ML20011F701
| ML20011F701 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 02/20/1990 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | Georgia Power Co, Oglethorpe Power Corp, Municipal Electric Authority of Georgia, City of Dalton, GA |
| Shared Package | |
| ML20011F702 | List: |
| References | |
| NPF-68-A-028, NPF-81-A-009 NUDOCS 9003070209 | |
| Download: ML20011F701 (12) | |
Text
_
/'
UNITED STATES 7
^
NUCLEAR RE@ULATORY COMMISSION i
,f msnmotow. o. c. mss
\\...../
j GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION l
MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA l
V0GTLE ELECTRIC GENERATING PLANT, UNIT 1 5
AMENDMENT TO FACILITY OPERATING LICENSE 7
Amendment No. 28 License No. NPF-68 1.
The Nuclear Regulatory Comission (the Comission) has found that:
The app (lication for amendment to the Vogtle Electric Generating Plant, A.
i Unit 1 the facility) Facility Operating License No. NPF-68 filed by the Georgia Power Company, acting for itself. Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated November 21, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
Thereisreasonableassurance(1)thattheactivitiesauthorizedby 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 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 Commission's regulations and all applicable requirements have been satisfied.
s t
P n*
grg-y w
w
-mm-+w
L i
I 2.
Accordingly, the license is hereby amended by page 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-68 is hereby amended to read as follows.
Technical Specifications and Environnental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 28 and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, are herehy incorporated into this license. GPC shall operate the i
facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall t
be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: February 20, 1990 n
9 4 -
F-
{
- l. !
'g UNITED STATES
- /
NUCLEAR REGULATORY COMMISSION n
g
\\,..... )!
a wasmwoTow. o. c. mss GEORGIA POWER COMPANY l
l OGLETHORPE POWER CORPORATION l
NUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA I
V0GTLE ELECTRIC GENERATING PLANT. UNIT 2' i
1 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No. 9 l
License No. NPF-81 1
1.
The Nuclear Regulatory Comission (the Comission) has found that:
1 A.
Theapp(licationforamendmenttotheVogtleElectricGeneratingPlant.
Unit 2 the facility) Facility Operating License No. NPF-81 filed by the Georgia Power Company, acting for itself Oglethorpe Power Corpo-ration, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated November 21, 1989, 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*
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.-
Thereisreasonableassurance(1)thattheactivitiesauthorizedby this arrendment can be conducted without endangering the health rnd safety of the public, and (ii) that such activities will be conoscted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; 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.
p
. \\
2.
Accordingly, the license is hereby amended by page 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-81 is hereby amended to read as follows:
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised I
throegh Amendment No. 9. and the Environmental Protection Plan contained in Appendix 8, both of which are attached hereto, are hereby incorporated into this license. GPC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
1 3.
This license amendment is effective as of its date of issuance and shall i
be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION l
(
l David B. Matthews, Director Project Directorate II-3 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Technical Specification Changes Date of Issuance: February 20, 1990 l
L
~
t l
~
V
t.
1 s
ATTACHMENT TO LICENSE AMENDMENT NO. 28
{
FACILITY OPERATING LICENSE NO. NPF-68 l,
AND LICENSE AKENDMENT NO. 9 FACILITY OPERATING LICENSE NO. NPF-81 i
i DOCKETS NOS. 50-424 AND $0-425 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the areas of change.
The corresponding overleaf pages are also provided to enintain document conpleteness.
Amended Page Overleaf Page 3/4 4-6 3/44-5 3/49-1 B3/4 4-1 B3/4 4-2 i
B3/4 9-1 B3/4 g-2 r
)
+
/
l j
REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS FILLED LIMITING CONDITION FOR OPERATION 3.4.1.4.1 At least one residual heat removal (RHR) train shall be OPERABLE and in operation *, and either:
i a.
Onc additional RHR train shall be OPERABLE **, or
}
b.
