ML20108E380
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- c-UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 JERSEY. CENTRAL POWER 6 LIGHT COMPANY DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION, UNIT NO. 1
' AMENDMENT TO PROVISIONAL-OPERATING LICENSE Amendment No. 11 License No. DPR-16 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
There is reasonabic assurance (i) that the activities authorized 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; and B.
The issuance of this amendment will not be inimien1 to the common defense and security or to the health and safety of the public.
2.
Accordingly, the license-is amended by a change to the Technical Specifications as indicated in the attachacnt to this licent.c amendment and Paragraph 3.B. of Provisional Operating License No. DPR-16 is herchy amended to read as follows:
"(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised, are hereby incorporated in the license.
The. licensee shall operate the facility in accordance with the Technical Specifications, as revised by issued changes thereto through Change No. 11".
- 3.. This license amendment is effective as of the date of its issuance.
FOR THE NUCLEAR REGULATORY COMISSION L
C W"
Georg r,' Chief Operating Reactors Branch #3 Division of Reactor Licensing
Attachment:
Change No. 11 to the Technical Specifications Date of. Issuance:
L.? 1 4;5 9605100216 960213 PDR FOIA DEKOK95-258 PDR
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CHANGES TO TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE NO. DPR-16
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DOCKET NO. 50-219 The proposed changes to the Technical SpecificatJons are shown on the attachedpagesand'areidentifiedbyavertical.}ineinthemargin..
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- a. 2-A 3.5 coNTA71-Applicability:
Applies to the operating status of the primary and secondary containment systems.
Objective:
To assure the integrity of the primary and secondary containir.cnt systems.
Specifications: A.
- Priraary Containment 1.
At any tice that thc nucicar systein is pressurized above 1
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. atmospheric or vorh is being donc which has the potential to drain the vessel, the suppression pool water volume and temperature shall be naintained within the f ollowing limits.
3 a.
Flaxinun water volunc - 92,000 it b.
liinimutn water volume - 82,000 iL
'c.
llaxismnn water ter@crature (1)
During normal power operation - 95 r (2)
During tenting which adds heat to the supprension pool, the water tenperature shall not exceed 30r.above the norma) power operation Jinit specified in (3) above.
In connect ion uf th cuch t et t in;,,
the pool tenperature nunt be reduced to belou the nomal puuor operation limit
'specified in (1) above wit hin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(3)
The react or'shall be sc rammed iron any operating condition if the pool tenperature y
reaches 110T.
Power operation shall not be resu;ied until the pool tc:r.perature is j
reduced below the nornal peuer operation j
lini.t speci fied in (1) above.
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(4)
During reactor isolation conditions, the reactor pressure vessel shall be depretsur-ized to less than ISO psig at noma) cooldown rates, if the pool temperature reachca 120F.
I 2.
Primary containment integrity sh511 be naintained at all times when the reactor is critical or then the reactor
' water temperature is above 212"r and fuel is in the reactor vessel except while perforcing low power physics
' tests at atmospheric pressure during or after refueling i
at power levcis not to exceed 5 1Nt.
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3.5-2 l
3.
Reactor Bui) ding to suppression chanber vacuun Breaker Sym f Except as specified in Specification 3.5 A.3.b below,
.a.
two reactor building to suppression chxnber vacuun breakers in cach Jine r. hall be operable at all times when. primary containment integrity is required.
The
~ set point of ' the dif ferential pressure instrumentation which actuates the air-operated vacuum brenhers shall not exceed 0.5 psid.
The vacuum breakers shall r.ove fron closed to fully open when subjected to a force equivalent'to not greater than 0.5 psid acting on the vacuun breaker, disc.
b.
Troin the time that one of the reactor building to suppression chamber vacuun breahers is unde or found to be inoperabic, the vacuun brenher.shall be locked closed and reactor operation is pernissib)e only during l
the succeeding seven days unless such vacuun breaher is made operabic sooner, provided that the procedure doet, 4
not violate primary containment integrity.
li the Jinits of Specificat ion 3.5./ 3.a are exceeded, c.
reactor shutdown chall be initiated and the reactor shall be in a cold shutdown condition ui thin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.
Pressure Supnression Chan+cr - Drvuell Yacuun Hrcahers a.
Uhen primary containment is required, all suppression chamber - drywell vacuun brechers shall be operable execpt during testing and as stated in Specification 3.5.A.4.b and c, helow.
Suppression chamber - dryvell vacuum breakers shall be considered operabic if:
(1) Th'c valve is denonstrated to open'from closed to fully open with the applied force at all salve positions not exceeding that equivalent to 0.5 psi acting on'the suppression chamber face of the valvo dick..
(2) The valve disk will close by gravity to'within not greater than 0.30 inch of any point on the seal surface of the disk when released af ter being opened by remote or manual means.
