ML20153A885
| ML20153A885 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 03/03/1988 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20153A888 | List: |
| References | |
| NUDOCS 8803210274 | |
| Download: ML20153A885 (25) | |
Text
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- cp oq'e, UNITED STATES
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NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555 o
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y *..4 PHILADELPHIA ELECTRIC COMPANY PUBLIC 5ERVICE ELECTRIC AND GAS COMPANY DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY POCKET N0. 50-277 PEACH ROTTOM ATOMIC POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.129 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 February 12, 1987 as supplemented on October 20, 1967, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commistion'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 (i) that the activities authorized by this amendment can be conducted without endangering the nealth 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 1
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Conmission'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 paraoraph 2.C(2) of Facility Operating License No. DPR-44 is hereby amended to read as follows:
8803210274 880303 PDR ADOCK 05000277 P
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(2)- Techn.ical Specifications The Technical Specifications contained in Appendices A~and 8, as revised through Amendment No.129, are hereby incorporated in the license.
PEC0 shall operate the facility:in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY CCt911SSION
/s/
Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 3, 1988 J
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(2)
Technical Specifications The Technical Specifications contained in Appendices A and 8, as revised through Amendment No.129, are herebv incorporated in the license.
PEC0 shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION J
Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 3, 1988 i
ATTACHMENT TO LICENSE AMENDMENT No.129 FACILITYOPERATINGLICENSENO.DPR-(4 DOCKET NO. 50-977 Replace the followino pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marcival lines.
Asterisk page is provided for document completeness.
Demove Insert 37 37 38 38 47 47*
48 48 61 61 62 62 63 63 90 90 i
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Amenilment tb. 25M,lO(MI 33 129
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.PBAPS Unit 2 3.1 BAq;S The reactor protection system automatically initiates a reactor scram to:
1.
Preserve t. c integrity of the fuc1 cladding.
2.
Preserve the integrity of the reactor coolant system.
3.
Minimize the energy which must be absorbed following a loss of coolant accident, and prevent inadvertant crit-icality.
This specification provides the limiting conditions for op-eration necessary to preserve the ability of the system to perform its intended function even during periods when in-strument channels may be out of service because of mainten-ance.
When necessary, one channel may be made.isoperable for brief intervals to conduct required functional tests and calibrations.
The reactor protectic, system is of the dual chinnel type (Reference subsection 7.2 FSAR).
The system is made up of two independent trip systems, each having two subchannels of tripping devices.
Each subchannel has an input from nt least one instrument channel which monitors a critical para-meter.
The outputs of the s'ubchannels are combined in a 1.out of 2 logic; i.e, an input signh1 on either one or both of the subchannels will cause a trip system trip.
The outputs of the trip systems are arranged so that a trip on both sys-tems is required to produce a reactor scram.
This system meets the intent o'f IEEE - 279 for Nuclear Power Plant Protection Systems.
The system has a reliability groater than that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 system.
With the exception of the Aver' age Power Range Monitor (APRM) channels, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation Valvo closure and the Turbine stop valvo closure, each subchannel has one instrument channel.
When the minimum condition for operation on 'tdur number of operable instrument channels per untripped protection trip system is met or if it cannot be met and the affected pro-taction trip system is placed in a tripped condition, the effectiveness of the protection system is preserved.
The APRM instrument channels are provided for each protec-tion trip system.
APRM's A and B operate contacts in one subchannel and APRM's C and E operate contacts in the other subchannel.
APMM's 3, D and F are arranged similarly in -
0:tober 1973
PBAPS UNIT 2 l
3.0 BASES (Cont'd) the other protection trip system.
Each protection trip system has one more APRM than is necessary to meet the minimum number l
required per channel.
This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration.
Additional IRM channels have also been provided to allow for bypassing of one such channel.
The, bases for the scraa setting for the IRM, APRM, high reactor pressure, reactor low water i
level, MSIV closure, generator load rejection, turbine stop valve closure and loss of condenser vacuum are discussed in Specification 2.1 and 2.2.
