ML20043H067
| ML20043H067 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/14/1990 |
| From: | Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20043H068 | List: |
| References | |
| NUDOCS 9006210534 | |
| Download: ML20043H067 (13) | |
Text
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10CFR 50.90 PH!! ADELPHIA ELECTRIC COMPANY NUCLEAR GROUP HEADQUARTERS 955 65 CHESTERBROOK BLVD.
WAYNE, PA 19087-5691 (n s) e40 sooo June 14, 1990 Docket Nos. 50-352 50-353
- l-License Nos. NPF-39 g
.b llUF -
' hi U.S. Nuclear Regulatory Commission I
( )y j
Attn: Document Control Desk Washington, DC 20555 i.
I I
SUBJECT:
Limerick Generating Station, Units 1 and 2
'3 Technical Specifications Change Request No. 90-02-0
Dear Sir:
Philadelphia Electric Company (PECo) hereby submits Technical Specifications Change Request (TSCR) No. 90-02-0, in accordance with 10 CFR 50.90, requesting an amendment to the Technical Specifications (TS) (Appendix A) of Operating License Nos. NPF-39 and NPF-85.
Information supporting this Change Request is contained in Attachment 1 to this letter, and the proposed 4l replacement pages are contained in Attachment 2.
This submittal requests changes to TS Section 3.3.1 " Reactor Protection System Instrumentation," and Section 3.3.6, " Control Rod Block Instrumentation," to eliminate the operability requirements for the Average Power Range Monitors (APRMs) in Operational Condition 5 except while performing a shutdown margin demonstration as described in TS Section 3.10.3.
If you have any questions regarding thr. matter, please contact us.
Sincerely yours, l1f*1 i
G. A. Hunger, Jr4 Manager Licensing Section Nuclear Engineering and Services
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Attachments cc:
T. T. Martin, Administrator, Region 1. USNRC
(;
T. J. Kenny, USNRC Senior Resident inspector, LGS T. M. Gerusky, Director, PA Bureau of Radiological Protection 3
=
TECHNICAL SPECIFICATION CHANGE REQUEST NO. 90-02-0
.,;CO MONWEALTH OF PENNSYLVANIA :
ss.
COUNTY OF CHESTER D. R. Helwig, being first duly sworn, deposes and says-That he is Vice President of Philadelphia Electric Company; the Applicant herein; that he has read the foregoing Application for Amendment of facility Operating Licenses to reduce the APRM operability requirement for OPCON 5 and knows the contents thereof; and the statements and matters set forth therein are true and correct to the best' of his knowledge, infomation and belief.
A Vice Preside t L
s Subscribed and sworn to-beforemethis/ day of j n w 1990.
v sh fiA%O b
JA4 a Notary Public NOTARIAL SEAL CATHERINE A. MENDE1 Notar' Public f
Trecyffrin Two., Chester County 1_
My Came. Nation factres Sect. 4.1993 i
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ATTACIMENT 1 LIMERICK GENERATING STATION Units 1 and 2 I
Docket No. 50-352 50-353 License No. NPF-39 5
NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST No. 90-02-0
" Proposed Changes to the Technical Specifications to Reduce ~the Average Power Range Monitor Operability Requirement During Operational Condition 5" l
Supporting Information for Changes - 10 pages I
Docket Nos.
50-352 50-352 License Nos. NPF-39 NPF-85 Philadelphia Electric Company, Licensee under Facility Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station (LCS), Units 1 and 2, respectively, hereby requests that the TS contained in Appendix A to the Operating Licenses be amended as proposed herein. The proposed changes to the TS are indicated by the vertical bars in the margins of the pages contained in.
We request the changes proposed herein to be effective by September 1, 1990, to allow their use during the next Unit I refueling outage scheduled to begin September 8, 1990.
This change request provides a discussion and description of the proposed TS changes, a safety assessment of the proposed TS changes, information supporting a finding of No Significant Hazards Consideration, and information supporting an Environmental Assessment.
Discussion and Description of the Proposed Changes LGS TS currently require that the neutron flux trips and control rod blocks of the Average Power Range Monitors (APRMs) be operable while in Operational Condition 5 (OPCON 5). This requirement restricts outage maintenance activities and requires surveillance and maintenance as necessary to maintain system operability. The primary reason for removing the ARPM operability requirement L
is to reduce critical path time by allowing maintenance activities to be performed on the Local Power Range Monitor (LPRM) strings (which input to the APRM circuitry) in conjunction with other refueling activities.
In addition, the proposed change will preclude the need for testing and maintenance to maintain system operability in OPCON 5 except when APRM operability is required t
for a shutdown margin demonstration.
We have determined that removing the APRM operability requirement while the plant is in OPCON 5 is acceptable since the APRMs are not necessary for safe operation because there are sufficients levels of protective controls designed to. prevent inadvertent criticality and fuel damage during refueling..The Intermediate Range Monitors (IRMs) Source Range Monitors (SRMs), Refueling Interlocks, and plant procedures, each provides protection which maintains the needed defense-in-depth and therefore precludes the need for the APRMs to be operable in OPCON 5.
However, the requirement for the APRNs to be opergdle during a shutdown margin demonstration when the mode switch is in Starta'p as allowed by TS Section 3.10.3 will remain unchanged. TS Section 3.10.3 is a l
Special Test Exception which allows operators to change the reactor mode switch l
from Refuel to Startup to perfons a shutdown margin demonstration. Therefore, we propose to add the following qualification to the OPCON 5 APRM TS operability requirements, " Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3."
This note is proposed to be added to TS Tables 3.3.1-1, " Reactor Protection System Instrumentation," and 4.3.1-1, " Reactor Protection System Instrumentation Surveillance Requirements "
and TS Tables 3.3.6.-1, " Control Rod Block Instrumentation," and 4.3.6-1
" Control Rod Block Instrumentation Surveillance Requirements." The additional i w
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Docket Nos.
50-352 50-352 License Nos. NPF-39 NPF-85 l
l<
space required to accommodate the proposed note "k" on page 3/4 3-8 requires item 8 be moved to page 3/4 3-7.
