ML20247E961
| ML20247E961 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/22/1989 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247E966 | List: |
| References | |
| NUDOCS 8904030279 | |
| Download: ML20247E961 (16) | |
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
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I PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELEU NIC COMPANY
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DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.141 License No. DPR-44 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Philadelphia Electric Company, et al. (the licensee) dated June 12, 1987 as amended on February 7, 1989 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I.
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 (1) 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 or 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.
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-44 is hereby amended to read as follows:
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Technical Specifications TheTechnical.SpecificationscontainedinAppendicesAandB,as revised through Amendment No. 141, are hereby incorporated in the license. PEC0 shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective prior to startup in Cycle 8.
FOR THE NUCLEAR REGULATORY COMMISSION v
Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: March 22, 1989
. r s
a ATTACHMENT TO LICENSE AMENDMENT NO. 141 FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50-277 l
l Replace the following pages of the Appendix A Technical Specifications with l
the enclosed pages. The revised areas are indicated by marginal lines, 1
Remove Insert 60 60 79 79 l
l 88 88 89 89 93a 93a i
_________m_________.-_._-____-._____________________m__________a____._____._______________m.__.______-____.___-___.______.___________-_______-_-_____________a
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.a PBAPS Unit 2 LIMITING ^,..NDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS F.
Surveillance Information l
F.
Surveillance Information Readouts Readouts The limiting conditions for Instrumentation shall be the instrumentation that calibrated and checked as provides surveillance in-indicated in Table 4.2.F.
formation readouts are given in Table 3.2.F.
l G.
Alternate Rod Insertion and l G.
Alternate Rod Insertion and Recirculation Pump Trip Recirculation Pump Trip The limiting conditions for Instrumentation shall be the instrumentation that functionally tested and initiates an Alternate Rod calibrated as indicated on Insertion scram and trips Table 4.2.G.
the reactor recirculation System logic shall be l
pumps to limit the conse-functionally tested as quences of a failure to scram indicated in Table 4.2.G.
during an anticipated transient are given in Table 3.2.G.
When in the RUN or STARTUP Mode, the required min bum number of instrument cr.nnels shall be operable with trip setpoints set consistent with the setting specified in Table 3.2.G.
The manual and automatic actuation logic, and actuation devices of both trip systems shall be operable when in the RUN or STARTUP Mode.
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Unit 2 I
PBAPS 3.2 BASES In addition to reactor protection instrumentation which initiates a reactor scram, protective instruments. tion has been provided which initiates action to mitigate the consequences of accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.
This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.
The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required even during periods when portions of such systems are out-of-service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.
When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
Some of the settings on the instrumentation that initiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.
Such instrumentation must be available whenever primary containment integrity is required.
The instrumentation which initiates primary system isolation is connected in a dual bus arrangement.
The low water level instrumentation set to trip at zero inches indicated level (538 inches above vessel zero) closes all isolation valves except those in Groups 1, 4 and 5.
Details of valve grouping and required closing times are given in Specification 3.7.
For valves which isolate at this level, this trip setting is adequate to prevent the core from being uncovered in the case of a break in the largest line assuming a 60 second valve closing time.
Required closing times are less than this.
The low-low reactor water level instrumentation is set to trip when reactor water level is minus 48 inches indicated level (490 inches above vessel zero).
This trip initiates HPCI, RCIC, Alternate Rod Insertion and trips the l recirculation pumps.
The low-low-low reactor water level instrumentation is set to trip when the reactor water level is minus 160 inches indicated level (378 inches above vessel zero).
This trip closes Main Steam Line Isolation Valves, Main Steam Drain Valves and Recirc Sample Valves (Group 1), activates the remainder of the CSCS subsystem, and starts "This page is effective upon completion of the ATWS Rule ARI/RPT Modification (Modification 865)."
Amendment No. III, 141.
Unit 2 PBAPS 3.2 BASES (Cont'd.)
The recirculation pump trip limits the consequences of an anticipated transient without scram (ATWS) event.
The response of the plant to this postulated event is within the bounds of study events given in General Electric Company Topical Report, NED0-10439, dated March, 1971.
An alternate rod insertion scram limits the consequences of a Reactor Protection System failure to scram during an anticipated transient.
The ARI/RPT System is electrically diverse from the RPS logic and actuation circuitry, which significantly reduces the potential for ATWS events caused by common mode electrical failures in RPS.
The f 9URPT system is required by 10 CFR 50.62.
In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200 degrees F.
Restoration of the main steam line tunnel ventilation flow momentarily exposes the temperature sensors to high gas temperatures.
