ML20080D946
| ML20080D946 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 12/29/1994 |
| From: | Mckee P Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20080D949 | List: |
| References | |
| NUDOCS 9501090266 | |
| Download: ML20080D946 (12) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION
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- WASHINGTON, D.C. 2066db001
,s.,.....j GPU NUCLEAR CORPORATION mn 1
JERSEY CENTRAL POWER & LIGHT COMPANY DOCKET NO. 50-219 0YSTER CREEK NUCLEAR GENERATING STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.176 License No. DPR-16 1.
The Nuclear Regulatory Commission (the Commission) has found that:
)
A.
The application for amendment by GPU Nuclear Corporation, et al..
i (the licensee), dated October 9, 1991, as supplemented March 9, April 27, and December 15, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Consission's rules and regulations set forth in 10 CFR Chapter I; 4
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; l
C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health l
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 e
requirements have been satisfied.
i t
9501090266 941229 PDR ADOCK 05000219 P
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' i 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated 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*
L (2)
Technical Specifications
~
The Technical Specifications contained in Appendices A and B, as revised through Amendment No.176, are hereby incorporated in the license. GPU Nuclear Corporation shall operate the facility in l
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 d
[h hT4 y
Phillip F. McKee, Director Project Directorate I-4 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation l
Attachment:
Changes to.the Technical Specifications Date of Issuance: December 29, 1994 i
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f ATTACHMENT TO LICENSE AMENDMENT NO.176 FACILITY OPERATING LICENSE NO. DPR-16 DOCKET NO. 50-219 Replace the following pages of the Appendix A Technical. Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the areas of change.
Remove Insert 3.1-1 3.1-1 3.1-2 3.1-2 3.1-3 3.1-3 3.1-4 3.1-4 3.1-5 3.1-5 3.1-6 3.1-6 3.1-7 3.1-7 3.10-2 3.10-2 3.10-3 3.10-3 F
i i
3.1 PROTECTIVE INSTRUMENTATION Apolicability: Applies' to the operating status of plant instrumentation which performs a protective function.
Ob.iective:
To assure the OPERABILITY of protective instrumentation.
Soecifications: A. The following operating requirements for plant protective instrumentation are given in Table 3.1.1:
1.
The reactor mode in which a specified function must be OPERABLE including allowable bypass conditions.
2.
The minimum number of OPERABLE instrument channels per OPERABLE trip system.
3.
The trip seitings which initiate automatic protective action.
4.
The action required when the limiting conditions for operation are not satisfied.
B.
1.
Failure of four chambers assigned to any one APRM shall make the APRM inoperable.
2.
Failure of two chambers from one radial core location in any one APRM shall make that APRM inoperable.
3.
Except during the performance of Technical Specification required LPRM/APRM surveillance, reactor power shall be reduced below the 80% rod line ar the corresponding RPS trip system shall be placed in the tripped condition, whenever all three of the following conditions exist:
1.
Reactor power is greater than 35%
-and-2.
More than one LPRM detector is bypassed or failed in the A level or the 8 level assigned to a single APRM channel
-and-3.
The diagonally opposite quadrant contains a single APRM channel with more than one bypassed or failed LPRM detector on the same axial level as the bypassed or failed detectors specified in (2) above.
0YSTER CREEK 3.1-1 Amendment No.: Ul,176
e C.
1.
Any two (2) LPRM assemblies which are input to the APRM system and are separated in distance by less than three (3) times the control rod pitch may not contain a combination of more'than three (3) inoperable detectors (i.e., APRM channel failed or bypassed, or LPRM detectors failed or bypassed) out of the four (4) detectors located in either the A and B, or the C and D levels.
2.
A Travelling In-Core Probe (TIP) chamber may be used as an APRM input to meet the criteria of 3.1.B and 3.1.C.1, provided the TIP is positioned in close proximity to one of the failed LPRM's.