Safety Assessment The proposed TS changes remove the requirement for APRM operability while the plant is in OPCON 5, except during shutdown margin demonstrations perfomed in accordance with TS Section 3.10.3.
To assess the impact on safety and the design bases accidents of the proposed change, we need to examine those systems and mechanisms which contribute to safe operation while the plant is in OPCON 5.
Each of these systems and mechanisms contribute to the defense-in-depth design and operation. We also will show that the current APRM operability requirement is unnecessary to maintain this defense-in-depth.
The Neutron Monitoring System (NMS) is composed of the following subsystems: SRM, IRM, LPRM, APRM, Rod Block Monitor, and Traversing Incore Probe. The purpose of the SRM, IRM, and APRM subsystems is to monitor local and core average neutron flux levels and provide trip signals to the Reactor Protection System (RPS) and control rod block portion of the Reactcr Manual Control System (RMCS) as required. The NMS provides local and core average power information to the reactor operator. The IRM and APRM are safety-related subsystems and provide safety functions.
The SRM subsystem is composed of four detectors that are inserted into the core during shutdown conditions. Although the subsystem is not safety-related.-
it is important to plant safety. The SRMs are required by TS to be operational in OPCON 5.
During refueling operations, the plant operators use the SRMs to
-ensure that neutron flux remains within an acceptable range. Also, plant operators can monitor the SRMs for increases in neutron flux which may indicate that the reactor is approaching criticality.
The IRM subsystem is composed of eight incore detectors that are inserted
.into the core. The IRM is a five-decade instrument with ten ranges that are l
1 ranged up during normal power increases. The IRMs are designed to monitor neutron flux levels at a local core location and provide protection against local criticality events caused by control rod withdrawal errors.
The IRMs monitor neutron flux levels from the upper portion of the SRM range to the lower portionoftheAPRMrange.
In terms of rated reactor power, the IRMs range from about 10 % of full reactor power to greater than 15% of full reactor power.
The IRMs provide control rod block and scram functions at 108 and 120, respectively, of a 125 division scale.
The APRMs do not have incore detectors of their own but receive input from the LPRM detectors which are located at various levels throughout the core.
The APRMs monitor core power from about 1% of full reactor power to 125% of full L
reactor power. 'The APRMs represent a core average power level while the IRMs and SRMs indicate a local power level.
In OPCON 5, the APRMs operate in the setdown mode to provide a control rod block and scram function at 12% and 15%
core average power, respectively.
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Docket Nos.
50-352 50-352 4
License Nos. NPF-39 NPF-85 The safety design bases of the IRM subsystem is to generate trip signals to prevent fuel damage resulting from anticipated or abnormal-operational transients that could possibly occur while operating in the intermediate power range. The safety design bases of the APRM subsystem is to generate trip signals in response to' average neutron flux increases in time to prevent fuel damage while the plant is in the operating power range. The independence and redundancy incorporated in the design of the IRM and APRM subsystems are consistent with the safety design bases of the NMS and RPS.
There are various levels of control to prevent inadvertent reactor criticality and fuel damage during refueling operations.
1)
Licensed plant operators are trained to operate equipment and follow approved procedures.
2)
Plant approved refueling and maintenance procedures specify core alteration steps.
3)
SRMs indicate the potential for reactor criticality and generate a control rod block signal on high neutron flux levels. When shutdown margin has not been demonstrated. TS Section 3.9.2 requires the shorting links be removed so that the SRMs will operate in the noncoincident scram mode to cause a' reactor scram as necessary.
4)
Refueling interlocks prevent the removal of more than one control rod-and prevent the insertion of fuel bundles into the core unless all control rods are fully inserted.
5)
The IRMs and APRMs provide an indication of local power and average power, respectively.
IRMs and APRMs will provide rod blocks and scram signals on high neutron flux levels.
The ARPMs are not necessary for safe operation of the plant during OPCON 5 i
because the IRMs will generate an RPS scram or control rod block if neutron flux increases to the applicable setpoint. The'IRMs are required by TS to be operational in OPCON 5.
The IRMs are a safety-related subsystem of the NMS and are designed to indicate and respond to neutron flux increases at local core locations. The APRMs are designed to monitor and respond (scram and/or control rod block) to a core average neutron flux level. The most likely reactivity p
insertion transient expected during refueling would be a core alteration type event, e.g., control rod withdrawal or fuel assembly insertion into the core. A core alteration event would result in a local core criticality transient readily
' detected by the IRMs and/or SRMs.
The IRM subsystem is designed and calibrated to respond to a neutron flux level that is significantly less than the flux level monitored by the APRMs.
For example, during refueling, when the IRMs are on their most sensitive range, the IRMs will generate a scram signal at less than 0.01% core average power while the APRMs will genrrate a scram signal at 15% core average power. The IRM.
Docket Nos, 50-352 50-352 4
License Nos. NPF-39 NPF-85 L
subsystem acts as a backup protection system to the Refueling Interlocks (RIs) during refueling.
RIs are required to be operational during refueling operations in OPCON 5.
They are not safety-related but are designed such that a single component l
failure does not cause an interlock failure. The purpose of the RIs is to restrict the movement of the control rods and the operation of the refueling equipment to rainforce operational procedures that prevent the reactor from becoming critical during refueling operations. RIs require that all control rods be fully inserted into the core prior to allowing reactor operators to select and withdraw a single control rod. Other RIs will prevent the withdrawal of a control rod if the fuel loaded refueling platform is over the core. Also, the RIs require an "all-rods-in" signal before allowing a fuel loaded refueling 3
platform to go over the core.
[
TS and plant operating procedures allow only one control rod to be withdrawn or removed at a time while the plant is in OPCON 5 and the mode switch is in " Refuel." The core loading pattern is designed to ensure that the core is l
subcritical by a specified margin with the most reactive control rod at the full out position. Withdrawal of one control rod would not cause criticality and the l
event would not register on the APRMs.
l The design of the control rod drive system reduces the probability of a l
control rod error during refueling. For example, the latching action of the collet finger assembly serves to block the index tube in place. The velocity limiter physically prevents the control blade from being removed from the core-with fuel in place.
