ML20113C220
| ML20113C220 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 06/26/1996 |
| From: | Dick G Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20113C223 | List: |
| References | |
| NPF-37-A-084, NPF-66-A-084 NUDOCS 9607010066 | |
| Download: ML20113C220 (9) | |
Text
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UNITED STATES g
,j NUCLEAR REGULATORY COMMISSION t
WASHINGTON, D.C. 20066 0001
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49.....,o COMONWEALTH EDISON COMPANY DOCKET NO. STN 50-454 BYRON STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 84 License No. NPF-37 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated September 16, 1994, as supplemented by letter i
dated January 31, 1996, 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 (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-37 is hereby amended to read as follows:
9607010066 960626 PDR ADOCK 05000454 P
(2)
Technical Soecifications The Technical Specifications contained in Appendix A as revised 1
through Amendment No. 84 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and is to be implemented within 30 days.
i FOR THE NUCLEAR REGULATORY COMMISSION Lit Y
Georg F. Dick, Senior Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 26, 1996
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4 UNITED STATES
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NUCLEAR REGULATORY COMMISSION j
WASHINGTON, D.C. 20eadMioo1 k
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es COMMONWEALTH EDISON COMPANY DOCKET NO. STN 50-455 BYRON STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 84 License No. NPF-66 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated September 16, 1994, as supplemented by letter dated January 31, 1996, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter 1; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-66 is hereby amended to read as follows:
1 (2)
Technical Soecifications The Technical Specifications contained in Appendix A (NUREG-1113),
as revised through Amendment No. 84 and revised by Attachment 2 l
to NPF-66, and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-37, J
dated February 14, 1985, are hereby incorporated into this license. Attachment 2 contains a revision to Appendix A which is hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and 4
the Environmental Protection Plan.
l 3.
This license amendment is effective as of the date of its issuance and is to be implemented within 30 days.
4 FOR THE NUCLEAR REGULATORY COMMISSION t
RC1 Geor F. Dick, Senior Project Manager Project Directorate III-2 l
Division of Reactor Projects - III/IV j
Office of Nuclear P,eactor Regulation
Attachment:
Changes to the Technical l
Specifications l
Date of Issuance: June 26, 1996
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1
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ATTACHMENT TO LICENSE AMENDMENT NOS. 84 AND 84 FACILITY OPERATING LICENSE NOS. NPF-37 AND NPF-66 DOCKET N05. STN 50-454 AND STN 50-455 Revise the Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages.
The revised pages are l
identified by the captioned amendment number and contain marginal lines l
indicating the area of change.
Remove Paaes Insert Paaes 3/4 3-1 3/4 3-1 3/4 3-14 3/4 3-14 8 3/4 3-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2 l
l l
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l l
l 4
1 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.
APPLICABILITY: As shown in Table 3.3-1.
ACTION:
As shown in Table 3.3-1.
SURVEILLANCE RE0VIREMENTS 4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and i
l the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.
4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be verified to be within its limit at least once per 18 months.
Each verification shall include at least one train such that both trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.
BYRON - UNITS 1 & 2 3/4 3-1 AMENDMENT N0 84
INSTRUMENTATION 3
SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in Table 4.3-2.
4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within the limit at least once per 18 months.
Each verification shall include at least one train such that both trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the " Total No. of Channels" Column of Table 3.3-3.
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i BYRON - UNITS 1 & 2 3/4 3-14 AMENDMENT NO. 24
3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and the Engineered Safety Features Actuation System instrumentation and interlocks ensures that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation, and 3) sufficient system functions capability is available from diverse parameters.
The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions.
The integrated operation of each of these systems is consistent with the assumptions used in the safety analyses.
The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability.
Specified surveillance intervals and surveillance and maintenance outage times have been determined in i
accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of 3ervice Times for the Reactor Protection Instrumentation System," and supplements to that report.
Surveillance intervals and out of service times were determined based on maintaining an appropriate level of reliability of tne Reactor Protection System and Engineered Safety Features instrumentation.
The Engineered Safety Features Actuation System Instrumentation Trip Setpoints specified in Table 3.3-4 are the nominal values at which the bistables are set for each functional unit. A Setpoint is considered to be adjusted consistent with the nominal value when the "as measured" Setpoint is within the band allowed for calibration accuracy.
To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which Setpoints can be measured and calibrated, Allowable Values for the Setpoints have been specified in Table 3.3-4.
Operation with Setpoints less conservative than the Trip Setpoint but within the Allowable Value is acceptable since an allowance has been made in the safety analysis to accommodate this error.
The methodology to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties.
Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.
Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not BYRON - UNITS 1 & 2 B 3/4 3-1 AMENDMENT N0. 84
i l.
INSTRUMENTATION BASES l
i REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM 1
l INSTRUMENTATION (Continued)
J met its allowance.
Being that there is a small statisitical chance that this will happen, an infrequent excessive drift is expected.
Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
The verification of response time at the specified frequencies provides assurance that the reactor trip and the engineered safety features actuation associated with each channel is completed within the time limit assumed in the safety analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable.
Response time may be verified by actual tests in any series of sequential, overlapping or total channel measurements, or by summation of allocated sensor response times with actual tests on the remainder of the channel in any series of sequential or overlapping measurements. Allocations for sensor response times may be obtained from:
(1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g., vendor) test measurements, or (3) utilizing vendor engineering specifications. WCAP-13632-P-A, Revision 2, " Elimination of Pressure Sensor Response Time Testing Requirements," provides the basis and methodology for using allocated sensor response times in the overall verification of the Technical Specifications channel response time. Ti.e allocations for sensor response times must be verified prior to placir.g the sensor in operational service and re-verified following maintenance that may adversely affect response time.
In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value.
