ML20195C805
| ML20195C805 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 06/02/1999 |
| From: | Clifford J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20195C812 | List: |
| References | |
| NUDOCS 9906080315 | |
| Download: ML20195C805 (34) | |
Text
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UNITED STATES s
j NUCLEAR REGULATORY COMMISSION 2
WASHINGTON, D.C. 3066H001
'+,.....,d GPU NUCLEAR. INC.
ANR JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 OYSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 208 License No. DPR-16 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by GPU Nuclear, Inc. et al., (the licensee), dated November 5,1998, as supplemented by a letter dated February 18,1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10
)
CFR ChapterI; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
9906080315 990602 PDR ADOCK 05000219 P
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2 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicsted in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-16 is hereby amended to read as follows:
(2) Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No208, are hereby incorporated in the license. GPU Nuclear, Inc. shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of the date of issuance, to be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION M6/d J mes W. Clifford, Chief, Section 2 Project Directorate 1 Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of issuance: June 2,1999 l
i
Fi l
ATTACHMENT TO LICENSE AMENDMENT NO. 208 FACILITY OPERATING LICENSE NO. DPR-16 l
l DOCKET NO. 50-219 l
i-Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages as indicated. The revised pages are identified by amendment number and contain marginallines indicating the areas of change.
BPRIDXR IDARd 2.3-1 2.3-1 l-2.3-2 2.3-2 2.3-3 2.3-3 2.3-4 2.3-4
. 3-5 2.3-5
^
t 2.3-6 2.3-6 2.3-7 2.3-7 2.3-8 3.1-9 3.1 -9 3.1-10 3.1 3.1-11 3.1-11
'3.1-12 3.1-12 3.1-13 3.1-13 3.1-14 3.1-14 3.1-15 3.1-15 3.1-16 3.1-16 3.1-16a 3.1-17 3.1-18 3.1-18 3.1 3.1-19 3.1-20 3.1-20 3.1 21 4.1-1 4.1-1 4.1-2 4.12 4,1 -3 4.1-3 4.1-4 4.1 -4 4.15 4.1-5 4.1-6 4.1-6 4.1-7 4.17 4.1 -8 4.1-9 4.1-10
.f L
s 1 l 2.3 '
LIMITING SAFETY SYSTEM SETTINGS Anolicability:
Applies to trip settings on automatic protective devices related to variables on which safety limits have been placed.
Obiective:
To provide automatic corrective action to prevent the safety limits from being exceeded.
.c Specification-Limiting safety system settings shall be as follows:
FUNCTION
. LIMITING SAFETY SYSTEM SETTINGS A.
- Neutron Flux, Scram A.I APRM When the reactor mode switch is in the Run position, the APRM lux scram setting shall be E.E S s[(0.90 x 10-6)W + 60.8) MFLPD with a maximum setpoint of 115.7% for core flow equal 6
to 61 x 10 lb/hr and greater, where:
S='
setting in percent of rated power W=
recirculation flow (Ib/hr)
FRP =
fraction of rated thermal power is the ratio of core thermalpower to rated thermalpower MFLPD =
maximurn fraction oflimiting power density where the limiting power density for each bundle is the design linear heat generation rate for that bundle.
De ratio of FRP/MFLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0 in which case the actual operating value will be used.
His adjustment may be accomplished by increasing the APRM gain and thus reducing the flow reference APRM High Flux Scram Curve by the reciprocal of the APRM gain change.
A.2.
IkM
$ 38.4 percent of rated neutron flux A.3 APRM Downscale 2 2% Rated Thermal Power coincident with IRM Upscale (high-high)orInoperative OYSTER CREEK-2.3-1
. AmendmentNo.: 74,36,A M, 208
R n-FUNCTION LIMITING SAFETY SYSTEM SETTINGS B.'-
Neutron Flux, Control Rod. The Rod Block setting shall be Block EM 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 the APRM scram trip apply.
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 actual operating value willbe used.
This adjustment may be accomplished by increasing the APRM gain and thus reducing the flow referenced APRM rod block curve by the reciprocal of the APRM gain change.
C.
Reactor High Pressure, 51060 psig Scram D.
Reactor High Pressure, 2 @ s 1085 psig.
Relief Valves Initiation 3 -@ s 1105 psig E.
Reactor High Pressure,
$1060 psig with time delay Isolation Condenser
$3 seconds Irdtiation F.
Reactor High Pressure, 4 @ 1212 psig 12 psi
. 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 OYSTER CREEK 2.3-2 Amendment No. N,76,H+,MO,46+,M, 208
(
FUNCTION LIMITING SAFETY SYSTEM SETTINGS
'~
1.
Reactor Low Water Level, 21l'5" above the top of the medve 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 Isolation Valve Closure K.
Reactor Low-Low Water Level, 27'2" above the top of the active fuel Core Spray Initiation L.
Reactor Low-Low Water Level, 27'2" above the top of the active Fuel with time Isolation Condenser Initiation delay 53 seconds M.
