ML19254F980

Tech Spec Change Request 75,revising Sections 1.0,2.1,3.4, & 3.7 to Facilitate Refueling Outage

ML19254F980
Person / Time
Site:	Oyster Creek
Issue date:	11/16/1979
From:	JERSEY CENTRAL POWER & LIGHT CO.
To:
Shared Package
ML19254F979	List: ML19254F978 ML19254F980
References
NUDOCS 7911200537
Download: ML19254F980 (19)

Text

JEFSEY CUJTRAL B7.iER & LIGilT COMPN 1Y OYSTER CREEK !!UCLEAR GDJERATI!G STATION Provisional Operating License I;o. DPR-16 Tecnnical Spec 1tication Change Request 11o. 75 Docket No. 50-219 Applicant submits, by this Technical Specification Change Pcquest No. 75 to the Oyster Creek Nuclear Gcnerating Station Technical Specifications, revised sections 1.0, 2.1, 3.1, 3.4, 3.5, and 3.7. ,

JEl; SLY CD TRAL POWER & LIGilT CCMPANY BY: ]7]M s V ~

Vice e fit STATE OF liEW JERSEY COUtTI'Y CF MORRIS Sworn and subscribed to before me on this - day off i/ v u <.t, 1979.

= / , . /' . /> /~ i s .

Notary Public

r. : Y u.:.> A. :3,;!s NOTARY PU2Lic CF MW JERSEY My Commission E4piras Aug. 16,1934 1351 .335 1316 2-91 7911200 5 3 4

r

-.,7. .- Jersey Central Power & Light Company luf (f ( )

Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960 (201)455-8200 November 16, 1979

'Ihe Honorable Fury Iou Smith Mayor of Iccey 'lownship '

P. O. Dox 475 Forked River, New Jersey 08731

Dear Mayor Smith:

Enclosed herewith is one copy of Technical Specification Change Eequest No. 75 for the Oyster Creek Nuclear Gcncrcting Station 'lechnical Specificationc.

These doctrnants were filed with the U. S. Nuclear Ecgulctory Commission on November 16 , 1979.

. Very truly yours, a

Ivan R. Finf ck, .

Vice Presid t la enclosure S7F2t2

~

l351 336 Jersey Central Power & Ught Company is a Member of the General Pubic Utilities System

UtIITED STATES OF AMERICA NUCLEAR REGUIATORY COtT4ISSION IN Tile fGTI'ER OF )

) DCCKET NO. 50-219 JEFSEY CENTRAL ICWER & LIQl? CQiPANY )

CERTIFIr. ATE OF SERVICE This is to certify thct a copy of Technical Specification Change Rc<Juest No. 75 for the Oyster Creek Nuclear Generating Station Technical Spccifications, filed with the U. S. Nuclear Regulatory Commission on November 16 , 1979, has this 16th day of Novcaber, 1979, been served on the Mayor of Lacey Tom ship, Ocean County, New Jersey by deposit in the United States mail addressed as follows:

The llonorable Mary Lou Smith Mcyor of Laccy 'Ibwnship P. O. Box 475 Forked River, New Jersey 08731 JERSEY CENTRAL POWER & LIGHT C04PAIW BY: jf) ,

" ' Vice Pres~ en DATED: November 16, 1979

^

.43 W 291 1351 .337

JERSEY CENTRAL POWER S LIGHT COMPANY OYSTER CREEK NUCLEAR GENERATING STATION PROVISIONAL OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Applicant hereby requests the Commission to change Appendix A to the License as follows:

1. Sections to be changed:

1.0, 2.1, 3.1, 3.4, 3.5, 3.7.

2. Extent of Changes:

The definitions of the shutdowm and refueling conditions are modified in Section 1.G. Unnecessary operability requirements on protective instru-mentation are deleted from Section 3.1. The operability requirements of the startup transformers are clarified in Sect. ion 3.7. Typographical

errors in Sections 2.1 and 3.5 are corrected. The electromatic relief valves' pressure relief function is allowed to be inoperable during ASME Code System hydrostatic pressure tests.

3. Changes requested:

The requested changes are on the attached revised Technical Specification

_ pages 1.0-1, 1.0-2, 2.1 -3, 3.1-6a , 3.1-7, 3.1-7a, 3.1-8, 3.1-9, 3.1-10, 3.1-11, 3.1- 12, 3.1-12a , 3. 4- 1b , 3. 5-2, 3. 7-1 and 3.7-2.

