ML19270A212

Tech Spec Change Request 92 Re Control Rod Drive Scram Discharge Vol Capability.Certificate of Svc Encl

ML19270A212
Person / Time
Site:	Oyster Creek
Issue date:	03/04/1981
From:	Finfrock I JERSEY CENTRAL POWER & LIGHT CO.
To:
Shared Package
ML19270A213	List: ML19270A212 ML19347D270
References
NUDOCS 8103110708
Download: ML19270A212 (9)

Text

TER-C5506-58 E - Jersey Central Power & Ugnt Company

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Mar *s:0wn tec Jesev 0 hs 201 539-6111 March 4 , 1981 The Honorable Henry Von Spreckelsen Mayor of I,acey Township P. O. Box 475 Forked River, New Jersey 08731

Dear Mayor Von Spreckelsen:

Enclosed herewith is one copy of Technical Specification Change Request No. 92 for the Oyster Creek Nuclear Generating Station Operating License.

This document was filed with the United States Nuclear Regulatory Ccmmission on March 4, 1981.

Very truly yours, All .

Ivan R. Fin , ek sr.

Vice Presid .t la Enclosure 8IJE'$e(

O3 I I o 7CT C49fr81 DoWrer S UQ'If Co*C8**y is a f.tember 3 The Ge*!egl P.tD3aC 1.htishe$ Sp5:em nklin Research Center A DMason of *he Fraruthn msutute

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3.1-17e TABIT 3.1.1 (GWD)

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• 71 1. The Interlock is not required dur1pg the start-up test program and demonstration of plant electrical h

85 output but shall be provided following these actions.

j. Not sequired below 40% of turbine rated steam flow.

A gg k. All four (4) drywell pressure Instrument channels may be made inoperable during the integrated primary M containment leakage rate test (See Specification 4.5), provided that priwary containment integrity is not

$ required and that no work is performed on the reactor or its connected s ptens which could result in g lowering the reactor water level to less than 4'8" above the top of the active fuel.

'4

1. Bypassed in IRH Ranges 8, 9, li 10

m. There is one time delay relay associated with each of two pumps.

n. One time delay relay per pump minst be operable.

o. There are two time delay relays associateil with each of two pumps.

tn 4 p. Two time delay relays per pump must be operable,

q. Manual Initiation of affected component can be accomplished after the automatic load sequencing is compt et eil.

r. Time delay starts after closing of containmen' spray i=>mp circuit breaker.

s. These functions not required to be operable with the reactor temperature less then 212'F and the vessel head removed or vented.

t. These functions may be Inoperable or bypassed when correspondlag portions in the same tore spray system logic train are Inoperable per Specification 3.4.A.

u. These funct ions are not required to be operable when primary containment integrity is not required to be maintained.

v. 1hese Snct ions not required to be uperable when the ADS is not respilted to be oreerable,

w. These functions must he operable only when irradiated fuel is in the fuel l col or reactor vessel b and secondary containment integrit y is r equi red per spec i ficat ion 3.5.B. $

o

y. The number of operable channels may be re lated to 2 per Specification 1.9 li and 17 2.

1he bypass function to permit scram reset in the shutdown or re(nel mode with control rod block muest be operahlo in this smide.

Amendment No. 41

R M5 TABl.E 4.3.1

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, g HINi>RM CillCK, CAlgRr:10N AND TTST FRI: QUI'NCY FOR PH(Tft:CTIVil INSTRIN!:NTATION

' s 5i pp Instrument Channel gie33 Ca. librate Test Remarks J Applies to Test and CallbrationJ_

,. m 8k 1. liigh Henctor Pressure NA I/3 mo. Note l By appilcation of test pressure

[n?r 2. liigh Drywell Pressure (scram) NA l/3 mo. Note ! " " " " "

R$

4

3. Inw Reactor Water I.evel I/d 1/3 umn. Note I

4. Inw-inw Water f.evel I/d 1/3 mo. Note i "

5. liigh Water I.evel in Scram Discharge' Volume (Scram) NA I/3 mo. Note i By varying level in switch columns