If the criteria of 3.1.B.2 or 3.1.C.1 cannot be met, POWER OPERATION may continue at up to rated power level provided a control rod withdrawal block is OPERATING or at power levels less than 61% of rated power until the TIP can be connected, positioned and satisfactorily tested, as long as Specification 3.1.B.1 and Table 3.1.1 are satisfied.
Bases:
The plant protection system automatically initiates protective functions to prevent exceeding established limits.
In addition, other protective instrumentation is provided to initiate action which mitigates the consequences of accidents or terminates operator control.
This specification provides the limiting conditions for operation necessary to preserve the effectiveness of these instrument systems.
Table 3.1.1 defines, for each function, the minimum number of OPERABLE instrument channels for an OPERABLE trip system for the various functions specified.
There are usually two trip systems required or available for each function.
The specified limiting conditions for operation apply for the indicated modes of operation.
When the specified limiting condition cannot be met, the specified Actions Required shall be undertaken promptly to modify plant operation to the condition indicated in a normal manner.
Conditions under which the specified plant instrumentation may be out-of-service are also defined in Table 3.1.1.
Except as noted in Table 3.1.1 an inoperable trip system will be placed in the tripped condition. A tripped trip system is considered OPERATING since by virtue of being tripped it is performing its required function. All sensors in the untripped trip system must be OPERABLE, except as follows:
1.
The high temperature sensor system in the main steam line tunnel has eight sensors in each protection logic channel.
This multiplicity of sensors serving a duplicate function permits this system to operate for twenty month nominal intervals without calibration. Thus, if one of the temperature sensors causes a trip in one of the two trip systems, there are severtl cross checks that would verify if this were a real one.
If not, this sensor could be removed from service.
However, a minimum of two of eight are required to be OPERABLE and only one of the two is required to accomplish a trip in a single trip syste:n.
0YSTER CREEK 3.1-2 Amendment No.: M/1,176
v
- 2..One APRM of the four in each.. trip system may be bypassed without tripping the trip system if_ core protection is maintain:d. Core' protection is maintained by the remaining three_APRM's in each trip system as discussed.in Section 7.5.1<8.7 of the Updated FSAR.
3.
One IRM channel in each of the two trip systems may be bypassed without compromising the effectiveness of the system.
There are few possible sources of rapid reactivity input to the system in the low power low flow condition. ' Effects of increasing pressure at zero w low void content are minor, cold water from sources available dehng i
startup is not much colder. than that already in the system, temperature coefficients are small, and control rod patterns are constrained to be uniform by operating procedures backed up by the rod worth minimizer. Worth of individual rods is very low in a uniform rod pattern.
Thus, of all possible sources of reactivity input, uniform control rod withdrawal is the most probable cause of significant power rise.
Because the flux distribution associated 1
with uniform rod withdrawals does not involve high local peaks, and because several rods must be moved to change power by a significant percentage of rated, the rate of power rise is very slow.
Generally the heat flux is in near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the scram level, the rate of power rise is no more than five percent of rated per minute, and three OPERABLE IRM instruments in each trip system would be more than adequate to assure a scram before the power could exceed the safety limit.
In many cases, if properly located, a single OPERABLE 1RM channel in each trip system would suffice.
4.
When required for surveillance testing, a channel is made inoperable.
In order to be able to test its trip function to the
_i final actuating device of its trip system, the trip system cannot' i
already be tripped by some other means such as a mode switch,
)
interlock, or manual trip.
Therefore, there will be times during the test that the channel is inoperable but not tripped.
For a two channel trip system, this means that full reliance is being placed on the channel that is not being tested. A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at i
least one OPERABLE channel in the same trip system is monitoring that parameter.
5.
Allowed outage times (A0T) to permit restoration of inoperable instrumentation to OPERABLE status are provided in Table 3.1.1.
A0Ts vary depending on type of function and the number of inoperable channels per function.