4. Discussion and Safety Evaluation:

Definitions 1.6 Shutdown Condition,1.7 Cold Shutdown and 1.11 Refuel Mode are being modified to reflect that they are applicable only when there is fuel in the reactor irssel. These definitions are used to define when various plant systems, equipment and instrumentation are required to be op(rable in order to assure that the plant is maintained in a safe con-dition. When the core has been unloaded into the spent fuel pool, no

equipment associated with reactor safety is required. A footnote has been added to Table 3.1.1 to require that Reactor Building isolation and Standby Gas Treatment System initiation on high radiation on the Reactor Building operating floor and on high radiation in the Reactor Building Ventilation Exhaust be operable when fuel is in the spent fuel pool or reactor vessel and secondary containment integrity is required per Specification 3.S.B.

A change on page 2.1-3 is being made to correct a typographical error.

The reference to " Reference 12" in the third paragraph should be to reference 10 (on page 2.1-5) .

Table 3.1.1 is being revised to delete or otherwise modify the operability requirements of various items of protective instrumentation in the shutdowm, refuel and startup modes.

I351 $38

-iW 3 /yt Technical Specification Change Request No. 7S

The requirements to have the following protective instrumentation functions operable during the shutdown or refuel modes with the reactor water less than 212*F and the vessel head removed or vented have been deleted:

A.2 High Reactor Pressure Scram A.7 High Radiation in Main Steamline Tunnel Scram A.8 Average Power Range Monitor Scram A.10 Main Steamline Isolation Valve Closure Scram B.2 High Flow in Main Steamline A Reactor Isolation B.3 High Flow in Main Steamline B Reactor Isolation B.4 High Temperature in Main Steamline Tunnel Reacter Isolation B.6 High Radiation in Main Steam Tunnel Reactor holation H.1 High Flow I. C. Steam Line - I. C. Isolation H.2 High Flow I. C. Condensate Line - I. C. Isolation I.1 High Radiation in Offgas Line - Offgas System Isolation K.4 APRM Upscale Pod Block The third paragraph of Note C on Table 3.1.1 is being deleted. With the above change to A.8 the paragraph is no longer necessary.

With the reactor in shutdowm or refuel, cooled down and vented, none of the above protective instrumentation functions provide any useful protection. The proposed change therefore in no way reduces the safety of the plant.

The requirements to have the following protective instrumentation functions operable are deleted during the indicated modes (Shutdown, Refuel, or Startup) when primary containment integrity is not required:

A.3 High Drywell Pressure Scram (Refuel and Startup)

E.1 High Drywell Pressure - Containment Spray (all 3 modes)

E.2 Low-Low Reactor Water Level - Cont. Spray (all 3 modes)

F.1 High Drywell Pressure - Primary Cont. Isolation (all 3 modes)

F.2 Low-Low Water Level - Primary Cont. Isolation (all 3 modes)

J.3 High Drywell Pressure - R. B. Isolation and SGTS Initiation (Shutdown and Refuel)

Specification 3.5. A.3 defines the need for primary containment integrity as follows:

" Primary containment integrity 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 low power physics tests at atmospheric pressure during or after refueling at power levels not to exceed 5 MWt."

Since the reactor is not pressurized, no accident could occur which would result is high drywell pressure. Thus the high drywell pressure signals perform no useful function and need not be operable. For the same nason, containment spray is not required, nor should the instrumentation that initiates containment spray be required. When primary containment integrity is not being maintained, a primary containment isolation signal would serve no useful purpose and need not be operable since isolation valves may be inoperable, hatches may he open, etc. Isolation is provided instead by the secondary containment, which would be isolated by high radiat.idh on the reactor building operating floor, high radiation in the reactor building ventilation exhaust, or low-low reactor water IcVel.

}f _] % _.

Technical Specification Change Request No. 75

Table 3.1.1 items D.1, 2 and 3 are modified so that the protective instrumentation functions that initiate core spray may be inoperable or bypassed when the core spray system is inoperable as allowed by Specifi-cation 3.4. A. This does not affect the safety of the plant, since the instrumentation serves no useful purpose when the core spray system is inoperable.

Table 3.1.1 items G.1, 2 and 3 are modified so that these protective functions are not required when the Automatic Depressurization System is not required to be operable. Specification 3.4.B requires the ADS to be operable when the reactor water temperature is greater than 212 F and pressurized :.bove 110 psig. When the ADS need:not be operable, the functions that initiate ADS serve no useful purpose.

Table 3.1.1 items K.1 and 2 are modified to allow two operable SRM channels rather than three as permitted by Specifications 3.9.E and F.

This change corrects an inconsistency that was overlooked when 3.9.E and F were issued in Amendment 23.

