Text

Proposed Tech Specs,Modifying Safety Limits & Surveillances of LPRM & APRM Sys & Related Bases to Ensure APRM Channels Respond within Necessary Range & Accuracy & to Verify Channel Operability
	ML20155J750
Person / Time
Site:	Oyster Creek
Issue date:	11/05/1998
From:	; GENERAL PUBLIC UTILITIES CORP.
To:	;
Shared Package
ML20155J687	List: ML20155J685; ML20155J714; ML20155J750;
References
	NUDOCS 9811120289
	Download: ML20155J750 (31)
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2.3 LIMITING SAFETY SYSTEM SETFINGS p Apolicability: Applies to trip settings on automatic protective devices related to variables on which l safety limits have been placed.

Obiective: To provide automatic corrective action to prevent the safety limits from being exceeded.

l Soecification: Limiting safety system settings 'shall be as follows:

FUNCTION LIMITING SAFETY SYSTEM SETTINGS )

t.

)

A. Neutron Flux, Scram I

A.1 'APRM When the reactor mode switch is in the Rnn position, i the APRM flux scram setting shall be i

EEP l S s [(0.90 x 10-6) W + 60.8) MFLPD l

[ with a maximum setpoint of 115.7% for core flow equal !

l to 61 x 106lb/hr and greater,

! i where: l S= setting in percent of rated power W= recirculation flow (Ib/hr)

FRP = fraction of rated thermal power is the ratio of core l thermal power to rated thermal power -

l MFLPD = maximum fraction oflimiting power density where the ]

I~ limiting power density for each bundle is the design -1 linear heat generation rate for that bundle. l L The ratio of FRP/MFLPD shall be set equal to 1.0 unless

! the actual operating value is less than 1.0 in which case j- the actual operating value will be used.

l This adjustment may be accomplished by increasing the APRM gain and thus reducing the flow reference APRM j l '

High Flux Scram Curve by the reciprocal of the APRM L gain change.

A.2 IRM s 38.4 percent of rated neutron flux j 1

I A.3 APRM Downscale 2 2% Rated Thennal Power coincident with IRM Upscale (high-high)orInoperative

OYSTER CREEK 2.3-1 Amendment No.: 71,75,111 L

9811120289 981105 s~

PDR ADOCK 05000219 p PDR _

FUNCTION LIMITING SAFETY SYSTEM SETFINGS B. . Neutron Flux, Control Rod The Rod Block setting shall be ]

Block- i l EBZ -l 4

S s [(0.90 x 10 ) W + 53.1] [MFLPD) with a maximum setpoint of 108% for core flow equal to 61 '

x 10' lb/hr and greater.

The definitions of S, W, FRP and MFLPD used above for j the APRM scram trip apply.

l The ratio of FRP to MFLPD shall be set equal to 1.0 unless ;

the actual operating value is less than 1.0, in which case the l L actual operating value will be used.

l This adjustment may be accomplished by increasing the APRM gain and thus reducing the flow referenced APRM l rod block curve by the reciprocal of the APRM gain L change.

l ,

C. Reactor High Pressure, $1060 psig .l Scram

! D. Reactor High Pressure, 2 @ s 1085 psig Relief Valves Initiation. 3 @ s 1105 psig E. Reactor High Pressure, s1060 psig with time delay Isolation Condenser s3 seconds Initiation I F. Reactor High Pressure, 4 @ 1212 psig 12 psi l Safety Valve Initiation 5 @ 1221 psig 112 psi !

G. Low Pressure Main Steam 2 825 psig (initiated in IRM Line, range 10) ;

MSIV Closure H. Main Steam Line Isolation $10% Valve Closure from full open Valve Closure, Scram 4

, OYSTER CREEK 2.3-2 Amendment No. 71,75,111,150,164,177

.v. -

FUNCTION LIMITING SAFETY SYSTEM SETTINGS

1. - Reactor Low Water Level, 211'5" above the top of the active fuel as indicated Scram under normal operating conditions

- J. Reactor Low-Low Water 27'2" above the top of the active fuel as indicated Level, Main Steam Line under normal operating conditions 3 Isolation Valve Closure 1

K. Reactor Low-Low Water Level, 27'2" above the top of the active fuel Core Spray initiation l I

L. Reactor Low-Low Water Level, 27'2" above the top of the active Fuel with time Isolation Condenser Initiation delay <3 seconds j i

M. Turbine Trip, Scram 10 percent turbine stop valve (s) closure from full . l open l

N. Generator Load Rejection, Scram Initiate upon loss of oil pressure from turbine acceleration relay O. DELETED P. Loss of Power

1) 4.16 KV Emergency Bus 0 volts with 3 seconds Undervoltage (Loss of 0.5 seconds time delay j Voltage) -

. 2) 4.16 KV Emergency Bus 3840 (+20V, -40V) volts Undervo!': age (Degraded 10 '10% (1.0) second time delay Voltage)

OYSTER CREEK . 2.3-3 Amendment No. 75,80,174,175 .

. - -. ~. -- -. -. ..- -~. -- .-

l 2.3 LIMITING SAFETY SYSTEM SETTINGS l

L Basesi i Safety limits have been established in Specifications 2.1 and 2.2 to protect the integrity of the l fuel cladding and reactor coolant system barriers, respectively. Automatic protective devices l

l have been provided in the plant design for corrective actions to prevent the safety limits from I

being exceeded in normal operation or operational transients caused by reasonably expected l single operator error or equipment malfunction. This Specification establishes the trip settings for these automatic protection devices, j l

t l The Average Power Range Monitor, APRMm, trip setting has been established to assure never reaching the fuel cladding integrity safety limit. The APRM system responds to changes in :

neutron flux. However, near the rated thermal power, the APRM is calibrated using a plant heat !

balance, so that the neutron flux that is sensed is read out as percent of the rated thermal power. ,

For slow maneuvers, such as those where core thermal power, surface heat flux, and the power I transferred to the water follow the neutron flux, the APRM will read reactor thermal power. For fast transients, the neutron flux will lead the power transferred from the cladding to the water due l to the effect of the fuel time constant. Therefore, when the neuron flux increases to the scram )

setting, the percent increase in heat flux and power transferred to the water will be less than the ;

l percent increase in neutron flux.

The APRM trip setting will be varied automatically with recirculation flow, with the trip setting i at the rated flow of 61.0 x 10' lb/hr of greater being 115.7% of rated neutron flux. Based on a complete evaluation of the reactor dynamic performance during normal operation as well as l expected maneuvers and the various mechanical failures, it was concluded that sufficient l protection is provided by the simple fixed scram' setting (2,3). However, in response to expressed beliefs (4) that variation of APRM flux scram with recirculation flow is a prudent measure to ensure safe plant operation, the scram setting will be varied with recirculation flow.

