ML20107D689
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UNITED ST ATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 Docket No. 50-219
-v Jersey Central Power 6 Light Company ATTN: Mr. I. R. Finfrock, Jr.
Vice President - Generation Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960
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Gentlemen:
' The Commission has' requested the Office of th'e Federal Register to publish the enclosed Notice of Proposed Issuance of an Amendment to Provisional Operating License No. DPR-16 for the Cyster Creek Nuclear Generating Station. The proposed amendment incluiles a change to the Technical Specifications as proposed by our letter of July 16, 1975.
and subsequently modified by a few mutually acceptable changes.
Your staff has indicated that the comments in your lettar of August 8, 1975, pre now resolved and that the Technical Specificaticus, as modified, are acceptable.
This amendment would incorporate:
(1) water temperature limits during any testing which adds heat to the suppression pool,,(2) suppression pool, water temperature limits requiring manual scram of the reactor,
.(3) suppression pool water temperature limits requiring reactor pressure vessel depressurization, (4) surveillance requirements to monitor water temperatures during operations which add heat to the suppression pool and (5)' external visual examinations of the suppression chambers following operations in which the pool temperatures exceed 160oF.
In addition to the limits on the temperature of the suppression chamber q
pool water, your operating procedures should define the operator action
..to be taken in the event a relief valve inadvertently opens or sticks open. This action would include:
(1) use of all available means to close the valve, (2) initiate suppression pool water cooling heat
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exchangers, (3) initiate reactor shutdown, and (4) if other relief l
valves are used to depressurize the reactor, their discharge shall
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be separated from that of the stuck-open relief valve to assure mixing and uniformity of energy insertion to the pool.
1 Copies of the proposed amendment, the related Safety Evaluation, and the Federal Register Notice are enclosed.
Sincerely,-
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Geo 9rge E j
r, Chief,
Operating Reactors Branch #a Division of Reactor Licensing Ii 9604180450 960213'
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DEKOK95-258 PDR
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r OCT 6 575 Jersey Central Power 6 Light Company 2 -
cc: w/ enclosures G. F. Trowbridge, Esquire The Honorable W. M. Mason Shaw, Pittman, Potts, Trowbridge 6 Madden Mayor,. Lacey Township Barr Building P. O. Box 475 910 17th Street, N. W.
Porked River, New Jersey 08731 Washington, D. C. 20006 J'ersey Central Power 6 Light Company Honorable Wm. F. Hyland ATTN:
Mr. Thomas M. Crimmins, Jr.
Attorney General Safety and Licensing.Abnager.
State of New Jersey GPU Service Corporation State House Annex 260 Cherry Hill Road Trenton, New Jersey 0S601 Parsippany, New Jersey 07054 1
Anthony Z. Roisman, Esquire Mr. Paul Arbesman Berlin, Roisman 6 Kessler Environmental Protection Agency' 1712 N Street, N. W.
Region II Office Washington, D. C. 20036 26 Federal Pla:a New York, New York 10007 Paul Rosenberg, Esquire Daniel Rappoport, Esquire Ocean County Library 2323 S. Broad Street 15 llooper Avenue Trenton, New Jersey 08610 Toms River, New Jersey OS753 Honorable Joseph W. Ferraro, Jr.
Deputy Attorney General
, State of New Jersey Division of Law - Room 316 Newark, New Jersey 07102 Burtis W. Horner, Esquire Stryker, Tams and Dill 55 Madison Avenue Morristown, New Jersey 07960 George F.'Kugler, Jr.
Attorney General State of New Jersey State House Annex Trenton, New Jersey 08625 d
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UNITED
- STATES
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NUCLEAR REGULATORY COMMISSION W ASHIN G TO N. D. C. 20 SS S JERSEY CENTRAL POWER S LIGHT COMPANY DOCKET NO. S0-219 OYSTER CREEK NUCLEAR GEN $ RATING STATION, UNIT NO. 1 PROPOSED AMENDMENT TO PROVISIONAL OPERATING LICENSE
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Amendment No.