On page 3.5-2, Specification 3.5.A.4.a, the first line; the reference should be to 3.5.A.4.b rather than to 3.5.A.3.b. This was a typographical error.

The last sentence of Specification 3.7.B (page 3.7-1 and 2) is being modified to more accurately express its original intent. When one startup transformer is out or' service, no engineered safety feature equipment associated with the remaining startup transformer may be out of service.

Section 3.4.B.1 is revised to allow the pressure relief function of the electromagnetic relief valves to be inoperable or bypassed during the system hydrostatic pressure test required by ASME Code Section XI, IS-500 at or near the end of each ten year inspection interval. This allowance is necessary since the hydrostatic test pressure is above the setpoint of the relief valves. The ADS function of the valves would be maintained-when the reactor water temperature is greater than 212 F and pressurized above 110 psig. Over-pressurization protection would continue to be provided by the safety valves.

g-  %

1351 340 Technical Specification Change Request No. 75

' 1.0-1 SECTION I DEFINITIONS The following frequently used terms are defined to aid in the uniform inter-prctation of the specifications.

1.1 OPERABLE A system or component shall be considered operable when it is capable of performing its required function in its required manner.

1.2. OPERATING Operating means that a system or component is performing its required function.

1.3 POWER OPERATION

' Power operation is any operation when the reactor is in the startup mode or' run code axcept when primary containment integrity is not required.

1.4 STARTUP MODE The reactor is in the startup mode when the reactor mode switch is in

'the startup mode position. In this mode, the reactor protection system scram trips initiated by condcaser low vacuum and main steam line isola-tion valve closure are bypassed when reactor pressure is less than 600 psig; the low pressure main steamline isolation valve closure is by-passed; the IRM trips for rod block an'd scram are operable; and the SRM trips for rod block are operable.

1.5 RUN MODE ,

The reactor is in the run mode when the reactor mode switch is in the run mode position. In this mode, the reactor protection system is energized with APRM protection and the control rod withdrawal interlocks are in service.

1.6 SHUTDOWN CONDITION The reactor is in a shutdown condition when the reactor mode switch is

'in the shutdown mode position and there is fuel in the reactor vessel. In

'this condition, the reactor is subcritical, a control rod block is initiated, all operable control rods are fully inserted, and specification 3.2-A is met.

1.7 COLD SHUTDOWN The reactor is'at cold shutdown when the mode switch is in the shut-down mode position, there is fuel in the reactor vessel, all operabic control rods are fully inserted, and the reactor coolant system maintained at less than 212*F and vented. -

1331 1341

+377297-~

1.0-2 1.8 PLACE IN SHUTDOWN CONDITION Proceed with and maintain an uninterrupted normal plant shutdown opera-tion until the shutdown condition is met.

1.9 PLACE IN COLD SHUTDOWN CONDITION Proceed with and maintain an uninterrupted normal plant shutdown opera-tion until the cold shutdown condition is met.

1.10 PLACE IN ISOLATED CONDITION Proceed with and maintain an uninterrupted normal isolation of the reactor from the turbine condenser system including closure of the main steam isolation valves.

1.11 REFUEL MODE The reactor is in the refuel mode when the reactor mode switch is in the refuel mode position.and there is fuel in the_ reactor .yess.e1. In this. mode.the refueling platform interlocks are in operation.

1.12 REFUELING OUTAGE For the purpose of designating frequency of testing and surveillance, a refueling outage shall mean a regularly scheduled refueling outage; however, where such outages occur within 8 months of the end of the previous refueling outage, the test or surveillance need not be per-formed until the next regularly scheduled outage. Following the firs t refueling outage ,the time between successive tests or surveillance shall not exceed 20 months.

1.13 PRIMARY CONTAINMEhT INTEGRITY Primary containment integrity means that the drywell and adsorption chamber are closed and all of the following conditions are satisfied:

A. All non-automatic primary containment isolation valves which are not required to be open for plant operation are closed.

B. At least one door in the airlock is closed and sealed.

C. All automatic containment isolation valves are operable or are secured in the closed position.

D. All blind flanges and manways are closed.

1.14 SECONDARY C0hTAINMENT IhTEGRITY Secondary containment integrity means that the reactor building is closed and the follo' ring conditions are met:

[

. 1351 342'

2.1-3 The design basis critical heat flux correlation is based on an interrelationship of reactor coolant flow and steam quality.

Steam quality is deternined by reactar power, pressure, and coolant inlet enthalpy which in turn is a function of feeuwater temperature and water level. This correlation is based upon experimental data taken over the entire pressure range of interest in a EWR, and the correlating line was determined by the statistical mean of the experimental data.

