ML20214U252
| ML20214U252 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 09/09/1986 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214U253 | List: |
| References | |
| TAC-60826, TAC-60827, NUDOCS 8609300555 | |
| Download: ML20214U252 (14) | |
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8 UNITED STATES
- 4' NUCLEAR REGULATORY COMMISSION 7
3 WASHINGTON, D. C. 20555
. e 5
~
9..v..9 ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO. 1 AVENDMENT TO FACILITY OPERATING LICENSE Amendment No. 65 License No NPF-2 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Alabama Power Comoany (the licenseel dated Februarv 7,1986, 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, as amended, the provisions of the Act, and the regulations of the Commission; C.
There is reasonable assurance: (1) that the activities authorized by this amendment can be conducted without endancering the health and safety of the public, and (111 that such activities will be conducted in compliance with the Commission's regulations; D.
The issuance of this license 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 Comission's regulations and all applicable requirements have i
been satisfied.
2.
Accordingly, the license is anended by changes to the Technical i
Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(?) of Facility Operating License No. NPF ? is hereby amended to read as follows:
khDO 8
P l
O 7
-2 (21 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 65, 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 and shall be implemented within 30 days of receipt of the amendment.
FOR THE NilCLEAR REGULATORY COMM SSION Lester S. Rubenstein, Director PWR Pro,iect Directorate #2 Division of PWR Licensing-A Office of Nuclear Reactor Pegulation
Attachment:
Changes to the Technical Specifications Date.of Issuance: September 9,1986 1
i i
+
ATTACHMEN' TO LICENSE AMENDMENT NO. 65 TO FACILITY OPERATING LICENSE NO. NPF-?
DOCKET NO. 50-348 Replace the following pages of the Appendix "A" Technical Specifica' ions t
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 Paces 3/4 d-2 3/4 4-2 R3/4 4-1 B3/4 4-1
n v
REACTOR COOLANT SYSTEM ATIQN 3/4.4 REACTOR COOLANT LOOPS AND COOLANT CIRC 3/4.4.1 STARTUP AND POWER OPERATIQN LIMITING CONDITION FOR OPERATION in All Reactor Coolant loops shall be in operat o.
3.4.1.1 MODES I and 2.*
APPLICABILITY:
Coolant loops in operation, be in ACTION:
With less than the above required Reactor at least HOT STAvDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
c SURVEILLANCE REQUIREMENT h ll be verified to be in The above required Reactor Coolant loops s at at least on l
operation and circulating reactor coo an 4.4.1.1 l
l I
i
- See Special Test Exception 3.10.4.
AMENDMENT NO. 26 3/4 4-1 NIT 1
i l
l l
HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 All three Reactor Coolant Loops listed below shall be OPERABLE and in l
cperation when the rod control system is operational or at least two Reactor Coolant Loops listed below shall be OPERABLE with one Reactor Coolant Loop in operation when the rod control system is disabled by opening the Reactor Trip Breakers or shutting down the rod drive motor / generator sets:*
1.
Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump, I
l 2.
Reactor Coolant Loop B and its associated steam generator and Reactor Coolant pump, l
s 3.
Reactor Coolant Loop C and its associated steam generator and l
Reactor Coolant pump.
l APPLICABILITY: MODE'3 i
ACTION:
l With less tnan the above required Reactor Coolant loops OPERABLE, l
a.
restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN witnin the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
I b.
With less tnan three Reactor Coolant loops in operation and the rod I
control system operational, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> open the Reactor Trip Breakers or shut down the rod drive motor / generator sets.
With no Reactor Coolant loops in operation, suspend all operations c.
involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required coolant loop to operation.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, shall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4.1.2.2 The required Reactor Coolant loop (s) shall be verified to be in operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.2.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 107 of wide range indication at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- All Reactor Coolant pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System l
boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
E Y-UNIT 1 3/4 4-2 AMENDMENT NO. 2f, 65
3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all Reactor Coolant loops in operation, and peintain DNBR above 1.30 during all norwel operations and In H0 DES 1 and 2 with one Reactor Coolant loop not in anticipated transients.
operation this specification requires that the plant be in at least fl0T STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
In MODE 3, three reactor coolant loops provide sufficient heat removal capability for removing core heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient decay heat removal capacity if a bank withdrawal accident can be prevented; i.e., by opening the Reactor Trip Breakers or shutting down the rod drive motor / generator When a bank withdrawal accident can be prevented, single failure sets.
considerations require that two loops be OPERABLE at all times.