Instrumentation sensing drywell pressure is provided to detect a loss of coolant accident and initiate the core standby cooling equipment.
A high drywell pressure scram is provided at the same setting as the core standby cooling systems (CSCS; initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality.
This instrumentation is a backup to the reactor vessel water level instrumentation.
i High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity is an l
indication of leaking fuel.
A scram is initiated whenever such 1
l radiation level exceeds fif teen times normal background.
The purpose of this scram is to limit fission product release so that i
10 CFR Part 100 guidelines are not exceeded.
Discharge of 1
excessive amounts of radioactivity to the site environs is prevented by the off-gas treatment system, which provides sufficient delay time to reduce fission product release rates to well below 10 CFR 20 guidelines.
4 A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.
Ref. paragraph 7.2.3.7 FSAR.
The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.
The APRM (Eigh flux in Start-up or Refuel) system provides protection against excessive power levels and short reactor periods in the start-up and intermediate power ranges.
The IRM system provides protection against short reactor porlods in these ranges.
The control rod drive scraa system is designed so that all of the i
water which is discharged from the reactor by a scram can be i
accommodated in the discharge, piping.
The scran discharge volume accommodates in excess of 50 gallons of water and is the low point in the piping.
No credit was taken for this volume in the design of the discharge piping as concerns Amendment No. pf, 129.
i.
i
<n YAste 3.2.4 PBAPS untt 2 3
E INSYauestNTATION THAT INITIAft1,RIMARY CONTAINutui ISOLATI0se k
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PBAPS Unit 2 NOTES FOR TABLE 3.2.A 1.
Whenever Primary Containment integrity is required by Section 3.7, there shall be two operable or tripped trip systems for each function.
2.
If the first column cannot be met for one of the trip systems, that trip system shall be tripped or the appropriate action listed below shall be taken:
A.
Initiate an orderly shutdown and have the reKctor in Cold Shutdown Condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have Main Steam Lines isolated within eight hours.
C.
Isolate Reactor Water Cleanup Syrten.
D.
Isolate Shutdown Cooling.
E.
Isolate Reactor Water Cleanup Filter Domineralisers unless the following provision is satisifed.
The RWCU Filter j
Demineral,iser may be used (the isolation overridden) to route the reactor water to the main condenser or waste surge. tank, with the high tamperature trip inoperable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided the water inlet temperature is monitored once per hour and confirmed to be below 180 degrees F.
P.
Isolate reedwater Flush System 3.
Instrument setpoint corresponds to 538 inches above vessel zero.
4.
Instrument setpoint corresponds to 378 inches above vessel zero.
5.
Two required for each steam line.
6.
These signals also start 85GTS and initiate secondary containment isolation.
7.
Only required in Run Mode (interlocked with Mode Switch).
l 8.
An alarm will be tripped in the control room to alert the control room operators to an increase in the main steam line tunnel radiation level.
AmendmentNo./,M,),)d. 129 I
63-
PBAPS l
Unit 2 l
l 3.2 BASES (Cont'd) the emergency diesel generators.
These trip level settings were chosen to be high enough to prevent spurious actuation but low enough to initiate C.9CS operation and primary system isolation so that post-accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded.
For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation and primary system isolation are initiated in time to meet the above criteria.
Reference paragraph 6.5.3.1 FSAR.
The high drywell pressure instrumentation is a diverse signal for malfunctions to the water level instrumentation and in addition to initiating CSCS, it causes isolation of Group 2 and 3 isolation valves.
For the breaks discussed above, this instrumentation will generally initiate CSCS operation before the low-low-low water level instrumentation; thus the results given above are applicable here also.
See Spec. 3.7 for Isolation Valve Closure Group.
The water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of all isolation valves except Groups 4 and 5.
Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.
The primary function of the instrumentation is to detect a break in the main steam line.
