ML20211A951
| ML20211A951 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 09/15/1997 |
| From: | Skay D NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20211A956 | List: |
| References | |
| NUDOCS 9709250028 | |
| Download: ML20211A951 (14) | |
Text
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t UNITED STATES
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NUCLEAR REGULATORY COMMISSION I
- %.. ;.. /j WASHINGTON. o.C. soteH001 COMMONWEALTH EDISON COMPANY DOCKET NO. 50-373 i
LASALLE COUNTY STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.120 License No. NPF-11 1
l.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment filed by the Comenwealth Edison i
Company (the licensee), dated July 1,1997, complies with the l
standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's 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 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 set forth in 10 CFR Chapter I; 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 Specifications as indicated in the enclosure to this license amendment and paragraph 2.c.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:
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Technical Snecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised j
.through Amendment No.120, and the Environmental Protection Plan contained in A>pendix B, are hereby incorporated in the license, 1
The licensee s1all opertte the facility in accordance with the i
Technical Specifications and the Environmental Protection Plan.
i L
3.
This license amendment is effective as of the date of its issuance and shall be implemented within 60 days.
i FOR THE-NUCLEAR REGULATORY COMMISSION wd.k 1
Donna N. Skay, Project nager Project Directorate !!!-2 3
Division of Reactor Projects - III/IV
^
Office of Nuclear Reactor Regulation l
Attachment:
t Changes to the Technical 4
Specifications i
Date of Issuance:
September 15,i1997 i.
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j e ATTACHMENT TO LICENSE AMENDMENT NO.120 FAtlllTY OPERATING LICENSE NO. NPF-11 QQC.ET _NO. 50-3]J Replace the following pages of the Appendix 'A' Technical Specifications with the enclosed pages. The revised pages are identified b contain a vertical line indicating the area of change. y amendment number and i
E!!0E JEiERI l-1 1-1 l
3/4 3-13 3/4 3-13 3/4 3-51 3/4 3-51 B 3/4 3-1 B 3/4 3-1 t
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1.0 DEFINITIONS Thefollowingtermsaredefinedsothatuniforminter$ncapitalizedtypeandretation of these fications may be achieved. The defined terms appear shall be applicable throughout these Technical Specifications.
All10N 1.1 ACV10N shall be that part of a Specification which prescribes remedial measures required under designated conditions.
AVERAGE PLANAR EXPOSURE 1.2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
AVERAGE PLANAR LINEAR HEAT GENERATION RATE 1.3 The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) shall be ap)11-cable to a specific planar height and is equal to the sum of the LliEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
CHANNEL CAllBRATION 1.4 A CHANNEL CAllBRATION shall be the adjustment as necessary of the channeloutputsuchthatitrespondswithint$enecessaryra,ngeand accuracy to known values of the parameter that the channel monitors.
The CHANNEL CAllBRATION shall encompass the entire channel including the CHANNEL FUNCTIONAL TEST. display, and trip functions, hannels with required sensor, alarm and shall include the Calibration of instrument c resistance temperature detector of an inplace qualitative assessm(RTD) or thermocouple sensors may consist ent of sensor behavior and normal calibration of the remaining adjustable devices in the. channel.
The CHANNEL CAllBRATION may be performed by means of any series of sequential, overlapping or total channel steps so that the entire channel is calibrated.
CHANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other measuring the s/or status derived from independent instrument channels indications and ame parameter.
CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:
a.
Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trips.
b.
Bistable channels - the injection of a simulated signal into the sensor to verify OPERABILITY including alarm and/or trip functions.
The CHANNEL FUNCTIONAL TEST may be performed by any series of sequential, overlapping, or total channel steps such that the entire channel is tested.
LA SALLE UNIT 1 1-1 Amendment No. 120
TABLE 3.3.2-1 (Continued)
VALVE GROUPS MINIMist OPERA 8tE APPLICA8tE OPERATED BY.
CHANNELS PER OPERATIO!!AL TRIP FUNCTION SIGNAL TRIP SYSTEM (b)
_C0lWITION.
ACTION 5.
RHR SYSTEM STEAM CONDENSING MODE ISOULTION a.
