ML20045E043
| ML20045E043 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 06/24/1993 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20045E044 | List: |
| References | |
| NUDOCS 9307010024 | |
| Download: ML20045E043 (12) | |
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' UNITED STATES e
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NUCLEAR REGULATORY COMMISSION I
WASHINGTON. D.C. 20565 0001
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GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON. GEORGIA DOCKET N0. 50-321 EDWIN 1. HATCH NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 186 License No. DPR-57 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment to the Edwin I. Hatch Nuclear Plant, Unit 1 (the facility) Facility Operating License No. DPR-57 filed by the Georgia Power Company, acting for itself, Oglethorpe Power Corporation, Municipal Electric Authority of Georgia, and City of Dalton, Georgia (the licensees), dated November 18, 1991, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended.(the Act), and the Commission's rules and regulations as set forth in 10 CFR Chapter I; i
f 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 (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 set forth in 10 CFR Chapter I; D.
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
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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. DPR-57 is hereby amended to' read as follows:
r Technical Soecifications The Technical Specifications contained in Appendices A and B, asL revised through Amendment No. 186, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective no later than 60 days from the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION i
David B. Matthews, Director Project Directorate 11-3 i
Division of Reactor Projects - I/II l
Office of Nuclear Reactor Regulation l
Attachment:
Technical Specification Changes Date of Issuance:
June 24, 1993 i
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ATTACHMENT TO LICENSE AMENDMENT N0.186 i
j FACILITY OPERATING LICENSE NO. DPR-57 i
j DOCKET NO. 50-3.21 i
Replace the following pages of the Appendix "A" Technical Specifications with i
the enclosed pages. The revised pages are identified by Amendment number.and j
contain vertical lines indicating the areas of change.
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Remove Paaes Insert Paces 11 Lii-i 3.2-1 3.2-1 3.2-23b 3.2-23b i
3.2-23c 3.2-23c j
3.2-49a 3.2-49a 3.2-49b.
3.2-49b 3.2-68 3.2-68 i
a 3.2-68a 3.2-68a 3
3.9-12 3.9-12 i
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LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS i
3.2.
PROTECTIVE INSTRUMENTATION (CONT')
4.2.
PROTECTIVE INSTRUMENTATION (CONT') 3.2-1 E.
Instrumentation Which E.
Instrumentation Which 3.2-1 l
Initiates or Controls the Initiates or Controls the LPCI Mode of RHR LPCI Mode of RHR F.
Instrumentation Which F.
Instrumentation Which 3.2-1 Initiates or Controls Initiates or Controls Core Spray Core Spray G.
Neutron Monitoring G.
Neutron Monitoring 3.2-1 Instrumentation Which Instrumentation Which Initiates Control Rod Initiates Control Rod Blocks Blocks H.
Radiation Monitoring H.
Radiation Monitoring 3.2-1 l
Systems Which Limit Radio-Systems Which Limit Radio-activity Release activity Release l
I.
Instrumentation Which Ini-I.
Instrumentation Which Ini-
.3.2-1 t1ates Recirculation Pump tiates Recirculation Pump Trip Trip J.
Instrumentation Which Mon-J.
Instrumentation Which Mon-3.2-1 itors Leakage Into The itors Leakage into The Drywell Drywell K.
Instrumentation Which K.
Instrumentation Which
'3.2-1' Provides Surveillance Provides Surveillance Information Information L.
Degraded Station Voltage L.
Degraded Station Voltage
.3.2-1 l
Protection Instrumentation Protection Instrumentation M.
(Deleted)
M.
(Deleted) 3.2-1 N.
Instrumentation Which Arms N.
Instrumentation Which Arms 3.2-1 Low Low Set S/RV System Low Low Set S/RV System I
3.3.
REACTIVITY CONTROL 4.3.
REACTIVITY CONTROL 3.3-1 A.
Core Reactivity Margin A.
Core Reactivity Margin 3.3-1 B.
Operable Control Rod 3.3-1 Exercise Requirements C.
Control Rod Drive System C.
Control Rod Drive System 3.3-2 D.
Minimum Count Rate for D.
Minimum Count Rate for 3.3-4 Rod Withdrawal-Rod Withdrawal E.
Rod Worth Inventory E.
Rod Worth Inventory 3.3-4 Determination Determination F.
Operation With a Limiting F.
Operation With a Limiting 3.3-5 Control Rod Pattern Control Rod Pattern dd n
HATCH - UNIT 1 11 Amendment 100. 186
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LIMITING CONDITIONS FOR 00ERATION SURVEILLANCE REQUIREMENTS 3.2 PROTECTIVE INSTRUMENTATION 4.2 PROTECTIVE INSTRUMENTATION 1
Acolicability Acolicability The Limiting Conditions for Operation The Surveillance Requirements apply to the plant instrumentation apply to the instrumentation which performs a protective function.
which performs a protective function.
Obiective Obiective The objective of the Limiting Condi-The objective of the Surveillance tions for Operation is to assure the Requirements is to specify the type operability of protective instrumen-and frequency of surveillance to tation.
be applied to protective instru-mentation.
Soecifications Soecifications The Limiting Conditions for Operation The check, functional test, and of the protective instrumentation af-calibration minimum frequency for fecting each of the following protec-protective instrumentation affect-tive actions shall be as indicated in ing each of the following protec-the corresponding LCO table.
tive actions shall be as indicated -
in the corresponding SR table.