License No. DPR-16 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.. There is reasonable 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 inimical 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 T.echnical Specifications as indicated in the attachment to this license amendment and Paragraph 3.B. of Provisional Operating License No. DPR-16 is hereby 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.
4 3.
This license p;aendment is effective as of the date of its issuance.
FOR Ti!E NUCLEAR REGULATORY CO.41ISSION
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George Lea, Chief l
Operating Reactors Branch #3 l
Division of Reactor Licensing i
Attachment, Change No. _ to the, j
Technical Specifications l
Date of Issuance:
.s
,e' PROPOSED
- CHANGES TO TECHNICAL SPECIFICATIONS PROVISIONAL OPERATING LICENSE NO. DPR-16 DOCKET NO. 50:219 a
The proposed changes to the Technical Specifications are shown on the attached pages and are identified by a vertical line in the margin.
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3.5-1 3.5 CONTAIl0iEN/
Applicability: Appi,1cs to the operating status of the primary and se,condary
' containment systems.
-Objective:
To assure the integrity of the primary and secondary centainment systems.
Specifications: A.
At any time that the nucicar system is pressurized above
. atmospheric or work is being done which has the potential to drain the vessel, the suppression pool water volume an'd temperature shall be maintained within the following limits.
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a.
Maximum water volume - 92,000 ft b.
Minimum water volume - 82,000 ft c.
Maximum water temperature (1) During normal power operation - 95 F (2)
During tenting which adds heat to the suppression pool, the water tenperature shall not exceed 10F above the normal power operation limit specified in (1) above.
In connecLjon with such tccting, the pool temperature must be reduced to
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below the nornal power operation limit specified in (1) above within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
(3)
The reactor shall be scramacd fron any operating condition if the pool ter.parature reaches 110F.
Power operation shall not be resuded until the pool temperature is reduced below the nornal pouer operation limit specified in (3) above.
(4) During reactor isolation conditions, the reactor precourc vessel shall be depressur-ized to less than 180 psig at normal cooldown rates if the pool temperature reaches 120F.
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2.
Primary containment intcgrity shall be maintained at all times when the reactor is critical or when the reactor water temperature is above 212 F and fuel is in the reactor vessel except while performing lo power physics tests at atmospheric pressure during or af ter refueling at power Icvels not to exceed 5 }Nt.
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3.5-2 3.
Reactor Building to Suppression Chamber vacuum Breaker System a.
Except as specified in Specification 3.5.A.3.b below, two reactor building to suppression chamber vacuum breakers in each.line shall 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 breakers shall not exceed 0.5 psid.
The, vacuum breakers shall move from closed to fully open when subjected to a force equivalent to no,e greator than 0.5 psid acting on the vacu' m bfeaker disc.
u b.
From the time that one of the reactor building to suppression chamber vacuum breahers is nade or found to be inoperabic, the vacuum breaker shall be locked closed and reactor operation is pernissibJe only during the succeeding seven days unicss such vacuun breaker is made operabic sooner, provided that the procedure doce not violate primary containment integrity.
c.
If the Jimits of Specification 3.5.A.3.a are exceeded, reactor shutdown shall be initiated and the reactor shall be in a cold shutdown condition within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.
Pressure Suppression Chanber - Drywell Vacuun Breakers a.
When primary containment is required, all suppression chamber - drywell vacuun brechers shall be operable except during testing and as stated in Specification 3.5. A.4.b and c, below.
- Suppression chamber - drywell vacuum breakers shall be considered operable if:
(1) The valve is denonstrated to open from closed to fully open with the applied force at all valve j
positions not exceeding that equivalent to 0.5 psi acting on the suppression chamber face of the valve disk.
(2) The valve disk will close by gravity to'within not greater than 0.10 inch of any point on the seal surface of the disk when released after 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 along the seal surface of.the disk.
4.
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3.5-3 b.
Two of the fourteen suppression chamber - drywell vacuum breakers may be inoperabic provided that they are secured in the closed position.
c.