In MODE 4, a single reactor coolant or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations Thus, if the reactor coolant loops require that at least two loops be OPERABLE.
are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
In MODE 5, single failure considerations require two RHR loops to be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be' within the capability of operator recognition and control.
The restrictions on starting a Reactor Coolant Pump with one or more Reactor Coolant System cold legs less than or equal to 310*F are provided to prevent Reactor Coolant System pressure trans'ents, caused by energy additions from the secondary system, which could exceed the'11mits of Appendix G to 10CFR The Reactor Coolant System will be protected against overpressure Part 50.
transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg terperatures.
FARLEY-UNIT 1 B 3/4 4-1 AMENDMENT NO. 2f, 63
... _ ~ ~
REACTOR COOLANT SYSTEM BASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent.the RCS from being pressurized above its Safety Limit of 2735 psig.
Each safety valve is d2 signed to relieve 345,000 lbs per hour of saturated steam at the valve set paint.
The relief capacity of a single safety valve is adequate to relieve cny overpressure condition which could occur during shutdown.
In the. event that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpres-surization.
In addition, the Overpressure Protection System provides a
/ diverse means of protection against RCS overpressurization at low temperatures.
Duringoperation,allpressuriz$fcodesafetyvalvesmustbeOPERABLEto 3
prevent the RCS from being pressurized above its safety limit of 2735 psig.
i The combined relief capacity of all of these valves is greater than the taximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump i
valves.
_ Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI cf the ASME Boiler and Pressure Code.
3/4.4.4 PRESSURIZER The limit on the maximum water vol.ume in the pressurizer assures that the parameter is maintaineg within the normal steady state envelope of operation assumed in the SAR.
The limit is consistent with the initial SAR assumptions.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance is sufficient to ensure that the parameter is restored to within its limit following expected transient operation. The maximum water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically rolid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant Pressure and establish natural circulation.
3/4.4.5 RELIEF VALVES (PORV's)
The power operated relief valves and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump.
Operation of the PORV's minimizes the undesirable spening of the spring-loaded pressurizer code safety valves.
Each PORV has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable.
l FARLEY-UNIT 1 B 3/4 4-2 AMENDMENT NO. 26
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og UNITED STATES 8
g NUCLEAR REGULATORY COMMISSION 5
' j WASHINGTON, D. C. 20555
\\....*/
ALARAMA POWER COMPANY DOCKET NO. 50-364
.10SEPH M. FARLEY NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 58 License No. NPF-8 1.
The Nuclear Regulatory Commission (the Comission) has found that:
A.
The application for amendment by Alabama Power Company (the licenseel dated February 7, 1986, 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, as amended, the provisions of the Act, and the regulations of the Commission; 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 Commission's regulations; D.
The issuance of this license 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 reoufrements 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:
?
-?-
(2)
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.58, 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 and shall be implemented within 30 days of receipt of the amendment.
FOR THE NilCLEAR REGULATORY COMMISSION ch W
lester. Rubenstein, Director PWR Project Directorate #P Division of PWP Licensing-A Office of Nuclear Reactor Regulation
Attachment:
Changes to tne Technical Specifications Date of Issuance: September 9,1986 i
i i
ATTACHMENT TO LICENSE AMENDMENT NO. 58 TO FACILITY OPERATING LICENSE NO. NPF-8 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 3/4 4-2 3/4 4-2 B3/4 4-1 R3/4 4-1
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION STARTUP AND POWER OPERATION LIMITING CONDITION FOR OPERATION 3.4.1.1 All Reactor Coolant loops shall be in operation.
I APPLICABILITY:
MODES I and 2.*
ACTION:
/
With less than the above required Reactor Coolant loops in operation, be in
[
at least HOT STANDBY within I hour.
SURVEILLANCE REQUIREMENT 4.4.1.1 The above required Reactor Coolant loops shall be verified to be in l
operation and circulating reactor coolantLat least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l i
l "See Special Test Exception 3.10.4.
FARLEY-UNIT 2 3/4 4-1 l
. -.., -.,.,.... ~,. - - - - - - _ - - _, - _. -. _ - - - - - - - -... _ - -.... -. -,. - - - - _ - _ _
REACTOR COOLANT SYSTEM HOT' STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 All three Reactor Coolant Loops listed below shall be OPERABLE and in operation when the rod control system is operational or at least two Reactor Coolant Loops listed below shall be OPERABLE with one Reactor Coolant Loop in operation when the rod control system is disabled by opening the Reactor Trip Breakers or shutting down the rod dr.ive motor / generator sets:*
~
1.