If an inoporable channel cannot be restored to OPERABLE status within the A0T, the channel or the associated trip system must be placed in the tripped condition.
Placing the inoperable channel in trip (or the associated trip system in trip) conservatively compensates for the inoperability and allows operation to continue. Alternatively, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where i
placing the inoperable channel in trip would result in a full scram) the Action Required must be taken.
0YSTER CREEK 3.1-3 Amendment No.: 3,lI6,UI,176 s/N/#
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i A0Ts discussed in 4 (6 h:;urs for surveillance) and 5 (repair A0Ts in i
Table 3.1.1, Notes nn, oo and pp) above have been determined in i
accordance with References 1 through 6 except for instrumentation in Table 3.1.1, Sections M and N.
Note kke-has been provided to specify a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> surveillance A0T for those instruments.
i Bypasses of inputs to a trip system other than the IRM and APRM bypasses are provided for meeting operational requirements listed in the notes in Table 3.1.1.
Note 'a' allows the "high water level in scram discharge volume" scram trip to be bypassed in the refuel mode.
In order to reset the safety system after a scram condition, it is necessary to drain the scram discharge volume to clear this scram input condition.
(This condition usually follows any scram, no matter what the initial cause might have been.)
In order to do this, this particular scram function can be bypassed only in the refuel position.
Since all of the control rods are completely inserted following a scram, it is permissible to bypass this condition because a control rod block prevents withdrawal as long as the switch is in the bypass condition for this function.
The manual scram associated with moving the mode switch to shutdown is used merely to provide a mechanism whereby the reactor protection system i
scram logic channels and the reactor manual control system can be energized.
The ability to reset a scram twenty (20) seconds after going into the SHUTDOWN MODE provides the beneficial function of relieving scram pressure from the control rod drives which will increase their expected lifetime.
To permit plant operation to generate adequate steam and pressure to I
establish turbine seals and condenser vacuum at relatively low reactor power, the main condenser vacuum trip is bypassed until 600 psig.
This bypass also applies to the main steam isolation valves for the same reason.
The action required when the minimum instrument logic conditions are not met is chosen so as to bring plant operation promptly to such a condition that the particular protection instrument is not required; or the plant is placed in the protection or safe condition that the instrument initiates.
This is accomplished in a normal manner without subjecting the plant to abnormal operations conditions.
The action and out-of-service requirements apply to all instrumentation within a particular function, e.g., if the requirements on any one of the ten scram functions cannot be met then control rods shall be inserted.
The trip level settings not specified in Specification 2.3 have been included in this specification.
The bases for these settings are discussed below.
The high drywell pressure trip setting is 1 3.5 psig. This trip,;;11 scram the reactor, initiate core spray, initiate primary containment isolation, initiate automatic depressurization in conjunction with low-low-low-reactor water level,. initiate the standby gas treatment system and isolate the reactor building.
The scram function shuts the core down during the loss-of-coolant accidents. A steam leak of about 15 gpm and a liquid leak of about 35 gpm from the primary system will cause drywell pressure to reach the scram point; and, therefore, the scram provides protection for breaks greater than the above.
0YSTER CREEK 3.1-4 ent No.:
High drywell pressure provides a second means of initiating the core spray to mitigate the consequences of loss-of-coolant accident.
Its trip setting of 13.5 psig initiates the core spray in time to provide adequate core cooling.
The break size' coverage of high drywell pressure was discussed above.
Low-low water level and high drywell pressure in addition to initiating core spray also causes isolation valve closure.
These settings are adequate to cause isolation to minimize the offsite dose within required limits.
It is permissible to make the drywell pressure instrument channels inoperable during performance of the integrated primary containment leakage rate test provided the reactor is in the COLD SHUTDOWN condition.
The reason for this is that the Engineered Safety Features, which are effective in case of a LOCA under these conditions, will still be effective because they will be activated (when the Engineered Safety Features system is required as identified in the technical specification of the system) by low-low reactor water level.*
The scram discharge volume has two separate instrument volumes utilized to detect water accumulation.