RHR Equipment Area A Temperature - High 8
1/RHR area I, 2, 3 22 b.
RHR Area Temperature -
High 8
1/RfR area 1, 2, 3 22 c.
RHR Heat Exchanger Steam Supply Flow - High 8
1 1, 2, 3 22 6.
RHR SYSTEM SHUTDOWN COOLING MODE ISOULTION a.
Reactor Vessel Water Level - Low, level 3 6
2 1,2,3 25 b.
Reactor Vessel (RHR Cut-in Permissive)
Pressure - High 6
1 1,2,3 25 c.
RHR Pump Suction Flow - High 6
1 1, 2, 3 25 d.
RHR Area Temperature -
High 6
1/RfR ama 1, 2, 3 25 e.
RHR Equipment Area AT - High 6
1/RHR area 1, 2, 3 25 B.
MANUAL INITIATION 1.
Inboard Valves 1,2,5,6,7 1/ group 1, 2, 3 26 4,3g5,6 1/ group 1, 2, 3 26 1
2.
Outboard Valves I g 3.
Inboard Valves 1/ group 1, 2, 3 and **,#
26 4.
Outboard Valves 4 (* "
1/ group 1, 2, 3 and **,#
26 5.
Inboard Valves 3,8,9
'1/ valve I, 2, 3 26 6.
Outboard Valves 333, 9 1/ valve 1, 2, 3 26 8
7.
Outboard Valve 8
I/ group I, 2, 3 26 LA SALLE - UNIT 1 3/4 3 Amendment No.120
- = _______
TABLE 3.3.6-1 CONTROL ROD WITIERAMAL BLOCK INSTEMENTATION
^
MINIMUM OPERA 8LE APPLICABLE CHANNELS PER OPERATIONAL TRIP FUNCTION TRIP FUNCTION COWITIONS ACTION 1.
ROD BLOCK MONITORfal a.
Upscale 2
1*
60
'b.
Inoperative 2
1*
60 c.
Downscale-2 1*
60
'2.
APB5 a.
Flow Blased Simulated Thermal Power-Upscale 4
1 61 b.
Inoperative 4
1, 2, 5 61 c.
Downscale-4 1
61 d.
Neutron Flux-High 4
2, 5 61 3.
SOURCE. RANGE MONITORS a.
Detector not full in(b) 3.
2 61 2
5 61 b.
Upscale (c) 3 2
61 2
5 61 c.
' Inoperative (c) 3 2
61 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 l
b.
Upscale 6
2, 5 61 c.
Inoperative 6
2, 5 61 d.
Downscale(e) 6 2, S 61 5.
SCRAM DISCHARGE VOLUE a.
Water Level-High 2
1, 2, 5**
62 b.
Scram Discharge Volume Switch in Bypass 1
5**
62 6.
RECIRCULATION FLOW UNIT a.
Upscale 2
1 62 b.
Inoperative 2
1 62 c.
Comparator 2
1 62 LA SALLE - UNIT 1 3/4 3-51 Amendment No.120
o 3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM !NSTRUMENMllDN The reactor protection system automatically initiates a reactor scram tot a.
Preserve the integrity of the fuel cladding.
b.
Preserve the integrity of the reactor coolant system, Minimize the energy which must be adsorbed following a loss-of-c.
coolant accident, and d.
Prevent inadvertent criticality.
This specification provides the limiting conditions for operation l
i necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.
The reactor protection system is made up of two independent tris systems.
There are usually four channels to monitor each parameter with two ciannels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279, 1971, for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System", March 1988, and MDE-83-0485 Revision 3. " Technical Specification Improvement Analysis for the Reactor Protection System for LaSalle County Station, Units 1 and 2", April 1991. The bases for the trip settings of the PRS are discussed in the bases for Specification 2.2.1.
When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintain's RPS trip capability.
The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the accident analysis. The RPS RESPONSE TIME acceptance criteria are included in plant Surveillance procedures. Only those functions with times assumed in the accident analysis are required to be response time tested.
As stated in Note
- of Table 3.3.1-2, Neutron detectors are exempt from response time testing.. In addition, for Functional Units 3 and 4, per Note if, l
the associated sensors are not required to be response time tested.
