ML20078K060
| ML20078K060 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 01/27/1995 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Philadelphia Electric Co |
| Shared Package | |
| ML20078K066 | List: |
| References | |
| NPF-39-A-087, NPF-85-A-049 NUDOCS 9502080376 | |
| Download: ML20078K060 (10) | |
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066 4 001 k...../
i PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352
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LIMERICK GENERATING STATION. UNIT 1 AMENDMENT'TO FACILITY OPERATING LICENSE Amendment No. 87 License No. NPF-39 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated August 22, 1994, 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 (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.
9502080376 950127 PDR ADDCK 05000352 P
PDR e
2.
Accordingly, the license is amended by changes to the Technical Specif' cations as indicated in the attachment to this license amendment, and phragraph 2.C.(2) of Facility Operating License No. NPF-85 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
87, are hereby incorporated into this license. Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR T LEAR REGULATORY COMMISSION Ag Nor John F. Stolz, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 27. 1995
4 ATTACHMENT TO LICENSE AMENDMENT NO. 97 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 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 Insert 1-2 1-2 3/4 1-19 3/4 1-19
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DEFINITIONS CORE ALTERATION 1.7 CORE ALTERATION shall be the movement of any fuel, sources, or reactivity i
control components within the reactor vessel with the vessel head removed i
and fuel in the vessel. The following exceptions are not considered to be CORE ALTERATIONS:
a)
Movement of source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement); and b)
Control rod movement, provided there are no fuel assemblies in the associated core cell.
Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.1.12.
Plant operation within these limits is addressed in individual specifications.
CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of the (GEXL) correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.
DOSE EOUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
DOWNSCALE TRIP SETPOINT (DTSP) 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.
E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.
EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc.
Times shall include diesel generator starting and sequence loading delays where applicable.
The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
LIMERICK - UNIT 1 1-2 Amendment No. 37. 66. 87
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REACTIVITY CONTROL SYSTEMS 3/4,1.5 STANDBY LIQUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system consisting of a minimum of two pumps and corresponding flow paths, shall be OPERABLE.
APPLICABILITY:
OPEPATIONAL CONDITIONS I and 2 ACTION:
a.
In OPERATIONAL CONDITION 1 or 2.
1.
With only one pump and corresponding explosive valve OPERABLE, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
With standby liquid control system otherwise inoperable, restore the system to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that:
1.
The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.
2.
The available volume of sodium pentaborate solution is at least 3160 gallons.
3.
The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.
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LIMERICK - UNIT 1 3/4 1-19 l
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UNITED STATES j
j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20606-0001 PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
49 License No. NPF-85 f
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated August 22, 1994, 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 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 tE. 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.
i
. 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. NPF-39 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
49, are hereby incorporated into this license.
Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR T UCLEAR REGULATORY COMMISSION g% 4 w
John F. Stolz, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: January 27. 1995
ATTACHMENT TO LICENSE AMENDMENT NO. 49 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 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 Insert 1-2 1-2 3/4 1-19 3/4 1-19
DEFINITIONS
-l CORE ALTERATION L
'1.7-
. CORE ALTERATION shall. be the movement of any fuel, sources, or reactivity control components within the reactor vessel with the vessel head removed and fuel in the vessel. The.following exceptions are not considered to be CORE ALTERATIONS:
a)
Movement of. source range monitors, local power range monitors, intermediate range monitors, traversing incore probes, or special moveable detectors (including undervessel replacement); and b)
Control rod movement, provided there are no fuel assemblies in the associated core cell.
Suspension of CORE ALTERATIONS shall not preclude completion of movement of a component to a safe position.
CORE OPERATING LIMITS REPORT 1.7a The CORE OPERATING LIMITS REPORT (COLR) is the unit-specific document that provides the core operating limits for the current operating reload cycle. These cycle-specific core operating limits shall be determined for each reload cycle in accordance with Specifications 6.9.1.9 thru 6.9.12.
Plant operation within these limits is addressed in individual specifications.
CRITICAL POWER RATIO 1.8 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly-which is calculated by application of the (GEXL) correlation to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.
DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, I-133, I-134, and I-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, " Calculation of Distance Factors for Power and Test Reactor Sites."
DOWNSCALE TRIP SETPOINT (DTSP) 1.9a The downscale trip setpoint associated with the Rod Block Monitor (RBM) rod block trip setting.
E-AVERAGE DISINTEGRATION ENERGY 1.10 E shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration, in MeV, for isotopes, with half lives greater than 15 minutes, making up at least 95% of the total noniodine activity in the coolant.
EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME 1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation set-point at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall include diesel generator starting and sequence loading delays where applicable. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
LIMERICK - UNIT 2 1-2 Amendment No. 4. 49 we -
REACTIVITY CONTROL SYSTEMS 3/4.1.5 STANDBY LIOUID CONTROL SYSTEM LIMITING CONDITION FOR OPERATION 3.1.5 The standby liquid control system, consisting of a minimum of two pumps and corresponding flow paths, shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2 ACTION:
a.
In OPERATIONAL CONDITION 1 or 2:
1.
With only one pump and corresponding explosive valve OPERABLE, restore one inoperable pump and corresponding explosive valve to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
With standby liquid control system otherwise inoperable, restore the system to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SMREElllANCE RE0VIREMENTS 4.1.5 The standby liquid control system shall be demonstrated OPERABLE:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by verifying that:
1.
The temperature of the sodium pentaborate solution is within the limits of Figure 3.1.5-1.
2.
The available volume of sodium pentaborate solution is at least 3160 gallons.
3.
The temperature of the pump suction piping is within the limits of Figure 3.1.5-1 for the most recent concentration analysis.
l LIMERICK - UNIT 2 3/4 1-19 Amendment No. 24. 49