ML20149H765
| ML20149H765 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 02/10/1988 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | Commonwealth Edison Co |
| Shared Package | |
| ML20149H767 | List: |
| References | |
| NPF-11-A-053, NPF-18-A-035 NUDOCS 8802220093 | |
| Download: ML20149H765 (12) | |
Text
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[ M EIG 'o UNITED STATES NUCLEAR REGULATORY COMMISSION e
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_ COMMONWEALTH EDISON COMPANY DOCKET NO. 50-373 LASAllE COUNTY STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 53 License No. NPF-11 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The applications for amendment filed by the Conconwealth Edison Company (the licensee), dated September 4,1987 and supplemented December 4,1987, complies with the standards and requirerents of theAtomicEnergyActof1954,asamended(theAct),andthe 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: (1)thattheactivitiesauthorizedby this amendrent 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 set forth in 10 CFR Chapter I; D.
The issuance of this amendrent will not be inimica'i to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendaent 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 enclosure to this license amendment and para-graph 2.C.(2) of the Facility Operating License No. NPF-11 is hereby amended to read as follows:
(2) Technical Specifications anc Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 53, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifi-cations and the Environmental Protection Plan.
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This amendment is effective 45 days after the date of issuance, FOR THE NUCLEAR REGULATORY COMMISSION Daniel R. Huller Director Project Directorate III-2 Division of Reactor Projects - III, IV, Y and Special Projects
Enclosure:
Changes to the Technical Specifications Date of Issuance: February 10, 1988 4
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i ENCLOSURE TO LICENSE AMENDMENT NO. 53 FACILITY 0PERATING _ LICENSE NO, NPF-11 DOCKET NO. 50-373 Replace the following page of the Appendix "A" Technical Specifications with the enclosed page. The revised page is identified by Amendment number and contains a vertical line indicating the area of change.
REMOVE INSERT 3/4 3-8 3/4 3-8 3/4 3-73 3/4 3-73 B 3/4 3-5 B 3/4 3-5 l
TABLE 4.3.1.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5
CHANNEL OPERATIONAL vi CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH f
FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED 7
8.
Scram Discharge Volume Water Level - High NA M
R 1, 2, S c
9.
Turbine Stop Valve - Closure NA M
R 1
10.
Turbine Control Valve Fast Closure Valve Trip System Oil Pressure - Low NA M
R*
1 11.
Reactor Mode Switch Shutdown Position NA R
NA 1,2,3,4,S 12.
Manual Scram NA M
NA 1,2,3,4,5 13.
Control Rod Drive a.
Charging Water Header w
Pressure - Low NA M
R 2, 5 D
b.
Delay Timer NA M
R 2, 5 (a) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) The IRM, and SRM channels shall be determined to overlap for at least 1/2 decades during each startup and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.
(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power levels calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER. The APRM Gain Adjustment Factor (GAF) for any channel shall be equal to the power value deter-mined by the heat balance divided by the APRM reading for that channel.
Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, adjust any APRM channel with a GAF > 1.02.
In addition, adjust any APRM channel within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, (1) if power is greater than or equal to 90% of RATED THERMAL POWER and the APRM channel GAF is
< 0.98, or (2) if power is less than 90% of RATED THERMAL POWER and the APRM reading exceeds the power E
value_ determined by the heat balance by more than 10% of RATED THERMAL POWER.
Until any required APRM adjustment has been accomplished, notification shall be posted on the reactor control panei.
5 (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.
=
P (f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH).
(g) Measure and compare core flow to rated core flow.
ui (h) This calibration shall consist of verifying the 6 1 1 second simulated thermal power time constant.
- Ihe specified 18-month interval may be waived for Cycle 1 provided the surveillance is performed during Refuel 1, which is to commence no later than October 27, 1985.
INSTRUMENTATION TRAVERSING IN-CORE PROBE SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.7.
The traversing in-core probe (TIP) system shall be OPERABLE with:
Movable detectors, drives and readout equipment to map the core in a.
the required measurement locations and b.
Indexing equipment to allow all required detectors to be calibrated in a common location.
APPLICABILITY: When the traversing in-core probe is used for:
- a.
Recalibration of the LPRM detectors, and
- b.
Monitoring the APLHGR, LHGR, MCPR, or MFLPD.
ACTION:
With one or more TIP measurement locations inoperable, required measure-a.
ments may be performed as described in 1 and 2 below, provided the reactor core is operating in an octant symmetric control rod pattern, and the total core TIP uncertainty for De present cycle has been measured to be less than 8.7 percent.
1.
TIP data for an inoperable measurement location may be replaced by data obtained from that string's redundant (symmetric) counterpart if the substitute TIP data was obtained from an operable measurement location.
2.
