ML20210K261
| ML20210K261 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 04/16/1986 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20210K267 | List: |
| References | |
| TAC-60293, TAC-60294, NUDOCS 8604280212 | |
| Download: ML20210K261 (18) | |
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UNITED STATES j
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ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT,llNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 63 License No. NPF-2 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Alabama Power Company (the licensee) dated November 27, 1985, 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 conforinity with the application, as amended, the provisions of the Act, and the regulations of the Commission;
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C.
There is reasonable assurance: (i) that the activities authorized by this amendment can be cor. ducted without endangering the health and safety of tSe public. and (ii) that such activities will be conducted in compliance with tne Comission's regulations; D.
The issuance of this license 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 accordan:e 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 attachment to this license' amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-2 is hereby amended to read as follows:
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6604280212 860416 '
PDR ADOCK 05000348 v
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(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 63, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION kt Lester S. R enstein, Director PWR Project Directorate #2 Division of PWR Licensing-A Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuaiice:
April 16,1986 O
ATTACHMENT TO LICENSE AMENDMENT NO. 63 TO FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of the Appendix "A" Techr.ical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical l'nes indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Remove Pages Insert Pages 1-2 1-2 3/4 4-23 3/4 4-23 B3/4 4-5 B3/4 4-5
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I 1.0 DEFINITIONS The defined terms of this section a;.near in capitalized type and are applicable throughout these Technical Specifications.
ACTION 1.1 ACTION shall be that part of a specification which prescribes remedia; measures required under designated conditions.
AXIAL FLUX DIFFERENCE 1.2 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals between the top and bottom halves of a two section excore,reutron detector.
CHANNEL CALIBRATION
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1.3 A CHANNEL CALIBRATION shall be the adjustment, as necessary, of the channel output such that it rasponds with the necessary range and accuracy to known values of the parameter which the channel monitors. The CHANNEL CALIBRATION shall encompass the entire channel including the sensor and alarm and/or trip. functions, and shall include the CHANNEL FUNCTIONAL TEST. The CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
CHANNEL CHECK 1.4 A CHANNEL CHECK shall be the qualitative assessment of channel behavior
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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
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measuring the same parameter.
CHANNEL FUNCTIONAL TEST 1.5 A CHANNEL FUNCTIONAL TEST shall be:
a.
Analog channels - the injection of a simulated signal into the channel as clos 2 to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.
b.
Bistable channels - the injection of a simulated signal into the sensor to verify OPERA 8ILITY including alarm and/or trig functions.
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FARLEY-UNIT 1 1-1 AMEN 0 MENT NO. 26
DEFINITIONS CONTAINMENT INTEGRITY 1.6 CONTAINMENT INTEGRITY shall exist when:
All penetrations required to be closed during accident conditions are a.
either:
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- 1) Capable of being closed by an OPERABLE containment eutomatic isolation valve system, or i
- 2) Closed by manual valves, blind flanges or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3, b.
All equipment hatches are closed and sealed, Each air lock is OPERABLE pursuant to Specification 3.6.1.3, c.
d.
The containment leakage rates are within the limits of Specification 3.6.1.2, and The setling mechanism associated with each penetration (e.g., welds, e.
bellows or 0-rings) is OPERABLE.
CONTROLLED LEAKAGE 1.7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
CORE ALTERATION 1.8 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement l
of a component to a safe conservative position.
. 4, DOSE EQUIVALENT 1-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131 I-132,1-133,1-134 and 1-135 actually The thyroid dose conversion factors used for this calculation shall be present.
those listed in Table E-7 of Regulatory Guide 1.109, Revision 1,1977.
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c.
FARLEY-UNIT 1 1-2 AMENDMENT NO. 63
-3/4.4.9 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.9 The specific activity of the primary coolant shall be limited to:
Less than or equal to 1.0 microcurie per gram DOSE EQUIVALENT I-131, a.
and b.
Less than or equal to 100/E microcurie per gram.
APPLICABILITY: MODES 3.,
2, 3, 4 and 5 ACTION:
MODES 1, 2 and 3*:
With the specific activity of the primary coolant greater than 1.0 a.
microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANU2Y with Tavg less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With the specific activity of the primary coolant greater than 100/E microcurie per gram, be in at least HOT STANDBY.with Tavg less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
- With T greater than or equal to 500'F.
avg 1
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FARLEY-UNIT 1 3/4 4-23 AMENDMENT NO. 63
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REACTOR COOLANT SYSTEM ACTION:
(Continued)
, MODES 1, 2, 3, 4 and 5:
a.
With th,e specific activi.ty of the primary coolant greater. than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of item da of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.
SURVEILLANCE REQUIREMENTS 4.4.9 The specific activity of the primary coolant shall be determined to be within the.11 mitt, by performance of the sampling and analysis program of Table 4.4-4.