The LGS Final Safety Analysis Report (FSAR) Section 15.4.1, " Rod Withdrawal Error - Low Power," evaluated the potential for a control rod removal error during refueling. The concern is potential inadvertent criticality due to the following events.
j 1.
Removal of the highest worth control rod.
2.
Withdrawal of a second control rod.
3.
Accidental insertion of fuel into a cell not controlled by a control rod blade.
The FSAR concludes that the above scenarios are adequately precluded by refueling interlocks, core design, and control rod hardware design. However, l
should operator errors, followed by equipment malfunctions, result in an inadvertent criticality event, necessary safety actions (control rod block or scram) will be taken prior to violation of a safety limit. The IRMs would provide a rod block or scram function as appropriate.
FSAR Section 15.9.2 " Approach to Operational Nuclear Safety," was reviewed I
to determine the safety requirements of the NMS during normal operations and operational transient conditions when the reactor is shutdown and the vessel
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Docket Nas. 50-352 50-352
+
License Nos. NPF-39 NPF-85 head removed. The intent of the review was to identify conditions when the APRMs and IRMs were required to be operable to mitigate unacceptable consequences of inadvertent operational or transient conditions. To bound the review, two Operating States in the FSAR were considered, i.e. Operating State A (vessel head removed, reactor shutdown, atmospheric pressure) and Operating State B (vessel head removed, reactor not shutdown, atmospheric pressure).
Since operating State A corresponds to the Refueling conditions with a single control rod withdrawn or removed, only this operating state was evaluated.
Of the various operating and transient events evaluated in the FSAR for Operating States A and B, only one event required a rod block function and eight other events required a reactor scram to avoid unacceptable consequences. 'There were two trips in Operating State A (applicable to Refueling - Shutdown), i.e.,
one control rod block and one reactor scram. The rod block was a result of one or more single active failures, not a high neutron flux signal from the NMS.
The only scram in Operating State A was a result of a manual or inadvertent operator action, not from the neutron monitoring system. Therefore, from a safety standpoint, the NMS (i.e., APRMs, IRMs) is not required to mitigate the undesirable operational or transient conditions evaluated by the FSAR while in OPCON 5 with the plant shutdown.
Section 7.2.1, " Description," of FSAR Section 7.2, " Reactor Trip System (Reactor Protection System (RPS)) Instrumentation and Controls." states that for the initial fuel load and during shutdown margin demonstration testing, high-t high trip contacts from each SRM are combined with.IRM and APRM trips to produce l
a noncoincident reactor NMS trip. Accordingly, the proposed change leaves intact the TS requirement that the APRMs be operable during shutdown margin testing. Otherwise, the FSAR does not require that the APRMs be available for s
OPCON 5 plant etnditions. The initial fuel loading represents a high reactivity case because th! fuel is unexposed. Also, the shutdown margin of the initial core cannot be demonstrated until the core is fully loaded.
The APPM subsystem also provides a reactor power signal to the Redundant Reactivity Iontrol System (RRCS). The RRCS is designed to mitigate the consequencer of an Anticipated Transient Without Scram (ATWS) event. The RRCS uses the APRM signal to detemine if the reactor has scrammed during a transient condition.
If the reactor has not shutdown by normal means to selected transient conditions, then the RRCS may initiate one of several protective systems.to mitigate a possible ATWS event. For example, a high reactor pressure or low reactor water level signal will initiate a timer within the RRCS.
If the timer times out and the APRMs are not downscale, then the RRCS will
. automatically initiate the Standby Liquid Control System (SLCS) to inject into the vessel.
Not requiring the APRM subsystem to be operable during refueling would i
remove the ARPM input into the RRCS. This does not cause c safety concern as the RRCS is not used during refueling operations and is not required to be operable during refueling by the TS, the FSAR, or by the ATWS design bases. The l
SLCS is, however, required by TS to be operable during control rod withdrawals while in OPCON 5.
The SLCS can be manually initiated if required; but the SLCS
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Docket N2s.- 50-352 50-352 License Nos. NPF-39 NPF-85 j
is not required to be automatically initiated during refueling. Thus, removing the APRM input to RRCS during refueling does not affect either the safe operation or refueling operation of the plant.
l Not discussed previously is a transient scenario which is not applicable to Operating State A, i.e., vessel head removed and reactor shutdown. However, for completeness, the event is discussed as follows. The transient assumes that a local inadvertent criticality event occurs at a core location near an IRM that is bypassed during refueling operations. The hypothetical question arises as to whether the APRM subsystem (if operable) would indicate and scram the control rods on a high neutron flux level before the operable IRMs would respond to the event. The answer is that a neutron flux transient would be observed by the IRMs before the APRM electronics would detect the event. The core coupling is such that a local criticality event would immediately be transmitted throughout the core and would be detected by the operable IRMs. The IRMs would be on scale l
before the APRMs detected the event because the IRMs are designed and calibrated l
to be more sensitive to neutron flux than the APRMs.
i In suianary, the APRMs are not necessary for safe operation of the plant while operating in OPCON 5 with the mode switch in " Refuel" for the following reasons.
o The IRMs are a safety-related subsystem of the NMS and are required by TS to be operable in OPCON 5.
The IRMs will generate an RPS Scram or control rod block if neutron flux increased =to the applicable setpoint.
o The IRMs and SRMs are designed and calibrated to be more sensitive to neutron flux than the APRMs.
o The IRMs are designed to monitor local core events while the APRMs provide a measure of core average power condition.
The IRMs can L
monitor and react to the most probable reactivity events expected during refueling, i.e., control rod withdrawal or fuel insertion.
o.
The IRMs would detect.and respond (control rod block or reactor scram) to an inadvertent criticality event before the APRMs would provide a trip function.
