ML20038A548
| ML20038A548 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/05/1981 |
| From: | Varga S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20038A549 | List: |
| References | |
| NUDOCS 8111130106 | |
| Download: ML20038A548 (19) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION o
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- p WASHINGTON, D. C. 20555
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ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 25 License No. NPF-2 1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Alabama Power Company (the licensee) dated September 25, 1981, 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 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.
8111130106 811005 DR ADOCK 05000
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Accordingly, the license is amended by changes to the T'echnical 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:
(2) Technical Scecifications The Technical Specifications contained in Appendices A and 3, as revised through Amendment No. 25.. 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 the date of its issuance.
F03THENUCL'RRrGULATORYCOMMISSION
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9 L
ar's,
he Operating Reactors Bra ch #1 Division of Licensing '
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 5,1981 l
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ATTACHMENT TO LICENSE AMENDMENT AMEN 0 MENT, N_0' 25 TO FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Revise Appendix A as follows:
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Remove Pages Insert Pages IV '
IV C.TI XI XVII XVII 3/4 3-23 3/4 3-23 3/4 3-53 3/4 3-53 3/4 3-54 3/4 3-54 3/4 3-55 3/4 3-55 t
7 3/4 4 3/4 4-6 3/4 4-6a B3/4 3-4 83/4 3-4 B3/4 4-2 B3/4 4-2 2
B3/4 4-2a 6-1 6 -1
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INDEX LIMITING CONDITIONS FOR OPERATION At:3_ SUP.VEILLAfiCE REQUIREMENTS SECTION Face M-3/4.2' P0k'ER DISTRIBUTION LIMITS 3/4.2'1 AXIAL FLUX DIFFERENCE...................................
3/42-1 3/4.2.2 H EAT FLUX HOT CHANNEL FACT 0R......................,....
3/4 2-5,
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3/4.2.3-NUCLEAR EllTHALPY HOT CHAtit;EL FACT 0R....................
3/4 2-9 t
3/4.2.4 QUADRAtiT POWER TILT.RATI0..............................
3/4 2-11 3/4.2,5 DNB PARAMETERS.........................................
3/4 2-13 3/4.-3 INSTRUMENTATION 3/,4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION....................
3/43-1 3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM.............
3/4 3-14 It!STRUMENTATION 3/4.3.3 MGri!TORING-INSTRUMENTATION 4
Radiation Monitoring...................................
3/4 3-35 Mova' l e Incore De tec tors...............................
3/4 3-39 a
m Seismic Instrumentation..................'.............
3/4 3-40 Mateorological Instrumentation.........................
3/4 3-43 R4 mote Shutdown Instrumentation........................
3/4 3-46 Chlori ne De tec tion Syst ems.............................
3/4 3-49 High Energy Line Break Isolation Sensors...............
3/4 3-50
- AccidentsMonitoring Instrumentation......................
3/4 3-53
'l Fire Detection Instrumentation.:.......................
3/4 3-55 3/4.4 REACTOR C00L' ANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS g
Normal Operation.......................................
G/4 4-1 I- (
3/4.4.2 S AF ETY VALV ES - SHU TD0WN..................'............ -
3/4 4-4 3/4.4.3 SAFETY VALVES --0DERATING..'............................
3/4 4-5 3/4.4.4 PRESSURIZER............................................
3/4 4 6 3/4.4.4a RELI EF VALVE 3... * * * +.
- a.............................
3/4 4-Aa ~l 3/4.4.5 STEAM GENERATORS...............
3/4 4-7 FARLEY - UNIT 1 IV AMEN 0 MENT NO. 25 i
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j INDEX BASES SECTION PAGE l
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3/4.3 INSTRUMENTATION i
3/4.3.1 PROTECTIVE INSTRUMENTATION..............................
B 3/4 3-1 I
3/4:3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION...............
B 3/4 3-1 l'
1 3/4.3.3 NONITORING INSTRUMENTATION..............................
B 3/4 3-1 l
I l
3/4.4 REACTOR C00'LANT SYSTEM 1
3/4.4.i nEAcToa C00 tant t00PS...........................
B 3/4 4-i l
3/4.4.2 and 3/4.4.3 SAFETY VAtVES...............................
B 3/4 4-1 j
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3/4.4.4 PRESSURI2ER........,.....................................
B 3/4 4-2.
3/4.4.4a RELIEF VALVES (P0RV's)................................... B 3/4 4-2 ]
j 3/4.4.5 STEAM GENERATORS........................................
B 3/4 4-2 l
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3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.......................... B 3/4 4-4 l
3/4.4.7 CHEMISTRY...............................................