For the worst case accident, main steam line break outside the drywell, a trip setting of 140% of rated steam flow in conjunction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel temperatures peak at approximately 1000 degrees F and release of radioactivity to the environs is below CFR 100 guidelines.
Reference Section 14.6.5 FSAR.
Temperature monitoring instrumentation is provided in the main steam line tunnel exhaust duct and along the steam line in the turbine building to detect leaks in these areas.
Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.
See Spec. 3.7 for Valve Group.
The setting is 200 degrees F for the main steam line tunnel detector.
For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.
High radiation monitors in the main steam line tunnel have teen l provided to detect gross fuel failute as in the control rod drop accident.
With the established setting of 15 times normal background, and main steam line isolation valve closure, fission product release is limited so that 10 CFR 100 guidelines are not exceeded for this accident.
Reference Section 14.6.2 PEAR.
Amendment No.
129 90-
gm
'o UNITED STATES
+["3v,( ",i NUCLEAR REGULATORY COMMISSION E
WASHINGTON, D. C. 20555
- %...../
PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GA5 COMPANY DELFARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-278 PEACH 530TTOM ATOMIC POWER STATION, UNIT NO. 3 AMENDMENT TO FACILITY OPERATING LICENSE, l
Amendment No. 132 License No. DPR-56 1.
The Nuclear Reaulatory Comission (the Comission) has found that:
A.
The application for amen &ent by Philadelphia Electric Company, et al. (the licensee) dated February 12, 1987 as supplemented on October 20, 1987, 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 fecility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to
'e health or safety of the public; and E.
The issuance of this amendmcnt is in accordance with 10 CFR Part 51 of the Comission'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-56 is hereby amended to read as follows:
, i (2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.132, are hereby _ incorporated in the license.
PEC0 shall operate the facility in accordance with the i
Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
/s/
Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II l
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 3, 1988 I
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. (2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.132, are hereby incorporated in the license.
PEC0 shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II
Attachment:
Changes to the Technical Specifications Date of Issuance: Parch 3,1988 l
c ATTACHMENT TO LICENSE AMENDMENT NO.132 FACILITY OPERATING LICENSE NO. OPR-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.
Asterisk page is provided for document completeness.
Remove insert 37 37 38 38 40 40 47 47*
48 48 61 61 62 6?
63 63 90 90 J
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PBAPS Unit 3 9
, NOTES POR TABLE 3.1.1 (Cont'd) 10.
The APRM downscale trip is automatically bypassed when the IRM instrumentation is operable and not high.
11.
An APRM will be considered operable if there are at least 2 LPRM inputs per level and at least 14 LPRM inputs of the normal complement.
12.
This equation will be used in the event of operation with a maximum fraction of lialting power density (MFLPD) greater thaa fraction of rated power (FRP), wheret FRP = 6. action of rated thermal power (3293 MWt).
MFLPD = maximum fraction of limiting
~
power density where the limiting power density is 13.4 KW/ft for all 8 x 8 fuel.
The ratio of FRP to MFLPD shall be set equal to 1.0 unless the actual operating value is less than the design value of 1.0, in which case the actual operating value will be used.
W=
Loop Recirculation flow in percent of design.
W is 100 for core flow of 102.5 million Ib/hr or greater.
Delta W =
the difference between two loop and single loop l
effective recirculation drive flow rate at the same core flow.
During single loop operation, the reduction in trip setting (-0.66 delta W) is accomplished by correcting the flow input of the flow biased Eigh Flux trip setting to preserve the original (two loop) relationship between APRM High Flus setpoint and recirculation drive flow or by adjusting the APRM Flux trip setting.
Delta W equals sero for two loop operation.
Trip level setting is in percent of rated power (3293 MWt).
13.
See Section 2.1.A.1.
l Amendment No. /I,M,M,J'I,M.133 40-
~~~
~
PBAPS Unit 3 3.3 assrs The reactor protection system automatically initiates a l
reactor scram to:
1.
Preserve the integrity of the fuel cladding.
2.