6. Inw-inw-law Wat er I.evel NA I/3 mo. Note 1 By application of test pressure

7. Iligh Flow in Main Steamline I/d " " " " "

I/3 aus. Note i

8. Inw Pressure in Rein NA l/3 mo. Note i " ." " " "

Steamline

9. Iligh tirywell Pressure I/d Note l " "

(Core Cooling) -

10. Main Steam isolation NA NA I/3 mo. By exercising valve Valve (Scram)

II. APRM l.evel NA I/ 3tl NA output adjitstment using operational type heat balance during power operation NIHl! 1: Initially once/mo., thereafter according to Fig. 4.1.1, with an interval not less than one month "

3 nor more than three nuint hs. L NOTl! 2: At least daily aluring reactor power operation, the reactor neut ron flux peaking fhctor shall be estimated anal t he I' low-s eferenced APHH scram and ruil block set t ings shall be o ljust ed, if necessary, as specified f,

h ui in Sect ion 2.3, Specificat ions (1) (a) and (2) (a). S ve

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= 19. Manual Scram Buttons NA NA l/3 sus c3 20 liigh Temperature Main NA Each refuel- Each refuel- Ilsing heat source box i Steamline Tunnel ing outage Ing outage

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• Ilsing built-in calibration equipment

22. Isolation Contlenser liigh NA I/3 uni 1/3 mo By application of test pressure Flow P (Steam and Water) g 23. Turbine Trip Scram NA Every w

e 3 months o

21 Generator looil Reject ton HA Every Every Scram 3 months ,1 uninths

25. Recirculation loop Flow NA Each Refuel- NA Py application of test pressure ing Outage

26. Inw Reactor Pressure NA Every Every By application of test pressure

. Core Spray Valve 3 months 3 months Permissive

27. Sc ram lis, charge Volume (Ro.1 Block) .

a) Water level high NA Each Refuel- Every 3 Dy varying level in switch column.

ing Outage aminths 61 N

li) Scram t rip bypass NA NA Each refuel-f ing outage y o

os t

• Calilirate lirior to startup anil poemal shutdown and thereafter check I/s and test 1/wk until no longer requirett. g

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b VN the specified limits, provide the require r ' protection. In the 1l7 1F.* analytical treatment of the transients milliseconos are f allowed between a neutros sensor reachf6 U he scram point and

,  ? 4 d the start of motion of the control rods J .is is adeounte and conservative when F parc.a to tac. typical time delay of N ut 210 mil.

seconds estimated from scram test results. 4pproximately the fr et millisecc:

of a fi'tf-these time intewals result from the sensor and circuit delays; chen the pil scram solenoid de-energizes. Approximately 120 milliseconds later, the controlq rod motion la estimatjEto actually begin. However, 200 milliseconds is conservatiyely .a-assumed for chi rise interval in the transienpalyses and this is _also included in t.

allowable scram m*~8pecitiec limits provice sufncient scram capaollity to N-insertion times of accommodate failure to scram of any one operable rod. This specification of,3-2-3. failure is in addition to any inoperable rods that exist in the j core, provided that those inoperable rods met the core reactivity J ' Specification 3.2.A, #

Control rods (8) which cannot be moved with control rod drive pressure are clearly indicative of an abnormal operating condition on the affected rods and are, therefore, considered to be inoperable, Inoperable rods are valved out of service to fix their position in the core and assure predictable behavior. If the rod is fully inserted and then valved out of servite, it is in a safe position of maximum contribution to shutdown reactivity. If ic is valved out of service in a non-fully inserted position, that pcsition is required to be consistent with the shutdown reactivity

' limitation stated in Specification 3.2.A. which assures the core can be shutdown-at all times with control rods.

Although there are many possible patterns of inoperable control rods which would meet this specification, the operator will be p'rovided with ,only a limited . number of predetermined patterns which allow him to continue operation with inoperable rods. The availability of allowable patterns to the operator assures that information for determining comp'11ance with the specification is immediately available to him at the time a control rod becomes inoperable and does not require relicnce on calculations at that time before compliance can'be determined.

The allowable inoperable rod patterns will be determined using infctmation obtained in the startup test program suppletented by calculations. During init'ial startup, the reactivity ccndition of the as-built core will be determined. Also, sub-critical '

patterns of widely separattd withdrawn control rods will be cbserved in the control rod sequences being used. The observa-tions, together with calculated strengths of the strongest control rods in these patterns wi"11 compriso .a set of allow-able separations of malfunctioning rods. Daring the fuel cycle, similar observatiora made during any cold shutdown esn be used to update and/or increase the allowable patterns.

The number of rods pernitted to be valved out of servie'c'could nklin Research Center A Dneman of The Frankhn insatuer

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