ML17158A173
| ML17158A173 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 03/08/1994 |
| From: | Chris Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17158A174 | List: |
| References | |
| NUDOCS 9403160282 | |
| Download: ML17158A173 (13) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHING TON, D.C. 20555-0001 P NNSYLVANIA POWER 5 IGHT COMPANY LLEGHENY ELECTRIC COOPERATIVE INC.
O~DCK T NO. 50-381 SUS U HANNA STEAM ELECTRIC STATION UNIT 1
AMENDMENT TO FACILITY OPERATING LICENS Amendment No.
f33 License No.
NPF-14 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for the amendment filed by the Pennsylvania Power 8
Light Company, dated October 8,
- 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission'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 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 set forth in 10 CFR Chapter I; 0.
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.
V403ib0282 940308 PDR ADOCK 05000387 p
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2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14 is hereby amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
133 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
PPKL 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 is to be implemented within 60 days of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION CPM Zrc8 Charles L. Miller, 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:
March 8, 1994
ATTACHMENT TO LICENSE AMENDMENT NO. 133 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The overleaf page is provided to maintain document completeness.*
REMOVE 1-1 1-2 INSERT
1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved.
The defined terms appear in capitalized type and 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 EXPOSURE 1.2 The AVERAGE BUNDLE EXPOSURE shall be equal to the sum of the axially averaged exposure of all the fuel rods in the specified bundle divided by the number of fuel rods in the fuel bundle.
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 applicable 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 CALIBRATION 1A A CHANNEL CALIBRATIONshall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values ofthe parameter which the channel monitors. The CHANNELCALIBRATIONshall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONALTEST.
Calibration of instrument channels with resistance temperature detector (RTD) or thermocouple sensors may consist of an in-place qualitative assessment of sensor behavior and normal calibration ofthe remaining adjustable devices in the channel. The CHANNELCALIBRATIONmay be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK i
, 1.5 A CHANNEL CHECK shall be the qualitative assessment of channel behavior during operation by observation.
rhis determination shall include, where possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUNCTIONALTEST 1.6 A CHANNEL FUNCTIONALTEST shall be:
- a. Analog channels
- the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITYincluding 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 FUNCTIONALTEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.
SUSQUEHANNA - UNIT 1 Amendment No
0 INITI N R
A T 1.7 CORE AI.TERATION shall be the addition. removal, relocation or movement of fuel, sources, or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.
Normal movement of the SRMs, IRMs, TIPs or special movable detectors is not considered a CORE ALTERATION.
Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.
R A IN IMIT R
T 1.7A The CORE OPERATING LIMITS REPORT is the Susquehanna SES Unit 1 specific document that provides 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 Specification 6.9.3.
Plant operation within these operating limits is addressed in individual specifications.
RITI A RATI 1.8 The CRITICALPOWER RATIO ICPR) shall be the ratio of that power in the assembly which is calculated by application of the appropriate correlationlsl to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.
VA 1-1 1
1.9 OOSE EQUIVALENTI-131 shall be that concentration of l-131, microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131, l-132, 1-1 33, l-131, and 1-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, CalcWation of Distance Factors for Power and Test Reactor Sites.
V A
I RATI 1.10 6 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, maldng up at least 95% of the total no~odine activity in the coolant.
1.11 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIMEshall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety functions, i.e., the valves travel to their required positions, pump discharge pressures reach their required values, etc. Times shall incjude 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.
F-1.12 The ENLHV~CLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIMEshall be that time interval to complete suppression of the electric arc between the fully open contacts of the recircWation pump circuit breaker from initial movement of the associated:
a.
Turbine stop valves, and b.
Turbine control valves.
This total system response time consists of two components, the instrumentation response time and the breaker arc suppression time. These times may be measured by any series of sequential, overlapping or total steps such that the entire response is measured.
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UNITED STATES NUCLEAR REGULATORY COIVIIVIISSION WASHINGTON. D.C. 20555.0001 PENNSYLVANIA POWER 8L L GHT COMPANY A L GHENY ELECTRIC COOPERATIVE INC.