(3) The position alarm system will annunciate in the control room if the valve is open more than 0.10 inch at any point a3ong the. scal surface of the disk.
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3.5-3 b.
Two of the fourteen suppression char.ber - dryve)1
, vacuum breakers nay be inoperable provided tirat they
'are secured in the closed position.
c.. One position alarn circuit for each 'operabic vacuu:s bren};cr may be inoperabic for up to'15 days 'providcd that cach operabic suppresnion chanhcr - drywell Vacuun breaker with one defective alern circuit is physica)1y verified to be closed in:.:ediately and daily during this period.
5.
Atter completion of the startup test progran and demonstration of plant electrical output, the prinary-containment atnosphere sha)1 be reduced to less than 5.0%.0 with nitrogen can uithin 24 hourn.after the 2
reactor node sc)cetor swit ch in placed in t he run r.odc.
j prinary contain::.cnt deinerting nay comence 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, prior to a scheduled shutdoun.
6.
If npecifications 3.5. A.1.n, b, c(1) and 3. S. A. ?
through 3.5. A.5 cannot be net,. react or :,hurdcan chall be initiated and the reactor chall be in the cold shutdown condillon wit hin 2/. hourn.
B.
Secondary Containnent 2.
Secondary containt::ent integrity r. hall he raaintained at all tirnen unlenn all of the follouing conditions are uct.
a.
The reactor is nuberitica) and Specifica tion 3.2. A in
- not, b.
The reactor is in the cold r.hutdown condition.
The.rcactor vessel head or the dryucil head are in c.
place.
d.
No work is being performed on the reactor or its connected systens in the reactor building.
No operati'ons are being periorned in, above, or c.
around the spent. fuel storage pool that could cause release of radioactive raaterials.
- 2..The standby gas tr'catnent systen shall be operabic at all tines when secondary containnent integrity is require 1
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except as specified by Specification ~3.5.11.3.
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3.5-4 iyc 1,=
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- 3..'One standby gas treatn.ent filter circuit ~may be' inoperabic for 7 days, when standby can treatment.
system operability is required, except during reactor startup, provided the remaining filter circuit is
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. proved'operabic' daily.
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If Specifications 3.5.B.2 and 3.5.B.3 are not net, reactor shutdown shall be initiated and. t.hc reactor shall' he '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 />'nnd the conditions of Specification 3.5.B.1 shall be not.
11ases :
Specifications are placed on the operating statun of'the con-
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tainment. nystems to assure their availability to control the relcane of any radioactivo notorial f rom irradiated f uel in t he cvent of an acciden't condition.
The primary'containuent tynt a (i)
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provides a barrier against uncontrolled release of fis.sion products l
i to the environs in the event of a hrenh in thc reactor coolant i
a systems.
Uhenever the reactor coolant unter temperature is above 212 r,
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f ailure of the reactor coolant. syste., veuld canne rapid expu)r. ion of '
the eco) ant from the reactor with an nr.r.ociated presourc rice in the primary. containment.
Primaty containeent in required, therefore, to cor.tain the thermal enctny of the expe))ed coolant n
and'finsion products which coufd he relc. sed f rom any f uci
- failures resulting from'the accident'.
If the reactor coolant is not above 232*P, there vou)d be no pressure rise in the G
containment.
In addition, the coolant cannot be expelled at a rate which could cause fuel failure to occur before the core spray J
system restores cooling t o the core.
Primary containuent is not needed while perfor. ming. low power physics tests r.ince the red vorth nininizer vould limit the worst case rod drop accident to
.1.5%Ak.
This amount of reactivity addition is insufficient to l
cause fuel damage.
The absorption chamb'er water volume provides the heat sinh for the reactor coolant systen energy released folleving the loss-of-coolant accident.
The core spray pumps and containment j
spray pumps are located tin the corner rooms and due to their
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proximity' to the torus, the ambient temperature in those rgs t
could' rise during the design basis accident.
Calculations l
made, assuming an. initial torus water tgmperature of 100*F c
.and'a ninimum water volume of 82,000 ft, indicate that too corner < room ambient. temperature would not exceed the core
- spray and containment spray pump motor operating tenperaturc limits,( and, therefore,- would not adversely af f ect the long
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3.5-4a 4
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term core cooling capability. The maximum water volume limit 4
allows for an operating range _without significantly affecting the accident analysos with respect to free air volume in thogg) absorption chamber.
For' example, the containment capability with a maximum water volume of 92,000 ft3 is reduced by not-more than 3.5% metal-water reaction below the capability with 82,000 ft3 Experimental data indicates'that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is maintained below 160oF during any period of relief valve operation with sonic conditions at the discharge exit.
Specifications have been placed on the envelope of reactor operating conditions so that~the reactor can be depressurized in a timely manner to avoid the regime of-potentially high suppression chamber loadings.