The high water level is based on the design that the water in the SDIV's, as detected by either set of level instruments, shall not be allowed to exceed 29.0 gallons; thereby, permitting 137 control rods to scram.
To provide further margin, an accumulation of not more than 14.0 gallons of water, as detected by either instrument volume, will result in a rod block and an alarm.
The accumulation of not more than 7.0 gallons of water, as detected in either instrument volume will result in an alarm.
Detailed analyses of transients have shown that sufficient protection is provided by other scrams below 45% power to permit bypassing of the turbine trip and generator load rejection scrams. However, for operational convenience, 40% of rated power has been chosen as the setpoint below which these trips are bypassed. This setpoint is coincident with bypass valve capacity.
A low condenser vacuum scram trip of 20 inches Hg has been provided to protect the main condenser in the event that vacuum is lost. A loss of condenser vacuum would cause the turbine stop valves to close, resulting in a turbine trip transient.
The low condenser vacuum trip provides a reliable backup to the turbine trip.
Thus, if there is a failure of the turbine trip on low vacuum, the reactor would automatically scram at 20 inches Hg.
The condenser is capable of receiving bypass steam until 7 inches Hg vacuum thereby mitigating the transient and providing a margin.
The settings to isolate the isolation condenser in the event of a break in the steam or condensate lines are based on the predicted maximum flows that these systems would experience during operation, thus permitting operation while affording protection in the event of a break.
The settings correspond to a flow rate of less than three times the 5
normal flow rate of 3.2X10 lb/hr.
Upon initiation of the alternate shutdown panel, this function is bypassed to prevent spurious isolation due to fire induced circuit faults.
0YSTER CREEK 3.1-5 Amendment No.: 13,18,I48,448 111,176 2Cerrectiom-11/30/87-
II ~
e "he setting of ten times the stack release limit for isolation of the air-ejector offgas line is to permit the operator to perform normal, i
immediate remedial action if-the stack limit is exceeded. The time necessary for this action would be extremely short when considering the annual averaging which is allowed under 10 CFR 20.106, and, therefore, would produce insignificant effects on doses to the public.
Four radiation monitors are provided which initiate isolation of the reactor building and operation of the standby gas treatment system.
Two monitors are located in the ventilation ducts, one is located in the area of the refueling pool and one is located in the reactor vessel head storage area.
The trip logic is basically a 1 out of 4 system. Any upscale trip will cause the desired action.
Trip settings of 17 mr/hr in the duct and 100 mr/hr on the refueling floor are based upon initiating standby gas treatment system so as not to exceed allowed dose rates of 10 CFR 20 at the nearest site boundary.
l 5
The SRM upscale of 5 x 10 CPS initiates a rod block so that the chamber l
can be relocated to a lower flux area to maintain SRM capability ag power is increased to the IRM range.
Full scale reading is 1 x 10 CPS.
i This rod block is bypassed in IRM Ranges 8 and higher since a level of 5 5
x 10 CPS is reached and the SRM chamber is at its fully withdrawn position.
The SRM downscale rod block of 100 CPS prevents the instrument chamber from being withdrawn too far from the core during the period that it is l
required to monitor the neutron flux.
This downscale rod block is also i
bypassed in IRM Ranges 8 and higher.
It is not required at this power levelsincegoodindicationexistsigtheIntermediateRangeandtheSRM will be reading approximately 5 x 10 CPS when using IRM Ranges 8 and higher.
\\
The IRM downscale rod block in conjunction with the chamber full-in position and range switch setting, provides a rod block to assure that the IRM is in its most sensitive condition before startup.
If the two latter conditions are satisfied, control rod withdrawal may commence even if the IRM is not reading at least 5%.
However, after a substantial neutron flux is obtained, the rod block setting prevents the chamber from being withdrawn to an insensitive area of the core.