An increase in the APRM scram trip setting would decrease the margin present before the fuel cladding integrity safety limit is reached. The APRM scram trip setting was determined by an analysis of margins required to provide'a reasonable range for maneuvering during operation.

Reducing this operating margin would increase the frequency of spurious scrams, which have an adverse effect on reactor safety because of the resulting thermal stresses. Thus, the APRM scram

! trip setting was selected because it provides adequate margin for the fuel cladding integrity safety l

l limit and yet allows operating margin that reduces the possibility of unnecessary scrams.

The scram trip setting must be adjusted to ensure that the LHOR transient peak is not increased for any combination of maximum fraction oflimiting inwer density (MFLPD) and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification l 2.3.A, when the. MFLPD is greater than the fraction of the rated power (FRP). the adjustment may be accomplished by increasing the APRM gain and thus reducing the flow referenced

APRM High Flux Scram Curve by the reciprocal of the APRM gain change.

i OYSTER CREEK 2.3-4 Amendment No. 71,75 l

l-1

For operation in the Startup mode while the reactor is at low pressure, the IRM range 9 High Flux scram setting of 12% of the rated power provides adequate thermal margin between the j maximum power and the safety limit of 18.3% of rated power to accommodate anticipated maneuvers associated with power plant startup. There are a few possible sources of rapid reactivity input to the system in the low power / low flow condition. Effects ofincreasing

pressure at zero or low void content are minor, because cold water from sources available during the startup is not much colder than that already in the system, temperature coefficients are small, .

and control rod sequences are constrained by operating procedures backed up by the rod worth ,

minimizer. Wwth ofindividual rods is very low in a constrained rod pattera. In a sequenced rod j withdrawal approach to the scram level, the rate of power rise is no more than five percent of the !

j rated per minute, and the IRM system would be more than adequate to assure a scram before the i power could exceed the safety limit.

l l

To continue operation beyond 12% of rated power, the IRMs must be transferred into range 10. ,

The Reactor Protection System is designed such that reactor pressure must be above 825 psig to successfully transfer the IRMs into range 10, thus assuring protection for the fuel cladding safety

limit. The IRM scram remains active until the mode switch is placed in the RUN position at l which time the trip becomes a coincident IRM upscale, APRM downscale scram.

The adequacy of the IRM scram was determined by comparing the scram level on the IRM range l 10 to the scram level on the APRMs at 30% of rated flow. The IRM scram is at 38.4% of rated j power while the APRM scram is at 52.7% of rated power. The minimum flow for Oyster Creek E is at 30% of rated power and this would be the lowest APRM scram point. The increased recirculation flow to 65% of flow will provide additional margin to CPR Limits. The APRM scram at 65% of rate flow is 87.1% of rated power, while the IRM range 10 scram remains at L 38.4% of rated power. Therefore, transients requiring a scram based on flux excursion will be

! terminated sooner with a IRM range 10 scram than with an APRM scram The transients l requiring a scram by nuclear instrumentation are the loss of feedwater heating and the improper l startup of an idle recirculation loop. The loss of feedwater heating transient is not affected by the L range 10 IRM since the feedwater heaters will not be put into service until after the LPRM l downscales have cleared, thus insuring the operability of the APRM system. This will be

l. administratively controlled. The improper startup of an idle recirculation loop becomes less l severe at lower power level and the IRM scram would be adequate to terminate the flux >

excursion.

ihe Rod Worth Minimizer is not required beyond 10% of rated power. The ability of the IRMs l to terminate a rod withdrawal transient is limited due to the number and location ofIRM ,

l detectors. An evaluation was performed that showed by maintaining a minimum recirculation flow of 39.65x10' lb/hr in range 10 a complete rod withdrawal initiated at 35% of rated power or i less would not result in violating the fuel cladding safety limit. Therefore, a rod block on the '

IRMs at less than 35% of rated power would be adequate protection against a rod withdrawal transient.

I l

t L

l' OYSTER CREEK 2.3-5 Amendment No.: 71

Reactor power level may be varied by moving control rods or by varying the recirculation flow rate. The APRM system provides a control rod block to prevent gross rod withdrawal at constant recirculation flow rate to protect against grossly exceeding the MCPR Fuel Cladding Integrity Safety Limit. This rod block trip setting, which is automatically varied with recirculation loop -

flow rate, prevents an increase in the reactor power level to excessive values due to control rod withdrawal. The flow variable trip setting provides substantial margin from fuel damage, assuming a steady-state operation at the trip setting, over the entire recirculation flow range. The margin to the safety limit increases as the flow decreases for the specified trip setting versus flow relationship. Therefore, the worst-case MCPR, which could occur during steady-state operation, is at 108% of the rated thermal power because of the APRM rod block trip setting. The actual power distribution in de core is established by specified control rod sequences and is monitored continuously by the incore LPRM system. As with APRM scram trip setting, the APRM rod block trip setting is adjusted downward if the maximum fraction oflimiting power density exceeds the fraction of the rated power, thus preserving the APRM rod block safety margin. As with the scram setting, this may be accomplished by adjusting the APRM gains.

The settings on the reactor high pressure scram, anticipatory scrams, reactor coolant system relief valves and isolation condenser have been established to assure never reaching the reactor coolant system pressure safety limit as well as assuring the system pressure does not exceed the range of the fuel cladding integrity safety limit. In addition, the APRM neutron flux scram and the turbine bypass system also provide protection for these safety limits, e.g., turbine trip and loss of electrical load transients (5). In addition to preventing power operation above 1060 psig, the pressure scram backs up the other scrams for these transients and other steam line isolation type transients. Actuation of the isolation condenser during these transients removes the reactor decay heat without further loss of reactor coolant thus protecting the reactor water level safety limit.

The reactor coolant system safety valves offer yet another protective feature for the reactor coolant system pressure safety limit since these valves are sized assuming no credit for other pressure relieving devices. In compliance with Section I of the ASME Boiler and Pressure Vessel Code, the safety valve must be set to open at a pressure no higher than 103% of design pressure, and they must limit the reactor pressure to no more than 110% of design pressure. The safety valves are sized according to the Code for a condition of main steam isolation valve closure while operating at 1930 MWt, followed by (1) a reactor scram on high neutron flux, (2) failure of the recirculation pump trip on high pressure, (3) failure of the turbine bypass valves to open, and (4) failure of the isolation condensers and relief valves to operate. Under these conditions, a total of 9 safety valves are required to turn the pressure transient. The ASME B&PV Code allows a il% of working pressure (1250 psig) variation in the lift point of the valves. This variation is recognized in Specification 4.3.