Curves are presented for two different pressures in Figure 2.1.1.

The upper curve is based on nominal operating pressure of 1035 pcia. The lower curve is based on a pressure of 1250 psia. In no case is reactor pressure ever expected to exceed 1250 psia becaus; of protection system settines well below this value and, therefore, the curves will cover all operating conditions with

, interpolation. For pressures between 600 psia (the lower end of the critical heat flux correlatica data) and 1035 psia, the upper curve is applicable with increased margin.

The power shape used in the calculation of Figure 2.1.1 is given in Table 3.2 of Reference 10 for a peak to average pcuer l of 1.5 with a peak location at the core nidplane (X/L = 0.5).

Table 3.2 further shows an axial power shape with an axial peak of the scto eagnitude but with a peak location abcve the core midplane (X/L = 0.65). These power shapes result in total peaking factors for each fuel type as shown in Specifi-cation 2.1.A.l. The total peaking factor for each fuel type

__ is to be less than that specified in fcccion 2.1. A.1 at ratec power. When operatinz bclow rated ocwer with hicher peakinc factors as duri: g control rod nanipulation or near end cf core life, applicability of the safety limit is assured by applying the reduction factors specified in 2.1.A.2.

The feedwater temperature assumed was the maxi =um design te=per-ature output of the feedwater heaters at the given pressures and flows (e.g., 334*F at 1035 psia and 100% flow). For any lower feedvater temperature, subcooling is increased and the curves provide increased cargin.

The water level assumed in the calculations was ten inches below the reactor low water level scram point (10'-7" above the tcp of the active fuel), which is the location of the botto= of the steam separator skirts. Of course, the reactor could not be operated in this condition. As long as the water level is above this point, the safety limit curves are applicable. As long as the water level is above the bottom of the steam separator skirts, the arount of carryunder would not be increased and the core inlet en thalpy would not be influenced.

The values of the parameters involved in Figure 2.1.1 can be determined from infor=ation available in the centrol room. Reactor pressure and flow are recorded and the APRM in-core _. nuclear ,

instrumentation is calibrated in terms of percent po e .

}

M 1 Y299'

- 3.1-6a and the requirements of Table 3.1.1 are met. In order to maintain reliability of core monitoring in that quadrant where an APRM is inoperable, it is permitted to remove the operable APRM from service for calibration and/or test provided that the same core protection is maintained by alternate means.

In the rare event that Travelling In-core Pro'aes (TIPS) are used to meet the requirements 3.1.B or 3.1.C, the licensee may perform an analysis o f substitute LPRM inputs to the APRM system using spare (non-APRM input) LPRM detectors and changc the APRM system as permitted by 10 CFR 50.59.

Under assumed loss-of-coolant accident conditions it is in-advisable to allow the simultaneous starting of emergency core cooling and heavy load auxiliary systems in order to minimize the voltage drop across the emergency buses and to protect against a potential diesel generator overload. The diesel generator load sequence time delay relays provide this protective function and are set accordingly. The repetitive accuracy rating of the timer mechanism as well as parametric analyses to evaluate the maximum acceptable tolerances for the diesel loading sequence timers were considered in the establishment of the appropriate load sequencing.

Manual actuation can be accomplished by the operator and is considered appropriate only when the automatic load sequencing has been completed. This will prevent simultaneous starting of heavy load auxiliary systems and protect against the potential for diesel generator overload.

Reference:

(1)

NEDO-10189 "An Analysis of Functional Common Mode Failures in GE BWR Protection and Control Instr-mentation", L. G. Frederick, et.al, July 1970.'

13-7P300- '

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1351 344'

TABLE 3.1.1 PROTECTIVE INSTRL?.EXTATION. REQUIREMENTS

.I

~

Min. No. of.

Reactor Modes Min. No. of Operable.

Operable or Instrument in which Function Must Bc Operable Operating. Channels Per (Tripped) Trip Operable Funct!.on Trip Setting Shutdown Refuel Startup Run Systems Trip Systems Action Required