For these LA SALLE - UNIT 1 B 3/4 3-1 Amendment No.
120
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UNITED STATES s
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NUCLEAR REGULATORY COMMISSION
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WASHINGTON, D.C. 306tH001 COMMONWEALTH EDISON COMPANY DOCKET NO. 50_.J11 LASALLE COUNTY STATION. UNIT 2 AMENDMENT TO FACI 'TY OPERATIN'i l_ICENSE Amendment No.105 License No. NPF-18 1.
The Nuclear Regulatory Cemission (the Comission) has found that:
A.
The application for amendment filed by the Comonwealth Edison Company (the licensee), dated July 1,1997, complies with the stanc'ards and requiremerts of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's 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 regulations of the Comission; C.
There is re:sonable assurance: (i) that the activities authorized by this amecdment can be conducted without endangering the health end safety of the public, and (ii) that such activities will be
[
conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D.
The issuance of this amendment ':111 not be ini:aical to the ccmon 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 Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amended to read as follows:
.m
i-2-
(2)
Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.105, and the Environmental Protection Plan contained in A)pendix B, are hereby incorporated in the license.
The licensee s1all 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 shall be implemented within 60 days.
FOR THE NUCLEAR REGULATORY C0tmISSION l0Ab Donna M. Skay, Project Wnager l
Project Directorata III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
September 15, 1997 a
ATTACHMENT TO LICENSE AMENDMENT NO. 105 FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO. 50-374 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the area of change.
REMOVE 1EEllI l-1 1-1 l
3/4 3-13 3/4 3-13 3/4 3-51 3/4 3-51 B 3/4 3-1 B 3/4 3-1
1.0 DEFINITIONS The following terms aro defined so that uniform inter $n capitalized type and retation of these speci-fications may be achieved. The defined terms appear l
t shall be applicable throughout these Technical Specifications.
ACTION 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.
AVERAGE PLANAR EXPOSURE 1.2 The AVERAGE PLANAR EXPOSURE shall be applicable to a specific planar height and is equal to the sum of the exposure of all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
AVERAGE PLANAR LINEAR HEAT GENERATION RATE l
1.3 The AVERAGE PLANAR LINEAR HEAT GENEPATION RATE (APLHGR) shall be appli-i cable to a specific planar height and is equal to the sum of the LINEAR HEAT GENERATION RATES for all the fuel rods in the specified bundle at the specified height divided by the number of fuel rods in the fuel bundle.
CHANNEL CALIBR6110H l
1.4 A CHANNEL CALIBRATION shall be the adjustment as necessary of the l
c%nnel output such that it responds within tbe necessary ra,nge and accuracy to known values of the parameter that the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel,d shall include the including the required sensor, alarm, display, and trip functions, an CHANNEL FUNCTIONAL TEST.
Calibration of instrument channels with resistance temperature detector of an inplace qualitative assessm(RTD) or thermocouple sensors may consist ent of sensor behavior and normal calibration cf the remaining adjustable devices in the channel. The CHANNEL CAllBRATION may be performed by means of a.ny series of sequential, over1&pping, or total channel steps so that the entire channel is calibrated.
QlANNEL CHECK 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation. This determination shall include, where possible, comparison of the channel indication and/or status with other indications and measuring the s/or status derived from independent instrument channels ame parameter.
CHANNEL FUNCTIONAL TEST 1.6 A CHANNEL FUNCTIONAL TEST shall be:
a.
Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions and channel failure trius.
b.
Gistable ch,6 s - the injection of a simulated signal into the sensor to venty OPERABILITY including alarm and/or trip functions.
The CHANNEL FUNCTIONAL TEST may be parformed by any series of sequential,d.
overlapping, or total channel steps such that the entire channel is teste LA SALLE - UNIT 2 1-1 Amendment No. 105
TABLE 3.3.2-1 (Continued)
ISOLATION ACTUATION INSTRUMENTATION VALVE GROUPS MINIMUM OPERABLE APPLICABLE OPERATED BY CHANNELS PER OPERATIONAL TRIP FUNCTION SIGNAL TRIP SYSTEM (b)
CONDITION ACTION l
S.
RHR SYSTEM STEAM CONDENSING MODE ISOLATION a.