One position alarm circuit for each operabic vacuum breaker may be inoperabic for up to 15 days provided that cach operabic suppression chamber - drywell s
vacuum breaker with one defective alarm circuit is physically verified to be closed immediately and daily during this period.
5,' "After completion of the startup test program and demonstration of plan: clectrical output, the prinary containment atmosphere shall be reduced to less than 5.0% 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 2
reactor mode selector switch is placed in the run code.
Primary containment deincrting nay commence 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s-prior to a scheduled shutdown.
6.
If specifications 3.5.A.1.a. b, c(1) and 3.5.A.2 through 3.5. A.5 cannot be net, reactor shutdoun shall be initiat.d and the reactor shall be in 1
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 />.
B.
Secondary containment integrity shall be maintained at all times unicss all of the following conditions are met.
1 The reactor is suberitical and Specification 3.2.A is a.
met.
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The reactor is in the cold shutdown condition.
c.
The reactor vessel head or the drywell head are in place.
d.
No work is being performed on the reactor or its connected systems in the reactor building.
No operations are being performed in, above, or e.
around the spent. fuel storage pool that could cause release of radioactive materials.
2.
The standby gas treatment system shallic operabic at all t'imes when secondary containment integrity is required except as specified by Specification.3.5.B.3.
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3.
One standby gas treatment filter circuit may be inoperable for 7 days, when standby gas treatmant, system operability is required, except during reactor startup, provided the remaining filter circuit is proved operable daily.
4.
If Specifications 3.5.B.2 and 3.5.B.3 are not met,
- reactor shutdown shall be initiated and the reactor shall be in the cold shutdown condition within 24
, hours *and the conditions of, Specification 3.5.B.1 shall be met.
Lses:
Specifications are placed on the operating status of the con-tainment systems to assure their availability to control the release of any radioactive material f rom irradiated fuel in the event of an acciden't condition.
The primary containment systen (1) provides a barrier against uncontrolled rc12ase of fission products to the environs in the event of a break in the reactor coolant systems.
Whenever the reactor coolant water tenperature is above 212 r, failure of the reactor coolant system would cause rapid expulsion of the coolant from the reactor with an associated pressure risc in the primary containment.
Primary containcent is required, therefore, to contain the thermal energy of the expelled coolant and fiscion products which could be released from any fuc1 i
failures resulting from the accident.- If the reactor coolant is ntt above 212*T, there would be no pressure rise in the containment.
In addition, the coolant cannot be expelled at a rate which could cause fuel failure to occur before the core spray system restores cooling to the core.
Priuary containment is not
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needed while performing low power physics tests since the rod worth minimizer would limit the worst caso rod drop accident to 1.5%Ak.
This amount of reactivity addition is insnfficient to cause fuci damaga.
The absorption chamber water volume _provides the heat sink for'the reactor coolant system energy released follouing the loss-of-coolan t accident.
The core spray pumps and. containment spray pumps are located in the corner rooms and due to their proximitytothetorus,theambienttemperatureinthosergops could rise during the design basis accident.
Calculations mado,-assuming an initial torus water tgnperature of 100*F and a minimum water volume of 82,000 ft indicate that tac corner room ambient temperature would not exceed the core spray and containnent spray pump motor operating
- temperaturc limits, and, therefore, would not adv,crsely af f ect the long m-e e
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y 3.5-4a
. term core cooling capability. The maximum water volume limit allows for an operating range without significantly affecting the accident analyses with respect to free air volume in the(8) absorption chamber.
For example, the containment capability with a maximum water volume of 92,000 ft3 is reduced by~not
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more than 3.5% metal-water reaction below the capability with 3
82,000 ft.
Experimental data indicates that excessive steam condensing loads can be avoided if the peak temperature of the suppression pool is.
maintained below 1600F 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 i
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and reactor building so that the containment external design pressure limits are not exceeded.
The vacuum relief system from the. reactor building to the pressure suppression chamber consists of two 100% vacuum relief breaker j
subsystems (2 parallel sets of '2 valves in series).