• A. Scram Insert control rods

1. Manual Scram X X X X 2 1

2. High Reactor **

X(s) X X 2 , 2 Pressure

3. liigh Drywell < 2 psig X(u) X(u) X 2 2 Pressure

4. Low Reactor ** X X X 2 2 Water Level

5. High' Water < 37 gal.  ; X(a) X X 2 2

' ~

Level in .

~~~

Scro- Dis-g charge Volume  ; ,

6. Low Condenser > 23" Hg X(b) X(b) X, n 2 2

,- Vacuum

1. ' 7. High Radiation <

10 x""gormal back- X(s) X X. 2 2 in Main Steam- ~ 8#

line Tunnel

' ~ ~

8. Average Power **

X(e,s) X(c) '

X(c)' 2 3 Range Monitor

_ (APRM)

9. Int rmediate ** X(d) X(d) 2 3

,s Range Monitor

( I101) .

c._~

O P

3.1-7s TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS Min. No. of Reactor Modes Min. No. of Operable in which Function Operabic or Instrument Must be Operable Operating Channels Per (Tripped) Trip Operable Action Function Trip Setting Shutdown Refuel Startup Run Systems Trip Systems Required *

10. Main Steamline **

X(b, s) X(b) X 2 4 Insert control Isolation Valve rods Closure

11. Turbine Trip ** X(j) 2 4 Scram

12. Generator ** X(j) 2 2 Load Rejectfon Scram X.

u 9

os .

sbd A  !

H

]

) 3.1-8 TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS (CONTD)

Min. No. of Min. No. of Operable Reactor Modes Operable or Instrument in Which Function Operating Channels Per Must Be Operable (Tripped) Trip Operable Action Function Trip Setting Shutdown Refuel Startup Run Systems Trip Systems Required

• B. Reactor Isolation Close main steam isolation valves

1. Low-Low Reactor ** X X X X 2 2 and close isola-Water Level tion condenser vent valves, or

. 2. High Flow in 3 120% rated X (s) X (s) X X 2 2 place in cold

-~^

Main Steam- shutdown condi-

}j line A tion

3. High Flow in 3 120% rated X (s) X (s) X X 2 2 u Main Steam-45= line B N

X (S) X (s) X X 2 2

4. High Tempera-3 Ambient at 0 ture in Main Power + 50 F Steamline f,, Tunnel

5. Low Pressure ** X 2 2 s in Main Steam-line X (s) X(s) X X 2 2

6. High Radiation 3 10X Normal in Main Steam Background Tunnel C. Isolation Condenser i

1. High Reactor ** X(s) X(s) X X 2 2 Place plant in i Pressure cold shutdown

2. Low-Low Reactor > 7 ' 2" above X(s) X(s) X X 2 2 l Water top of active fuel ) )

)

~

. 1 3.1-9 TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS (CONTD1 Min. No. of Min. No. of Operable Reactor Modes Operable or Instrument in Which Function Operating Channels Per Must be Operable (Tripped) Trip Operable Action Function Trip Setting Shutdown Refuel Startup Run Systems Trip Systems Required

• D. Core Spray Consider the respective core

1. Low-Low Reactor ** X(t) X(t) X(t) X 2 2 spray loop in- l Water Level operable, & com-ply with Spec.3.4

2. liigh Drywell < 2 psig X(t) X(t) X(t) X 2(k) 2(k)

P ressure

3. Low Reactor > 285 psig X(t) X(t) X(t) X 2 2 u Pressure (valve

& permissive)

"E . Containment Spray

1. High Drywell j2psig X(u) X(u) X(u) X 2(k) 2(k) Considerthecon-l 3 Pressure tainment spray Co loop inoperable

2. Lo.-Low Reactor [7'2"above X(u) X(u) X(u) X 2 2 and comply with l Water Level top of Spec. 3.4 active fuel F. Primary Containment Isolation Isolate contain-h 1. liigh Drywell [2psig X(u) X(u) X(u) X 2(k) 2(k) mentorplaceinl Pressure cold shutdown

' condition

2. Low-Low Reactor ~ 7 ' 2" ab ove X(u) X(u) X(u) X 2 2 O Water Level top of active fuel

~G. ' Automatic Depressurization

/ See note h

1. 111gh Drywell < 2 psig X(V) X (V) X (v) X 2(k) 2(k)

Pressure

. i 3.1-10 .

TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS (CONTD),

Min. No. of Reactor Modes Min. No. of Operable in Which Function Operable or Instrument Must Be Operable Operating Channels Per (Tripped) Trip Operable Action Function Trip Setting Shutdown Refuel Startup Run Systems Trip Systems Requireo*

See note h

2. Low-Low-Iow > 4' 8" above X(v) X(y) X(v) X 2 2 Reactor Water top of Level active fuel

3. AC Voltage NA X(V) X 2 2 Present auto depressurization on loss of AC power. See note i II . Isolation Condenser Isolation Isolate Affected,

1. Iligh Flow Steam < 20 psig 6 P X(s) X(s) X X 2 2 isolation con- l

? Line denser, comply Ln 2. liigh Flow Con- with Spec. 3.8

[27"O P 11 20 X(s) X(s) X X 2 2 densate Line I offgas System Isolation Isolate reactor a 1. liigh Radiation <10 x Stack X(s) X(s) X X 1 2 or trip the l 4 in Offgas Line Release limit inoperable in-(e) (See 3.6-A.1) strument channel J. Reactor Building Isolation and Isolate Reactor Standby Gas Treatment System Bldg.6 Initiate Initiation

l. liigh Radiat. ion < 100 Mr/llr X(w) X (w) X X 1 1 Standby Cas Treat-ment Systen, or l

~

Reactor Building Manual Surveill-h Operation Floor ance for not more

[ 2. Reactor Bldg.

Ventilation

< 17 Mr/Hr Xw) X (w) X X 1 1 han 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (total for all in-l