The withdrawal of only one control rod in OPCON 5 is permitted by the O
"one-rod-out" interlock while in " Refuel." The core is designed to be subcritical with one rod out.
o The withdrawal of a second control rod or inadvertent addition of a fuel bundle in OPCON 5 is precluded by refueling interlocks, refueling procedures, and administrative controls.
The APRMs will still be required to be operational during a shutdown o
margin demonstration performed in OPCON 5 (a special test exception intheTS)..-
Docket Nos, 50-352 50-352 License Nos. NPF-39 NPF-85 o
The SRMs are required to be operational in OPCON 5.
The transient analysis discussed in the FSAR does not require the o
ARPMs to be operational in OPCON 5-to mitigate an undesirable operational or transient condition.
The proposed TS changes will not represent a change in the plant as described in the LGS FSAR.
FSAR Sections 7.1.2, 7.2, 7.6.1, 7.6.2, 7.7.1. 15.4 and 15.9 were reviewed in making this determination.
In conclusion, monitoring of neutron flux levels, administrative controls.
plant procedures, refueling interlocks, and SRM and IRM protective ftMares provide and maintain the defense-in-depth design and operation which precludes the need for the APRMs to be operable in OPCON 5 with the mode switch in
" Refuel."
Information Supporting a finding of No Significant Hazards Consideration We have concluded that the proposed changes to the LGS TS, which reduce the APRM operability requirement for OPCON 5 do not constitute a significant hazards consideration.
In support of this determination, an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 are provided below.
1)
The proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.
Not requiring APRMs to be operational in OPCON 5 will not increase the probability of inadvertent reactor criticality during refueling operations.
RIs, NMS (SRMs. IRMs), and procedural restrictions provide assurance that inadvertent criticality does not occur due to the simultaneous withdrawal p
or removal of two control rods or due to the inadvertent insertion of a fuel bundle into a core location with a control blade removed.
The FSAR Section 15.4.1 discusses the potential for a control rod withdrawal error during refueling and start-up operations. The discussion concludes that the withdrawal of one control rod does not require a safety action because the total worth of one control rod is not sufficient to cause criticality. The attempted withdrawal of two control rods, assuming an operator error and a single active failure, would result in a control rod block initiated by the RIs. The safety-related IRM subsystem, which is required by TS to be operable while in OPCON 5 is designed to generate _a rod block or reactor scram on high neutron flux and is therefore a backup L
protective system for the RIs during refueling.
The FSAR Section 15.9.2 discusses two potential transient conditions during refueling with the' reactor shutdown. The first is a control rod withdrawal error during Refueling which is terminated by a control rod block. The l 1
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Docket N:s.
50-352.
50-352 4.
License Nos. NPF-39 NPF-85 second is a manual or inadvertent reactor transient caused by multiple operator errors or equipment failures which is mitigated by a reactor scram signal. Although neither of these two events is assumed to be mitigated by the MS, the MS subsystems (i.e., SRMs. IRMs) are available and operable to generate a control rod block or scram signal if required during refueling.
The safety-related IRM subsystem of the MS is required by TS to be operable during OPCON 5 to support the safety design bases of the NMS and RPS. The SRM is not a safety-related subsystem but is important to plant safety and is required by TS to be operable in OPCON 5.
The SRM subsystem provides the plant operator with neutron flux levels from startup conditions to the IRM operating range. The SRMs and IRMs are designed to respond to local core conditions and would indicate and respond (control rod block or scram) to an accident condition to mitigate the transient.
Thus, the APRMs are not necessary to be operable in OPCON 5.
The proposed l
TS change will not alter the current requirements that the APRMs be l
operable during shutdown margin demonstrations in OPCON 5 when the mode switch is in Startup.
The proposed TS change would reduce the APRM operability requirement in-OPCON 5 and would not affect the FSAR evaluation of the inadvertent criticality due to the withdrawal or removal of the highest worth control rod or due to the insertion of fuel bundles in uncontrolled cells. The FSAR concludes that the RIs and plant procedures provide assurance that inadvertent criticality does not occur during refueling.
The FSAR also presented two potential transients during refueling with the reactor shutdown. One transient is terminated by a control rod block and the other by a reactor scram. Neither transient-requires the NMS-to mitigate the event. Removing the APRM operability requirement in OPCON 5 would not increase the consequences of either transient since the IRMs will still be operable in OPCON 5 to generate an RPS scram or control rod' block-if neutron flux increased to the-applicable setpoint. The IRMs function as a backup' protective system to RIs-during refueling.
The consequences of an accident will not be increased by the proposed TS change because of the existing lines of defense which prevent an inadvertent criticality event during refueling, e.g., administrative restrictions, refueling procedures, licensed plant operators. SRMs, RIs, and IRMs. Furthermore, should the number of operable IRM or SRM channels be less than that required by TS, the TS require that core alteration activities be suspended and all insertable control rods be inserted into the core.
l Therefore, the proposed changes do not result in an increase in the i
probability or consequences of an accident previously evaluated.
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Docket Nos, 50-352 50-352 License Nos. NPF-39 NPF-85 2):
The proposed charnes do not create the possibility of a new or different kind of acB dert dron any accident previously evaluated.
The proposed changes to the TS will remove the APRM operability requirement while in OPCON 5 (except for shutdown margin demonstration testing);
however. the SRMs and IRMs will still be required to be operable in OPCON 5.
The IRMs are safety-related and are designed to detect and respond to increases in neutron flux within the local core regions. Any inadvertent increases in neutron flux during refueling would originate at a local core location, i.e., rod withdrawal error or fuel bundle insertion. TS require IRM operability and will generate an RPS scram or control rod block if neutron flux increased to the setpoint. Therefore, removing the APRM operability requirement in OPCON 5 would would not effect any safety-related equipment or equipment-luportant to safety.
The APRMs provide core power infors.ation to the control room operator and also provides trip signals to the RMCS and RPS as required. Also, the APRMs provide an input signal to the P4CS. The absence of an APRM input 4
signal will not affect these systems during refueling operations.