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FARLEY-UNIT 1 3/4 4 24 AMENDMENT N0.,57 J
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REACTOR COOLANT SYSTEM BASES 3/4.4.8 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity of the Reactor Coolant System.
The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady Sttte Limits.
The surveillance requirements provide adeauate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.9 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture accident in conjunction with an assumed steady state primary-to-secondary steam generator leakage rate of 1.0 GPM. The values for the limits on specific j?
activity represent limits based upon a parametric evaluation by the NRC of typical site locations.
These values are conservative in that specific site parameters of the Farley site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microCuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
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O FARLEY-UNIT 1 B 3/4 4-5 AMENUMENT NO. 63 i
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j REACTOR COOLANT SYSTEM BASES 1
Reducing T,y to less then 500*F prevents the release of activity should a steam generator tube rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on,1odine spiking will be used to assess the parameters associated with spiking phenomena. A reduction in
)
frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
4 3/4.4.10 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G.
l 1)
The reactor coolant temperature and pressure and system heatup and cooldown j
rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3.
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a)
Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.
Limit lines for cooldown rates between those presented
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may be obtained by interpolation.
b)
Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-i ductile failure only.
For normal operation, other inherent plant j
characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
2)
These limit lines shall be calculated periodically using methods provioed cbelow.
3).
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70*F.
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FARLEY-UNIT 1 5 3/4 4-6 AMENDMENT N0. 26 1
1
p f.ag'o UNITED STATES g
NUCLEAR REGULATORY COMMISSION n
WASHINGTON, D. C. 20555
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ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amen 9 ment No. 54 License No. NPF-8 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Alabama Power Company (the licensee) dated November 27, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amer.ded (the Actl and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Comission;
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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 license 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 attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-8 is hereby amended to read as follows:
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Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 54, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This. license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION p-efb. 210vb Lester S.
ubenstein, Director PWR Project Directorate #?
1 Division of PWR Licensing-A i
Office of Nuclear Reactor Regulation j
Attachment:
Changes to the Technical 1
Specifications Date of Issuance: April 16, 1986 Me 1
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ATTACHMENT TO LICENSE AMEN 0 MENT NO. 54 TO FACILITY OPERATING LICENSE NO. NPF-8 1
DOCKET NO. 50-364 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by amendment number and contain vertical lines indicating the area of change.
The corresponding overleaf pages are also provided to maintain document completeness.
Remove Pages Insert Pages 1-2 1-2 3/4 4-23 3/4 4-23 B3/4 4-5 B3/4 4-5 W
W W
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0
1.0 DEFINITIONS The defined terms of this section appear in capitalized type and are applicable throughout these Technical Specifications.
l ACTION 1.1 ACTION shall be that part of a specification which prescribes remedial measures required under designated conditions.
AXIAL FLUX DIFFERENCE 1.2 AXIAL FLUX DIFFERENCE shall be the difference in normalized flux signals i
between the top and bottom halves of a two section excore neutron detector.
CHANNEL CALIBRATION I,
1.3 A CHANNEL CALIBRATION shall 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 CHANNEL
- l CALIBRATION shall encompass the entire channel including the sensor and alarm j
and/or trip functions, and shall include the CHANNEL FUNCTIONAL TEST. The l
CHANNEL CALIBRATION may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
j C_
l CHANNEL CHECK 1.4 A CHANNEL CHECK shall be the qualitative assessment of channel behavior j
during operation by observation.
This determination shall include, where i_
possible, comparison of the channel indication and/or status with other indications and/or status derived from independent instrument channels 4
measuring the same parameter.
CHANNEL FUNCTIONAL TEST l
1.5 A CHANNEL FUNCTIONAL TEST shall be:
a.
Analog channels - the injection of a simulated signal into the l
channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions.
- b.. Bistable channels - the injection of a simulated signal lpto the sensor to verify OPERABILITY including altre and/or trip;, functions..
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FARLEY-UNIT 2 1-1 i
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DEFINITIONS 3
CONTAINMENT INTEGRITY 1.6 CONTAINMENT INTEGPITY shall exist when:
a.
All penetrations required to be closed during accident conditions are either:
- 1) Capable of besng closed by an OPERABLE containment automatic isolation valve systen, or
- 2) Closed by manual valves, blind flanges or deactivated automatic valves secured in their closed positions, except as provided in Table 3.6-1 of Specification 3.6.3, b.
All equipment hatches are closed and sealed, c.
Each air lock is OPERABLE pursuant to Specification 3.6.1.3, d.
The containment leakage rates are within the limits of Specification 3.6.1.2, and The sealing mechanism associated with each penetration (e.g., welds, e.
bellows or 0-rings) is OPERABLE.
CONTROLLED LEAXAGE 1.7 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.
CORE ALTERATION 1.8 CORE ALTERATION shall be the movement or manipulation of any component within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position.