Operation of either subsystem will maintain the containment external pressure 1ess 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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Suppression Chamber Surveillance 1.
At least once per day 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 each major refueling outage.
4 3.
Whenever heat from relief valve operation is being added to the suppression pool, the pool temperature shall be. continually monitored and also observed until,the heat addition is terminated.
4.
Whenever 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 examination of the suppression chamber shall be made before resuming normal power operation.'
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4.5-6b s,
Basist' The primary containment prcoperational test. pressures r.re based upon the calculated pt'imary containr.cnt pressure res-ponse in the event of a loss-of-coolant accident, The peak drywc11 pressure would bc 33 psic which would rapidly reduce to 20 psig within 100 seconds following the pipc break.
The total tino the drywell pressure would be above 35 psig is calculated to be about 7 toconds.
Following the pipe break absorption char.ber pressure risas to 20 psig within 8 seconds, equalizes with dryvell pressure c.t 25 psig within 60 seconds and thercafter rapidly decays with the drywell pressure decay. (1) e The design pres'surc0 of the dryvc11 an absorption chamber are 62 psig and 35 psig, respnctively 2)
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d rate is 0.57./ day at a pressure of 35 psig.
hdou s Po i a ove, the prosruro responso of the d- -
chamberfollowinP,anaceidentwould'b[wc1~1 d
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60 soconds.
1:ased on the calculated primary cont-d r"" -
pressure respo.uc discu.sacd abovo and *!n rpti a chambor design prcosure,,b".
ary containtz.cnt prepperational test
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prcusures ucro chosen.
Also, be. sed on the primary contain-
.non t pressure respanso ar.d the fact that the d ry.iell au"s obsorption chaaber function as a unit ti 4
cl$cin ment v11] be tested as a unit rather that te i
ual conponents separately.
v The design basis loss-of-cool'nt accident was evalua*.ed a-a the primary containment maximum allowabic accident 'eak rate of. 0%/dr.y at 35 psig.
Tiic. analysis sho 'd tk with this Acak rate and a standby gas crea:mont syster, I
ter efficiency o: 90 p,ercent for halo;;cnc 95% fo-particulaten, and assuming the {1ssion product release fractions stated in TID-14844. t1ic.aximu. total whold bod o
passing cloud dose is about 10 rem and *he maximum total p
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After the containment oxygen concentration has.been reduced to meet the specification initfally, the containment atmosphere is maintained above atmospheri'c pre'ssure by the primary containment inerting system.
This system supplies nitrogen makeup to the containment so that the very slight Icakage from the containment is replaced by nitrogen, further reducing the oxygen concentration.
In addition, the oxygen concentration is con-
.tinuous y recor ed and high oxygen conenceration is annunciated.
Therefore, l
d a weekly check of oxygen concentration is ndequate.
This system also provides capability for determining if there is gross leakage from the containment.
The drywell cxterior was coated with Firebar D prior to concrete pouring The Firebar.D separated the drywell steci plate from during construction.
the concrete.
After installation, the drywell liner was heated and expanded
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, to compress the Firebar D to supply a gap between the steel drywell and the concrete.
The gap prevents contact of thc' drywell wall with the concrete l
wh'ich might cause excessive local stresses during drywc11 expansion in.a loss-of-coolant accident.
The surveillance program is being conducted to demonstrate that'the Firebar D will maintain its integrity and not deteriorate throughout plant life.
The surveillance any deteriora-tiontendencyofthematerial.{gyquencyisadequate.todetect 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.
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Beenuse of the large volunc and thermal capac1[y of' the suppression peol, the volume and temperature normally changes very slowly and tonicering these parancters daily is sufficient to establish any temperature trer.ds.
3y requiring the suppression pool temperature to be continually monitored and also observed during periods of significant heat addition, the tennerature trends will be closely followed so that appropriate action can b'c 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 structural discontinuitics in the vicinity of the relici valve discharge sin,cc these are expected to be the points,of highest stress.
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