ML20125D347
| ML20125D347 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek (DPR-16-A-086, DPR-16-A-86) |
| Issue date: | 06/07/1985 |
| From: | Zwolinski J Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20125D349 | List: |
| References | |
| NUDOCS 8506120325 | |
| Download: ML20125D347 (7) | |
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WASHINGTON, D. C. 20555
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2 GPU NilCLEAR CORPORATION AND JERSEY CENTRAL POWER & LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment.No. 86 License No. DPR-16 1.
The Nuclear Repulatory Comission (the Comission) has found that:
A.
The application for amendment by GPU Nuclear Corporation and Jersey Central Power and Light Company (the licensees) dated July 19, 1984, as supplemented by agreement noted in meeting minutes of February 22, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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.
Thereisreasonableassurance(i)thattheactivitiesauthorized 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
8506120325 850607 ADOCK O 29 gDR
. 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. 86, 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 T NUCLEAR REGULAT_0RY OPMISSION
)
1 o
John A. Zwolinski, Chief Operating Reactors Branch #5 Division of Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 7, 1985.
J ATTACHMENT TO LICENSE AMENDMENT NO.86 PROVISIONAL OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain vertical lines indicating the area of change.
REMOVE INSERT 3.5-3 3.5-3 3.5-5 3.5-5 3.5-7 3.5-7 4.5-10 4.5-10 0
3.5-3 b.
Two of the fourteen suppression chamber - drywell vacuum breakers may be inoperable provided that they are secured in the closed position.
c.
One position alarm circuit for each operable vacuum breaker may be inoperable for up to 15 days provided that each operable suppression chamber - drywell vacuum breaker with one defective alarm circuit is physically verified to be closed immediately and daily during this period.
6.
After completion of the startup test program and demonstration of plant electrical output, the primary containment atmosphere shall be reduced to less than 4.0% 0, with nitrogen gas within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the reactor mode selector switch Ts placed in the run mode, primary contaiment deinerting may commence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to a scheduled shutdown.
7.
If specifications 3.5. A.I.a.
b, c(1) and 3.5. A.2 through 3.5. A.5 cannot be met, reactor shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
8.
Shock Suppressors (Snubbers) a.
During all modes of operation except cold shutdown and refuel, all safety related snubbers listed in Table 3.5.1 shall be operable except as noted 3.5.A.8.b. c and d below.
b.
From and after the time that a snubber is determined to be inoperable, continued reactor operation is permissible only.during the succeeding 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> unless the snubber is sooner made operable or replaced.
c.
If the requirements of 3.5.A.8.a and 3.5.A.8.b cannot be met, an orderly shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.
~d.
If a snubber is determined to be inoperable while the reactor is in the shutdown or refuel mode, the snubber shall be made operable or replaced prior to reactor startup.
e.
Snubbers may be added to safety related systems without prior License Mendment to Table 3.5.1 provided that a revision to Table 3.5.1 is included with the next License Mendment request.
9.
Drywell-Suppression Chamber Differential pressure a.
Differential pressure between the drywell and suppression chamber shall be maintained within the acceptable operating range shown on Figure 3.5-1 within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the reactor mode selector switch is placed in the run mode.
The differential pressure may be reduced to less than the range shown on Figure 3.5-124 hours prior to a scheduled shutdown.
The differential pressure may be decreased to less than the required value for a maximum of four hours during required operability testing of the drywell-pressure suppression chamber vacuum breakers.
AmendmentNo.f,k.k,86
L 3.5-5 j
importantly, the accessibility of the valve lever arm and position reference external to the valve. The fail-safe feature of the alarm l
circuits assures operator attention if a line fault occurs.
Conservative estimates of the hydrogen produced, consistent with the core cooling system provided, show that the hydrogen air mixture resulting from a loss-of-coolant accident is considerably below the flammability limit and hence it cannot burn, and inerting would not be needed. However, inerting of the primary containment was included in the proposed design and operation.
The 5% oxygen Ifmit is the oxygen concentration limit stated by the American Gas Association for whichcombustionwillnotoccur.(gdrogen-oxygenmixturesbelow The 4% oxygen limit was established by analysis of the Generation and Mitigation of(12)
Combustible Gas Mixtures in Inerted HWR Mark I Containments.
To preclude the possibility of starting up the reactor and operating a long period of time with a significant leak in the primary system, leak checks nust be made when the system (k N07 near rated temperature and pressure.
It has been shown that an acceptable margin with respect to flammability exists without containment inerting.
Inerting the primary containment 4
provides additional margin to that already considered acceptable.
Therefore, permitting access to the drywell for the purpose of leak checking would not reduce the margin of safety below that considered adequate and is judged prudent in terms of the added plant safety offered by the opportunity for leak inspection. The 24-hour time to provide inerting is judged to be a reasonable time to perform the operation and establish the required 0 limit.