.s Exhaust struments under CD 3. liigh Drvvell < 2 psig Xu) X (u) X X 1(k) 2(k) J) in any 30-day l LO P res sure period.

I 4. Low Low Reactor 1 7'2" above X X X X 1 2 Water Level top of active fuel

. \

s.

~

l C3 '

W 3.1-11 N (E 3)

TADI.E 3.1.1 PROTECTIVE INSTittlHENTATION REQllIREMENTS (CONTD) bd , Min.No.of g Min. No. of Operable Reactor Modes Operable or Ins t rumen t in Which Function Operating Channels Per U Hunt Be' Operable (Tripped) Trip Operable Function Trip Setting Shutdown Refuel Startup Run Systems Trip Systems Action Required

• Rod Block No control rod withdrawals

1. 3Rt Upscalc < 5 x 105 cps X x(l) 1 3 (Y) permitted l

2. SR!I Downscale 1 100 cps ( ) X X(1) 1 3 (Y) _!

3. IIUI Devnscale > 5/125 fullacale(g) X ". X 2 3

4. APRM Upscale ** X (s) X X 2 .3 (c)

5. APRH Dounscale 1 2/150 fullocale X 2 3(c) 6 ., IIOf Upscale .s 108/125 fullacale X X 2 3 Condenser Vacuum ' Insert control g y 1solotton rods i

1. High Radiation d 10 x Normal During Startup and 2 2 in Mafn Steam Background .run when vacuum Tunnel pump 10 operating _

. Diesel Cenerator Time delay after Ioad Sequence Timers energiz. of relay,

,n . Contaircent Spray 40 sec + 15% X X X X 2(m) 1(n) Consider contaitunent

. Pump . .

spray loop inoperable g( nnd comply with Spec. 3.4.C(See Note q) 1351 $ r,0 c+,

ON

3.1-? ' .

TABLE 3.111 (COIIT'I).) -

Action required'vhen minimum conditions for-operation are not satisfied. Also Permissible to trip inoperable trip system.

When necessary to conduct tests and calibrations, one channel may be made inoperable for up to one hour per month without tripping'its trip systen.

• • c ce Specification 2.3 for Limiting Safety System Settings, m gg Notes; C 6 0

a. 23150 Pernissibic to bypass, with control rod block, for reactor protection system reset in refuel mode. __

~

b. Pernissible to bypass below 600 psig in refuel and startup nedes. EE50 c.

One (1) APRM in cach operable trip system may be bypassed or inoperable 'provided the rcquirements of specification 3.1.C and 3.10.D are satisfied.

bypa'ssed except as noted below or permitted by noteTwo APRM's in the same quadrant shall not be concurrently ?pC/

-' l (sa LJ7 Any one APRM may be removed from service for up to one hour for test or calibration without inserting .h

--' trips in its trip system only if the remaining operable APRM's meet the requirements of specification 3.1.B.1 and no control rods are moved outward during the calibration or test. During this short period, q,, the requirements of specifications 3.'1.B.2, 3.1.C and 3.10.D need not be net, m

d. The (11U1) shc11 he inserted and operable until the AritM's .tre operabic and reading at least 2/150 full scale.

e .' Air ejector iso.lation valve closure tiac delay shall not exceed 15 ninutes.

f. l'nless SRM chonbers are fully inserted.

g. 1 ot a[1plicable uhen I101 on louest range.

\

%. One i ns t rux.cnt channel in each trip system may be inoperable provided the circuit which it operates in the tjs t rip system is placed in a simulated tripped condition. If repairs cannot be completed within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> the j ref ac t<>r : ball be placed in the cold shutdown condit len.

W If more than one instrument channel in any trip system becernes inoperahlc the reactor shall be placed in thc' cold shutdown conditien. Relief valve controllers shsil not he bypasced for r' ore than 3 hou n (total Line for all contro]Jers) in any 30-day 9[ pr:rlad an i only ode r el ief valve control ler r ay be bypasne.I' at a time.

• TABLE 3.1.1 (CON ' D .)

i. The interlock is not required during the start-up test program and demonstration of plant electrical output but shall be provided following these actions.

j. Not required below 40's of turbine rated steam flow.

k. All four (4) drywe 1 pressure instrument channels may be made inoperable during the integrated primary containment leakage rate test (See Specification 4.5), provided that primary containment integrity is not required and tha,t no work is performed on the reactor or its corinected systems which could result in lowering the reactor water level to less than 4'8" above the top of the active fuel.

T

1. Bypassed in IRF! Ranges 8, 9, 6 10. 6 9 6 9

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

n. One. time delay relay per pump must be operable, y

o. Th'e re are two time delay relays associated with each of two pumps.

[5d c9

p. Two time delay relays per pump must be operable.

( r4 g q. blanual initiation of affected component can be accomplished after the automatic load sequencing is g

- completed. -

b

u. r. Time delay starts after closing of containment spray pump circuit breaker.

If N s, These functions not required to be operable with the reactor emperature less than 212*F and the vessv1 head removed or vented.