OYSTER CREEK 2.3-6 Amendment No.: 71,75,111,150

The low pressure isolation of the main steam line at 825 psig was provided to give protection against fast reactor depressurization and the resulting rapid cool-down of the vessel. The low-pressure isolation protection is enabled with entry into IRM range 10 or the RUN mode. In addition, a scram on 10% main steam isolation valve (MSIV) closure anticipates the pressure and flux transients which occur during normal or inadvertert isolation valve closure. Bypass vf the MSIV closure scram function below 600 psig is permitted to provide sealing steam and allow the

, establishment of condenser vacuum. Advantage is taken of the MSIV scram feature to provide protection for the low-pressure portion _of the fuel cladding integrity safety limit. To continue operation beyond 12% of rated power, the IRM's must be transferred into range 10. Reactor pressure must be above 825 psig to successfully transfer the IRM's into range 10. Entry into range 10 at less than 825 psig will result in main steam !ine isolation valve closure and MSIV closure scram. This provides automatic scram protection for the fuel cladding integrity safety limit which allows a maximum power of 25% of rated at pressures below 800 psia. Below 600 psig, when the MSIV closure scram is bypassed, scram protection is provided by the IRMs.

Operation of the reactor at pressure lower than 825 psig requires that the mode switch be in the STARTUP position and the IRMs be in range 9 or lower. The protection for the fuel clad integrity safety limit is provided by the IRM high neutron flux scram in each IRM range. The IRM range 9 high flux scram setting at 12% of rated power provides adequate thermal margin to the safety limit of 25% of rated power. There are few possible significant sources of rapid reactivity input to the system through IRM range 9: effects ofincreasing pressure at zero and low void content are minor; reactivity excursions from colder makeup water, will cause an IRM high flux trip; and the control rod sequences are constrained by operating procedures backed up by the rod vorth minimizer. In the unlikely event of a rapid or uncontrolled increase in reactivity, the IRM system would be more than adequate to ensure a scram before power could exceed the safety limit. Furthermore, a mechanical stop on the IRM range switch requires an operator to pull up on the switch handle to ,

pass through the stop and enter range 10. This provides protection against an inadvertent entry into j range 10 at low pressures. The IRM scram remains active until the mode switch is placed in the RUN position at which time the trip becomes a vincident IRM upscale, APRM downscale scram.

The low level water level trip setting of I l'5" above the top of the active fuel has been established to assure that the reactor is not operated at a water level tielow that for which the fuel cladding ;

integrity safety limit is applicable. With the scram set at this point, the generation of steam, and ;

thus the loss ofinventory is stopped. For example, for a loss of feedwater flow a reactor scram at the valoe indicated and isolation valve closure at the low-low water level set point results in more !

than 4 feet of water remaining above the core after isolation (6).

During periods when the reactor is shut down, decay heat is present and adequate water level must i be maintained to provide core cooling. Thus, the low-low level trip point of 7'2" above the core is l provided to actuate the core spray system (when the core spray system is required as identified in Section 3.4) to provide cooling water should the level drop to this point.*

The turbine stop valve (s) scrum is provided to anticipate the pressure, neutron flux, and heat flux increase caused by the rapid closure of the turbine stop valve (s) and failure of the turbine bypass system.

OYSTER CREEK 2.3-7 Amendment No.195

Correction 11/30/87

ne generator load rejection scram is provided to anticipate the rapid increase in pressure and neutron flux resulting from fast closure of the turbine control valves to a load rejection and failure of the turbine bypass system. This scram is initiated by the loss of turbine acceleration relay oil pressure. He timing for this scram is almost identical to the turbine trip.

The undervoltage protection system is a 2 out of 3 coincident logic relay system designed to shift emergency buses C arid D to on-site power should normal power be lost. There is a separate 2 out of 3 coincident logic relay designed to shift emergency buses to on-site power shou:d normal power be degraded to an unacceptable level. He trip points and time delay settings have been selected to assure an adequate power source to emergency safeguards systems in the event of a total loss of normal power or degraded conditions which would adversely affect the functioning of engineered safety features connected to the plant emergency power distribution system.

The APRM downscale signal insures that there is adequate Neutron Monitoring System protection if the reactor mode switch is placed in the run position prior to APRMs coming on scale. With the reactor mode switch in run, an APRM downscale signal coincident with an associate IRM Upscale (High-High) or Inoperative signal generates a trip signal. This function is not specifically credited in the accident analyses but it is retained for overall redundancy and diversity of the RPS.

References (1) FDSAR, Vo ume 1, Section VII-4.2.4.2 (2) FDSAR, Amendment 28, Item III.A-12 (3) FDSAR, Amendment 32, Question 13 (4) Letters, Peter A. Morris, Director, Division of Reaction Licensing, USAEC, to John E.

Logan, Vice President, Jersey Central Power and Light Company (5) FDSAR, Amendment 65, Section B.XI (6) FDSAR, Amendment 65, Section B.IX OYSTER CREEK 2-3.8 Amendment No. 80,175,195

%4 1 M

, . TABLE 3.1.1

. Sheet I of 13 PROTECTIVE INSTRUMENTATION REQUIREMENTS Minimum Number of Minimum Number of.

- Reactor Modes in Which OPERABLE or Instrument Channels Per Function Must Be Operable OPERATING [ tripped] OPERABLE Trio Settina Shutdown Refuel - Startuo Run Trio Systems . Trio System . _A_ption Reauired*

Function

- A. Scram X X X X 2 1 ' Insert

1. Manual Scram control

" X 2 2(an) . rods

2. High Reactor Pressure X(s) X(II)

3. High Drywell Pressure $ 3.5 psig X(u) X(u) X 2 2(nn)

4. Low Reactor Water Level

" X X X '2 2(nn)

5. o. High Water Level in $ 29 gal. X(a) X(z) X(z) 2 2(nn)

Scram' Discharge Volume North Side

b. High Water Levelin 5 29 gal. X(a) X(z) X(z) 2 2(nn)-

Scram Discharge Volume South Side

6. Low Condenser Vacuum > 20 in. hg. X(b) X 1 3(mm)(nn)

7. DELETED .

OYSTER CREEK 3.1-9 ,

Change: 4,8 Amendment No.: 20,44,63,79,112,130,131,149,162,169,171 .