The momentary temperature increase can cause an unnecessary main steam line isolation and reactor scram.
Permission is provided to increase the temperature trip setpoint to 250 degrees F for 30 minutes during restoration of ventilation system to avoid an unnecessary plant transient.
The Emergency Aux. Power Source Degraded Voltage trip function prevents damage to safety-related equipment in the event of a sustained period of low voltage.
The voltage supply to each of the 4kV buses will be monitored by undervoltage relaying.
With a degraded voltage condition on the off-site source, the undervoltage sensing relays operate to initiate a timing sequence.
The timing sequence provides constant and inverse time voltage characteristics.
Degraded voltage protection includes:
(1) An instantaneous relay (ITE) initiated at 90% voltage which initiates a 60-second time delay relay and a 6 second time delay relay.
The 6-second time delay relay requires the presence of a safety injection signal to initiate transfer; (2) An inverse time voltage relay (CV-6) initiated at 87% voltage with a maximum 60 second delay and operates at 70% voltage in 30 seconds; and (3) An inverse time voltage relay (IAV) initiated at approximately 60% voltage and operates at 1.8 seconds at zero volts.
When the timing sequence is completed, the corresponding 4kV emergency circuit breakers are tripped and the emergency buses are transferred to the alternate source. The 60-second timing sequences were selected to prevent unnecessary transfers during motor starts and to allow the automatic tapchanger on the startup transformer to respond to the voltage condition.
The 6-second timing sequence is necessary to prevent separation of the emergency buses from the off-site source during motor starting transients, yet still oe contained within the time envelope in FSAR Table 8.5.1.
Amendment No. S2, 97,113,141
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DOCKET NO. 50-278 PEACH BOTTOM ATOMIC POWER STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.143 License No. DPR-56 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Philadelphia Electric Company, et al. (the licensee) dated June 12, 1987 as amended on February 7, 1989, complies with the standards and requirements of the Atomic Energy Ac.t of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I.
B.
The facility will operate in conformity with the application, the l
provisions of the Act, and the rules and regulations Of the Comission; C.
There is reasonable assurance (1) that the activities autht;rized by this amendment can be conducted without endangering the hea P.h and safety of the public, and (ii) that such activities will be canducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health or safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all appit.ble requirements have been satisfied.
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-56 is hereby amended to read as follows:
e 2
(2)
Technical Specifications l
The Technical Specifications contained in Appendices A and B, as rev', sed through Amendment No.143, are hereby incorporated in the license.
PEC0 shall operate the _ facility in accordance with the Technical Specifications.
3.
This license amendment is effective prior to startup in Cycle 8.
FOR THE NUCLEAR REGULATORY COMMISSION
\\
L Walter R. Butler. Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 22, 1989 l
l
ATTACHMENT TO LICENSE AMENDMENT NO.143 FACILITY OPERATING LICENSE NO. DPR-56 DOCKET NO. 50-278 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines.
Remove Insert 60 60 79 79
'88 88 89 89 93a 93a
+
Unit 3 PBAPS LIMITIWC CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS F.
Surveillance Information F.
Surveillance Information Readouts Readouts The limiting conditions for Instrumentation shall be the instrumentation that calibrated and checked as provides surveillance in-indicated in Table 4.2.F.
formation readouts are given in Table 3.2.F.
l G.
Alternate Rod Insertion and G.
Alternate Rod Insertion and Recirculation Pump Trip Recirculation Pump Trip The limiting conditions for Instrumentation shall be the instrumentation that functionally tested and initiates an Alternate Rod calibrated as indicated on Insertion scram and trips Table 4.2.G.
the reactor recirculation System logic shall be pumps to limit the conse-functionally tested as quences of a failure to scram indicated in Table 4.2.G.
during an anticipated transient are given in Table 3.2.G.
When in the RUN or STARTUP Mode, the required minimum number of instrument channels shall be operable with trip setpoints set consistent with the setting specified in Table 3.2.G.
The manual and automatic actuation logic, and actuation devices of both trip systems shall be operable when in the RUN or STARTUP Mode.
- 60 Amendment No.143 i
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PBAPS Unit 3 3.2 BASES
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In addition to reactor protection instrumentation which initiates a reactor scram, protective instrumentation has been provided which initiates action to mitigate the consequences of' accidents which are beyond the operator's ability to control, or terminates operator errors before they result in serious consequences.
This set of specifications provides the limiting conditions of operation for the primary system isolation function, initiation of the core cooling systems, control rod block and standby gas treatment systems.