The secondary ~ side water level of at least two steam generators shall i
be greater than 17% of wide range (LI-0501, LI-0502, LI-0503, LI-0504).
3 APPLICABILITY:
MODE 5 with reactor coolant loops filled ***.
ACTION:
a.
With one of the RHR trains inoperable or with less than the required steam generator water level, immediately initiate corrective action j
to return the inoperable RHR train to OPERABLE status or restore the required steam generator water level as soon as possible, b.
With no RHR train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR train to operation.
)
SURVEILLANCE REQUIREMENTS I
4.4.1.4.1.1 The secondary side water level of at least two steam generators when required shall be determined to be within limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.4.1.2 At least one RHR train shall be determined 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 />.
"The RHR pump may be deenergized 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.
- 0ne RHR train may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR train is OPERABLE and in operation.
- A reactor coolant pump shall not be started unless the secondary water temperature of each steam generator is less than 50'F above each of the l
Reactor Coolant System cold leg temperatures.
i V0GTLE UNITS - 1 & 2 3/4 4-5
~ _.
~k l
REACTOR COOLANT SYSTEM COLD SHUTDOWN - LOOPS NOT FILLED LIMITING CONDITION FOR OPERATION l
)
3.4.1.4.2 Two residual heat removal (RHR) trains shall be OPERABLE
- and at I
least one RHR train shall be in operation.** Reactor Makeup Water Storage Tank (RWST) discharge valves (1208-04-175,1208-04-176#,1208-04-177# and j
i l
1208-04-183) shall be closed and secured in position.
APPLICABILITY: MODE 5 with reactor coolant loops not filled.
ACTION:
With less than the above required RHR trains OPERABLE, immediately I
a.
initiate corrective action to return the required RHR trains to OPERABLE status as soon as possible.
b.
With no RHR train in operation, suspend all operations involving a i
reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR train to operation.
With the Reactor Makeup Water Storage Tank (RWST) discharge valves c.
(1208-U4-175, 1208-U4-176#, 1208-04-177#, and 1208-U4-183) not closed l
and secured in position, immediately close and secure in position the RWST discharge valves.
SURVEILLANCE REQUIREMENTS 4.4.1.4.2.1 At least one RHR train shall be determined 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 />.
4.4.1.4.2.2 Valves 1208-U4-175, 1208-U4-176#, 1208-U4-177#, and 1208-U4-183 l
shall be verified closed and secured in position by mechanical stops at least once per 31 days.
- 0ne RHR train may be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing provided the other RHR train is OPERABLE and in operation.
- The RHR pump may be deenergized 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 l
below saturation temperature.
i
- RWST discharge valves 1208-U4-176 and 1208-U4-177 may be open under administrative control provided the Reactor Coolant System is in compliance with the SHUTDOWN MARGIN requirements of Specification 3.1.1.2 and the high flux at shutdown alarm is OPERABLE with a setpoint of 2.30 times background in accordance with Note 9 of Table 4.3-1.
V0GTLE UNITS - 1 & 2 3/4 4-6 Amendment No. 28 (Unit 1)
Amendment No. 9 (Unit 2)
i u
3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION LIMITING CONDITION FOR OPERATION j
3.9.1 The boron concentration of all filled portions of the Reactor Coolant System and the refueling canal shall be maintained uniform and sufficient to ensure that the more restrictive of the following reactivity conditions are met:
A K,ff of 0.95 or less, or a.
b.
A boron concentration of greater than or equal to 2000 ppe, r
Additionally, valves 1208-04-175, 1208-04-177#, 1208-U4-183, and 1208-04-176#'
l shall be closed and secured in position.
APPLICABILITY: MODE 6.
ACTION:
With the requirements of a. and b. above not satisfied, immediately a.
suspend all operations involving CORE ALTERATIONS or positive reactivity changes and initiate and continue boration at greater than or equal to 30 gpm of, a solution containing greater than or equal to 7000 ppm boron or its equivalent until K,ff is reduced to less than or equal to 0.95 or the boron concentration is restored to greater r
than or equal to 2000 ppm, whichever is the more restrictive.
b.