- .,. . . n ..,-,,.n, -n.--~

~

TABLE 3.1.1 '

Sheet 2 of13 PROTECrlVE INSTRUMENTATION REQUIREMENTS Minimum Number of Minimum Numberof Reactor Modes in Which OPERABLE or Instrument Channels Per Function Must Be Operable OPERATING [ tripped]

Function Trio Settine OPERABLE Shutdown Refuel Startup Run Trio Systems Trio System Action Reauired*

8. Average Power Range **

X(c,s) X(c) X(c) 2 Monitor (APRM) 3(nn)

9. Intermediate Range Monitor "

X(d) X(d) 2 3(nn)

(IRM)

10. Main Steamline Isolation "

X(b,s) X(b) X 2 Valve Closure 4(nn)

11. Turbine Trip Scram "

X(j) 2 4(nn)

12. Generator Load Rejection **

Scram X(j) 2 2(nn) i

13. APRM Downscale/IRM ' "

Upscale X(c) 2 3(nn)

B. Reactorisolation

1. Low-Low Reactor Water " X X X X 2 Level 2(oo) Close Main Steam Isolation Valves and Close Isolation 1 Condenser Vent Valves

2. High Flow in Main or, s 120% rated X(s) X(s) X X 2 Steamline A 2(oo) PLACE IN COLD SHUTDOWN

! OYSTER CREEK 3.1-10 1

i

TABLE 3.1.1 Sheet 2 of13 PROTECTIVE INSTRUMENTATION REQUIREMENTS Minimum Number of Minimum Number of Reactor Modes in Which OPERABLE or_ Instrument Channels Per Function Must Be Operable OPERATING [trippedj OPERABLE Trio Setting Shutdown Refuel Startup Run Trio Systems Trio System Action Recuired' Function

8. Average Power Range

- X(c,s) X(c) X(c) 2 3(nn)

Monitor (APRM)

9. Intermediate Range Monitor " X(d) X(d) 2 3(nn)

(IRM)

10. Main Steamline Isolation

- X(b,s) X(b) X 2 4(nn)

Valve Closure

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

12. Generator Load Rejection ** X(j) 2 2(nn)

Scram

' ** _2 3(nn)

13. APRM Downscale/IRM _X(c)

Upscale B. ReactorIsolation

" X X X X 2 2(oo) Close Main Steam

1. Low-Low Reactor Water Isolation Valves and Level '

Close Isolation Condenser Vent Valves ,

or, s; 120% rated X(s) X(s) X X 2 2(oo) PLACE IN COLD

2. High Flow in Main Steamline A SHUTDOWN i

OYSTER CREEK 3.1-10 r

_ _ = _ _ _ . . . _ . _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ _ = _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ . _ - _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ . _ _ _ . _ _ _

Amendment No.: 44,171 .

TABLE 3.1.1 Sheet 3 of 13 PROTECTIVE INSTRUMENTATION REQUIREMENTS .

Reactor Modes Minimum Number of Minimum Number of in which Function OPERABLE or Instrument Channels Per -

Must Be OPERABLE OPERATING [ tripped] OPERABLE Trio Setting Shutdown - Refuel Startup Run Trio Systems Trio System Action Reauired*

Function

3. High Flow in Main Steam- $12ure rated X(s) X(s) X X 2 2(oo) line B

4. High Temperature in Main $ Ambient at Pow ~er X(s) X(s) X X 2 2(oo)

Steamline Tunnel + 50 F

5. Low Pressure in Main

" X(cc) X 2 2(oo)

Steamline

6. DELETED C. Isolation Condenser Initiation

" X(s) X(ll) X 2 2(pp) PLACE IN COLD

1. liigh Reactor Pressure X(s)

X X 2 2(pp) SilUTDOWN

2. Low-Low Reactor >7'2" above TOP of X(s) X(s)

CONDITION Water Level ACTIVE FUEL D. Core Spray X(t) X(t) X 2 2(pp) Consider the

1. Low-Low Reactor X(t) respective Water Level core spray loop

2. High Drywell Pressure s 3.5 psig X(t) X(t) X(t) X 2(k) 2(k)(pp) inoperable and -

comply with X 2 2(pp) Spec 3.4

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

(valve permissive)

OYSTER CREEK 3.1-11 -

Amendment No.: 44,717^,112,131,169,i?1

TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS Sheet 4 of 13 Reactor Modes Minimum Number of Minimum Number of in which Function OPERABLE or Instrument Channels Must Be OPERABLE OPERATING [ tripped] Per OPERABLE Shutdown - defuel Startup Trio Systems Trio System Action Reouired' Function Trio Settine Run E. Containment Soray Comply with Technical Specification 3.4 F. Primary Containment Isolation

1. High Drywell Pressure s 3.5 psig X(u) X(u) X(u) X 2(k) 2(k)(oo) Isolate contalment or PLACEIN COLD

2. Low-Low Reactor 2 7'2" above TOP of X(u) X(u) X(u) X 2 2(oo) SHUTDOWN CONDITION Water Level ACTIVE FUEL G. Automatic Depressurization

1. High Drywell Pressure s 3.5 psig X(v) X(v) X(v) X 2(k) 2(k) See note h

2. Low-Low-Low 24'8" above TOP of X(v) X(v) X(v) X 2 2 See note h Reactor Water Level ACTIVE FUEL

3. Core Spray Booster >21.2 psid X(v) X(v) X(v) X Note i Note i See note 1 Pump d/p Pennissive H. Isolation Condenser Isolation (See Note hh)

1. High Flow Steam Line s 20 psig P X(s) X(s) X X 2 2(oo) Isolate affected Isolation Condenser !

s 27" P H2O X(s) X X 2 -2(oo) comply with Spec

2. High Flow Condensate X(s) 3.8. See note dd ,

Line I

OYSTER CREEK 3.1-12 Amendment No.: 44,79,108,112,160,171,190,195 Chmge 4; Correction: 5/11/84 i

_ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ . _ _ _ _ _ . _ _ - . _ . _ _ _ _ _ __m_ _ _ _ _ _ _ . _ _ _ _ _ . . . _ _ - _ _ _ _ -_ . . . . . - _

j TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS Sheet 5 ef13 Reactor Modes ' Minimum Number of - ' Minimum' Number o' f-

-in which Function OPERABLE or - Instrument Channels Per! -

Must Be OPERABLE OPERATING [tnpped] OPERABLE Startup Trio Systems Trio System ' . Action Reauired*

Function Trio Settina Shutdown Refuel R!!!! ,

. I. Offnas System Isolation X 2(ii) See noteji

1. High Radiation In s 2000 mrem /hr X(s) X(s) X 1(ii)

Offgas Line(e)

J. Reactor Building Isolation and Standby Gas Treatment System Initiation .

Isolate Reactor X .X- 1 Building and

1. High Radiation s 100 mR/hr ' X(w) X(w) 1

. Initiate Standby .

Reactor Building Gas Treatment - !

Operating Floor System or Manual' X X 1 Surveillance for . i

2. ReactorBuilding s 17 mR/hr X(w) X(w) 1 not more than 24 Ventilation Exhaust Hours (Total for X X 0 ' 2(k) all %struments

3. High Drywell Pressure s 3.5 psig X(u) X(u) '

underJ)in any X X 2 30-day pened.

4. Low-Low Reactor 2 7'2" above TOP of X X Water Level ACTIVE FUEL q K. Rod Block

.)