2 Snubbers are designed to prevent unrestrained pipe motion under dynamic loads as might occur during an earthquake or severe transient, while allowing normal thermal motion during startup and shutdown. The consequence of an inoperable snubber is an increase in the probability of structural damage to piping as a result of a seismic or other event initiating dynamic loads.
It is, therefore, required that all snubbers required to protect the primary coolant system or any other safety system or component be i
operable during reactor operation.
All safety related hydraulic snubbers are visually inspected for overall integrity and operability.
The I
AmendmentNo,f,86 l
t.
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3.5-7 When secondary containment is not maintained, the additional restrictions on operation and maintenance give assurance that the probability of inadvertent releases of radioactive material will be minimized. Maintenance will not be performed on systems which connect to the reactor vessel lower than the top of the active fuel unless the system is isolated by at least one locked closed isolation valve.
The standby gas treatment system (6) filters and exhausts the reactor building atmosphere to the stack c'uring secondary containment isolation conditions, with a minimum release of radioactive materials from the reactor building to the environs.
Two separate filter trains are provided each having 100%
capacity.(6)
If one filter train becomes inoperable, there is no imediate threat to secondary containment and reactor operation may continue while repairs are being made. Since the test interval for this system is one month (Specification 4.5),
the time out-of-service allowance of 7 days is based on considerations presented in the Bases in Specification 3.2 for a one-out-of-two system.
References:
FDSAR, Volume I, Section V-1 FDSAR, Volume I, Section V-1.4.1 FDSAR, Volume I, Section V-1.7 Licensing Application Amendment 11, Questic, III-25 FDSAR, Volume I, Section V-2 FDSAR, Volume I, Section V-2.4 Licensing Application, Amendment 42 Licensing Application. Amendment 32, Question 3 Robbins, C. H., " Tests on a Full Scale 1/48 Segment of the Humboldt Bay Pressure Suppression Contalment,"
GEAP-3596, November 17, 1960.
(10) Bodega Bay Prelininary Hazards Summary Report, Appendix I, Docket 50-205, December 28, 1962.
(11) Report H. R. Erickson, Bergen-Paterson to K. R. Goller, NRC, October 7, 1974.
Subject:
Hydraulic Shock Sway Arrestors.
(12) General Electric NEDO-22155 " Generation and Mitigation of Combustible Gas Mixtures in Inerted BWR Mark I Contairments" June 1982.
In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed on August 2,1976, which demonstrated a factor of safety of at least two for the weakest element in the suppression chamber support system. The maintenance of a drywell-suppression chamber differential pressure within the range shown on Figure 3.5-1 with a suppression chamber water level corresponding to a downcomer submergence range of 3.0 to 5.3 feet will assure the integrity of the suppression chamber when subjected to post-LOCA suppression pool hydrodynamic forces.
Amendment No. h, h, }$. 86
y 4.5.10 i
After the containment oxygen concentration has been reduced to meet the specification initially, the containment atmosphere is maintained above atmospheric pressure by the primary containment inerting system. This j
1ystem supplies nitrogen makeup.to the' containment'so that the very L
i slight leakage from the containment is replaced by nitrogen, further reducing the oxygen concentration. In addition, the oxygen concentration j
is continuously recorded and high oxygen concentration is annunciated.
l Therefore, a weekly check of oxygen concentration is adequate. This system also provides capability for determining if.there is gross leakage from the containment.
i The drywell exterior was coated with Firebar D prior to concrete pouring i
during construction. The Firebar D separated the drywell steel plate from the concrete. After installation, the drywell liner was heated and expanded to. compress the Firebar D to supply a gap between the steel drywell and the concrete. The gap prevents contact of the drywell wall with the concrete which might cause excessive local stresses during-drywell expansion in,a loss-of-coolant accident. The surveillance i
program is being conducted to demonstrate that the Firebar D will maintain its integrity and not deteriorate throughout plant life. The, surveillance frequgggy is adequate to detect any deterioration tendency of the material.
ta; The operability of the instrument line flow check valves are demonstrated
, to assure isolation capability for excess flow and to assure the operability of the instrument sensor when required.
Because of the large volume and thennal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends. By requiring the suppression pool temperature to be continually monitored and also observed during periods of significant heat addition, 2
the temperature trends will be closely followed so that appropriate l
action can be taken. The requirement for an external visual examination following any event where potentially high loadings could occur provides assurance.that no significant damage was encountered. Particular attention should be focused on structura1' discontinuities in the 'icinity.
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of the relief valve discharge since these are expected to be the points of highest stress.
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ChangeNo.M, Amendment No.
Amendment No. 86 1
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