5 HALON SYSTDt3 4-75 4.18.6 HOSE STATIONS 4-76

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4.19 OTSG TUBE INSERVICE INSPECTION 4-77 4.19.1 STEAM GENERATOR SAMPLE SELECTION AND INSPECTION METHODS 4-77 4.19.2 STEAM GENERATOR TUBE SAMPLE SELECTION AND INSPECIION 4-77 4.19.3 INSPECTION FREQUENCIES 4-79 4.19.4 ACCEPTANCE CRITERIA 4-80 4.19.5 REPORTS 4-81 4.20 REACTOR BUILDING AIR TREATMENT 4-86 5 DESIGN FEATURES 5-1 5.1 SITE 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTDi 5-3 5.3 REACTOR 5-4 5.3.1 REACTOR CORE 5-4 5.3.2 REACTOR COOLANT SYSTDI 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 NEW FUEL STORAGE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 111

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