Removing the APRM operability in OPCON 5 will not affect the response of safety-related equipment as previously evaluated in the FSAR. The proposed changes to the TS do not affect any safety-related equipment or equipment important to safety.
Removing the APRM operability requirement during refueling will eliminate the APRM input into the RRCS. However, the RRCS is not in service during refueling. The automatic injection mode of the SLCS'is not required to be operational during refueling. SLCS can be manually operated, if required, independent of the RRCS.
The proposed changes to the TS would remove the APRM operability requirement during refueling operations. TS require IRM operability and L
will generate an RPS scram or control rod block if neutron flux increased to the applicable setpoint.
No new types of accidents would be introduced since the SRMs and IRMs are available and required to be operable in OPCON 5.
Both SRMs and IRMs would-indicate and provide a control rod block or scram signal, as appropriate, to an increase in neutron flux to mitigate a transient event. Furthermore, should the number of operable IRM or SRM channels be less than that required by TS, the TS require that core alteration activities be suspended L
and all insertable control rods be inserted into the core.
Finally, the APRMs do not have functions which can cause an accident condition.
Therefore, the proposed TS changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.,
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y Dockst Nos. 50-352 50-352 License Nos. NPF-39 NPF-85 (3) The proposed changes do not involve a significant reduction in a margin of safety.
For the reasons discussed in items 1 and 2 above and because the TS Bases do not discuss or require APRM operability during OPCON 5. Refueling, the proposed (S changes do not involve a significant reduction in a margin of safety.
Infomation Supporting an Envirormental Assessment An environmental assessment is not required for the changes proposed by this Change Request because the requested changes conform to the criteria for
" actions eligible for categorical exclusion" as specified in 10 CFR 51.22(c)(9).
The requested changes will have no impact on the environment. The proposed changes do not involve a significant hazards consideration as discussed in the preceding section. The proposed changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite.
In addition, the proposed changes do not involve a significant increase in individual or cumulative occupational radiation exposure.
Conclusion The Plant Operations Review Committee and the Nuclear Review doard have reviewed these proposed changes to the TS and have concluded that they do not involve an unreviewed safety question, or a significant hazards consideration, and will mt endanger.the health and safety of the public.,_
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ATTACHMENT 2 LIMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos.
50-352 50-353 License Nos. NPF-39 MPF-85 TECHNICAL SPECIFICATION CHANGE REQUEST No.- 90-02-0 List of Attached Pages-Unit l' Unit 2 1.
3/4 3-2 3/4 3-2 3/4 3-5 3/4 3-5 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 3/4 3-58 3/4 3-58 3/4 3-59 3/4 3-59
'3/4 3-61 3/4 3-61 3/4 3-62 3/4 3-62 l
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_ _.... 4 TABLE-3.3.1-Ik
. REACTOR PROTECTION SYSTEM. INSTRUMENTATION APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS
[
FUNCTIONAL UNIT CONDITIONS
'PER TRIP SYSTEM (a)
' ACTION.
1.
Intermediate Range Monitors (b).
a.
Neutron Flux - High 2
-3 1
1 3, 4
.3 2
5(c) 3(d)
'3 b.
Inoperative 2
3 1
3, 4.
3 2
5 3(d) 3 2.
Average Power Range Monitor (*):
a.
Neutron Flux - Upscale, Setdown 2
2 1
3 2
2 --
5(c)(1) 2(d) 3 l
b.
Neutron Flux - Upscale
- 1) Flow Biased.
1 2
4
- 2) High Flow Clamped 1
2 4
c.
Inoperative 1, 2 -
2 1
3 2
2 5(c)(1) 2(d) 3 I
d.
Downscale 1(g) 2 4
3.
Reactor Vessel Steam Dome Pressure - High 1,2(f) 2..
1 4.
Reactor Vessel Water Level - Low, Level 3
- 1, 2 2
.1 5.
Main Steam Line Isolation Valve -
Closure 1(g) 1/ valve 4~
LIMERICK - UNIT 1 3/4 3-2
- -e.
e
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,_.e.
F TABLE 3.3.1-1-(Continued)
,[
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)
A' channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for
~
required surveillance without placing the-trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
( b) -_ This function _shall be automatically bypassed when the reactor mode switch is in the Run-position and the associated APRM is not downscale.
(c)
The " shorting links" snall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn
- and shutdown margin demonstrations performed per Specification 3.10.3.
(d)-
The noncoincident NMS reactor trip function logic is such that all channels j
go to both trip systems. Therefore, when the " shorting links" are removed, i
the Minimum OPERABLE Channels Per Trip System is 4 APRMs, 6 IRMs'and 2 SRMs.
i (e)
An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.
?
'(f)
This function is not required to be OPERABLE when the reactor pressure i
vessel head is removed per Specification 3.10.1.
7 (g)
This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
(h) '
This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.
(i)
.With any control rod withdrawn. Not applicable to control rods removed per i
Specification 3.9.10.1 or 3.9.10.2.
(j)
This function shall be automatically bypassed when turbine first stage i
pressure is equivalent to a THERMAL POWER of less than 30% of RATED THERMAL POWER.
(k)'
Also actuates the EOC-RPT system.
(1)
Required to be OPERABLE only prior to and during shutdown margin i
demonstrations as performed per Specification 3.10.3.
-GNot required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
LIMERICK - UNIT 1 3/4 3-5
TABLE 4.3.1.1-1
-[
-REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE' REQUIREMENTS CHANNEL.-
OPERATIONAL CHANNEL FUNCTIONAL-CHANNEL ~
CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK ~
TEST CALIBRATION (U)
SURVEILLANCE REQUIRED 1.
a.
Neutron Flux - High 5/U.S(b)
S/U(c),W R
2 S
W(j)
R 3, 4, 5 b.
Inoperative N.A.
W(j)
N. A. --
2, 3, 4,.5 2.
Average Power Range Monitor (0 :
a.
Neutron Flux -
S/U.S(b)
S/U(c),W-SA 2
Upscale, Setdown S
~
W(j)
SA 3, 5( )
g.
b.