X -2 No control rod s5x105cps

- X(1) 1

1. SRM Upscale withdrawals i

-2 pennitted

2. SRM Downscale 2100 cps (f) X- X(I) 1 t

i 3. IRM Downscale 2 /125 fullscale(g) X X 2 -3 I

l

^

OYSTER CREEK 3.1-13 t

Amendment No.: 44,72,75,79,91,112,171,191 Chr.nge: 4

-)

_ . - . . - _ . . . . _ . . .-. -. . - . . . - . - - . . - . . - . .-- ....-c . ~ _ . -- . _ _ _ _ - _ _ _ . _ _ _ _ _ _ _ _ - .

' TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS -

Sheet 6 of13 Reactor Modes Minimum Numberof Minimum Number of l in which Function OPERABLE or Instrument Channels -

Must Be OPERABLE OPERATING [ tripped] _ Per OPERABLE '

Trio Setting Refuel . Startup Run Trio Systems ' Trio System Action Reauired* '

Shutdown

' Function .

K. Rod Block (Cont'd.)

4. APRM Upscale " ' X(s) X X 2 3(c)

5. APRM Downscale 22/150 fullscale X 2 3 (c)

- 6. --IRM Upscale s 108/125 fullscale X X 2 3' s 14 gallons X(z) . X(z) X(z) I 1per

7. a) Water Level High Instrument Scram Discharge '

Volume Volume North X(z) X(z) X(z) 1 1 per b) WaterLevelHigh .s 14 gallons Instrument -

Scram Discharge Volume Volume South L. Condenser Vacuum Pump Isolation 1

Deleted M. Diesel Generator Time Delay after

. Load Sequence energization of relay Timers ,

X' X X X 2(m) 1(n)(kk) Consider the. l

1. CRD Pump 60 seei 15%

pump inoperable and comply with Spec 3.4.D (See j note q) t OYSTER CREEK 3.1-14 Amendment No.: 44,63,160,169

TABLE 3.1.1 PROTECTIVE INSTRUMENTATION REQUIREMENTS Sheet 7 of 13 Reactor Modes Minimum Number of Minimum Number of ,

in which Function OPERABLE or Instrument Channels ,

Must Be OPERABLE . OPERATING [ tripped] PerOPERABLE .

Startup Rp!! Trio Systems Trio System Action Reauired* -

Function Trio Settina Shutdown Refuel '

M. Diesel Generator Load Sequence Timers (Cont'd.) ,

X 2(pXkk) . . Consider the

2. Service Water Pump 120 seci 15% X X X 2(o) pump inoperable (SKIA) and comply - ,

(SK2A) within 7 days 10 sec. i 15% (S** ""* @

(SK7A)

(SK8A)

X 1(n)(kk) Consider the !

3. Reactor Building - 166 sec i 15% X X X 2(m) pump inoperable ,

Closed Cooling Water and comply i Pump (bb) within 7 days .- !

(See note q) .

N. Loss of Power t

a. 4.16 KV Emergency **

X(ff) X(ff) : X(fi) X(f1) 2 1(kk) ,

Bus Undervoltage (Loss of Voltage) ,

2 3(kk)~ See note ce

b. 4.16 KV Emergency " X(ff) X(ff) X(ff) X(ff)

Bus Undervoltage ,

(Degraded Voltage)

O. Containment Vent and Purne Isolation X 1 Isolate vent &

1. Drywellliigh s 74.6 R/hr X(u) X(u) X(u) 1 purge pathways or ;

Radiation - PLACE IN COLD SHUTDOWN :

CONDITION .

OYSTER CREEK 3.1-15.

Amendment No.: 15,44,60,80,160,171,195

e. . +- e.- , -

TABLE 3.1.1 (CONTD)

Sheet 8 of 13

Action required when minimum conditions for operation are not satisfied. Also permissible to trip inoperable trip system. A channel may b inoperable status for up to six hours for required surveillance without placing the trip system in the tripped condition provided at least o; instrument channel in the same trip system is monitoring that parameter. ];

" See Specification 2.3 for Limiting Safety System Settings.

l i

Notes:

t

a. Permissible to bypass, with control rod block, for reactor protection system reset in REFUEL MODE. !

b. Permissible to bypass below 600 psig in REFUEL and STARTUP MODES.

c. One (1) APRM in each OPERABLE trip system may be bypassed or inoperable provided the requirements o Two APRM's in the same quadrant shall not be concurrently bypassed except as noted below or permitted by note. j i

j Any one APRM may be removed from service for up to six hours for test or calibration without inserting trips in its trip system only if the rt 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 th short period, the requirements of Specifications 3.1.B.2,3.1.C and 3.10.C need not be met.

d. The IRMs shall be inserted and OPERABLE until the APRMs are OPERABLE and reading at least 2/150 full scale. ,

e. Offgas system isolation trip set at _<2,000 mrem /hr. Air ejector isolation valve closure time delay shall not exceed 15 minutes.

i

f. Unless SRM chambers are fully inserted.

g. Not applicable when IRM on lowest range.

h. With one or more instrument channel (s) inoperable in one ADS trip system, place the relay contact (s) for the inoperable initiation signal in the tripped condition within 4 days, or declare ADS inoperable and take the action required by Specification 3.4.B.3.

I With one or more instrument channel (s) inoperable in both ADS trip systems, restore ADS initiation capability in at least one trip system within I hour, declare ADS inoperable and take the action required by Specification 3.4.B.3. ,

OYSTER CREEK 3.1-16 5

Amendment No.: 44,75,108,110,171,190,191,195

TABLE 3.1.1 (CONTD) .

Sheet 9 of 13 Relief valve controllers shall not be bypassed for any more than 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months (total time for all controllers)'in any 30-day period and only one relief valve controller may be bypassed at a time.

i. With two core sorav systems OPERABLE:

1. A maximum of two core spray booster pump differential pressure (d/p) switches may be inoperable provided that the switches are fu opposing .

ADS trip systems [i.e., only: either RV-40 A&D or RV-40 BAC). Place the relay contacts associated with the inoperable d/p switch (es) in the energized position, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the inoperable d/p switch (es) within 8 days, or declare ADS iroperable_and take the action req by Specification 3.4.B3; or,

2. If two inoperable d/p switches are in the same ADS trip system [i.e., RV-40 A&B or RV-40 CAD), place the relay contacts associated with the inoperable d/p switch (es) in the de-energized position, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Restore the inoperable d/p switches within 4 days, or declare' ADS inoperable and take the action required by Specification 3.4.B3.

With only one core sprav system OPERABLE:

If one or more d/p switches become inoperable in the OPERABLE core spray system, declare ADS inoperable and take the action required by Specification 3.4.B.3.