DOSE EQUIVALENT I-131 1.9 DOSE EQUIVALENT I-131 shall be that concentration of I-131 (microcurie / gram) which alone would produce the same thyroid dose as the quantity and isotopic mixture of I-131,1-132, I-133,1-134 cnd I-135 actually present. The thyroid dose conversion factors used for this calculation shall be those listed in Table E-7 of Regulatory Guide 1.109, Revision 1,1977.
F FARLEY-UNIT 2 1-2 AMEN 0 MENT NO. 54
REACTOR COOLANT SYSTEM 3/4.4.9 SPECIFIC ACTIVITY LIMITING CONDITION FOR OPERATION 3.4.9 The specific activity of the primary coolant shall be' limited to:
Less than or equal to 1.0 microcurie per gram DOSE EQUIVALENT I-131, a.
and b.
Less than or equal to 100/E microcurie per gram.
APPLICABILITY: hbDES1,2,3,4and5 4
ACTION:
MODES 1, 2 and 3*:
With the specific activity of the primary coolant greater than 1.0 a.
microcurie per gram DOSE EQUIVALENT I-131 for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> during one continuous time interval or exceeding the limit line shown on Figure 3.4-1, be in at least HOT STANDBY with Tavg less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> b.
With the specific activity of the primary coolant greater than 100/E microcurie per gram, be in at least HOT STANDBY with Tavg less than 500*F within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
i:
- With T greater than or equal to 500*F.
avg e
5 FARLEY-UNIT 2 3/4 4-23 AMENDMENT NO. 54
REACTOR COOL ANT SYSTEM ACTION: (Continued)
MODES 1, 2, 3, 4 and 5:
a.
With the specific activity of the primary coolant greater than 1.0 microcurie per gram DOSE EQUIVALENT I-131 or greater than 100/E microcuries per gram, perform the sampling and analysis requirements of item 4a of Table 4.4-4 until the specific activity of the primary coolant is restored to within its limits.
SURVEILLANCE REQUIREMENTS 4.4.9 The specific activity of the primary toolant shall be determined to be
, within the limits by performance of.the sampling and analysis program of Table 4.444.
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FARLEY-UNIT 2 3/4 4-24 AMENDMENT 110. 49 i
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REACTOR COOLANT SYSTEM BASES 3/4.4.8 CHEMISTRY The limitations on Reactor Coolant System chemistry ensure that corrosion of the Reactor Coolant System is minimized and reduces the potential for Reactor Coolant System leakage or failure due to stress corrosion. Maintaining the chemistry within the Steady State Limits provides adequate corrosion protection to ensure the structural integrity of the Reactor Coolant System over the life of the plant. The associated effects of exceeding the oxygen, chloride and fluoride limits are time and temperature dependent. Corrosion studies show that operation may be continued with contaminant concentration levels in excess of the Steady State Limits, up to the Transient Limits, for the specified limited time intervals without having a significant effect on the structural integrity, of the Reactor Coolant System. The time interval permitting continued operation within the restrictions of the Transient Limits provides time for taking corrective actions to restore the contaminant concentrations to within the Steady State Limits.
The surveillance requirements provide adequate assurance that concentrations in excess of the limits will be detected in sufficient time to take corrective action.
3/4.4.9 SPECIFIC ACTIVITY The limitations on the specific activity of the primary coolant ensure that the resulting 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> doses at the site boundary will not exceed an appropriately small fraction of Part 100 limits following a steam generator tube rupture i
accident in conjunction with an assumed steady state primary-to-secondary steam 11 generator leakage rate of 1.0 GPM. The values for the limits on specific activity represent limits based upon a parametric evaluation by the NRC of i
typical site locations. These values are conservative in that specific site parameters of the Farley site, such as site boundary location and meteorological conditions, were not considered in this evaluation.
, l'.
The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity greater than 1.0 microCuries/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure 3.4-1, accommodates possible iodine spiking phenomenon which may occur following changes in THERMAL POWER.
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FARLEY-UNIT 2 B 3/4 4-5 AMENDMENT NO. 54 4
1 e, - -,-
BASES Reducing T,yg to less than 500*F prevents the release of activity should
]
a stear generator tube rupture since the saturation pressure of the primary j
coolant is below the lift pressure of the atmospheric steam relief valves.
The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in sufficient time to take corrective action.
Information obtained on iodine spiking will be used j
to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.
3/4.4.10 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G.
1)
The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2 and 3.4-3 for the first full power service period.
~~
a)
Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.
Limit lines for cooldown rates betwe<en those presented may be obtained by interpolation.
b)
Figures 3.4-2 and 3.4-3 define limits to assure prevention of non-ductile failure only.
For normal operation, other inherent plant characteristics, e.g., pump heat addition and pressurizer heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.
s 2), These limit lines shall be calculated periodically using methods provided below.
3)
The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70'F.
y FARLEY-UNIT 2 8 3/4 4-6
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