B 3/4 4-5
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3/4.4.8 SPECIFIC ACTIVITY.......................................
B 3/4 4-5 t
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3/4.4.9 PRESSURE / TEMPERATURE LIMITS.................~...........
B 3/4 4-6 i
3/4.4.10 STRUCTURAL INTEGRITY...................................
B 3/4 2
FARLEY - UNIT 1 XI AMENDMENT NO. 25
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INDEX ADMINISTRATIVE CONTROLS SECTION PAGE 6.1 RESPONSIBILITY..............................................
6-1 I
6.2 ORGANIZATION 0ffsite.....................................................
6-1 Fa c i l i ty S ta f f..............................................
6-1 Sh i ft Techn i cal Advi sor.....................................
6 -1 l
6.3 FACILITY STAFF QUALIFICATIONS...............................
6-5 6.4 TRAINING.'...................................................
6-5 6.5 REVIEW AND AUDIT 6.5.1 PLANT OPERATIONS REVIEW COMMITTEE (PORC)
Function..................................................
6-5 Co m p o s i t i o n...............................................
6-5 Alternates................................................
6-5 Meeting Frequency.........................................
6-6 Quorum....................................................
6-6 Re s p on si bi l i ti e s......................... z................
6-6 Auchority.................................................
6-7 Records...................................................
6-7 I
6.5.2 NUCLEAR OPERATIONS REVIEW BOARD (NORB)
Function..................................................
6-7 Composition...............................................
6-8 Al t e r n a t e s................................................
6-5 FARLEY - UNIT 1 XVII AMENDMENT NO. 25
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TABLE 3.3-4 g
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- Q ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS a
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2 FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES c
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1.
SAFETY INJECTION, TUR81NE TRIP AND 1l FEEDWATE:1 ISOLATION a.
Manual Initiation Not Applicable Not Applicable i
b.
Automatic Actuation Logic tiot Applicable Not Applicable c.
Containment Pressure--High 1 4.0 psig 1 4.5 psig l
d.
Pressurizer Pressure--Low
> 1850 psig
> 1840 psig i
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Differential Pressure i 100 psi i 112 psi Between Steam Lines--High wh j
f.
Steam Line Pressure--Low
> 585 psig steam line
> 575 psig steam line f
pressure pressure I
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INSTRUMENTATION ACCIDENT MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.3.8 The accident tr.onitoring instrumentation channels shown in Table 3.3-11 shall be OPERABLE.
APPLICABILITY: MODES 1, 2 and 3.
ACTION:
a.
With the number of OPERABLE accident monitoring channe'is less than the Required Number of channels shown in Table 3.3-11, restore the inoperable channel 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 />.
b.
With the number of OPERABLE accident monitoring channels less than
- the Minimum Channels OPERABLE requirements of Table 3.3-11; restore the inoperabia channel (s) to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 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 />.
The provisions of Specification 3.0.4 are not applicable.
c.
SURVEILLANCE REOUIREMFNTS 4.3.3.8 Each accident monitoring instrumentation channel shall be demonstrated OPERABLE by performance of the CHANNEL CHECK and CHANNEL CALIBRATION operations at the frequencies shown in Table 4.3-7.
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FARLEY-UNIT 1 3/4 3-53 AMENDMENT N0. 25 i
TABLE 3.3-11 D
ACCIDENT MONITORING INSTRUMENTATION A
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REQUIRED HINIMUM
--4 NUMBER CilANNELS INSTRUMENT OF CilANNELS OPERABLE 1.
Reactor Coolant Outlet Temperature-T,,gt-Wide Range 2
1 2.
Reactor Coc,lant Inlet Temperature-T
-Wide Range 2
1 Cold 3.
Reactor Coolant Pressure-Vide Range
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2 1
4.
Steam Generator Water Level-Vide Range or Narrow Range 2/ steam generator 1/stea: generator 5.
Refueling Water Storage Tank Water Level 2
1
~6.
Containment Pressure 2
1 7
Pressurizer Water Level 2
1 8.
Steam Line Pressure 2/ steam generator 1/ steam generator
- 9.
Auxiliary Feedwater Flow Rate 2
1
- 10.
Reactor Coolant System Subcooling Margin Monitor 2.
1
- 11.
PORV Position In'dicator 1/ valve 1/ valve R
- 12.
PORV Dlock Valve Position Indicator 1/ valve 1/ valve 2
- 13.
Safety Valve Position Indication (One channel is position 2/ valve
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indicator and one channel is discharge temperature) 1/ valve o
~*Not applicable if the associated block valve is in the closed position.