Turbine Trip, Scram 10 percent turbine stop valve (s) closure from full open 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 secondsi Undervoltage (Loss of 0.5 seconds time delay Voltage) 2)
4.16 KV Emergency Bus 3840 (+20V,-40V) volts Undervoltage (Degraded 10110% (1.0) second time delay Voltage)
OYSTER CREEK 2.3-3 Amendment No. 76,40, M4, +75, 208
m 2.3 LIMITING SAFETY SYSTEM SETilNGS Bassa:
p
(
Safety limits have been established in Specifications 2.1 and 2.2 to protect the integrity of the l
t fuel cladding and reactor coolant system baniers, respectively. Automatic protective devices have been provided in the plant design for corrective actions to prevent the safety limits from l
being exceeded in normal operation or operational transients caused by reasonably expected l
single operator enor or equipment malfunction. This Specification establishes the trip settings for these automatic protection devices.-
The Average Power Range Monitor, APRMN, 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 transfened to the water follow the neutron flux, the APRM will read reactor thermal power. For fast transients, the neutron flux will lead the power transfened frorp the cladding to the water due to the effect of the fuel time constant. Therefore, when the neuron nux increases to the scram setting, the percent increase in heat flux and power transferred to the water will be less than the percentincreasein neutron flux.
The APRM trip setting will be varied automatically with recirculation flow, with the trip setting 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 eW maneuvers and the various mechanical failures, it was concluded that sufHeient 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 recirculatiori 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 limit and yet allows operating margin that reduces the possibility of unnecessary scrams.
The scram trip setting must be adjusted to ensure that the LHGR transient peak is not increased for any combination of maximum fraction oflimiting power density (MFLPD) and reactor core thermal power. The scram setting is adjusted in accordance with the formula in Specification 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.
OYSTER CREEK :
2.3-4 Amendment No.W,M, 208 -
l l
F i For operation in the Startup mode while the reactor is at 1:w pressure, the IRM range 9 High Flux scram setting of 12% of the rated power provides adequate thermal margin between the maximum power and the safety limit of 18.3% of rated power to accommodate anticipated maneuvers associated with power plant startup. Here 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. Worth ofindividual rods is very low in a constrained rod pattern. In a sequenced rod withdrawal approach to the scram level, the rate of power rise is no more than five percent'of the rated per minute, and the IRM system would be more than adequate to assure a scram before the power could exceed the safety limit.
To continue operation beyond 12% of rated power, the IRMs must be transferred into range 10.
He 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. De IRM scram remains active until the mode switch is placed in the RUN position at 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 10 to the scram level on the APRMs at 30% of rated flow, ne IRM scram is at 38.4% of rated power while the APRM scram is at 52.7% of rated power. The minimum flow for Oyster Creek 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 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. De transients
- requiring a scram by nuclear instrumentation are the loss of feedwater heating and the improper startup of an idle recirculation loop. The loss of feedwater heating transient is not affected by the range 10 IRM since the feedwater heaters will not be put into service until after the LPRM downscales have cleared, thus insuring the operability of the APRM system. This will be administratively controlled. The improper startup of an idle recirculation loop becomes less severe at lower power level and the IRM scram would be adequate to terminate the flux excursion.
The Rod Worth Minimizer is not required beyond 10% of rated power. The ability of the IRMs to' terminate a rod withdrawal transient is limited due to the number and location ofIRM 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 less would not result in v!clating the fuel cladding safety limit. Herefore, a rod block on the IRMs at less than 35% of rated power would be adequate protection against a rod withdrawal transient..
l OYSTER CREEK 2.3-5 Amendment No.: 'M, 208
Reactor power level may be varied by moving control rods or by varying the recirculati:n f1:w 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 dwiw.cs for the specified trip setting versus flow relationship. Therefore, the worst-case MCPR, which could o'ccur during steady-state operation, is at 108% of the rated thennal power because of the APRM rod block trip setting. The actual
- power distribution in the 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 j
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 aie -
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 it,olation condensers and relief valves to operate. Under these conditions, a total of 9 safety valves are reqaired to tum the pressure transient. The ASME B&PV Code allows a il% of working i
pressure (1250 psig) variation in the lift point of the valves. This variation is reccgnized in Specification 4.3.
a i
1 OYSTER CREEK 2.3-6 Amendment No.: M,-75, !!!,150, 208
[
The 1:w pressure isolati:n cf the main steam line et 825 psig was provided to give protection against fast reactor depressurization and the resulting rapid cool-down of the vessel. He 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 inadvertent isolation valve closure. Bypass of 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 line 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. Dere 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 worth 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 ci~.;or to pull up on the switch handle to pass through the stop and enter range 10. This provides protection against an inadvertent entry into l
range 10 at low pressures. De IRM scram remains active until the mode switch is placed in the RUN position at which time the trip becomes a coincident IRM upscale, APRM downscale scram.
