ML19248D441

Amend 39 to License DPR-16,adding Limiting Safety Sys Setting for Automatic Initiation of Isolation Condenser & Assembly Average Power Void Relationship for All Types in Reactor

ML19248D441
Person / Time
Site:	Oyster Creek
Issue date:	07/23/1979
From:	Ziemann D Office of Nuclear Reactor Regulation
To:
Shared Package
ML19248D442	List: ML19248D441 ML19248D445 ML19248D448
References
TASK-16, TASK-RR NUDOCS 7908160169
Download: ML19248D441 (8)

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UNITED STATES

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g NUCLEAR REGULATORY COMMISSION gQ,, 7 WASHINGTON. D. C. 20555 5

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JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-210 0_YSTER CREEK NUCLEAR GENERATING STATION, UNIT NO.1 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 39 License No. DPR-16 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Jersey Central Power & Light Company (the licensee) dated May 23, 1979, as supported by letter dated May 19, 1979, complies with the standards and requirements of the Atomic Energy Act gf 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 appligation, the provisions of the Act, and the rules and regulations of the Comission; C.

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 Comission's regulations; D.

The issuance of this amencment 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 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 3.B of Provisional Operating License No.

OPR-16 is hereby amended to read as follows:

B.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 39, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

2 This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION m

Dennis L. Ziemann,.ief Operating Reactors Branch #2 Division of Operating Reactors Attachnent:

Changes to the Technical Specifications Date of Issuance: July 23,1979 O

ATTACHMENT TO LICENSE AMENDMENT NO. 39 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 areas of change.

REMOVE INSERT 2.3-3 2.3-3 3.1-8 3.1-8 3.10-1 3.10-1 3.10-2 3.10-2 3.10-9 3.10-9 (Figure 3.10-1) 3.10-10 (Figure 3.10-2) 9 O

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LIM ~TINO SA?l'"! STST._. SCINSS nNC !ON

7) Lov Pressure Mats Stes= l'ne, 2S25 psig MSIV Closure

8) Main Stea= Line Isolatien valve 5 10: Valve Closure fre= full Closure, Scra=

open

9) Reac:er lov L' ate: Level, scra=

1 11',5" above the top of the ac:ive fuel as i=dicated u= der cer=al operating cenditie:

10) Lasc:or Lev-Lev Water Level, E 7',2" above the top of the Main' Stean Line isolation Valve ac:ive fuel as indicated under ner=al operating conditiens.

Closure,

11) Rese:c: Lev-Lov Wa:er Level, E 7'2" above the tcp of the ac:ive fuel.

Cere Spray Inizia:1en

12) Reac::: Lev-Lev Water Level, E 7 '2" above the tcp of the Iscla:1cn Cendense 7 d datic:

active fuel with ti=e delay 13 seconds.

13) Turbine Trip Scra=

10 percent turbine step valve (s clesure fren full open.

I

14) Gener :c Lead Rej ecti:n Scra=

Initiate upcn less of oil pressure fic: turbi=e ac:eleratics relay.

3 ASIS:

Saf e:y li=its have been established,in Specifi:atiens 2.1 and 2.2

o prote:: the in:egrity of the fuel cladding cr.d reactor ccelan:

system barriers.

Aut==atic pro:e::ive devices have been previded in the plant design :c take correc:ive action to prevent the saf ety li:1:s frc= being exceeded in neraal cperatien or operational

ansients caused by reasenable expected single operater errer c:

equip =ent =alfunctics.

This Specifica icn establishes :he trip se::ings for :hese au:cca:1c protection devices.

The Average ? cue Range Moniter, A?RM('), trip setting has been established to assure never reaching the fuel cladding integrity safe:y 14 ':.

The APRM syste= responds to changes in neu:ren flux.

Hevever, near rated ther=al power the APRM is calibrated, using a plan: hea: balance, so ths: the neutr:n flux that is sensed is read cu: as percent of ra:ed :her=al pcVer.

Ic: slev =aneuvers, those where core :her=al pever, surf ace hes flux, and the pcuer transf erred to the va:e fellev the neu:ren flux, the APEX will read reactor :her=al power.

Ter fast transients, the neu::en ficx vill lead the power

ansferred frc: the cladding := the water due to the eff ect of the fuel :'=e constan:. Theref ore, when the neu ::n flux increases :o the scra= se::ing, the percen: increase in heat fl=r. and pcVer transf erred t the water vill be less han the per:en: i= crease in neu:ren flux.

The APRM trip se::ing vill be varied aunc ::ically.vi h recircula:ier 3

flev vi:h :he : rip se::ing at rated fiev 61.0 x 10 lb/hr er

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1ased en a cc plete grea:e: being 113. 7% c f ra: ed ne ----

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• TABLE 3.11 PROTECTIVE IllSTRUlfEllTATIO!! REQUIllDfDITS (C0llTD)

(J D.

r.n Hin. flo. of j

Hin. !!o. of Opersble Reactor flodes Goerable or Instrument in Which Function Operating Channels Per flu s t Be Operabic (Tripped) Trip Operable Action Function Trip Setting Shutdown Refuel Startup Run Systems Trip Systems Requiredd Close mnin steam D.