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^^Not applic.ible il the block v.ilve is verilied in Liut closed po.iLiosi and power resiuived.
- Specification effective 90 days following the return to power following the third refueling outage.
TABLE 4.3-7
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f ACCIDENT MONITORING INSTRUMENTATION SURVElllANCE REQUIREMENTS 9
E CHANNEL CllANNEL Q..
1HSTRUMENT CHECK CALIBRATION s
l Reactor Coolant Outlet Temperature-T a
M R
il t Wide Range 2.
Reactor Coolant Temperature-T
-Wide Range M
R Cold 3.
Reactor Coolant Pressure-Wide Range M
R 4.
Steam Generator Water level-. Wide Range or Harrow Range M
R S.
Refueling Water Storage Tank Water Level M
R w
3 6.
Containment Pressure M
R 7.
Pressurizer Water Level M
R 8.
Steam Line Pressure
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M R
! 9.
Auxiliary Feedwater Flow Rate M
R
- 10.
Reactor Coolant System Subcooling g
Margin Monitor M
R 2
E
- ^11.
PORV Position Indicator M
R E
'#^^12.
PORV Block Valve Position Indicator M
R
- 13.
Safety Valve Position Indicator M
R
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~ 1 0t applicable.if the associated block valve is in the closed position.
- Not, applicable if the block valve is verified in the closed position and power removed.
Specification effective 90 days following the return to power foll6 wing the third refueling outage.
REACTOR COOLANT SYSTEM 3/4.4.4 PRESSUNIZER l
LI!!TTING CONDITION FOR ODERATION -
3.4.4 The pressurizer shall be OPERA 3LE with at lean 125 kw of pressurizer heaters and a water voluma of less than or,cqucl to 563 (63.5% indicated) cubic feet.*
l APPLICABILITY: MODES 1, 2, and 3.
ACTION:
With the pressurizer inoperabic due to an inoperable acergency a.
power supply to the pressurizer heaters either restore the inoperable emergency power supply within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least 110T STAND 3Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN uithin the following 6 hourt.
b.
With the pressurizer otherwise inoperable be in at less:
HOT STAND 3Y with the reactor trip breche.rs open within 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> and in HOT SHUTDCUN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE PJ. QUI 1JJ1ENTS 4.4.4.1 The pressurizer water voluma chall be determined t: be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
- 4.4.4.2 The emergency power supply for 'the pressurizer heaters shall be deuonstrated OPERABLE at least once por 18 months by transferring p: var from the normel to the e,crgen'cy power supply and energizing the hen:crs.
- 4.4.4.3 The capacity of each of the above. required grou'ps of pressurizer.,
heaters shall be verified by measuring circuit current at least once per 92 days.
- Limit not applicable during either a THERMAL POUE.R ramp ghan;e in excess of 5% RATED TIIEm!AL POWER per minute or a DIEMIAL ?G%ER step change in ejcess of 10% of RATED THEPMAL POWER.
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- Specification effective 95 dgys following the return to power following the third refueling outage.
FARLF.Y-UNIT 1 3/4 4-6 AMENDMENT N0. 25
REACTOR COOLANT SYSTDi 3/4.4.4.a RELIEF VALVES LIMITING CONDITION FOR OPERATION
APPLICABILITY: MODES 1, 2, and 3.
ACTION:
With one or more PORV(s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either a.
restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve (s);
otherwise, be in at least HOT STAh*DBY-vichin the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
b.
With on'e or more, block valve (s) inoperable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the block valve (s) to OPERABLE status or close the block valve (s) and remove power from the block valve (s);
otherwise, be in at least HOT STANDBY within the'next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.
The provisions of Specification 3.0.4 are not applicable.
c.
SURVEILLANCE REQUIREMJNTS
- '4.4.4.a.1 Each PORV shall be denonstrated OPERABLE at least once per 18 months by performance of a CHANNEL CALIBRATION and operating the valve through one cycle of full travel.
- 4.4.4.a'.2 Each block valve shall be demonstrated OPERABLE at least once per 92 days by operating the valve through one complete cycle of full travel unless the block valve is closed with the' power removed in order to meet the ACTION requirements of a. above.
1
- Specification effective 90 days following the return to power following the third refueling outage.
- Implementation of this specification is deferred until resoluti6n of the Unit 1 Technical Specification upgrade, t
FARLEY-UNIT 1 3/4 4-6a AMENDMENT N0. 25 k
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W BASES 3/4.3.3.7 HIGH ENERCY LINE BREAK ISOLATION SENSORS The high energy. line break isolation sensors are designed to mit.i-gate the consequences of the discharge of steam and/or water to the affected room and other lines and systems contained therein. In addi-tion, the sensors will initiate signals that will alert the operator to bring the plant to a shutdown condition.