The APRM downstale setting of 2 2/150 full scale is provided in the RUN MODE to prevent control rod withdrawal without adequate neutron monitoring.
High flow in the main steamline is set at 120% of rated flow.
At this setting the isolation valves close and in the event of a steam line break limit the loss of inventory so that fuel clad perforation does not i
The 120% flow would correspond to the thermal power so this l
occur.
^
would either indicate a line break or too high a power.
i Temperature sensors are provided in the steam line tunnel to provide for closure of the main steamline isolation valves should a break or leak i
occur in this area of the plant.
The trip is set at 50*F above ambient i
temperature at rated power.
This setting will cause isolation to occur i
for main steamline breaks which result in a flow of a few pounds per minute or greater.
Isolation occurs soon enough to meet the criterion of no clad perforation.
0YSTER CREEK 3.1-6 Amendment No.: 7,7,169,I71,176 01-/5/7kr-n/5/71-
l The loc-low-loc water level trip point is set at 4'8" above the top of
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the active fuel and will prevent spurious operation of the automatic i
reitef. system.
The_ trip point established will initiate the automatic depressurization system in time to provide adequate core cooling.
i Specification 3.1.B.1 defines the minimum number of APRM channel.-inputs required _to permit accurate average core power monitoring.
Specification 3.1.B.3 defines APRM channel input operability requirements in order to ensure a sufficient APRM response to regional power oscillations.. Specifications 3.1.B.2 and 3.1.C.1 further define the distribution of the OPERABLE chambers to provide monitoring of local power changes that might be caused by a single rod withdrawal. Any nearby, OPERABLE LPRM chamber can provide the required input for average i
core monitoring. A Travelling Incore Probe or Probes can be used temporarily to provide APRM input (s) until LPRM replacement is possible.
1 Since APRM rod block protection is not required below 61% of rated power, as discussed in Section 2.3, limitina Safety System Settinas, operation may continue below 61% as long as Specification 3.1.B.1 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
-l maintained by alternate means.
I In the rare event that Travelling In-core Probes (TIPS) are used to meet i
the requirements 3.1.B or 3.1.C the licensee may perform an analysis of substitute LPRM inputs to the APRM system using spare (non-APRM input)
LPRM detectors and change the APRM system as permitted by 10 CFR 50.59.
5 Under assumed loss-of-coolant accident conditions and certain loss of offsite power conditions with no assumed loss-of-coolant accident, it is inadvisable to allow the simultaneous starting of emergency core cooling i
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 t
appropriate load sequencing, i
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.
Also, the Reactor Building Closed Cooling Water and Service Water pump circuit breakers will trip whenever a loss-of-coolant accident condition exists.
This is justified by Amendment 42 of the Licensing Application which determined that these pumps were not required during this accident condition.
l 5
Ghanget-6 OYSTER CREEK 3.1-7 Amendment No.: 9,15,Ilf,fff,176 01/4/80-i
i.
C.
Minimum Critical Pow r Ratio (MCPR)
]
During steady state power operation thh MINIMUM CRITICAL POWER RATIO (MCPR) shall be, equal to or greater than the MCPR limit as specified in the COLR.
The MCPR limit for each cycle as identified in the COLR shall be greater than or equal to 1.47.
When APRM status changes due to instrument failure (APRM or LPRM input failure), the MCPR requirement for the degraded condition shall be met within a time interval of eight (8) hours, provided that the control l
rod block is placed in operation during ihis interval.
For core flows other than rated, the nominal value for MCPR shall be i
increased by a factor of k, where k is as shown in the COLR.
g g
[
If at any time during power operation it is determined by normal surveillance that the limiting value for MCPR is being exceeded for reasons other than instrument failure, action shall be initiated to restore operation to within the prescribed limits.