Ile ac t or Iso lat ion isointion valves 1.

I nu-l.ou Reactor A*

X X

X X

2 2

and close isola-tion condenser tJ a t c r 1. eve l r

um vent valveo, or 120% rated X

X X

X 2

-7 2

pinco in cold

2. Ilich Flow in 3

S, g

-^

L.

shutdown condi-itain Stonm-line A r.7-tion

3. Iligh Flov in j 120% rated

.X X

X X

2 2

Itain Stenm-line B

4. liigh Tempe rn-3 Ambient at X

X X

X 2

'2 Lore in ifnin Powe r + 50*F Stcanline Tunnel AA X

2 2

5. l.ov Pro nurn in ifain Stenm-line

6. Iligh Radiation j 10X llormal X

X X

X 2

2 in liain Steam Bac kgr ound Tunnel C.

Isolnt lon Ccudent r "a

X X

X 2

2

'1c :e lent ir 1.

high I'enctor ol*, utJoen licos'ne na itt on X

X X

(

2 2

l 7

I.<o-lou Runciir t

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3.10-1 3.10 CORI LIMITS frJlj ability-Applies to core conditions required to meet the Final Ac-ceptance Criteria for Emergency Core Cooling Perfor=ance.

Ob j e' t rie :

To assure conformance to the peak clad temocrature linita-tions during a postulated loss-of-coolant accident as speci-fied in 10 CFR 50.46 (January 4, 1974) and to assure confor-mance to the 17.2 kW/ft. (for 7 x 7 fuel) and 14.5 KV/ft.

(for 8 x 8 fuel) operating limit.s for local linear heat generation rate.

Caa-i'icetion:

A.

Average Planar LHCR During power operation, the average linear heat generation rate (LHGR) of all the rods in any fuel assembly, as a func-rien of avera;e planar exposure, at any axial location shall not exceed the maximum average planar LHGR (MAPLHGR) limit shown in Figure 3.10-1.

If at any time during power operation it is determined by normal surveillance that the limiting value for APLHGR is being exceeded, action shall be initiated to restore operation to wit!.in the prescribed limits.

If the APLHGR is not returned to within the prnscribed limits within two (2) hours, action shall be initiated to bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

During this period surveillance and corresponding acti'on shall continue urtil reactor operation is within the prescribed limits at which time power operation may be continued.

E.

Local LiiCR During power operation, the linear heat generation rate (LHCR) of any rod in any fuel assembly, at any axial location shall not exceed the maximum allowable LHCR as calculated by the following equation:

LHCR < LHGR

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(3 -L max d P oi Whc. : LECF'd " "E AP .= Maximum Power Spiking Penalty P LT = Total Core Length - 144 inches L = Axial position above bot tom of core Ame-dment Nc. M (( 39 wea

2 4v. and 3.10-2 Fuel Tvoe LECR P/? d II 17.2 .032 IIII 17.2 .046 III? 17.2 .033 V 14.5 .033 V3 14.5 .039 If at any ti=e during operatien it is determined by nor=al surveillance that the itaiting value for LEGR is being exceeded, action shall be initiated to restore operation to within the prescribed li=its. If the LEGR is not returned to within the prescribed limits withia :vo (2) hours, action shall be initiated to bring the reacect :o the cold shutdown condition within 36 hours. During this period, surveillance and correspending action shall continue until reactor operatien is within the prescribed it=1ts at which time power operatica =ay be centinued. C. Asse=bly Averaged ?cuer Void Relationship During power operation, the asse=bl/ average vrid fracci.n and asse=bly power shall be such tha: the folleving relations ip is satisified: 1-V7 (PR x FCP ) 'Jhere : V7 = Bundle average beid fraction ?R = Assembly radial power factor Fractional core power (relative to 1930 MW:) FCP = 3 = Pcus: Va t: li=it The Ai=iting values ;f "3" for each fuel type are shewn in the table below: Fuel Tvre(s) 3 I, II, III .365 .-ii Ir.t;, T. - v.;. s ~ V, V3 .332 D. Mini =u= Critical ?cwer Ratio (MCPR) During steady state pcuer operatien, MC?R shall be g: ater :han or equal to the following: ARPM Status MC?R Li=it 1. If any tvc (2) LPFy. asse=blies which 1.64 are input to the A?RM syste= and are separated in distance by less than Amendment No. JE, 24, 33, 3a c.. ,9 Ns.Ne$ % u'5

HGURE 3.10- 1 MXUIUM ALLOMBLE AVi GE PLAT;4R LIllEAR llEAT CE!IERM M tl R!.TE 14 . _i _.. _g. ~ J ~ i. ._.____.___________' 4 ~

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