Preserve the integrity of the reactor coolant system.
3.
Minimize the energy which must be absorbed following a loss of coolant accident, and provent inadvertant crit-icality.
This specification provides th'e limiting conditions for op-eration necessary to preserve the ability of the system to perform its intended function even during periods when in-strument channels may be out of service because of mainten-ance.
When necessary, one channel may be made Leoproble f6r brief intervals to conduct required functional tests and calibrations.
The reactor protection system is of the dual channel type (Ref erence subsection 7.2 FSAR).
The system is made up of two independent trip systems, each having two subchannels of tripping devices.
Each subchannel has an input from at least one instrument channel which monitors a critical para-meter.
The outputs of the subchannels are combined in a 1.out of 2 logic; i.e, an input signal on either one or both of the subchannels will cause a trip system trip.
The outputs of the trip systems are arranged so that a trip on both sys-tems is required to produce a reactor scram.
This system meets the intent of IEEE - 279 for Nuclear Power i
Plant Protection Systems.
The system has a reliability groater than that of a 2 out of 3 system and somewhat less than that of a 1 out of 2 system.
With the exception of the Average Power Range Monitor (APRM) channcis, the Intermediate Range Monitor (IRM) channels, the Main Steam Isolation valvo closure and the Turbine Stop Valve closure, each subchannel has one instrument channel.
When the minimum condition for operation on'the number of i
operable instrument channels per untripped protection trip system is met or if it cannot be met and the affected pro-tection trip system is placed in a tripped condition, the l
effectiveness of the protection system is preserved.
The APRM instrument channols are provided for each protec-tion trip system.
APRM's A and E operate contacts in one subchannel and APRM's C and E operate contacts in the other subchannel.
APRM's 5, D and F are arranged similarly in l
i cutober 1973
- - -,- l
PBAPS Unit 3 l
3.0 BASES (Cont'd) l the other protection trip system.
Each protection trip system has one more APRM than is necessary to meet the minimum number required per channel.
This allows the bypassing of one APRM per protection trip system for maintenance, testing or calibration.
Additional IRM channels have also been provided to allow for bypassing of one such channel.
The. bases for the scram setting
)
for the IRM, APRM, high reactor pressure, reactor low water i
level, MSIV closure, generator load rejection, turbine stop valve closure and loss of condenser vacuum are discussed in Specification 2.1 and 2.2.
Instrumentation sensing drywell pressure is provided to detect a loss of coolant accident and initiate the core standby cooling j
equipment.
A high drywell pressure scram is provided at the same setting as the core standby cooling systems (CSCS) initiation to minimize the energy which must be accommodated during a loss of coolant accident and to prevent return to criticality.
This instrumentation is a backup to the reactor vessel water level instrumentation.
High radiation levels in the main steam line tunnel above that due to the normal nitrogen and oxygen radioactivity is an indication of leaking fuel.
A scram is initiated whenever such l radiation level exceeds fifteen times normal background.
The purpose of this scram is to limit fission product release so that 10 CPR Part 100 guidelines are not exceeded.
Discharge of excessive amounts of radioactivity to the site environs is prevented by the off-gas treatment system, which provides sufficient delay time to reduce fission product release rates to well below 10 CPR 20 guidelines.
A reactor mode switch is provided which actuates or bypasses the various scram functions appropriate to the particular plant operating status.
Ref. paragraph 7.2.3.7 FSAR.
The manual scram function is active in all modes, thus providing for a manual means of rapidly inserting control rods during all modes of reactor operation.
The APRM (High flux in Start-up or Refuel) system provides protection against excessive power levels and short reactor periods in the start-up and intermediate power ranges.
The IRM system provides protection against short reactor periods in these ranges.
4 The control rod drive scram system is designed so that all of the water which is discharged from the reactor by a scram can be accommodated in the discharge piping.
The scran discharge volume accommodates in excess of 50 gallons of water and is the low point in the piping.