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t. These functions may be inoperabic or bypassed when corresponding portions of the core spray system are inoperabic per Specification 3,4,A.

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

CD v. These functions not required to be operable when the ADS is not rcquired to be operabic.

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w. These functions must be operabic only when irradiated fuel is in the fuel pool or reactor vessel and i secondary containnient' integrity is required per specification 3.5.B.

y. The number of operabic channels may be reduced to 2 per Specification 3.9-E and F.

3.4-lb that the pump and any necessary valves can be started or operated from the control room or from local control stations ar.o the torus is r:cchanically intact.

B. Automatic Depressurization System

1. Five electromatic relief valves of the automatic depressurization system shall be operabic when the reactor water temperature is greater than 212*F and pressurized above 110 psig, except as specified in 3.4.B.2. The automatic pressure relief function of these valves (but not the automatic depressurization function) may be inoperable or bypassed during the system hydrostatic pressure test required by ASME Code Section XI, IS-500 at or near the end of each ten year inspection interval.

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4. _ Reactor Building to Suppression Chamber Vacuum Breaker System

a. Except as specified in Specification 3.5.A.4.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 differential 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 of not greater than 0.5 psid acting on the vacuum breaker disc.

b. From the time that one of the reactor building to suppression chamber vacuum breakers is made or found to be inoperable, the vacuum breaker shall be locked closed and reactor operation is permissible only during the succeeding seven days unless such vacuum breaker is made operable sooner, provided that the procedure does not violate primary containment integrity.

c If the limits 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 />.

5. Pressure Suppression Chamber - Drvwell Vacuum Breakers

a. When primary containment is required, all suppression chamber - drywell vacuum breakers 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 demonstrated to open from closed to fully open with the applied force at all valve 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.

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3.7-1 3.7 AUXILIARY ELECTRICAL POWER Applicability: Applies to the operating status of the auxiliary electrical power supply.

Objective: To assure the operability of the auxiliary electrical power supoly.

Specification: A. The reactor shall not be made critical unless all of the following requirements are satisfied:

1. The following buses or panels energized.

a. 4160 volt buses 1C and 1D in the turbine building switchgear room.

b. 460 volt buses lA2,1B2, lA21,1B21 vital MCC 1A2 and 1B2 in the reactor building switch r ar room: lA3 and 1B3 at the intake structure; 1A21A, 1B21A, lA21B, and 1B21B and isolation valve MCC 1AB2 on 23' 6" elevation in the reactor building; 1A24 and 1B24 at the stack.

c. 208/120 volt panels 3, 4, 4A, 4B, 4C and VACP-1 in the reactor building switchgear room,

d. 120 volt protection panel 1 and 2 in the cable room,

e. 125 volt DC distribution centers A and B, and panel D in the battery room; isolation valve motor control center DC-1 on 23'6" elevation in reactor building and panel E in the cable room.

f. 24 volt D.C. power panels A and B in the cable room.

2. One 230 KV line is fully operational and switch gear and both startup transf omers are energized to carry power to the station 4160 volt AC buses and carry power to or away from the plant.

3. An additional source of power consisting of one of the following is in service connected to feed the appropriate plant 4160 V bus or buses:

a. A second 230 KV line fully operational.

b. One 34.5 KV line fully operational.

4. The station batteries are available for normal service and a battery charger is in service for each battery, except one battery and associated charger may be removed from service as required for surveillance testing as set forth in Specification 4.7.B.

B. The reactor shall be placed in the cold shutdown condition if the availability of power falls below that required by Specification A above, except that the reactor may remain

in operation for a period not to exceed 7 days in any 30 i

_ day period if a startup transformer is out of service. I l __h

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D "" D D D' 3.7-2 o c Ju o 2 k} ft _m s-None of the engineered safety feature equipment fed by the remaining transformer may be out of service.