J. Not required below 40% of rated reactor thermal power. .

e

k. All four (4) drywell pressure instrument channels may be made inoperable during the integrated primary containment leakage rate t' st (See Spe 4.5), provided that the plant is in the COLD SHUTDOWN condition and that no work i.s performed on the reactor or its connected systems w result in lowering the reactor water level to less than 4'8" above the TOP OF THE ACTIVE FUEL. '

l. Bypass in IRM Ranges 8,9, and 10.

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

n. One time delay relay per pump must be OPERABLE.

OYSTER CREE'C 3.1-17 Amendment Nr. 171,184,190

TABLE 3.1.1 (CONTDF Sheet 10 of 13 .

o. There are two time delay relays associated with each of two pumps. One timer per pump is for sequence starting (SKI A, SK2A) and one timer per pump 7 is for tripping the pump circuit breaker (SK7A, SK8A).

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

J

q. Manual initiation of affected component can be accomplished after the automatic load sequencing is completed.

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

. s. Dese functions not required to be OPERABLE with the reactor temperature less than 212*F and the vessel head removed or vented or during REACT VESSEL PRESSURETESTING.

~

t. These functions may be inoperable or bypassed when corresponding portions in the same core spray system logic train are inoperab!e per Specificati 3.4.A.

u. These fu.netions not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required to be maintained. :

v. These functions not required to be OPERABLE when the ADS is not required to be OPERABLE. ,

w. These functions must be OPERABLE only when irradiated fuel is in the fuel pool or reactor vessel and SECONDARY CONTAINMENT 1NTEGRITY is ,

required per Specification 3.5.B.

y. Deleted

[

z. He bypass function to permit scram reset in the SHUTDOWN or REFUEL MODE with control rod block must be OPERABLE in this mode.

Pump circuit breakers will be tripped in 10 seconds i 15% during a LOCA by relays SK7A and SK8A.

aa. '

i bb. Pump circuit breakers will trip instantaneously during a LOCA.

cc. Only applicable during STARTUP MODE while OPERATING in IRM range 10.

I 3.1-18 ;

OYSTER CREEK .

Amendment No.: 15,44,60,63,71,72,108,120,171 i

t

e TABLE 3.i.1 (CONTD)

Sheet 11 of13 dd. ' If an isolation condenser inlet (steam side) isolation valve becomes or is made inoperable in the open position during the RUN MODE comply with Specification 3.8.E. If an AC motor-operated outlet (condensate return) isolation valve becomes or is made inoperable in the open position during the RUN MODE comply with Specification 3.8.F.

ee. With the number of OPERABLE channels one less than the Minimum Number of OPERABLE Instrument Channels per OPERABLE Trip System, operation may proceed until performance of the next required Channel Functional Test provided the inoperable channel is placed in the tripped condition within I hour.

fr. This function is not required to be OPERABLE when the associated sefety bus is not required to be energized or fully OPERABLE as per applicable sections of these Technical Specifications.

gg. Deleted hh. The high flow trip function for "B" Isolation Condenser is bypassed upon initiation of the alternate shutdown panel. This prevents a spurious trip of the

. Isolation Condenser in the event of fire induced circuit damage.

ii. Instrument shall be OPERABLE during main condenser air ejector operation except that a channel may be taken out-of-service for the purpose of a check, calibration, test, or maintenance without declaring it inoperable.

jj. With no channel OPERABLE, main condenser offgas may be released to the environment for as long as 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months provided the stack radioactive noble gas monitor is OPERABLE. Otherwise, be in at least SHUTDOWN CONDITION within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

kk. One channel may be placed in an inoperable status for up to two hours for required surveillance without placing the trip system in the tripped condition.

II. This function not required to be OPERABLE with the reactor vessel head removed or unbolted. f i

mm. " Instrument Channel" in this case refers to the bellows which sense vacuum in each of the three condensers (A, B, and C), and " Trip System" refers to vacuum trip systems I and 2.

l OYSTER CREEK 3.1-19 Amendment No.: 80,91,108,112,130,162,171

TABLE 1.1.1 (CONT'D)

Sheut 12 cf13 nn. With one required channel inoperable in one Trip System, within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , restore the inoperable channel or place the inoperable channel and/or that Trip .

System in the tripped ^ condition.

With two or more required channels inoperable:

1. Within one hour, verify sufficient channels remain OPERABLE or tripped' to maintain trip capability, and

2. Within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , place the inoperable channel (s) in one Trip System and/or that Trip System'^ in the tripped condition ^, and

3. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , restore the inoperable channels in the other Trip System to an. OPERABLE status or tripped'.

Otherwise, take the Action Required.

An inoperable channel or Trip System need not be placed in the tripped condition where this would cause the Trip Function to occur. In these :

cases, if the inoperable channel is not restored to OPERABLE status within the required time, the Action Required shall be taken.

^^ This action applies to that Trip System with the most inoperable channels; if both Trip Systems 1 we the same number ofinoperable channels, the action can be applied to either Trip System.

oo. With one required channel inoperable in o_ne Trip System, either

1. Place the inoperable channel in the tripped condition within

a. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for parameters common to Scram Instrumentation, and

b. 24 he for parameters not common to Scram Instrumentation.

or I

2. Take the Action Required.

OYSTER CREEK 3.1-20 .

Amendment No.: 171

.. .. .. .= .. ..

TABLE 3.1.1 (CONTD)

Sheet 13 of13 With one required channel inoperable in both Trip Systems,

1. Place the inoperable channel in one Trip System in the tripped condition within one hour, and

2. a. . Place the inoperable channel in the remaining Trip System in 'he tripped condition within (1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for parameters common to Scram Instrumentation, and

,(2) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for parameters not common to Scram Instrumentation.

or

b. Take the Action Required.

pp. With one or more required channels inoperable per Trip System: ,

1.

For one channel inoperable, within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months place the inoperable channel in the tripped condition or take the Action Required.

2. With more than one channel inoperable, take the Action Required.

3.1-21 OYSTER CREEK Amendment No.: 171

__. - _ _ _ = _ _ _ _ _ _ _ _ _ _ _ _ _ - - - - - _ _ _ _ - _ - - - _ - _ - . _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ _ _ _ - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ _ _ _ _ = .

SECTION 4 SUREVEILLANCE REOUIREMENTS 4.1 - PROTECTIVE INSTRUMENTATION Aoolicability: Applies to the surveillance of the instrumentation that performs a safety function. i Objective: To specify the minimum frequency and type of surveillance to be applied to the safety instnamentation.

Specification: Instrumentation shall be checked, tested, and calibrated as indicated in Tables 4.1.1 and 4.1.2 using the definitions given in Section 1.

1 l

l l

l 1

i l

OYSTER CREEK 4.1-1 Amendment No.: 171 l

r.