One example where time response could be affected is replacing the sensing assembly of a transmitter.
The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being
)
exceeded.
If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once l
the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) feedwater isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, and (11) essential service water pumps start and automatic valves position.
BYRON - UNITS 1 & 2 B 3/4 3-2 AMENDMENT N0. 84
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/t UNITED STATES i3
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,g NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2008H001
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COPfiONWEALTH EDISON COMPANY D_0CKET NO. STN 50-456 BRAIDWOOD STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 76 License No. NPF-72 1.
The Nuclear Regulatory Comission (the Comission) has found that:
j A.
The application for amendment by Commonwealth Edison,ompany (the r
licensee) dated September 16, 1994, as supplemented by letter dated January 31, 1996, 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 Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with t'ne Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-72 is hereby amended to read as follows:
=
'A
],
(2)
Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 76 and the Environmental Protection Plan contained in Appendix B, both of which were attached to License No. NPF-72, dated July 2,1987, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and is to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION Ramin R. Assa, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV 1
Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: June 26, 1996
e'
-4 UNITED STATES l
l
,j NUCLEAR REGULATORY COMMISSION l
o WASHINGTON, D.C. 20066 0001 COMiONWEALTH EDISON COMPANY DOCKET NO. STN 50-457 BRAIDWOOD STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 76 License No. NPF-77 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Comonwealth Edison Company (the licensee) dated September 16, 1994, as supplemented by letter dated January 31, 1996, 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 1; B.
The facility 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 (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-77 is hereby amended to read as follows:
1 l
(2)
Technical Soecifications The Technical Specifications contained in Appendix A as revised through Amendment No. 76 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are hereby incorporated into this license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance and is to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION f'h Ramin R. Assa, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications 4
Date of Issuance: June 26, 1996 i
ATTACHMENT TO LICENSE AMENDMENT NOS. 7s AND 76 FACILITY OPERATING LICENSE NOS. NPF-72 AND NPF-77 DOCKET NOS. STN 50-456 AND STN 50-457 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
Remove Paaes Insert Paaes 3/4 3-1 3/4 3-1 3/4 3-14 3/4 3-14 B 3/4 3-2 B 3/4 3-2
)
3/4.3 INSTRUMENTATIM j
3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION l
1 LIMITING CONDITION FOR OPERATION e
3.3.1 As a minimum, the Reactor Trip System instrumentation channels and interlocks of Table 3.3-1 shall be OPERABLE.
APPLICABILITY: As shown in Table 3.3-1.
l ACTION:
As shown in Table 3.3-1.
i SURVEILLANCE REQUIREMENTS i
4.3.1.1 Each Reactor Trip System instrumentation channel and interlock and the automatic trip logic shall be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.
2 l
4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function i
shall be verified to be within its limit at least once per 18 months.
Each verification shall include at least one train such that both trains are verified at least once per 36 months and one channel per function such that i
all channels are verified at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the " Total No. of Channels" column of Table 3.3-1.
j i
4 BRAIDWOOD - UNITS 1 & 2 3/4 3-1 AMENDMENT NO. 76
INSTRUMENTATION SURVEILLANCE REQUIREMENTS 4.3.2.1 Each ESFAS instrumentation channel and interlock and the automatic i
actuation logic and relays shall be demonstrated OPERABLE by the performance of the ESFAS Instrumentation Surveillance Requirements specified in i
Table 4.3-2.
4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function shall be verified to be within the limit at least once per 18 months. Each verification shall include at least one train such that both trains are verified at least once per 36 months and one channel per function such that all channels are verified at least once per N times 18 months where N is the total number of redundant channels in a specific ESFAS function as shown in the
" Total No. of Channels" Column of Table 3.1-3.
i BRAIDWOOD - UNITS 1 & 2 3/4 3-14 AMENDNENT NO. 76
INSTRUMENTATION BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
The methodologi to derive the Trip Setpoints is based upon combining all of the uncertainties in the channels.
Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and rack instrumentation utilized in these channels are expected to be capable of operating within the allowances of these encertainty magnitudes. Rack drift in excess of the Allowable Value exhibits the behavior that the rack has not met its allowance.
Being that thare is a small statisitical chance that this will happen, an infrequent excessive drift is expected.
Rack or sensor drift, in excess of the allowance that is more than occasional, may be indicative of more serious problems and should warrant further investigation.
The verification of response time at the specified frequencies provides assurance that the reactor trip and the engineered safety features actuation associated with each channel is completed within the time limit assumed in the safety analyses.
No credit was taken in the analyses for those channels with response times indicated as not applicable. Response time may be verified by actual tests in any series of sequential, overlapping or total channel measurements, or by summation of allocated sensor response times with actual tests on the remainder of the channel in any series of sequential or overlapping measurements. Allocations for sensor response times may be obtained from:
(1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g., vendor) test measurements, or (3) utilizing vendor engineering specifications.
WCAP-13632-P-A, Revision 2, " Elimination of Pressure Sensor Response Time Testing Requirements," provides the basis and methodology for using allocated sensor response times in the overall verification of the Technical Specifications channel response time. The allocations for sensor response times must be verified prior to placing the sensor in operational service and re-verified following maintenance that may adversely affect response time.
In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. One example where time response could be affected is replacing the sensing assembly of a transmitter.
The Engineered Safety Features Actuation System senses selected plant parameters and determines whether or not predetermined limits are being exceeded.
If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is e.ompleted, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident: (1) Safety Injection pumps start and automatic valves position, (2) Reactor trip, (3) feedwater isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, (10) containment cooling fans start and automatic valves position, and (11) essential service water pumps start and automatic valves position.
BRAIDWOOD - UNITS 1 & 2 8 3/4 3-2 AMENDMENT N0. 76
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