ML20035C558
| ML20035C558 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 03/30/1993 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | GPU Nuclear Corp, Jersey Central Power & Light Co, Metropolitan Edison Co, Pennsylvania Electric Co |
| Shared Package | |
| ML20035C559 | List: |
| References | |
| NUDOCS 9304080113 | |
| Download: ML20035C558 (7) | |
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NUCLEAR _ REGULATORY COMMISSION 7,;
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-!.t METROPOLITAN EDISON COMPANY:
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JERSEY CENTRAL POWER & LIGHT COMPANY l
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GPU NUCLEAR CORPORATION f
q DOCKET NO. 50-289-q THREE MILE ISLAND NUCLEAR STATION. UNIT NO. 1
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AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.172 License No. DPR-50L 1.
The Nuclear Regulatory Commission (the Commission) has found that:
j A.
The application for amendment by GPU Nuclear Corporation,.et 'al.
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(the licensee), dated November 12, 1992, complies with the standards and requirements of the Atomic Energy Act' of:1954, as amended :(the Act), and the Commission's rules and regulations-set forth in-10 CFR Chapter I;-
i B.
The facility will operate in conformity with the. application, the j
provisions of the Act,- and the rules and regulations of. the l
Commission; j
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There is reasonable assurance (i) that the activities authorized by _
1 this amendment can be conducted without endangering the health and
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safety of the public, and (ii) that such. activities will be conducted in compliance with the Commission's regulations;~
q D.
The issuance of this amendment will not be inimical to the common i
defense and security or to the health and safety of the public; and E.
Tha. issuance of this amendment is in accordance with 10 CFR'Part 51 of the_ Commission's' regulations _ and all applicable _ requirements have
.j been satisfied..
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PDR ADDCK 05000289 1
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2.
Accordingly, the license is amended by changes to the :.cnaical Specifications as ind.cated in the attachment to this licer;se amendment, and paragraph 2.c.(2) of Facility Operating License No.
DPR-50 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendix A, as j
revised through Amendment No.172, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in accordance with the Technical. Specifications.
3.
This license amendment is effective as of its date of issuance, to be l
implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
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Johd F;.Stolz, Director Prpjec'tDirectorateI-4 4
Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 30, 1993 t
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ATTACHMENT TO LICENSE AMENDMENT NO.172 FACILITY OPERATING LICENSE NO. OPR-50 r
DOCKET NO. 50-289 l
,i Replace the following pages of the Appendix A Technical Specifications with l
the attached pages.
The revised pages are identified by an amendment number and contain vertical lines indicating the area of change.
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Remove Insert d
P 4-11 4-11 r
4-13 4-13 4-52 4-52 4-52a 4-52a f
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4.2 REACTOR COOLANT SYSTEM INSERVICE INSPECTION AND TESTING Acolicability This technical specification applies to the inservice inspection (ISI) and inservice testing (IST) of the reactor coolant system pressure boundary and portions of other safety oriented system pressure boundaries.
Ob.iective The objective of the ISI and IST programs is to provide assurance of.the l
continuing integrity of the reactor coolant system while at the same time minimizing radiation exposure to personnel in the performance of inservice inspections and tests.
P Specification 4.2.1 ISI of ASME Code Class 1, Class 2, and Class 3 components shall be-l performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the NRC.
4.2.2 IST of ASME Code Class 1, Class 2 and Class 3 pumps and valves shall
'l be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(f), except where specific written relief has been l
granted by the NRC.
4.2.3 (Deleted) 4.2.4 The accessible portions of one reactor e.oolant pump motor flywheel assembly will be ultrasonically inspected within the first ISI period, two reactor coolant pump motor flywheel assemblies within the first two ISI periods and all four by the end of the 10 year inspection interval. However, the U.T. procedure is developmental and will be used only to the extent that it is shown to be meaningful. The extent of coverage will be limited to those areas of the flywheeJ which are accesrible without motor disassembly, i.e., can be reached through the access ports. Also, if radiation levels at the lower access ports are prohibitive, only the upper access ports will be used.
4-11 Amendment No. 15, 77, SJ, pp, 7J, 118,172
Bases Specifications 4.2.1 and 2 ensure that inservice inspection of ASME Code Class 1, 2 and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves will be performed in accordance with a periodically updated version of Section XI of the ASME Boiler and Pressure Vessel Code and Addenda as required by 10 CFR 50, paragraphs 55a(f) and 55a(g).
Relief from l'
any of the above requirements has been provided in writing by the NRC and is not a part of these technical specifications.
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4-13 Amendment No. 29, SJ, 57, Order dtd. 4/20/81,71, Corr. Ltr..dtd. JJ/2/SJ, Reissued 3/20/85, JJp,157,172
a 4.9 DECAY HEAT REMOVAL CAPABILITY - PERIODIC TESTING l
l Aeolicability Applies to the periodic testing of systems or components which function to remove decay heat.
Obiective l
To verify that systems / components required for decay heat removal an'e capable of performing their design function.
Soecification 4.9.1 Emergency Feedwater System - Periodic Testing (Reactor Coolant System Temperature greater than 250*F.)
4.9.1.1 Verify each EFW Pump is tested in accordance with the requirements and acceptance criteria of the ASME Section XI Inservice Test Program.
4.9.1.2 During testing of the EFW System when the reactor is in STARTUP, HOT STANDBY or POWER OPERATION, if one steam generator flow path.
is made inoperable, a dedicated qualified individual who is in communication with the control room shall be continuously stationed at the affected EFW local manual valves. On instruction from the Control Room Operator, the individual shall realign the valves from the test mode to.their operational alignment.
4.9.1.3 At least once per 31 days, each EFW System flowpath valve from both CSTs to the OTSGs via the motor driven pumps and the turbine driven pump shall be verified to be in the required status.
4.9.1.4 On a refueling interval basis:
a) Verify that each EFW pump starts automatically upon receipt of 1
an EFW test signal, b) Verify that each EFW control valve responds upon receipt of an EFW test signal.
c) Verify that each EFW control valve responds in manual control l
from the control room and remote shutdown panel.
4.9.1.5 Prior to start-up, following a refueling shutdown or a cold shutdown greater than 30 days, conduct a test to demonstrate that the motor driven EFW pumps can pump water from the condensate tanks to the Steam Generators.
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4-52 Amendment No.
7S, JJP; 124,172
o' 4.9.1.6 Acceptance Criteria These tests shall be considered satisfactory if control board indication and visual observation of the equipment demonstrates that all components have operated properly except for the tests required by Specification 4.9.1.1.
4.9.2 Decay Heat Removal Capability - Pegiodic Testing (Reactor Coolant System Temperature 250*F or less).
4.9.2.1 On a daily basis, verify operability of the means for decay heat removal required by specification 3.4.2 by observation of console status indication.
- These requirements supplement the requirements of 4.5.2.2 and 4.5.4.
Bases ASME Section XI specifies requirements and acceptance standards for the testing of nuclear safety related pumps.
The quarterly EFW pump test frequency specified by the ASME Section XI Code will be sufficient to verify that the turbine-driven and both motor-driven EFW pumps are operable.
Compliance with the normal acceptance criteria assures that the EFW pumps are operating as expected. The surveillance requirements ensure that the overall EFW System functional capability is maintained.
Daily verification of the operability of the required means for decay heat removal ensures that sufficient decay heat removal capability will be maintained.
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Amendment No.
7S,JJP,124,172