RHR Equipment Area A Temperature - High 8
1/RHR area 1, 2, 3 22 j
b.
RHR Area Temperature -
High 8
1/RHR area 1, 2, 3 22 r
l c.
RHR Heat Exchanger Steam Supply Flow - High 8
1 1, 2, 3 22 6.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION a.
Reactor Vessel Water Level - Low, Level 3 6
2 1, 2, 3 25 b.
Reactor Vessel (RHR Cut-in Permissive)
Pressure - High 6
1 1, 2, 3 25 c.
RHR Piemp Suction Flow - High 6 1
1, 2, 3 25 d.
RHR Area Temperature -
High 6
1/RHR area 1, 2, 3 25 e.
RHR Equipment Area AT - High 6 1/RHR area 1, 2, 3 25 B.
MANUAL INITIATION l
1.
Inboard Valves 1,2,5,6,7 1/ group 1, 2, 3 26 4*){,,5,6 1 group 1, 2, 3 26 1
I 2.
Outboard Valves g
I group 1, 2, 3 and **,#
26 3.
Inboard Valves 4.
Outboard Valves 4(*"')
I group 1, 2, 3 and **,#
26 5.
Inboard Valves 3,8,9 1 valve 1, 2, 3 26 33, 9 1 valve 1, 2, 3 26 6.
Outboard Valves 3
8 7.
Outboard Valve 8
1/ group 1, 2, 3 26 LA SALLE - UNIT 2 3/4 3-13 Amendment No. 105
TA BLE 3.3.6-1 CONTROL ROD WITIENAWAL BLOCK INSTRtMENTATION MINIMUM OPERABLE APPLICABLE CHANNELS PER OPERATIONAL.
TRIP FUNLTION TRIP FUNCTION CONDITIONS ACTION l
1.
R00 BLOCK MONITORfal a.
Upscale 2
1*'
60 b.
Inoperative 2-1*
60 -
c.
Downscale 2
1*
60
-1 E* O I
a.
Flow Blased Simulated Thermal Power-Upscale 4
1 61 b.
Inoperative 4
1, 2, 5 61 c.
Downscale 4
1 61 j
d.
Neutron Flux-High 4
2, 5 61 l
3.
SOURCE liANGE MONITORS i
a.
Detector not full in(b) 3 2
61 2
5 61 b.
Upscale (c) 3 2
61 2
5 61 c.
Inoperative (c) 3 2
61 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 l
b.
Upscale 6
2, 5 61 c.
Inoperative-6 2, 5 61 d.
Downscale(e) 6 2, 5 61 5.
a.
Water level-High 2
1, 2, 5**
62 b.
Scram Discharge Volume Switch in Bypass 1
5**
62 6.
RECIRCULATION FLOW UNIT a.
Upscale 2
1 62 b.
Inoperative 2
1 62 c.
Comparator 2
1 62-LA SALLE - UNIT 2' 3/4 3-51 Amendment No.105
O i
3/4.3 INSTRUMENTATION BASES 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:
a.
Preserve the integrity of the fuel cladding.
b.
Preserve the integrity of the reactor coolant system.
l c.
Minimize the energy which must be adsorbed following a loss-of-l coolant accident, and i
d.
Prevent inadvertent criticality.
l This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.
The reactor protection system is made up of two independent trip systems.
There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279, 1971, for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined In accordance with NEDC-30851P-A, " Technical Specification Improvement Analyses for BWR Reactar Protection System", March 1988, and MDE-83-0485 Revision 3, " Technical Specification Improvement' Analysis for the Reactor Protection System for LaSalle County Station, Units 1 and 2", April 1991. The bases for the trip settings of the RPS are discussed in the bases for Specification 2.2.1.
When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains RPS trip capability.
The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the accident analysis. The RPS RESPONSE TIME acceptance criteria are included in plant Surveillance procedures. Only those functions with times assumed in the accident analysis are required to be response time tested.
As stated in Note
- of Table 3.3.1-2, Neutron detectors are exempt from response time testing.
In addition, for Functional Units 3 and 4, per Note if, I
the associated sensors are not required to be response time tested.
For these LA SALLE - UNIT 2 B 3/4 3-1 Amendment No. 105
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