TIP data for an inoperable measurement location may be replaced by data obtained from a 3-dimensional BWR core simulator code normalized with available operating measurements, provided the total number of simulated channels (measurement locations) does not exceed:
a)
All channels of a single TIP machine, or b)
A total of five channels if more than one TIP machine is
- involved, b.
Otherwise, with the TIP system inoperable, suspend use of the system for the above applicable monitoring or calibration functions.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
c.
SURVEILtANCE REQUIREMENTS 4.3.7.7 The traversing in-core probe system shall be demonstrated OPERABLE by normalizing each of the above required detector outputs within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to use for the above applicable monitoring or calibration functions.
"Only the detector (s) in the required measurement location (s) are required to be OPERABLE.
LA SALLE - UNIT 1 3/4 3-73 Amendment No. 53
INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Continued) 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION i
The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.
This capability is con-sistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "THI-2 Lessons.
Learned Task Force Status Report and Short-Term Recommendations."
3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown.
At these power levels, reactivity additions should not be made without this flux level information available to the operator. When the inter-mediate range monitors are on scale adequate information is available without the SRMs and they can be retracted.
3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe (TIP) system with the l
specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core.
The specification allows use of substituted TIP data from symmetric channels if the control rod pattern is symmetric since the TIP data is adjusted by the plant computer to remove machine dependent and power level dependent bias.
The source of data for the substitution may also be a 3-dimensional BWR core simulator calculated data set which is normalized to available real data.
Since uncertainty could be introduced by the simulation and normalization process, an evaluation of the specific control rod pattern and core operating state must be performed to ensure that adequate margin to core operating limits is maintained.
3/4.3.7.8 AMMONIA DETECTION SYSTEM The OPERABILITY of the ammonia detection system ensures that an accidental ammonia release will he detected promptly and the necessary protective actions will be automatically initiated to provide protection for control room person-nel.
Upon detection of a high concentration of ammonia, the control room emergency ventilation system will automatically be placed in the recirculation mode of operation to provide the required protection. The detection systems required by this specification are consistent with the recommendations of Regulatory Guide 1.78 "Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release."
3/4.3.7.9 FIRE DETECTION INSTRUMENTATION OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.
This capa-bility is required in order to detect and locate fires in their early stages.
Prompt detection of fires will reduce the potential for damage to safety-related i
LA SALLE - UNIT 1 B 3/4 3-5 Amendment No. 53 f
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UNITED STATES 8
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COMMONWEALTH EDISON COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION. UNIT 2 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No. 35 License No. NPF-18 1.
The Nuclear Regulatory Comission (the Commission) has found that:
i A.
The application for amendment filed by the Comonwealth Edison Company (the licensee), dated September 4,1987 and supplemented December 4, 1987, complies with the 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 Comisstun'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 l
E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of l
the Comission's regulations and all applicable requirements have been satisfied.
j 2.
Accordingly, the license is amended by changes to the Technical Specifi-cations as indicated in the enclosure to this license amendment and para-graph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amended to read as follows:
(2) Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendnent No. 35, ano the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specift-cations and the Environmental Protection Plan.
5
O 2-3.
This amendment is effective 45 days af ter the date of issuance.
FOR THE NUCLEAR REGULATORY C0tNISSION Daniel R. Muller, Director Project Directorate III-2 Division of Rear. tor Projects - III, IV, Y and Special Projects
Enclosure:
Changes to the Technical Specifications Date of Issuance: February 10, 1988
ENCLOSURE TO LICENSE AMENDMENT NO. 35 FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO. 50-374 Replace the following page of the Appendix "A" Technical Specifications with the enclosed page.
The revised page is identified by Arendsent number and contains a vertical line indicating the area of change.
REMOVE INSERT 3/4 3-8 3/4 3-8 3/4 3-73 3/4 3-73 8 3/4 3-5 B 3/4 3-5 T.
9
TABLE 4.3.1.1-1 (Continued)
[
REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL OPERATIONAL E
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED e
E 8.
Scram Discharge Volume Water G
Level - High NA M
R 1, 2, 5 9.
Turbine Step Valve - Closure NA M
R 1
10.
Turbine Control Valve fast Closure Valve Trip System Oil Pressure - Low NA M
R 1
11.
Reactor Mode Switch Shutdown Position NA R
NA 1,2,3,4,5 12.
Manual Scram NA M
NA 1,2,3,4,5 I
13.
Control Rod Drive l
a.
Charging Water Haader Pressure - Low NA M
R 2, 5
}
b.
Delay Timer NA M
R 2, 5 Y
(c) Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) The IRM, and SRM channels shall be determined to overlap for at least 1/2 decades during each startup and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days.
(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power levels calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER. The APRM Gain Adjustment Factor (GAF) for any channel shall be equal to the power value determined by the heat balance divided by the APRM reading for that channel.
Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, adjust any APRM channel with a GAF > 1.02.
In addition, adjust any APRM channel within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, (1) if power is greater than or equal to 90% of RATED THERMAL POWER and the APRM channel GAF is < 0.98, or (2)
T if power is less than 90% of RATED THERMAL POWER and the APRM reading exceeds the power value determined by the M
heat balence by more than 10% of RATED THERMAL POWER. Until any required APRM adjustment has been acccarglished, i
notification shall be posted on the reactor control panel.
3 (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.
5 (f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH).
~
(g) Measure and compare core flow to rated core flow.
(h) This calibration shall consist of verifying the 6 1 1 second simulated thermal power time constant.
1 TRAVERSING IN-CORE PROBE SYSTEM LIMITING CONDITION FOR OPERATION 3.3.7.7.
The traversing in-core probe (TIP) system shall be OPERABLE with:
a.
Movable detectors, drives and readout equipment to map the core in the required measurement locations and b.
Indexing equipment to allow all required detectors to be calibrated in a common location.
APPLICABILITY: When the traversing in-core probe is used for:
- a.
Recalibration of the LPRM detectors, and
- b.
Monitoring the APLHGR, LHGR, MCPR, or MFLPD.
ACTION:
a.
With one or more TIP measurement locations inoperable, required measure-mentsmay be performed as described in 1 and 2 belew, provided the reactor core is operating in an octant symmetric control rod pattern, and the total core TIP uncertainty for the present cycle has been measured to be less than 8.7 percent.
i 1.
TIP data for an inoperable measurement location may be replaced by data obtained from that string's redundant (symmetric) counterpart if the substitute TIP data was obtained from an operable measurement location.
2.
TIP data for an inoperable measurement location may be replaced by data obtained from a 3-dimensional BWR core simulator code normalized i
with available operating measurements, provided the total number of 1
simulated channels (measurement locations) does not exceed:
ij a)
All channels of a single TIP machine, or l
b)
A total of five channels if more than one TIP machine is
[
- involved, b.
Otherwise, with the TIP system inoperable, suspend use of the system for the above applicable monitoring or calibration functions.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.7 The traversing in-core probe system shall be demonstrated CPERABLE by I
normalizing each of the above required detector outputs within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> prior to use for the above applicable monitoring or calibration functions.
"Only the detector (s) in the required measurement location (s) are required j
j to be OPERABLE.
t 4
LA SALLE - UNIT 2 3/4 3-73 Amendment No. 35 t
~
.v INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Continued) 3 3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.
This capability is con-sistent with the recommendations of Regulatory Guide 1.97, "Instrumentation for Light Water Cooled Nuclear Power Plants to Assess Plant Conditions During and Following an Accident," December 1975 and NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations".
3/4.3.7.6 SOURCE RANGE MONITORS The source range monitors provide the operator with information of the status of the neutron level in the core at very low power levels during startup and shutdown.
At these power levels, reactivity additions should not be made without this flux level information available to the operator. When the inter-mediate range monitors are on scale adequate information is available without the SRMs and they can be retracted.
3/4.3.7.7 TRAVERSING IN-CORE PROBE SYSTEM The OPERABILITY of the traversing in-core probe (TIP) system with the specified minimum complement of equipment ensures that the measurements 4
obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core.
The specification allows use of substituted TIP data from symmetric channels if the control rod pattern is symmetric since the TIP data is adjusted by the plant computer to remove machine dependent and power level dependent bias.
The source of data for the substitution may also be a 3-dimensional BWR core simulator calculated data set which is normalized to avedlable real data.
Since uncertainty could be introduced by the simulation and normalization process, an evaluation of the specific control rod pattern and core operating state must be performed to ensure that adequate margin to core operating limits is maintained.
3/4.3.7.8 AMMONIA DETECTION SYSTEM The OPERABILITY of the ammonia detection system ensures that an accidental ammonia release will be detected promptly and the necessary protective actions will be automatically initiated to provide protection for control room per-sonnel.
Upon detection of a high concentration of ammonia, the control rcom emergency ventilation system will automatically be placed in the recirculation mode of operation to provide the required protection.
The detection systems required by this specification are consistent with the recommendations of Regula-tory Guide 1.78, "Assumptions for Evaluating the Habitability of a Nuclear Power Plant Control Room During a Postulated Hazardous Chemical Release."
3/4.3.7.9 FIRE DETECTION INSTRUMENTATION
~
OPERABILITY of the fire detection instrumentation ensures that adequate warning capability is available for the prompt detection of fires.
This capability is required in order to detect and locate fires in their early LA SALLE - UNIT 2 B 3/4 3-5 Amendment No. 35