-Neutron Flux - Upscale
- 1) Flow Biased 5,0(g)
S/U(c),W W(d)(e),SA 1
- 2) High Flow Clamped S
S/U(c),W W(d)(e),SA 1
c.
Inoperative N.A.
W(j)
N.A.
1, 2, 3, 5(k) l-d.
Downscale S
W SA 1
3.
Reactor Vessel Steam Dome Pressure - High S
M R
1,2(h) 4.
Low, Level 3 S
M R
1, 2 5.
Main Steam Line Isolation Valve - Closure N.A.
M R
1 6.
Main Steam Line Radiation -
High S-M R
1,2(h) 7.
Drywell Pressure - High S
M R
1, 2 8.
Scram Discharge Volume Water Level - High 1,2,5((I) a.
Level Transmitter S'
M R
1, 2, 5 I) b.
Float Switch N.A.
M R
LIMERICK - UNIT 1 3/4 3-7
~
~_
_... =...
L:. ;
a TABLE 4.3.1.1-1 (Continued)
~
P r-
= -
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS
~
CHANNEL OPERATIONAL.
CHANNEL FUNCTIGNAL CHANNEL-CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 9.
Turbine Stop Valve - Closure N.A.
M' R
1
~
- 10. Turbine Control Valve Fast Closure, Trip 011 Pressure - Low N.A.
M R
1 11.
Reactor Mode Switch Shutdown Position N.A.
R N.A.
1, 2, 3, 4,~5
- 12. Manual Scram N.A.
M N.A.
1, 2,~3, 4, 5-(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after -
entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.
(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER. Adjust the.APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER. Any APRM channel gain adjustment made in compliance with Specification 3.2.2 shall not be included in determining the absolute difference.
(e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.
(f) The LPRMs shall be calibrated ot~1 east.once per 1000 effective full power hours (EFPH) using the TIP system.
(g) Verify measured core flow (total' core flow) to be greater than or equal to established core flow at the existing loop flow (APRM % flow). During the.startup test program, data shall be recorded-for the parameters listed to provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with.
the criteria listed shall commence upon the conclusion of-the startup test program. -
(h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(1) With any control rod withdrawn. Not applicable.to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(j)
If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position..
(k) Required to be OPERABLE.only prior to and during shutdown margin. demonstrations as performed per Specification 3.10.3.
LIMERICK'- UNIT 1 3/4 3-8
ETABLE 3.3 6-1
.c-
~ CONTROL-ROD BLOCK INSTRUMENTATION
^
.a=^
MINIMUM APPLICABLE OPERABLE CHANNELS ~
OPERATIONAL TRIP.FUNCTIDH PER TRIP FUNCTION-CONDITIONS ACTION
- 1. ROD BLOCK MONITOR (a)
~
a.
Upscale--
2 1*
60 b.
Inoperative 2-1*
60.
c.
Downscale-2 1*
60 2.
~
a.
Flow Biased Neutron Flux -
Upscale 4
1 61 b.
Inoperative-4 1,2,5(f) 61 c.
Downscale 4
1 61 g
d.
Neutron Flux - Upscale, Startup 4
2,5(f) 61 3.
SOURCE RANGE MONITORS *** (b) a.
Detector not full in 3
2 61 2
5 61 b.
Upscale (c) 3 2
61
'2 5
61 Inoperative (c) 3-2
'61 c.
2' 5
61 d.
Downscale(d) 3 2
61 2
5 61 4.
INTERMEDIATE RANGE MONITORS a.
Detector not full in 6
2,5
'61 b.
Upscale 6
2,5 61 Inoperati 6
2,5 61 c.
Downscaleg d.
6 2,5 61 5.
Water Level-High 2
1,2,5**
62.
f 6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW l
a.
Upscale 2
1 62 b.
Inoperative 2
1 62 l
c.
Comparator.
2 1
62 7.
REACTOR MODE SWITCH SHUTDOWN POSITION 2
3,4 63 LIMERICK - UNIT 1 3/4-3-58 l
i.-
TABLE 3.3.6-1~(Continued)
' CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS ACTION 60 - Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3.
ACTION 61 With the number of OPERABLE channels one or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in'the tripped condition within one hour, u
e ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the-tripped condition within one hour.
ACTION 63 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, initiate a rod block.
NOTES With THERMAL POWER > 30% of RATED THERMAL POWER.
With more than one control rod withdrawn.
Not applicable to control rods-l removed per Specification 3.9.10.1 or 3.9.10.2.
These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs are in the core.
(a) The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30% of RATED THERMAL POWER.
(b) This function shall be automatically bypassed if detector count rate is
> 100 cps or the IRM channels are on range 3 or higher.
(c)--ThisfunctionisautomaticallybypassedwhentheassociatedIRMchannels are on range 8 or higher.
L (d) This function is automatically bypassed when the IRM channels are on range 3 or higher.
(e).This function is automatically bypassed when the IRM channels are on range 1.
i (f) Required'to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.
LIMERICK - UNIT 1 3/4 3-59
m.
~
- TABLE 4.3.6-1 s.
CONTROL ROD BLOCK' INSTRUMENTATION SURVEILLANCE REQUIREMENTS
- CHANKEL.
OPERATIONAL-CHANNEL-
-FUNCTIONAL' CHANNEL CONDITIONS FOR WHICH
~
-TRIP FUNCTION.
CHECK TEST CALIBRATION (a)
SURVEILLANCE REQUIRED ~
l
~
1.
R0D BLOCK MONITOR
.S/U(b)(c),(c) b)(c)
(c)'
a.
Upscale N.A.
'l*
S/U(
N.A.
1*
b.-
Incperative N.A.
c.
Downscale N.A.
S/U(b)(c)' (c)
SA 1*
2.
APRM a.
Flow Biased Neutron Flux -
S/U((b),Mb) M SA
~1 Upscale
.N.A.
N.A.
'1, 2, 5***-
b.
Inoperative N.A.
S/U(b),M c.
Downscale-N.A.
S/U SA 1
d.