With valves 1208-U4-175, 1208-U4-177#, 1208-U4-183. and 1208-U4-176#
l not closed and secured in position, immediately close and secure in position.
SURVEILLANCE REQUIREMENTS k
- 4. 9.1.1 The boron concentration of the Reactor Coolant System and the refueling canal shall be determined by chemical analysis at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
- 4. 9.1. 2 Valves 1208-U4-175, 1208-U4-177#, 1208-U4-183, and 1208-U4-176# shall l
be verified closed and secured in position by mechanical stops at least once u
i per 31 days.
- RMWST discharge valves 1208-U4-176 and 1208-U4-177 may be open under administrative control provided the Reactor Coolant System is in compliance with the requirements of Specification 3.9.1 and the high flux at shutdown alarm is OPERABLE with a setpoint of 2.30 times background.
For the purpose of this Specification, the high flux at shutdown alarm will be demonstrated OPERABLE pursuant to Specification 4.9.2.
V0GTLE UNITS - 1 & 2 3/4 9-1 Amendment No.28 (Unit 1)
Amendment No.9 (Unit 2) 1 l
Y,
)/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 antici-pated 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 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by opening the Reactor Trip System breakers.
In MODE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR train provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at least two trains / loops (either RHR or RCS) be OPERABLE.
In MODE 5 with reactor coolant loops not filled, a single RHR train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RHR trains be OPERABLE.
The locking closed of the required. valves, except valves 1208-U4-176 and 1208-U4-177 for short periods of time to maintain chemistry control, in Mode 5 (with the loops not filled) precludes the possibility of uncontrolled boron dilution of the filled portion of the Reactor Coolant System. These actions prevent flow to the RCS of unborated water in excess of that analyzed.
These limitations are consistent with the initial conditions assumed for the boron dilution accident in the safety analysis.
The operation of one reactor coolant pump (RCP) 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 an RCP with one or more RCS cold legs less than or equal to 350*F are provided to prevent RCS pressure transients, caused by energy additions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50.
The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50'F above each of the RCS cold leg temperatures.
.V0GTLE UNITS - 1 & 2 B 3/4 4-1 Amendment No. 28 (Unit 1)
Amendment No. 9 (Unit 2)
-m
-,-i.
-=
-,,,,---w
~-,,,.
-+___,,_w m m
__.,_,____,______.,,._,_.,____u__
i L
BASES r
3/4.4.2 SAFETY VALVES i
The pressurizer Code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.
Each safety valve is designed to relieve 420,000 lbs per hour of saturated steam at the valve Setpoint.
The relief capacity of a single safety valve is adequate to relieve any overpressure condition which could occur during shutdown.
i are OPERABLE In the event that no safety valves reliefcapabIlityandwillpreventRCSoverpressurization.an operating RHR train, c In addition, the Cold Overpressure Protection System provides 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 maximum surge rate resulting from a complete loss-of-load assuming no Reactor trip until the first Reactor Trip System Trip Setpoint 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.
During shutdown conditions in Mode 5 only one pressurizer code safety is required for overpressure protection.
In lieu of an actual operable code safety valve an unisolated and unsealed vent pathway of equivalent size can be taken as synonymous w(i.e. a direct unimpaired opening) ith an OPERABLE code safety.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
3/4.4.3 PRESSURIZER The 12-hour periodic surveillance is sufficient to ensure that the param-eter is restored to within its limit following expected transient operation.
1 The maximum water volume 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 OlERABLE enhances the capability of the plant to control Reactor Coolant System n essure and establish natural circulation.
1 l
V0GTLE UNITS - 1 & 2 B 3/4 4-2 i
i I
w
--m
-v,--.-
w e-,--
~
3/4.9s REFUELING OPERATIONS 1
l BASES I
3/4.9.1 BORON CONCENTRATION The limitations on reactivity conditions during REFUELING ensure that:
(1) the reactor will remain suberitical during CORE ALTERATIONS, and (2) a uniform boron concentration is maintained for reactivity control in the water volume having direct access to the reactor vessel.
