ML20211J674
| ML20211J674 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 10/27/1986 |
| From: | Donohew J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20211J681 | List: |
| References | |
| NUDOCS 8611110219 | |
| Download: ML20211J674 (5) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
- a WASHINGTON, D. C. 20555
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GPU NUC_ LEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO PROVISIONAL OPERATING I_ICENSE Amendment No.110
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License No. DPR-16 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by GPU Nuclear Corporation and Jersey Central Power and Light Company (the licensees) dated September 5,1986, complies with the standards and require-ments 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; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorized
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by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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. 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to.this license amendment and Paragraph 2.C.(2) of Provisional Operating License No. DPR-16 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.110, 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 the date of its issuance.
FOR THE NUCLEAR REGULATORY COMMISSION h.
. Donohew, Jr., P@ ject Manager
( BWR Project Directorate #1 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 27, 1986 1
ATTACHMENT TO LICENSE AMENDMENT N0.110 PROVISIONAL OPERATING LICENSE NO. DPR-16 f
DOCKET N0. 50-219 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.
REMOVE' INSERT 3.1-3 3.1-3 3.1-16 3.1-16 7,,a-
g 4.
When required for surveillance testing, a channel is made inoperable.
In order to be able to test its trip function to the final actuating device of its trip _ system, the trip system cannot already be-tripped by some other means such as a mode switch, interlock, or manual-trip.
Therefore, there will be times during the test that the channel is inoperable but not tripped.
For a two channel trip system, this means that full reliance is being placed on the channel that is not being tested.
A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one operable channel in the same trip ~ system is monitoring that parameter.
Bypasses of inputs to a trip system other than the IRH and APRM bypasses are provided for meeting operational requirements listed in the notes in Table 3.1.1.
Note a allows the "high water level in scram discharge volume" scram trip to be bypassed in the refuel mode.
In order to reset the safety system after a scram condition, it is necessary to drain the scram discharge volume to. clear this scram input condition.
(This condition usually follows any scram, no matter what the initial cause might have been.) In order to do this, this particular scram function can be bypassed only in the refuel position.
Since all of the 1
control rods are completely inserted following a scram, it is permissible to bypass this condition because a control rod block prevents withdrawal as long as the switch is in the bypass condition for this function.
The manual scram associated with moving the mode switch to shutdown.is used cerely to provide a mechanism whereby the reactor protection system scram logic channels and the reactor manual control system can be energized. The ability to reset'a scram twenty (20) seconds after going into the shutdown mode provides the beneficial function of relieving scram pressure from the control rod drives which will increase their expected lifetime.
To permit plant operation to generate adequate steam and pressure to establish turbine seals and condenser vacuum at relatively low reactor power, the main condenser vacuum trip is bypassed until 600 osig.
This bypass also applies to the main steam isolation valves for the same reason.
The action required when the minimum instrument logic conditions are not met is chosen so as to bring plant operation promptly to such a condition that the 1
0YSTER CREEK 3.1-3 Amendment No.: 20, 110 2/4/77
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3 TABLE 3.1.1 (CONTD)
- Action required when minimum conditions for operation are not satisfied. Also permissible to trip inoperable trip system. When necessary to conduct tests and calibrations, one channel may be made i
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inoperable for up to two hours per Technical Specification required surveillance without tripping its W
trip system.
- See Specification 2.3 for Limiting Safety System Settings.
Notes:
l a.
Permissible to bypass, with control rod block, for reactor protection system reset in refuel mode.
b.
Permissible to bypass below 800 psia in refuel and startup modes.
l c.
One (1) AP M in each operable trip system may be bypassed or inoperable provided the requirements of specification 3.1.C and 3.10.C are satisfied. Two APRM's in the same quadrant shall not be concurrently bypassed except as noted below or permitted by' note.
Any one APM may be removed from service for up to one hour for test or. calibration without inserting trips in its trip system only if the remaining operable APRM's meet the requirements of specifica-
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tion 3.1.8.1 and no control rods are moved outward during the calibration or test. During this short period, the requirements of specifications 3.1.B.2, 3.1.C and 3.10.C need not be met.
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d.
The IRM shall be inserted and operable until the APRM's are operable and reading at least 2/150 full scale.
f Air ejector isolation valve closure time delay shall not exceed 15 minutes.
e.
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f.
Unless SRM chambers are fully inserted.
'i Not applicable when IRM on lowest range.
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g.
One instrument channel in each trip system may be inoperable provided the circuit which it operates in
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g the trip system is placed in a simulated tripped condition.
If repairs cannot be completed 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 the cold shudown condition.
If more than one instrument channel in any ii
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trip system becomes inoperable, the reactor shall be placed in the cold shutdown condition.
Relief valve
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controllers shall not be bypassed for more than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (total time for all controllers) in any 30-day 2:
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period and only one relief valve controller may be bypassed at a time.
!l 1.
The interlock is not required during the start-up test program and demonstration of plant electrical l
output but shall be provided following these actions.
Not required below 40% of turbine rated steam flow.
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All four (4) drywell pressure instrument channels may be made inoperable during the integrated primary
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containment leakage rate test (See Specification 4.5), provided that the plant is in the cold shutdown
_,g condition and that no work is performed on the reactor or its connected systems which could result in
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lowering the ' reactor water level to less than 4'8" above the top of the active fuel.
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Bypass in IRM Ranges 8, 9, and 10.
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