Neutron Flux - Upscale, Startup N.A.
S/U(b),M SA 2,
5***
l 3.
SOURCE RANGE MONITORS S/U(b)
N.A.
2, 5 a.
Detector not full in N.A.
S/U(b),W b.
Upscale N.A.
S/U(b),W SA 2, 5 c.
Inoperative N.A.
S/U(b),W N.A.
2, 5-d.
Downscale N.A.
,W SA 2, 5 4.
INTERMEDIATE RANGE-MONITORS S/U((b),Wb) W N.A.
2, 5 a.
Detector not full.in N.A.
b.
Upscale N.A.
S/U(b),
SA 2, 5 c.
Inoperative N.A.
S/U N.A.
2, 5 S/U(b),W d.
Downscale N.A.
,W SA 2, 5 5.
Water Level-High N.A.
M R
1, 2, 5**
6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW S/U(b),M'b) M.
SA.
1-a.
Upscale N.A.
b.
Inoperative N.A.
S/U(b),M N. A.-
.I c.
Comparator N.A.
S/U SA 1
7.
REACTOR MODE SWITCH SHUTDOWN POSITION N.A.
R N.A.
3, 4 LIMERICK - UNIT 1
-3/4 3-61
.=
TABLE 4.3.6-1 (Continued)-
c.
CONTROL ROD-BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS
-TABLE NOTATIONS (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
(c) Includes reactor! manual control multiplexing system input.
With THERHAL POWER > 30% of RATED THERMAL POWER.
With more than one control rod withdrawn. Not_ applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2
- Required to be OPERABLE only prior to and during shutdown margin
~i demonstrations as performed per Specification 3.10.3.
')
k l
r l '
L LIMERICK - UNIT 1 3/4 3-62
<," v TABLE 3.3.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION I
APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)
ACTION L
Intermediate Range Monitors (b).
a.
Neutron Flux -liigh 2
3 1
3, 4 3
2 5(c) 3(d) 3 b.
Inoperative 2
3 1
3, 4 3
2 5
3(d) 3 2.
Average Power Range Mon? tori'):
a.
Neutron Flux - Upscale, Setdown 2
2 1
3 2
2 5(c)(1) 2(d) 3 l
b.
Neutron Flux - Upscale
- 1) Flow Biased 1
2 4
- 2) High Flow C1cmped 1
2 4
c.
Inoperative I, 2 2
1 3
2 2
5(c)(1) 2(d) 3 d.
Downscale 1(g) 2 4
3.
Reactor Vessel Steam Dome Pressure - High 1,2(f) 2 1
4.
Reactor Vessel Water Level - Low, Level 3 1, 2 2
1 5.
Main Steam Line Isolation Valve -
Closure 1(g) 1/ valve 4
LIMERICK - UNIT 2 3/4 3-2
TABLE 3.3.1-1 (Continued)
' d)
{
' REACTOR PROTECT!0N SYSTEM INSTRUMENTATION
. TABLE NOTATIONS (a).
A N :nnel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
(b)
This function shall be automatically bypassed when the reactor mode switch is in the Run position and the associated APRM is not downscale, j
i (c)
The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control-rod is withdrawn
- and shutdown margin 1
demonstrations performed per Specification 3.10.3.
(d)
The noncoincident NHS reactor trip function logic is such that all channels go to both trip systems. Therefore, when the " shorting links" are removed, the Minimum OPERABLE Channels Per Trip System is 4 APRMs 6 IRMs and 2 SRMs.
(e)
An APRM channel is inoperable if there are less than 2 LPRM inputs per level f
or less than 14 LPRM inputs to an APRM channel.
(f)
This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
t (g)
This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
(h)
This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required, t
(i)
With any control rod withdrawn. Not applicable to control rods removed per
[
Specification 3.9.10.1 or 3.9.10.2.
(j)
This function shall be automatically bypassed when turbine first stage pressure is equivalent to a THERMAL POWER of less tM n 30% of RATED-THERMAL POWER.
(k)-
Also actuates the EOC-RPT system.
(1)
Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.
1
.*Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
L LIMERICK - UNIT 2 3/4 3-5
TABLE 4.3.1.1-1 O
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS I
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH-FUNCTIONAL UNIT CHECK TEST CALIBRATION (a)
SURVEILLANCE REQUIRED 1.
a.
Neutron Flux - High S/U,5(b)
S/U(c),W R
2 5'
W(j)
R 3,4,5 b.
Inoperative N.A.
W(j)
N.A.
2,3,4,5 2.
Average Power Range Monitor (f):
a.
Neutron Flux -
S/U.S(b)
S/U(c),W SA 2
Upscale, Setdown S
W(j)
SA 3,5(k)
[
b.
Neutron Flux - Upscale
- 1) Flow Biased 5,0(g)
S/U(c),W W(d)(e),SA 1
- 2) Hfgh Flow Clamped S
S/U(c),W W(d)(e),SA 1
c.
Inoperative N.A.
W(j)
N.A.
1,2,3,5(k) l d.
Downscale S
W SA 1
3.
Reactor Vessel Steam Dome Pressure - High S
M R
1,2(h) 4.
Low, Level 3 S
M R
1, 2 5.
Main Steam Line Isolation Valve - Closure N.A.
M R
1 6.
Main Steam Line Radiation -
High S
M R
1,2(h) 7.
Drywell Pressure - High S
M R
1, 2 8.
Scram Discharge Volume Water Level - High 1,2,5((I)I) a.
Level Transmitter 5
M R
b.
Float Switch N.A.
M R
1, 2, 5 LIMERICK - UNIT 2 3/4 3-7
.o ;
o TABLE 4.3.1.1-1 (Continued) o REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERAT.ONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED.
I 9.
Turbine Stop Valve - Closure N.A.
M R
I
- 10. Turbine Control Valve Fast Closure, Trip 011 Pressure - Low N.A.
M R
1
- 11. Reactor Mode Switch Shutdown Position N.A.
R N.A.
1,2,3,4,5
- 12. Manual Scram N.A.