I 4.1 PROTECTIVE INSTRUMENTATION l Bases Surveillance intervals are based on reliability analyses and have been determined in accordance with General Electric Licensing Topical Reports given in References 1 through 5. :

The functions list'ed in Table 4.1.1 logically divide into three groups: '

e

a. On-off sensors that provide a scram function or some other equally important function.

f

- b. Analog devices coupled with a bi-stable trip that provides a scram function or !

some other vitally important function. !

c. Devices which only serve a useful function during some restricted mode of operation, such as startup or shutdown, or for which the only practical test is one i that can be performed only at shutdo'wn.

]

Group (b) devices utilize an analog sensor followed by an amplifier and bi-stable trip circuit. The ,

sensor and amplifier are active components and a failure would generally result in an upscale ,

' signal, a downscale signal, or no signal. These conditions are alarmed so a failure would not go

- undetected. The bi-stable portion does need to be tested in order to prove that it will assume its j tripped state when required. I i

Group (c) devices are active only during a given portion of the operational cycle. For example, the IRM is inactive during fhll-power operation and active during startup. Thus, the only test that is significant is the one performed just prior to shutdown and startup. The condenser Low Vacuum trip can only be tested during shutdown, and although it is connected into the reactor protection system, it is not required to protect the reactor. Testing at each REFUELING OUTAGE is adequate. The switches for the high temperature main steamline tunnel are not accessible during normal operation because of their location above the main steam lines.

Therefore, after initial calibration and in-place OPERABILITY checks, they will not be tested between refueling shutdowns. Considering the physical arrangement of the piping which would ,

allow a steam leak at any of the four sensing locations to affect the other locations, it is considered that the function is not jeopardized by limiting calibration and testing to refueling outages.

The logic of the instrument safety systems in Table 4.1.1 is such that testing the instrument channels also trips the trip system, verifying that it is OPERABLE. However, certain systems require coincident instrument channel trips to completely test their trip systems. Therefore, Table 4.1.2 specifies the minimum trip system test frequency for these tripped systems. This assures that all trip systems for protective instrumentation are adequately tested, from sensors through the j trip system. ;

OYSTER CREEK 4.1-2 Amendment No.: 171 a .. , - ,. - - - . .- ,

IRM calibration is to be performed during reactor startup. The calibration of the IRMs during startup will be significant since the IRMs will be relied on for neutron monitoring and reactor protection up to 38.4% of rated power during a reactor startup.

To ensure that the APRMs are accurately indicating the true core average power, the APRMs are j calibrated to the reactor power calculated from a heat balance. Limiting Safety System Settings '

(LSSS) 2.3.A.1 allows the APRMs to be reading greater than actual thermal power to compensate for localized power peaking. When this adjustment is made, the requirement for the !

I absolute difference between the APRM channels and the calculated power to indicate within 2% l RTP is modified to include any gain adjustments required by LSSS 2.3.A.I. ,

4 LPRM gain settings are determined from the local flux profiles measured by the Traversing Incore Probe (TIP) System. This establishes the relative local flux profile for appropriate representative j input to the APRM System. The 1000 MWDff Frequency is based on operating experience with i

LPRM sensitivity changes.

1 General Electric Licensing Topical Report NEDC-30851P-A (Reference 1), Section 5.7 indicates !

that the major contributor to reactor protection system unavailability is common cause failure of i the automatic scram contactors. Analysis showed a weekly test interval to be optimum for scram I contactors. The test of the automatic scram contactors can be performed as part of the Channel l Calibration or Test of Scram Functions or by use of the subchannel test switches.

References:

(1) NEDC-3085' P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection Syste.m."

(2) NEDC-30936P-A, "BWR Owners' Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)," Parts 1 and 1 (3) NEDC-30851P-A, Supplement 1, " Technical Specification Improvement Analysis for BWR Control Rod Block '

Instrumentation."

(4) NEDC-30851P-A, Supplement 2, " Technical Specification i Improvement Analysis for BWR Isolation Instrumentation Common to RPS and ECCS Instrumentation."

(5) NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR isolation Actuation Instrumentation."

i OYSTER CREEK 4.1-3 AMENDMENT NO.: 71,171

- wr - , , 4 m .- .. ,, .~

TABLE 4.1.1 Page1of6 MINIMUM CHECK. CAllBRATION AND TEST FREOUENCY FOR PROTECTIVE INSTRUMENTATION Instrument Channel Check Calibrate Test Remarks (Apolies to Test & Calibration)

1. High Reactor Pressure 1/d Note 3 1/3 mo.

2. High Drywell Pressure (Scram) N/A - 1/3 mo. 1/3 mo. By application of test pressure ,

3. Low Reactor Water Level 1/d Note 3 1/3 mo.

4. Low-Low Water Level 1/d Note 3 1/3 mo.

5. High Water Level in Scram Discharge Volume

a. Digital N/A 1/3 mo. 1/3 mo. By varying level in sensor columns N/A

b. Analog Note 3 1/3 mo.

6. Low-Low-Low Water Level N/A I/3 mo. 1/3 mo. By application of test pressure

7. High Flow in Main Steamline 1/d 1/3 mo. 1/3 mo. By application of test pressure

8. Low Pressure in Main Steamline N/A l/3 mo. 1/3 mo. By application of test pressure

9. High Drywell Pressure (Core Cooling) 1/d 1/3 mo. 1/3 mo. By application of test pressure i 10. Main Steam Isolation Valve (Scram) N/A N/A l/3 mo. By exercising valva OYSTER CREEK 4.1-4 Change: 7, Amendment No.: 152,171

_.m..._ ._.__ . __m._ _ . ___ _ _ _ _ . . _ _ _ . _ _ _ _ _ . _ _ . _ _ _ _ _ _ _ _ _ _

+-e

.. . . _ _ _ _ . . . _ _ _ .. . _ . . . . . _ _ _ . _ _ _ _ . - , - - - - - _ _ . - - _ _ - . . - - _ - _ _ - - - - - . - - - - - . - - _ ~ . -

1

TABLE 4.1.1 J

Page 2 of 6

, MINIMUM CHECK. CALIBRATION AND TEST FREOUENCY FOR PROTECTIVE INSTRUMENTATION

Instrument Channel Check Calibrate Test Remarks (Applies to Test & Calibration)

~

l1. APRM Leve! N/A 1/3d N/A Verify the absolute difference between the APRM channels and the calculated power is s 2% rated thermal power [plus any gains required by LSSS 2.3.A.1].

l APRM Scram Trips Note 2 1/3 mo. 1/3 mo. Using built-in calibration equipment 4

- Flow biased neutron flux - high during POWER OPERATION i e Fixed neutron flux-high or inop I . Downscale

. 12. APRM Rod Blocks Note 2 1/3 mo. 1/3 mo. Upscale and downscale

13. DELETED

14. High Radiation in Reactor Building Operating Floor 1/s 1/3 mo. 1/3 mo. Using gamma source for calibration Ventilation Exhaust 1/s 1/3 mo. 1/3 mo.

j. 15. High Radiation on Air Ejector Off-Gas 1/3 mo. 1/3 mo. Using built-in calibration equipment

! 1/s Channel Check -

j . I/mo. Source check -

4 1/24 mo. . Calibration according to established i station calibration procedures IG4 mo. NMea i ' OYSTER CREEK . 4.1-5 Change: 7, Amendment No.: 108,141,171

% r

-- -- - . - - - . - _ . . . - - - , . . . - - -.-_.---2-m v- e % -=ssL._- - _ . _ . _ - _.

