ML20113D760
| ML20113D760 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 06/26/1996 |
| From: | Assa R NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20113C223 | List: |
| References | |
| NUDOCS 9607030310 | |
| Download: ML20113D760 (8) | |
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.A UNITED STATES f-
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NUCLEAR REGULATORY COMMISSION s
WASHINGTON, D.C. 3068dH1001 COMONWEALTH EDISON COMPANY DOCKET 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:
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 (i.he Act) and the Commission'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 ol' 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 arendment is in accordance with 10 CFR Part 51 of the Commisston'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:
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9607030310 960626 ADOCK0500g4 DR
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(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 COMISSION 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 Date of Issuance:
June 26, 1996 i
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION 5*
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COMONWEALTH 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 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 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 (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 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:
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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 1
l 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
'h Ramin R. Assa, Project Manager l
Project Directorate III-2 l
Division of Reactor Projects - III/IV l
Office of Nuclear Reactor Regulation l
Attachment:
Changes to the Technical Specifications Date of Issuance: June 26, 1996 l
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- s 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
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3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION i
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LIMITING CONDITION FOR OPERATION 1
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.
J ACTION:
i As shown in Table 3.3-1.
SURVEILLANCE REQUIREMENTS 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.
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.
BRAIDWOOD - UNITS 1 & 2 3/4 3-1 AMENDMENT N0. 76 l
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INSTRUMENTATION i
SURVEILLANCE RE0VIREMENTS j
l 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 l
Table 4.3-2.
4.3.2.2 The ENGINEERED SAFETY FEATURES RESPONSE TIME of each ESFAS function l
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 i
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 j
" Total No. of Channels" Column of Table 3.3-3.
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i BRAIDWOOD - UNITS 1 & 2 3/4 3-14 AMENDMENT NO. 76 l
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INSTRUMENTATION l,
BASES REACTOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued)
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 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 l
serious problems and should warrant further investigation.
The verification of response time at the specified frequencies provides
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assurance that the reactor trip and the engineered safety features actuation l
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 l
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 I
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
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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) l i
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 l
and automatic valves position, and (11) essential service water pumps start and l
automatic valves position.
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BRAIDWOOD - UNITS 1 & 2 B 3/4 3-2 AMENDMENT N0. 76 l
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