M N.A.
1,2,3,4,5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for a least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.
(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER. Any APRM channel gain adjustment made in compliance with Specification 3.2.2 shall not be included in determining the absolute difference.
(e)
This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.
(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) using the TIP system.
(g) Verify measured core flow (total core flow) to be greater than or equal to established core flow at the existing loop flow (APRM % flow). During the startup test program, data shall be recorded for the parameters listed to provide a basis for establishing the specified relationships. Comparisons of the actual data in accordance with the criteria listed shall commence upon the conclusion of the startup test program.
(h) This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(1) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(j)
If the RPS shorting links are required to be removed per Specification 3.9.2, they may be reinstalled for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance. During this time, CORE ALTERATIONS shall be suspended, and no control rod shall be moved from its existing position.
(k) Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.
LIMERICK - UNIT 2 3/4 3-8
TABLE 3.3.6-1
- [
CONTROL ROD BLOCK INSTRUMENIATION MINIMUM APPLICABLE OPERABLE CHANNELS OPERATIONAL TRIP FUNCTION PER TRIP FUNCTION CONDITIONS ACTION
- 1. ROD BLOCK MONITOR (a) a.
Upscale 2
1*
60 b.
Inoperative 2
1*
60 c.
Downscale 2
1*
60 2.
APRM a.
Flow Biased Neutron Flux -
Upscale 4
1 61 l
1,2,5(f) 61 b.
Inoperative 4
c.
Downscale 4
1 61 2,5(I) 61 l
d.
Neutron Flux - Upscale, Startup 4
SOURCE RANGE MONITORS *** (b) 3.
a.
Detector not full in 3
2 61 2
5 61 b.
Upscale (c) 3 2
61 2
5 61 Inoperative (c) 3 2
61 c.
2 5
61 d.
Downscale(d) 3 2
61 2
5 61 4.
INTERMEDIATE RANGE MONITORS a.
Detector not full in 6
2,5 61 b.
Upscale 6
2,5 61 c.
Inoperatiye 6
2,5 61 d.
Downscalete) 6 2,5 61 3.
Water Level-High 2
1,2,5**
62 6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.
Upscale 2
1 62 b.
Inoperative 2
1 62 c.
Comparator 2
1 62 7.
REACTOR MODE SWITCH SHUTDOWN POSITION 2
3,4 63 LIMERICK - UNIT 2 3/4 3-58 q
o.-
.O TABLE 3.3.6-1(Continued)
CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS ACTION 60 Declare the RBM inoperable and take the ACTION required by Specification 3.1.4.3.
ACTION 61 With the number of OPERABLE channels one or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the tripped condition within one hour.
ACTION 62 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel in the tripped condition within one hour.
ACTION 63 With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip function requirement, initiate a rod block.
NOTES With THERMAL POWER > 30% of RATED THERMAL POWER.
With more than one control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
These channels are not required when sixteen or fewer fuel assemblies, adjacent to the SRMs. are in the core.
(a) The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30% of-RATED THERMAL POWER.
(b) This function shall be automatically bypassed if detector count rate is
> 100 cps or the IRM channels are on range 3 or higher.
(c)' This function is automatically bypassed when the associated IRM channels are on range 8 or higher.
(d) This function is automatically bypassed when the IRM channels are on range 3 or higher.
(e) This function is automatically bypassed when the IRM channels are on range 1.
(f)-RequiredtobeOPERABLEonlypriortoandduringshutdownmargin demonstrations as performed per Specification 3.10.3.
LIMERICK - UNIT 2 3/4 3-59 i
4
lABLE 4.3.6-I c,
o CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS o
CHANNEL OPERATIONAL CHANNEL
' FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION (a)
SURVEILLANCE REQUIRED 1.
ROD BLOCK MONITOR S/U((b)(c)' (c) b)(c)
(c) a.
Upscale N.A.
SA 1*
b.
Inoperative N.A.
S/U(b)(c)
(c)
N.A.
1*
c.
Downscale N.A.
S/U SA 1*
2.
APRM a.
Flow Biased Neutron Flux -
Upscale N.A..
S/U(D) M SA 1
S/U(b),
N.A.
1, 2, 5***
l b.
Inoperative N.A.
S/U(b),M c.
Downscale N.A.
M SA 1
d.
Neutron Flux - Upscale, Startup N.A.
S/U(b),M SA 2, 5***
I 3.
SOURCE RANGE MONITORS S/U((b),W b) W a.
Detector not full in N.A.
N.A.
2, 5 b.
Upscale N.A.
S/U SA 2, 5 S/U(b),Wb),W c.
Inoperative N.A.
N.A.
2, 5 S/U(
SA 2, 5 d.
Downscale N.A.
4.
INTERMEDIATE RANGE MONITORS S/U(b)'y b) a.
Detector not full in N.A.
N.A.
2, 5 S/U(
b.
Upscale N.A.
S/U(b)'y 2, 5 c.
Inoperative N.A.
S/U(b)'y N.A-2, 5 d.
Downscale N.A.
,y 3
5.
Water Level-High N.A.
M R
1, 2, 5**
6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW b)
S/U((b),M a.
Upscale N.A.
SA 1
b.
Inoperative N.A.
S/U(D),M M
N.A.
I c.
Comparator N.A.
S/U SA 1
7.
REACTOR MODE SWITCH SHUIDOWN POSITION N.A.
R N.A.
3, 4 LIMERICK - UNIT 2 3/4 3-61
TABLE 4.3.6-1 (Continued) es CONTROL ROD BLOCK' INSTRUMENTATION SURVEILLANCE REQUIREMENTS j
i j
-tr l
TABLE NOTATIONS r
(a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
s (b) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
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(c) Includes reactor manual-control multiplexing system input.
With THERMAL POWER >~30% of RATED THERMAL POWER.
- ' With more than one control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2
' *** Required to be OPERABLE only prior to and during shutdown margin demonstrations as performed per Specification 3.10.3.
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LIMERICK - UNIT 2 3/4 3-62 t
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