3/4.3.3.8 ACCIDENT MONITORING INSTRt?ENTATION g
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2he OPERABILITY of the accideut monitoring ins'trumentation ensures that sufficient information is availdble for selected plant paraceters to monitor and assess these variables following an accident.
3/4.3.3.9 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 stages. Prompt detection of fires will reduce the potential for damage *to safety related equipment and is an integral element in the overall facility fire protection program.
In the event that a portion of the fire detection instrumentation is inoperable, the establishment of frequent fire patrols in the affected areas is required to provide detection capability until the inoperable instrumentation'is restored to OPERABILITY.
e FARLEY-UNIT 1 B 3/4 3-4 AMENDMENT NO. 25
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4 REACTOR COOLANT SYSTEM BASES 3/4.4.2 and 3/4.4.3' SAFEIT VALVES (Continued) than the maximum surge rate resulting from a complete loss of load assuming no reactor trip until the first Reactor Protective Syste= 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 powek operated '
relief valves or steam dump valves.
Demonstration of the safety valves lif t setting will occur only during I
shutdown and will be performed in accordance with the provisions of
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Section XI of the ASFE Boiler and Pressure Vessel Code.
3/4.4.4 PRESSURIZER The limit on the maximum water vo'lume in the pressurizer assures that
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the parameter is maintained within the nor=al steady state envelope cf 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 thn parameter is restored to within its limit following
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expected transient operation. The maximum 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 nu=ber of pressurizer heaters be OPERABLE assures that the plant will be able to establish natural uirculation.
3/4.4.4.a REIIEF VAIVES (PORV'S)
The power operated relief valves and steam bubble function to relieve
.RCS pressure daring.all design transients up to and including the desigt step load decrease with steam dump. Operaton of the PORV's minimizes :e
. undesirable opening of the spring-loaded pres'surizer code safety valve 6.
Each PORV has a remotely operated 'olock valve to provide a positive s'.utoff capability should a relief valve become inoperable.
3/4.4.S ' STEAM CEPERATORS.
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The Surveilfance Requirements for Inspection of the steam generato tubes ensure that the structural integrity of this portion of the RCS will-be maintained.
The program for inservice inspection of steam generater
-tubes is based' on a modification of Regulatory Guide 1.83, Revision 1.
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Inservice inspection of steam generator tubing is essential in order to maintain surveillance of the conditions of the tubes in the event that there is evidence of sachanical damage or progressive degradation due to design, FARLEY, UNIT 1 B 3/4 4-2 AMENDMENT NO. 25 f _-
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REACTOR COOLAMT SYSTEM BASES 3/4.4.5 ~ STEAM CENERATORS '(Continued) manufacturing errors, or inservice conditions that. lead,to corrosion.
Inservice inspect
.n of steam generator tubing also provides a means of charac'terizing tht-nature and cause of any tube degradation so that corrective measures can be taken.
I The plant is expected to be operated in a manner such that the se:ondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes.
If the secondary coolant chemistry is not maintained.tithin these li::its, localized corrosion may likely result in stress corrosion cracking.
The extent of cracking during ' plant operation would be limited by the i
FARLEY-UNIT 1 3'3/4 4-2a AMENDMENT N0. 25
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s-6.0 ADMINISTRATIVE CONTROLS 6.1 RESPONSIBILITY 6.1.1 The Plant Manager shall be responsible for overall facility operation and shall delegate in writing the sue. cession to this responsibility during his absence.
6.2 ORGANIZATION OFFSITE 6.2.1 The offsite organizat. ion for facility management and technical support shall be as shown on Figure 6.2-1.
FACILITY STAFF 6.2.2 The Facility organization shall be as shown on Figure 6.2-2 and:
a.
Each on duty shift shall be composed of at least the minimum shift crew composition shown in Table 6.2-1.
b.
At least one licensed Operator shall be in the control room when fuel is in the reactor.
c.
At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.
d.
An individual qualified in radiation protection procedures shall be on site when fuel is in the reactor, e.
All CORE ALTERATIONS shall be directly supervised by either a licensed Sen!or Reactor Operator or Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsi-bilities during this operation.
f.
A Fire Brigade of at least 5 members shall be maintained onsite at all times. The Fire Brigade shall not include 3 members of the minimum shift crew tacessary for safe shutdown of the unit or any personnel reouiced for other essential functions during a fire emergency.
6.2.3 SHIFT TECHNICAL ADVISOR 6.2.3.1 The Shift Technical Advisor shall serve in an advisory capacity to the Shift Supervisor primarily in the assessment of accident and transient occurrences.