The objectives of the Specifications are (i) to assure the effectiveness of the protective instrumentation when required even during periods when portions of-such systems are out-of-service for maintenance, and (ii) to prescribe the trip settings required to assure adequate performance.
When necessary, one channel may be made inoperable for brief intervals to conduct required functional tests and calibrations.
Some of the settings on the instrumentation that frnitiate or control core and containment cooling have tolerances explicitly stated where the high and low values are both critical and may have a substantial effect on safety.
The set points of other instrumentation, where only the high or low end of the setting has a direct bearing on safety, are chosen at a level away from the normal operating range to prevent inadvertent actuation of the safety system involved and exposure to abnormal situations.
Actuation of primary containment valves is initiated by protective instrumentation shown in Table 3.2.A which senses the conditions for which isolation is required.
Such instrumentation must be available whenever primary containment integrity is required.
The instrumentation whi d initiates primary system isolation is connected in a dual bus arrangement.
The low water level instrumentation set to trip at zero inches indicated level (538 inches above vessel zero) closes all isolation valves except those in Groups 1, 4 and 5.
Details of valve grouping and required closing times are given in Specification 3.7.
For valves which isolate at this level, this trip setting is adequate to prevent the core from being uncovered in the case l of a break in the largest line assuming a 60 second valve closing time.
Required closing times are less than this.
The low-low reactor water level instrumentation is set to trip when reactor water level is minus 48 inches indicated level (490 inches above vessel zero).
This trip initiates HPCI, RCIC, Alternate Rod Insertion and trips the l recirculation pumps.
The low-low-low reactor water level instrumentation is set to trip when the reactor water level is minus 160 inches indicated level (378 inches above vessel zero).
This trip closes Main Steam Line Isolation Valves, Main Steam Drain Valves and Recirc Sample Valves (Group 1), activates the remainder of the CSCS subsystem, and starts "This page is effective upon completion of the ATWS Rule ARI/RPT Modification (Modification 865)."
Amendment No. 115, 143.______ _--_____ - - _ -
~ >
b PBAPS Unit 3 3.2 BASES (Cont'd.)
The recirculation pump trip limits the consequences of an anticipated transient without scram (ATWS) event.
The response of the plant to this postulated event is within the bounds of study events given in General Electric Company Topical Report, NED0-10439, dated March, 1971.
An alternate rod insertion scram limits the consequences of a Reactor Protection System failure to scram during an anticipated transient.
The ARI/RPT System is electrically diverse from the RPS logic and actuation circuitry, which significantly reduces the potential for ATWS events caused by common mode electrical failures in RPS.
The ARI/RPT system is required by 10 CFR 50.62.
In the event of a loss of the reactor building ventilation system, radiant heating in the vicinity of the main steam lines raises the ambient temperature above 200 degrees F.
Restoration of the main steam line tunnel ventilation flow momentarily exposes the temperature sensors to high gas temperatures.
The momentary temperature increase can cause an unnecessary main steam.line isolation and reactor scram.
Permission is provided to increase the temperature trip setpoint to 250 degrees F for 30 minutes during restoration of ventilation system to avoid an unnecessary plant transient.
The Emergericy Aux. Power Source Degraded Voltage trip function prevents damage to safety related equipment in the event of a sustained period of low voltage.
The voltage supply to each of the 4kV buses will be monitored by undervoltage relaying.
With a degraded voltage condition on the off site source, the undervoltage sensing relays operate to initiate a timing sequence.
The timing sequence provides constant and inverse time voltage characteristics.
Degraded voltage protection includes:
(1) An instantaneous relay (ITE) initiated at 90% voltage which initiates a 60-second time delay relay and a 6 second time delay relay.
The 6-second time delay relay requires the presence of a safety injection signal to initiate transfer; (2) An inverse time voltage relay (CV-6) initiated at 87% voltage with a maximum 60 second delay and operates at 70% voltage in 30 seconds; and (3) An inverse time voltage relay (IAV) initiated at approximately 60% voltage and operates at 1.8 seconds at zero volts.
a When the timing sequence is completed, the corresponding 4kV emergency circuit breakers are tripped and the emergency buses are transferred to the alternate The 60-second timing sequences were selected to prevent unnecessary source.
transfers during motor starts and to allow the automatic tapchanger on the startup transformer to respond to the voltage condition.
The 6-second timing sequence is necessary to preveMt separation of the emergency buses from the off-site source during motor starting transients, yet still be contained within the time envelope in FSAR Table 8.5.1.
l Amendment No. 81, 99, 217, 143
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