The purpose of the vacuum relief valves is to equalize the pressure between the drywell and suppression, chamber and suppression chamber and reactor building so that the containment external design pressure limits are not exceeded.
The vacuum relief. system from the reactor building to the pressurc suppression chamber consists of two 100's vacuum relief breaker subsystems (2 parallel' sets of 2 valves in series).
Operation of cither subsystem will maintain the containment external pressure less-than the external design pressure; the external design pressure of the drywell is 2 psi; the external design pressure of the suppression chamber is 1 psi (FDSAR Amendment 15,Section II).
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4.5-6a l
P.
Suppression' Chamber Surveillance 1.
At least once per dap the suppression chamber water level and temperature and pressure suppression system pressure shall be checked.
2.
A visual inspection of the suppression chamber interior, including water line regions, shall be made at cach major refueling outage.
3.
Whenever heat from relief valve ' operation is being added to the su'ppression pool, the pool temperature shall be continually monitored and also observed until the heat addition is terminated.
4.
b'henever operation of a relief valve is indicated and the suppression pool temperature reaches 160F or above while the reactor primary coolant system pressure is greater than 180 psig, an external visual examinction of the suppression chamber shall be made before resuming normal power operation.
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4.5-6b 4
Basis:
The primary containment preoperational test pressures are based upon the calculated primary containment pressure res-ponse in the event of a loss-of-coolant accident, The peak drywell pressure would be 38 psic.which would rapidly reduce to 20 psig stithin 100 seconds following the pipe break.
The total tice :he dr>vell pressure would bc_above 35'psig is calculated to be about 7 seconds.
Following the pipe break absorption chamber pressure rises to 20 psig within 8 seconds, equalizer. with drywell pressure at 25 psig within 60 seconds and thercatter rapidly decays with the drywell pressure decay. (1)
The design pressures of the drywell an{2) absorption chanber are 62 psic and 35 psic, respectively.
The desicr. Icak rate is 0.5!.*/ day at a presourc of 35 psig.
As pointed out above, the pressure responso of the drywell n.d absorption chamber followinp, an accident would be the s.e.m after about 60 seconds, haucd on the calculated primary contain=0r.t l
pressure respoasc discussed above ar.d the nbsorption chanber design pressure, primary contaitur.cnc prepperational test press.ures were cho:;en.
Also, bcced on the primary contain-ment pressure rerpanso ar.d the fact that thu drywell and absorption chanber function as a unit, the prinary con:31n-ment will be tested as a unit rather thar, testing the indi-vidua) conponents separately.
1hc design basis loss-of-cool' ant accident was evaluated at the primary containment maxirc.un allowabic accident leak rate of 1 0%/. day at 35 psig.
The analysis shoved that with this leak rate and a standby gas creatacnt system filter efficiency of 90 p'ercent for halogens, 95% for particulates, and assuuing the fission product release fractions stated in TID-14844, the maximum total whold body passing cloud dose is about 10 rem and the maximum total e
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Af ter the'containnent oxygen concentration has been reduced to neet the specification initially, the containment atmosphere is maintained above This atmospheric pressure by the primary containment inerting system.
system supplies nitrogen makeup to the containment so that the very slight Icakage from the containment is replacod by nitrogen, further reducing the oxygen-concentration.
In addition, the oxygen concentration is con-tinuously recorded and high oxygen conenetration is annunciated.
Therefore, a wechly check of oxygen concentration is' adequate.
This system also provides capability. for determining if there is gross leakage from the containment.
The dryuell exterior was coated with Firebar'D prior to concrete pouring during construction.
The Firebar D separated the drywell steel plate from After installatio'n, the drywell liner was heated and cxpanded the concretc.-
to co:apress the Firebar D to supply a gap between the steel drywell and the The cap prevents contact of the*dryvc11 wall with the concrete concretc.
which might cause excessive local. stresses during drywell expansion in a loss-of-coolant. accident.
Tbc surveillance progra:a is being conducted to denonstrate that the Firebar D will maintain'its integrity and not deteriorate throughout plant life.
The survei1]ance d
detect any det eriora-tiontendencyofthematerial.ggquencyisaequate.to
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line flow check valves are demonstrated to The operability of the instrument asuurc isolation capability for excess flow and to assure the operability of the instrument sensor when' required.
Eccause of the large volume and thermal capacity of the supprest.fon pool, the volu:ac and t emperature normally changes very slowly and r.onitoring these to est ab]$ sh any temperature t rer.ds.
By parameters daily is sufficient requirJng the suppression pool temperature to be continua 11,. conitored and the temperature also observed during periods of significant heat addition trendu will be closely followed so that appropriato action can be tanen.
The i
for an external visual examination following any event where requirement potentially high loadings ceuld occur provides assurance that no significant
'l damage was encountered.
p:rticular attention should be focused on structural discontinuities in the vicinity of the relici valvo discharge since these are Ij l
cxpected t o be the points of highest stress.
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