/mendment No. 4 25 FARLEY - UNIT 1 6-1
MINIMUM SHIFT CREW COMPOSITION #
LICENSE APPLICABLE MDES CATEGORY 1, 2, 3 & 4 5&6 SOL 1
1*
OL 2
1 Non-Licensed 2
1 STA la None
- Does not include the licensed Senior Reactor Operator or Senior Reactor Operator' Limited to Fuel Handling, supervising 00P.E ALTERATIONS.
- Shift crew compo'sition (including an individual qualified in radiatien protection procedures) may be less than the minimum req.irements for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accomodate unexpected absence of on duty shift crew members provided imediate action is taken to restore the shift crew composition to within the minimum requirements of Table 6.2-1.
a] Individual may fill the same position on Unit 2.
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4 FARLEY - UNIT 1 54 AMENDMENT NO. 25
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_6. 3 FAcTyTY STAFF QUALIFICATIONS -
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1 6.3.1 Each ee:-ber of the facility staff shall meet or c:::ced the mind :u:
O qualificat.fons of ANST N18.1-1971 for comparable positions and the supple ental requiraments specified in Sections A and C of Encionure 1 of ' the March 2*,1930 MRC Ictter to all licensecs, except for (1) the Cheatst-/ an:! Health Physi..s -
1 Supervisor who shall meet or exceed the qualifications. cf Regulatory Guide 1.8, Sepreeber 1975.
6.4 TPJilHII;G
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6.4.1 A retraining and replacement training program for the facility staff shall be maintained under the direction of the Training Supervisor and shall meet or exceed the requirements and recocaendttions of Section 5.5 of ANSI 1118.1-1971 and Appendix "A" of 10 CFR Part 55 I
6.4.2 A training program for the Fire Brigade shall ~ be maintained under
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the direction of the Training Supervisor and shall meet or exceed the requirements of Section 27 of the NFPA Code-1976, except for Fire Brigade training sessions which shall be held at least quarterly.
6.5 P.EVIElf AND AUDIT
- 6. 5.1 PLAMT ODERATIONS REVIEW COMMITTEE (PORC)
FUNCTION 1
}.
related to nuclear safSty.
6.5.1.1 The PORC shall function to advise the Plant Managar on all matters COMPOSITIOR 6.5.1.2 The PORC shall be composed of the:
Chairman:
Plant Manager Vice Chairman:
Assistant Plant !lanager Member:
Technical Superintendent liember:
Operations Superintendent Ne.?.ber: (Non-voting)
Plant Quality Assurance Er.gir.eer
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Member:
Maintenance Superintendent Al.TERf!ATES 6.5.1.3 All alternate members shall be appointed in writing by the PORC Chairman to serve on a temporary basis; heuever..no r.cre than one altar-nate shall participate as voting members in PORC activities at any or:e time.
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- The Minimur.: qualifications requirement for the Ch.cmistry and Health Physics Supervisor shall become effective when the initial incumbent in this position is replaced.
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FARLEY - UNIT T 6-5 AMENDMENT NO. 25 G*
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ADf11NTSTRATIVF CONTR0!.S
- 6. 8. 3 The following' programs shall be maini.ained:
a.
Primary Coolant Sources Outside Containment l
A prograta to reduce leakage from those portions of systems *cytside containment that could contain highly radioactive fluids during a serious traasient or accident tn as low as practical levels.
The i
systems include recirculation portions of the containment spray, safety injection and chemical and volume control systems, the we.ste gas system, the Reactor Coolant. sampling system, tia residual heat removal system, and the containment atmosphere sacpling system.
The program shall include the following.
(i) Preventative maintenance and periodic visual inspection recuire-ments, and (ii) Integrated leak test requirements for each system with the exception of the waste gas system and the containment at:r. spWe sampling system which are " snoop" test:'d at refueling cycle intervals.or less.
b.
In-Plant. Radiation Monitoring A program which vill ensure the capability to accurately detercine the airborne iodine concentration in certain plant areas where personnel may be present under accid.mt conditions.
This progran shall include the following:
(i) Training of personnel,
~
(ii) Procedures for monitoring, and (iii) Provisions for maintenance of sampling and analyses equipment.
c, Backup th?thod for Determining Subc,colina Marcin A progrcm which will ensure the capability to accurately monitor the Reactor Coolant System Subcooling margin.
This prograra shall include the training of personnel and the procedures for monitoring.
- Specification effective 90 days fellowing the retum to power following the third refueling outage.
FARLEY UNIT 1 6-13a AMENDMENT NO. 25 ae he.
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