Reactor Coolant !.ocp A and its associated steam generator and Reactor Coolant pump, 2.
Reactor Coolant Loop B and its associated steam generator and Reactor Coolant pump, 3.
Reactor Coolant loop C and its associated steam generator and Reactor Coolant pump.
APPLICABILITY: MODE 3 ACTION:
With less than the above required Reactor Coolant loops OPERABLE, a.
restore the required loops to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in 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 />, b.
With less than three Reactor Coolant loops in operation and the rod control system operational, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> open the Reactor Trip Breakers or shut down the rod drive motor / generator sets, With no Reactor Coolant loops in operation, suspend all operations c.
involving a reduction in boron concentration of the Reactor Coolant f
System and immediately initiate corrective action to return the I
required coolant loop to operation.
SURVEILLANCE REQUIREMENTS
=
4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, snall be determined to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.
4.4.1.2.2 The required Reactor Coolant loop (s) shall be verified to be in operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.2.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 10% of wide range indication at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- All Reactor Coolant pumps may be de-energized for up to I hour provided (1) no operations are permitted that would cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10*F below saturation temperature.
FARLEY-UNIT 2 3/4 4-2 AMENDMENT N3. 53
3[4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all Reactor Coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients.
In MODES 1 and 2 with one Reactor Coolant loop not in operation this specification requires that the plant be in at least' HOT STANDBY within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
In MODE 3, three reactor coolant loops provide sufficient heat removal capability for removing core heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient decay heat removal capacity if a bank withdrawal accident can be prevented; i.e., by opening the Reactor Trip Breakers or shJtting down the rod drive motor / generator sets. When a bank withdrawal accident can be prevented, single failure considerations require that two loops be OPERABLE at all times.
In MODE 4, a single reactor coolant or RHR loop provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.
In MODE 5, single failure considerations require two RHR loops to be OPERABLE.
The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity charige rate associated with boron reduction will, therefore, be within the capability of operator recognition and control.
The restrictions on starting a Reactor Coolant Pump with one or more Reactor Coolant System cold legs less than or equal to 310*F are provided to prevent Reactor Coolant System pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10CFR Part 50. The Reactor Coolant System will be protected against overpressure transients and will not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures.
FARLEY-UNIT 2 B 3/4 4-1 AMENDMENT NO.58
~
8ASES 3/4.4.2 and 3/4.4.3 SAFETY VALVES
' The pressurizer code safety valves operate to prevent the RCS from.being Each safety valve is pressurized above its Safety Limit of 2735 psit.
1 345,000 lbs per hour of saturated steam at the valve set designed to relieve The relief capacity of a single safety valve is adequate to relieve point.
In the event cny overpressure condition which could occur during shutdown.
that no safety valves are OPERABLE, an operating RHR loop, connected to the RCS, provides overpressure relief capability and will. prevent RCS overpres-surization.
In addition, the Overp essure Protection System provi, des a diverse means of protection against RCS overpressurization at low temperatures.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2735 psig.
The combined relief capacity of all of these valves is greater than the saximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective System trip set point is reached (i.e., no credit is taken for a direct reactor trip on the loss of load) and also assuming no operation of the power operated relief valves or steam dump valves.
Demonstration of the safety valves' lif t settings will occur only during I
shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
l 3/4.4.4 PRESSURIZER The limit on the maximum water volume in the pressurizer assures that the I
parameter is maintainey within the normal steady state envelope of operation The limit is consistent with the initial SAR assumptions.
assumed in the SAR.
The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance is sufficient to ensure that the parameter The is restored to within its limit following expected transient operation.
maximuc water volume also ensures that a steam bubble is formed and thus the RCS is not a hydraulically solid system. The requirement that a minimum number of pressurizer heaters be OPERABLE enhances the capability of the plant to control Reactor Coolant Pressure and establish natural circulation.
3/4.4.5 RELIEF VALVES (PORV's)
The power operated relief valves and sttam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORV's minimizes the undesirable Each PORV has a opening of the spring-loaded pressurizer code safety valves.
remotely operated block valve to provide a positive shutoff capability should l
a relief valve beccme inoperable.
8 3/4 4-2 Amendment No. 13 FARLEY-UNIT 2
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