C. Standby Diesel Generators

1. The reactor shall not be made critical unless both diesel generators are operable and capable of feeding their designated 4160 volt buses.

2. If one diesel generator becomes inoperable during power operation, repairs shall be initiated i= mediately and the other diesel shall be operated at least one hour every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at greater than 20% rated power until repairs are completed. The reactor may remain in operation for a period not to exceed 7 days in any 30-day period if a diesel generator is out of service. During the repair period none of the engineered safety features normally fed by the operational diesel generator may be out of service or the reactor shall be placed in the cold shut-down condition.

3. If both diesel generators become inoperable during power operation, the reactor shall be placed in the cold shut-down condition.

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4. For the diesel generators to be considered operable there shall be a minimum of 14,500 gallons of diesel fuel in the standby diesel generator fuel tank.

Bases: The general objective is to assure an adequate supply of power with at least one active and one standby source of power available for operation of equipment required for a safe plant shutdown, to maintain the plant in a safe shutdown condition and to operate the required engineered safety feature equipment following an accident.

AC power for shutdown and operation of engineered safety feature equipment can be provided by any of four active (two 230 KV and two 34.5 KV lines) and either of two standby (two diesel generators) sources of power. Normally all six sources are available.

However, to provide for maintenance and repair of equipment and still have redundancy of power scurces the requirement of one active and one standby source of power was established. The plant's main generator is ..ot given credit as a source since it is not available during shutdown. The plant 125V DC power is normally supplied by two batteries, each with an associated charger. A third charger is available to supply either battery.

These chargers are active sources and supply the normal 125V DC requirements with the batteries as sta rdby sources.

In applying the minimum requirement of one active and one standby source of AC power, since both 230 KV lines are on the same set of towers, either one or both 230 KV lines are considered as a single active scurce. ,

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Jersey Central Power & Light Company

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• LL Q td 91 Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960 (201)455-8200 November 16, 1979 Director of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555

Dear Sir:

1

Subject:

Oyster Creek Generating Station Docket No. 50-219 ,

Technical Specification Change Request No. 75 In accordance with 10 CFR 50.59, Jersey Central Power & Light Company, owner and operator of the Oyster Creek Nuclear Generating Station, Provisional Operating License No. DPR-16, request changes to Appendix A of that license.

' Ibis Technical Specification Change will revise sections 1.0, 2.1, 3.1, 3.4, 3. 5, and 3.7. Tncsc cnanges will facilitate the epcoming refueling outoge (scheduled to start January 5, 1979) by removing technically unnecessary operability requirements and surveillance.

This Technical Specification Change Request has been reviewed and approved by the Station Manager, the Plant Operations Review Committee and an Independent Safety Review Group in accordance with Section 6.5 of the Oyster Creek Technical Specifications.

We have determined that this submittal is Class III per 10 CFR 170.22 and have enclosed a check for $4000 pursuant to that section.

Very truly yours, Ivan R. Fin ock, r.

Vice Presi nt la Enclosure 43M '*3-1351 357 Jersey Central Power & Light Company is a Member of the General Pubhc Utihties System

JEFSEY CEA' TRAL PQiER & LIGHT CQ4PANY OYSTER CREEK NUCLEAR GENERATING STATICN Provisional Operating License No. DPR-16 Tecnnical Specification Change Request No. 75 Docket No. 50-219 Applicant submits, by this Technical Specification Change Request No. 75 to the Oyster Creek Nuclear Generating Station Technical Specifications, revised sections 1.0, 2.1, 3.1, 3.4, 3.5, and 3.7.

JERSEY CD' ITAL ' R & LIGHT OCMPANY BY: 7 V ' ' ~ "V1'c~

Me f nt STNIE OF NEW JERSEY COUtTIY CF MORRIS Sworn and subscribed to before me on this t' // day of /!v<. . 4 <. ,

1979.

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Notary Public Th : '1/,. re;3 NOTARY PUit!C CF Nrw y;43ey My Commhslon Lp;res An, J$,39g 1-374 3C T351 358-