TABLE 4.1.1 Page 3 of 6 -

MINIMUM CliECK. CALIBRATION AND TEST FREOUENCY FOR PROTECTIVE INSTRUMENTATION -

Instrument Channel Check Calibrate Test Remarks (Applies to Test & Calibration) .

16. IRM Level N/A Each startup N/A ,

IRM Scram * *

Using built-in calibration equipment '

17. IRM Blocks N/A Prior to startup Prior to startup Upscale and downscale and shutdown and shutdown

18. Condenser Low Vacuum N/A 1/24 mo. 1/24 mo.

19. Manual Scram Buttons N/A N/A 1/3 mo. '

20. High Temperature Main Steamline Tunnel N/A l/24 mo. Each refueling Using heat source box ;

outage

21. SRM * *

Using built-in calibration equipment

22. Isolation Condenser fligh Flow AP (Steam & Water) N/A 1/3 mo. 1/3 mo. By application of test pressure

23. Turbine Trip Scram N/A N/A l/3 mo. ,

24. Generator Load Rejection Scram N/A I/3 mo. 1/3 mo.

25. Recirculation Loop Flow N/A 1/24 mo. N/A By application of test pressure i

i

26. Low Reactor Pressure Core Spray Valve Permissive N/A 1/3 mo. 1/3 mo. By application of test pressure OYSTER CREEK 4.1-6 Amendment No.: 71,144,193

TABLE 4.1.1 Page 4 of 6 MINIMUM CHECK. CALIBRATION AND TEST FREOUENCY FOR PROTECTIVE INSTRUMENTATION Instmment Channel Check Calibrate Test Remarks (Applies to Test & Calibrationi

27. Scram Discharge Volume (Rod Block) a) Water level high N/A Each refueling 1/3 mo. Calibrate by varying level in sensor outage column b) Scram Trip bypass N/A N/A Each refueling outage

28. Loss of Power a) 4.16 KV Emergency Bus Undervoltage 1/d 1/24 mo. 1/mo.

(Loss of Voltage) b) 4.16 KV Emergency Bus Undervoltage 1/d 1/24 mo. 1/mo.

(Degraded Voltage)

29. Drywell High Radiation N/A Each refueling Each refueling outage outage

30. Automatic Scram Contactors N/A N/A 1/wk Note 1

31. Core Spray Booster Pump Differential Pressure N/A 1/3 mo. 1/3 mo. By application of a test pressure OYSTE R CREEK 4.1-7 Amendment No.: 63,80,116,141,144,152,171,190 1 - _ _ _ _ _ - . _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

k TABLE 4.1.1 Page 5 of 6 MINIMUM CHECK. CALIBRATION AND TEST FREOUENCY FOR PROTECTIVE INSTRUMENTATION Instrument Channel Check Calibrate Test Reinarks (Applies to Test & Calibration)

32. LPRM Level a) Electronics N/A l/12 mo. 1/12 mo.

b) Detectors N/A Note 4 N/A-

Calibrate prior to startup and nonnal shutdown and thereafter check 1/s and test 1/wk until no longer required. ;

Levend N/A = Not Applicable 1/s = Once per shift 1/d = Once per day :

1/3d = Once per 3 days; ,

1/wk = Once per week

l/mo. = Once per month 1/3 mo. = Once every 3 months; l/20 mo. = Once every 20 months ;

l/24 mo. = Once every 24 months ;

OYSTER CREEK 4.1-8 Change: 5,7, Amendment No.: 171 L

TABLE 4.1.1 i Page 6 of 6 MINIMUM CHECK. CALIBRATION AND TEST FREOUENCY FOR PROTECTIVE INSTRUMENTATION

. NOTE 1: Each automatic scram contactor is required to be tested at least once per week. When not tested by other means, the weekly test can '!

be performed by using the subchannel test switches. ;

NOTE 2: At least daily during reactor POWER OPERATION, the reactor neutron flux peaking factor shall be estimated and flow-referenced APRM scram and rod block settings shall be adjusted, if necessary, as specified'in Section 2.3 Specifications A.1 and A.2. j NOTE 3: Calibrate electronic bistable trips by injection of an external test current once per 3 months. Calibrate transmitters by application of test - ,

pressure once per 12 months. l NOTE 4: Perform LPRM detectors calibration every 1000 MWD /MT Average Core Exposure The following notes are only for item 15 of Table 4.1.1:

A channel may be taken out of service for the purpose of a check, calibration, test or maintenance without declaring the channel to be inoperable.

[

a. The thannel Test shall also demonstrate that control room alarm annunciation occurs if any of the fillowing conditions exists: ,

, 1) Instrument indicate . measured levels above the alarm setpoint. :

2) Instrument indicabs a downscale failure.

3) Instnament controls not set in operate mode. ,

4) Instrument electrical power loss.

l i

t Y

L OYSTER CREEK 4.1-9 ;

Change: 5,7, Amendment No.: 71,80,95,108,171 i

---,--w- w -- - - , m- %

TABLE 4.1.2 MINIMUM TEST FREOUENCIES FOR TRIP SYSTEMS Trip System Minimum Test Frequency

1) Dual Channel (Scram) Same as for respective instmmentation in Table 4.1.1

2) Rod Block Same as for respective I instrumentation in Table 4.1,1 J

3) - DELETED DELETED I

4) ' Automatic Depressurization Each refueling outage l each trip system, one at a time

5) MSIV Closure Each refueling outage j each closure logic circuit independently '

(1 valve at a time)

6) _ Core Sorav 1/3 mo and each refueling outage I each trip system, one at a time '

l

7) Primary Containment Isolation, Each refueling outage i each closure circuit independently l (1 valve at a time)

8) Refueling Interlocks Prior to each refueling operation

9) holation Condenser Actuation and Isolation Each refueling outage each trip circuit independently (1 valve at a time)

10) Reactor Building Isolation and SGTS Initiation Same as for respective instmmentation in Table 4.1.1

11) Condenser Vacuum Pump Isolation Prior to each startup

12) Air Elector Offgas Line Isolation Each refueling outage

13) Containment Vent and Purge Isolation 1/24 mo.

OYSTER CREEK 4.1-10 Amendment No.: 108,116,144,160,171,193 l

-

ML20155J750

Text

References:

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