If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, action shall be initiated to bring the reactor to the cold shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. During this period, surveillance and corresponding action shall continue until reactor operation is within the prescribed limit at which time power operation may be continued.
t Baseet f
The Specification for averaga planar LHGR assures that the peak l
cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200'T limit specified in
[
The analytical methods and assumptions used in evaluating the fuel design limits are presented in FSAR Chapter 4.
LOCA analyses are performed for each fuel design at selected exposure points to determined APLHGR limits that meet the PCT and maximum oxidation limits of 10 CFR 50.46.
The analysis is performed using GE calculational models which are consistent with the requirements of 10 CFR 50, Appendix K.
The PCT following a postulated LOCA is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is not strongly influenced by the rod to rod power distribution within an assembly.
Since expected location variations in power distribution within a fuel assembly affect the calculated peak clad temperature by less than 120*F relative to the peak temperature for a typical fuel design, the limit on the average linear heat generation rate is sufficient to assure that calculated temperatures are below the limits specified in 10 CFR 50.46.
OYSTER CREEK 3.10-2 Amendment No.: 48/, 7$/,
111, 146, 444, 176 t
e e
n
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.- ~
-i i
Th3 maximum av0rtg3 plcnte LHGR limite for tha various fuol
. types: currently;being used~are provided in the COLR. The
[
'I MAPLHGR limits for both fivelloop and four-loop operation; i
with.the~ idle. loop unisolated are shown.- Four-loop
{
operation with the idle loop isolated (suction, discharge-
.j and discharge bypass valves closed) requires that a MAPLHGR-I multiplier of 0.98 be applied to all fuel types.
-i Additional requirements for isolated loop operation ars' '
[
given in Specification 3.3.F.2.
Fuel _ design evaluations are performed to demonstrate that the cladding 14 plastic strain and other fuel design limits i
are not exceeded during anticipated operational occurrences for operation with LHGRs up to the operating limit LHGR.'
1 The analytical methods and assumptions used in evaluating the anticipated operational occurrences to establish the l
operating limit MCPR are presented'in the FSAR, Chapters 4,
'I 6 and 15 and in Technical Specification 6.9.1.f.'
To assure' that the-Safety Limit MCPR is not exceeded during any moderate frequency transient event, limiting transients l
have been analyzed to determine the largest reduction in Critical Power Ratio (CPR). -The types of transients evaluated are pressurization, positive reactivity insertion and coolant temperature decrease. The operational MCPR limit is selected to provide margin'to accommodate i
transients and uncertainties in monitoring the core operating state, manufacturing, and in the critical power correlation itself. This limit is derived by addition of' f
the CPR for the most limiting transient to the safety limit MCPR designated in Specification 2.1.
r A lower bound of 1.47 has been established'for the 1
operating limit MCPR value to provide sufficient margin to the MCPR safety limit in the event of reactor thermal-1 hydraulic instability. The 1.47 limit will be considered against the minimum operating CPR limit based on reload transient and accident analysis. The higher of core i
stability or reactor transient and accident determined MCPR will be used to determine the cycle operating limit.
The APRM response is used to predict when the rod block occurs in the analysis of the rod withdrawal error transient. The transient rod position at the rod block and corresponding MCPR can be determined. The MCPR has been i
evaluated for different APRM responses which would result from changes in the APRM status as a consequence of i
bypassed APRM channel and/or failed / bypassed LPRM' inputs.
'I The steady state MCPR required to protect the minimum i
transient CPR for the worst case APEM status condition (APRM Status 1) is determined in the rod withdrawal error transient analysis. The steady state MCPR values for APRM status conditions 1, 2, and 3 will be evaluated each cycle.
For those cycles where the rod withdrawal error transient is not the most severe transient the MCPR Value for APRM status conditions 1, 2, and 3 will be the same and I
be equal to the limiting transient MCPR value.
t OYSTER CREEK 3.10-3 Amendment No. II, II9', IIf176 i
,n s
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