No credit was taken for this volume in the 4
design of the discharge piping as concerns 4
Amerdment No. # 132 43
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NOTES POR TABLE 3.2.A 1.
Whenever Primary Containment integrity is required by Section 3.7, there shall be two operable or tripped trip systeas for each function.
2.
If the first column cannot be met for one of the trip systems, that trip system shall be tripped or the appropriate action listed below shall be taken:
A.
Initiate an orderly shutdown and have the reactor in Cold Shutdown Condition in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
B.
Initiate an orderly load reduction and have Main Steam Lines isolated within eight hours.
C.
Isolate Reactor Water Cleanup System.
D.
Isolate Shutdown Cooling.
E.
Isolate Reactor Water Cleanup Filter Domineralizers unless the following provision is satisifed.
The RWCU Filter Domineraliser may be used (the isolation overridden) to route the reactor water to the main condenser or waste surge tank,-
with the high temperature trip inoperable for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, provided the water inlet temperature is monitored once per hour and confirmed to be below 180 degrees F.
P.
Isolate Feedwater Flush System 3.
Instrument setpoint corresponds to 538 inches above vessel zero.
l 4.
Instrument setpoint corresponds to 378 inches above vessel sero.
5.
Two required for each steam line.
6.
These signals also start SBGTS and initiate secondary containment isolation.
7.
Only required in Run Mode (interlocked with Mode Sw* ?ch).
1 1
l 8.
An alarm will be tripped in the control room to alert the control room oprators to an increase in the main steam line tunnel radiation level.
1 I
Amendment Ib. pf,%,}d,)2I, 132 - - - - -.
i l
PBAPS Unit 3 3.2 BASES (Cont'd) the emergency diesel generators.
These trip level settings were chosen to be high enough to prevent spurious actuation but low enough i
to initiate CSCS operation and primary system isolation so that post-accident cooling can be accomplished and the guidelines of 10 CFR 100 will not be exceeded.
For large breaks up to the complete circumferential break of a 28-inch recirculation line and with the trip setting given above, CSCS initiation and primary system isolation are initiated in time to meet the above criteria.
Reference paragraph 6.5.3.1 FSAR.
The high drywell pressure instrumentation is a diverse signal for malfunctions to the water level instrumentation and in addition to initiating CSCS, it causes isolation of Group 2 and 3 isolation valves.
For the breaks discussed above, this instrumentation will generally initiate CSCS operation before the low-low-low water level instrumentation; thus the results given above are applicabl6 here also.
See Spec. 3.7 for Isolation Valve Closure Group.
The water level instrumentation initiates protection for the full spectrum of loss-of-coolant accidents and causes isolation of all isolation valves except Groups 4 and 5.
Venturis are provided in the main steam lines as a means of measuring steam flow and also limiting the loss of mass inventory from the vessel during a steam line break accident.
The primary function of the instrumentation is to detect a. break in the main steam line.
For the worst case accident, main steam line break outside the drywell, a trip setting of 140% of rated steam flow in conjunction with the flow limiters and main steam line valve closure, limits the mass inventory loss such that fuel is not uncovered, fuel temperatures peak at approximately 1000 degrees F and release of radioactivity to the environs is below CFR 100 guidelines.
Reference Section 14.6.5 FSAR.
Temperature monitoring instrumentation is provided in the main steam line tunnel exhaust duct and along the steam line in the turbine building to detect leaks in these areas.
Trips are provided on this instrumentation and when exceeded, cause closure of isolation valves.
See Spec. 3.7 for Valve Group.
The setting is 200 degrees F for the main steam line tunnel detector.
For large breaks, the high steam flow instrumentation is a backup to the temperature instrumentation.
High radiation monitors in the main steam line tunnel have been l provided to detect gross fuel failure as in the control rod drop accident.
With the established setting of 15 times normal background, and main steam line isolation valve closure, fission product release is limited so that 10 CPR 100 guidelines are not exceeded for this accident.
Reference section 14.6.2 FSAR.
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Amendment tb.
132.
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