DOCKET NO. 50-388 SUS UEHAN A STEAM ELECTRIC STATION UNIT 2 AMENDMENT TO FACI ITY OPERATING LIC NS Amendment No. 102 License No. NPF-22 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for the amendment filed by the Pennsylvania Power 8 <-
Light Company, dated October 8,
- 1993, complies with the standards and
=
requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission'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 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 set forth in 10 CFR Chapter I; 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.
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2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No.
NPF-22 is hereby amended to read as follows:
(2) Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 102 and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
PP&L 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 is to be implemented within 60 days of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
March 8, 1994 Charles L. Miller, Director Project Directorate I-2
.Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
ATTACHMENT TO LICENSE AMENDMENT NO.
CILITY OP RATING ICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The overleaf page is provided to maintain document completeness.*
REMOV 1-1 1-2 INSERT 1-1 1-2*
1.0 DEFINITIONS The following terms are defined so that uniform interpretation of these specifications may be achieved.
The defined terms appear in capitalized type and shall be applicable throughout these Technical Specifications.
ACTtOM 1.1 ACTION shall be that part of a Specification which prescribes remedial measures required under designated conditions.
AVERAGE EXPOSURE 1.2 The AVERAGE BUNDLE EXPOSURE shall be equal to the sum of the axially averaged exposure of all the fuel rods in the specified bundle divided by the number of fuel rods in the fuel bundle.
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, applicable 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 CALIBRATION 1.4 A CHANNEL CALIBRATIONshall be the adjustment, as necessary, of the channel output such that it responds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNELCALIBRATIONshall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the CHANNEL FUNCTIONALTEST.
Calibration of instrument channels with resistance temperature detector {RTD) or thermocouple sensors may consist of an in-place qualitative assessment of sensor behavior and normal calibration ofthe remaining adjustable devices in the channel. The CHANNELCALIBRATIONmay be performed by any series of sequential, overlapping or total channel steps such 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 indications and/or status derived from independent instrument channels measuring the same parameter.
CHANNEL FUN TIONALTEST 1.6 A CHANNEL FUNCTIONALTEST shall be:
- a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITYincluding 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 CHANNELFUNCTIONALTEST may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is tested.
SUSQUEHANNA - UNIT 2 Amendment No. g1 102
OEFINITI N R
A ATI 1.7 CORE ALTERATIONshall be the addition. removal, relocation or movement of fuel, sources, or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel
~ Normal movement of the SRMs, IRMs. TIPs or special movable detectors is not considered a CORE ALTERATION.
Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.
R P RATIN IMIT R
RT 1.7A The CORE OPERATING LIMITS REPORT is the Susquehanna SES Unit 2 specific document that provides 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 Specification 6.9.3.
Plant operation within these operating limits is addressed in individual specifications.
RITI A W R RATI 1.8 The CRITICALPOWER RATIO ICPR) shall be the ratio of that power in the assembly which is calculated by application of the appropriate correlationlsi to cause some point in the assembly to experience boiling transition. divided by the actual assembly, operating power.
IVA NT I-1 1
1.9 DOSE EQUIVALENT1-131 shall be that concentration of 1-131, microcuries per gram, which alone woukl produce the same thyroid dose as the quantity and isotopic mixture of I-131, I-132, l-133, l-l34, and 1-135 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.
~ V A
I N
ATI 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 non~odine activity in the coolant.
1.11 The EMERGENCY CORE COOLING SYSTEM {ECCS) RESPONSE TIMEshall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety functions, 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.
TR 1.12 The ENDAF~CLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIMEshall be that time interval to complete suppression of the electric arc between the fully open contacts of the recircubition pump circurt breaker from initial movement of the associated:
a.
Turbine stop valves, and b.
Turbine control valves.
This total system response time consists of two components, the instrumentation response time and the breaker arc suppression time. These times may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
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