The locking closed of the required valves, except valves 1208-U4-176 and 1208-U4-177 for short periods of time to maintain chemistry control, during refueling operations precludes f
the possibility of uncontrolled boron dilution of the filled portions of the j
These actions prevent flow to the RCS of unborated water in excess of that analyzed.
These limitations are consistent with the initial conditions assumed for the Boron Dilution Accident in the safety The Boron concentration value of 2000 ppm or greater ensures a K j
analysis.
of 0.95 or less and includes a conservative allowance for calculational df uncertainties of 100 ppm of boron.
3/4.9.2 INSTRUMENTATION The OPERABILITY of the Source Range Neutron Flux Monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
3/4.9.3 DECAY TIME The minimum requirement for reactor suberiticality prior to movement of irradiated tval assemblies in the reactor vessel ensures that sufficient time het r%Psed to allow the radioactive decay of the short-lived fission product.$, ibis decay time is consistent with the assumptions used in the safety w 4 4as.
3/4.9.4 CONTAINMENT BUILDING PENETRATIONS l
The requirements on containment building penetration closure and OPERABILITY ensure that a release of radioactive material within containment will be restricted from leakage to the environment.
The OPERABILITY and closure restrictions are sufficient to restrict radioactive material release from a fuel element rupture based upon the lack of containment pressurization potential while in the REFUELING MODE.
3/4.9.5 COMMUNICATIONS The requirement for communications capability ensures that refueling station personnel can be promptly informed of significant changes in the facility status or core reactivity conditions during CORE ALTERATIONS.
V0GTLE UNITS - 1 & 2 B 3/4 9-1 Amendment No.28 (Unit 1)
Amendment No.9 (Unit 2)
F i
i i
REFUELING OPERATIONS BASES 3/4.9.6 REFUELING MACHINE The OPERABILITY requirements of the refueling machine and auxiliary hoist ensure th6t:
(1) The refueling machine will be used for the movement of fuel assemblies and/or rod control cluster assemblies (RCCA) or thimble plug assemblies, and the auxiliary hoist will be used for the movement of control rod drive shafts.
(2) the refueling machine will have sufficient load capacity to lift a fuel assembly and/or a rod control cluster assembly or thimble plug assembly, and the auxiliary hoist will have sufficient load capacity to lift a control rod drive shaft and attached RCCA, and i
(3) the core internals and reactor vessel are protected from excessive lifting force in the event they are inadvertently engaged during lifting operations.
L 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE AREAS The restriction on movement of loads in excess of the nominal weight of a fuel and control rod assembly and associated handling tool over other fuel assemblies in the storage pool ensures that in the event this load is dropped:
(1) the activity release will be limited to that contained in a single fuel assembly, and (2) any possible distortion of fuel in the storage racks will not result in a critical array.
release assumed in the safety analyses.This assumption is consistent with the activ 3/4.9.8 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION The requirement that at least one residual heat removal (RHR) train be in operation ensures that:
(1) sufficient cooling capacity is available to remove decay heat and maintain the water in the reactor vessel below 140'F as required during the REFUELING MODE, and (2) sufficient coolant circulation is maintained through the core to minimize the effect of a boron dilution incident and prevent boron stratification.
The requirement to have two RHR trains OPERABLE when there is less than 23 feet of water above the reactor vessel flange ensures that a single failure of the operating RHR train will not result in a complete loss of residual heat removal capability. With the reactor vessel head removed and at least 23 feet of water above the reactor pressure vessel flange, a large heat sink is avail-able for core cooling.
Thus, in the event of a failure of the operating RHR train, adequate time is provided to initiate emergency procedures to cool the core.
V0GTLE UNITS - 1 & 2 B 3/4 9-2
--