De low level water level trip setting of Il'5" above the top of the active fuel has been established to assure that the reactor is not operated at a water level b'elow 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 value 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 be maintained to provide core cooling. Bus, the low-low level trip point of 7'2" above the core is 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.*
ne turbine stop valve (s) scram 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.-495,-303, 208
- Correction 11/30/87
The 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. The 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 i
emergency buses C and D to on-site power should normal power be lost. There is a separate 2 out i
of 3 coincident logic relay designed to shift emergency buses to on-site power should normal power be degraded to an >===;e-ble level. The trip points and time deh.y 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 aff'ect 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.
1 References (1).
FDSAR, Volume 1, Section VII-4.2.4.2 i
(2)-
FDSAR, Amandment 28, Item III.A-12
-(3)
FDSAR, Amandment 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, Amandment 65, Section B.XI J
'(6)-
FDSAR, Amendment 65, Section B.IX 4
OYSTER CREEK 2-3.8'
- AmendmentNo.10,173,its, 208
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SECTION 4 SUREVEILLANCE REOUIREhENTS 4.1 PROTECTIVE INSTRUMENTATION i
- Apolicability: Applies to the surveillance of the instrumentation that performs a safety function.
)
i Qbiective To specify the minimum frequency and type ofsurveillance to be applied to the safety instrumentation.
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 OYSTER CREEK 4.1-1 Amendment No.:+7+, 208 i
I
n 4.1 PROTECTIVE INSTRUMENTATION Bases-l Surveillance intervals are based on reliability analyses and have been determined in accordance with General Electric Licensing Topical Reports given in References I through 5.
'The functions listed in Table 4.1.1 logically divide into three groups:
a.
On-off sensors that provide a scram function or some other equally important l
function.
I b.
Analog devices coupled with a bi-stable trip that provides a scram function or some other vitally important function.
1 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 l
that can be performed only at shutdown.
1 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 L
undetected. The bi-stable portion does need to be tested in order to prove that it will assume its j
tripped state when required.
' Group (c) devices are active only during a given portion of the operational cycle. For example, the IRM is inactive during full-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 j-Vacuum trip can only be tested during shutdown, and although it is connected into the reactor l
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 j
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 tr;p systems for protective instrumentation are adequately tested, from sensors through the trip system.
OYSTER CREEK 4.1-2 Amendment No.:4M, 208
[
!u
IRM calibration is to be performed during reactor startup. The calibration :f the IRMs during startup will be significant since the IRMs will be relied on for neutron montoring 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 powu, trie APRMs are 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 ebmpensate for localized power peaking. When this adjustment is made, the requirement for the absolute difference between the APRM channels and the calculated power to indicate within 2%
RTP is modified to include any gain adjustments required by LSSS 2.3.A.1.
{
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.
input to the APRM System. The 1000 MWD /T Frequency is based on operating experience with LPRM sensitivity changes.
l General E.ectric Licensing Topical Report NEDC-30851P-A (Reference 1), Section 5.7 indicates 7
that the major contributor to reactor protection system unavailability is common cause failure of the automatic scram contactors. Analysis showed a weekly test interval to be optimum for scram contactors. The test of the automatic scram contactors can be performed as part of the Channel
' Calibration or Test of Scram Functions or by use of the subchannel test switches.
References-(1)
. NEDC-30851P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System."
j (2)
NEDC-30936P-A, "BWR Owners' Group Technical Specification Improvement Methodology (With Demonstration for BWR ECCS Actuation Instrumentation)," Parts 1 and 2.
(3)
NEDC-30851P-A, Supplement 1, " Technical Specification Improvement Analysis for BWR Control Rod Block Instrumentation."
(4)
NEDC-30851P-A, Supplement 2, " Technical Specification Improvement Analysis for BWR isolation Instrumentation i
Common to RPS and ECCS Instrumentation."
j (5)
NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR Isolation Actuation Instrumentation."
1 OYSTER CREEK 4.1-3 AhENDMENT NO.: 7+;W+, 208
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TABLE 4.1.2 MINIMUM TEST FREOUENCIES FOR TRIP SYSTEMS Trip System Minimum Test Freauency I)
Dual Channal (Scram)
Same as for respective instrumentation in Table 4.1.1 2)
Rod Block Same as for respective instrumentation in Table 4.1.1 3)-
DELETED DELETED 4)
Automatic Depreeenrintion Each refueling outage each trip system, one at a time
~
5)
MSIV Closure Each refueling outage each closure logic circuit independently (1 valve at a time) 6)
Core Sorav' 1/3 mo and each refueling outage each trip system, one at a time 7)
Primary Containment Isolation, Each refueling outage each closure circuitindependently (1 valve at a time) 8)
Refueling Interlocks Prior to each refueling operation 9)
Isolation Condenser Actuation and Isolation Each refueling outage each trip circuit independently (1 valve at a time) 10)
Reactor Buildina Isolation and SGTS Initiation Same as for respective instmmentation in Table 4.1.1 11)
Condenser Vacuum Pumo Isolation Prior to each start'up 12)
Air Elector Offnas Line Isolation Each refueling outage 13)
Containment Vent and Purne IsolalinD 1/24 mo.
OYSTER CREEK 4.1-10 Amendment No.: 108,116,!'d,!60,l!,193, 208 i