Protective Action LCO Table SR Table A. Initiates isolation Actuation 3.2-1 4.2-1 B. Initiates or Controls HPCI 3.2-2 4.2-2 C. Initiates or Controls RCIC 3.2-3 4.2-3 D. Initiates or Controls ADS 3.2-4 4.2-4 E. Initiates or Controls the 3.2-5 4.2-5 LPCI Mode of RHR F. Initiates or Controls Core
-3.2-6 4.2-6 Spray G. Initiates Control Rod Blocks 3.2-7 4.2-7 H. Limits Radioactivity Release' 3.2-8 4.2-8
!. Initiates Recirculation Pump 3.2-9 4.2-9 Trip J. Monitors Leakage Into the 3.2-10 4.2-10 Drywell K. Provides Surveillance 3.2-11 4.2-11 Information L. Degraded Station Voltage 3.2-12 4.2-12 Protection Instrumentation M. (Deleted) 3.2-13 4.2-13 N. Arms the Low Low Set S/RV 3.2-14 4.2-14 System HATCH - UNIT 1 3.2-1
- k ndment b. W Ja 8 bg
TABLE 3.2-12 a
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DEGRADED STATION VOLTAGE PROTECTION INSTRUMENTATION 8
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E Action to be Taken Q
Required Channels if the Number of Ref. No, Instrurnent Operable Required Trip Setting Required Operable
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Channele To Tno Channels le Not Met 1
4.16 kw Emergency Bue 2/ Bus 2/Bue greater then or equel to 2800 (c)
Undervoltage Reley volte. At 2800 volte time deley (t.oes of Voltage will be less then or equel to Corutitiord 6.5 sec.
2 4.16 kw Emergency Bus 2/ Bus 2/ Bus greater then or equel to 3280 (c)
Undervoltage Relay waste. At 3280 volte time deley (Oegraded Voltage will be less then or equel to Conditiord 21.5 sec.
Y NOTES FOR TABLE 3.2-12 w
7 e.
The column enthled 'Ref. No.* is only for convermence so that e one-temne relationship con be established between home in Table 3.2-12 are items in Table 4.2-12.
b.
This instrumenteelon le repaired to be operable during toector etettup. power operation, and hot etwatdown.
o.1 With the number of operable channele one lose then the required operable channels, operation may proceed uned perfeemence of the next requwod instrument functional test provided a trip signalis pieced in the LOSP lock-out relay logic for the -,
_l'- inoperable channel r+
o.2. One instrument channel may be Inoperable for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to perform required surveillences pdor to entedng other appecable actione.
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I 44 HATCH - UNIT 1 3.2-23c Amen,d, ment No. 186 i
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h TABt.E 4.2-12
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' DEGRADED STATION VOLTAGE PROTECTION INSTRUMENTATION
- 1.
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Instrument Functional Instrument
- a:y Ref. No.
Instrument instrument Check Test Minimum Calibretion 1
lb)
Mn6 mum Freauency Frequenev Mirumum Freauency 1
4.16 kV Emergency Bue N/A Oncehnonth once/ operating Undervoltage Reley cycle (Loos of Voltage Condition) 2 4.16 kV Emergency Bue N/A Oncehnonth Once/ operating Undervoltage Reley cycle (Degraded Voltage Condition) w NOTES FOR TABLE 4.2-12 s
The column entided *Ref. No."le ordy for converwence so that e one-to-one relationship con be established between items in Table 3.2-12 end items in Table 4.2-12.
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Surveillence of tNo instrumentation le required during reactor startup, power operation, and hot shutdown.
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HATCH - UNIT 1 3.2-49b heendment No.186
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BASFS FOR LIMITING rnunlTIONS FOR OPERATION 3.2.J.4.
Scintillation Detector For Monitorino Radiciodine (Continued) level reading is indicative of a leak in the nuclear system process.
i barrier in the primary containment. A sample.that is continuously drawn
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from the primary containment is collected on an iodine filter and monitored by a gamma sensitive scintillation detector. Radiation levels-are read out by a log rate meter and recorded on a. strip chart located in the control room. A high radiation level alam and a. failure alarm are also provided and are annunciated in the control room. Also, a high-low flow alam is annunciated in the control room.
j 5.
GM Tubes for Monitorina Noble Gases.
A set of GM tubes contained in an instrument rack are used to monitor the release of noble gases in the drywell and torus. A high radiation level reading is indicative of a leak in the nuclear system process barrier in the primary containment. A sample that's continuously drawn from the.
l primary containment is passed through a shielded saeple chamber which
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contains the beta sensitive GM tubes. Radiation levels are read out by a l
log rate meter and recorded on a strip chart located in the control room.
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A high radiation level alam and failure alam are provided and are -
j annunciated in the control room. Also, a high-flow alam is annunciated
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in the control room.
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Instrumentation Which Provides Surveillance Infomation'(Table 3.2-11) i i
For each parameter monitored, as listed in Table 3.2-11, there are two.
I channels of instrumentation except for the control rod positions indicating.
system and the Post-Accident Effluent Monitors..- By comparing readings between the two channels, a near continuous surveillance of instrument -
perfomance is available. - Any significant deviation in readings will
-i initiate an early recalibration, thereby maintaining the quality of the' instrument readings, j
The hydrogen and oxygen analyzing systems consist of two redundant, separate l
systems and are each capable of analyzing the hydrogen and oxygen content of.
l the drywell-torus simultaneously. They are designed to be completely-
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testable at both the analyzer rack and in the control room. With an oxygen i
concentration of less than 45 by volume,' a flamaable mixture with hydrogen l
1s not possible.
L.
Dearaded Station Voltaos Protection Instrumentation (Table 3.2-12) l.
The undervoltage relays shall automatically initiate the disconnection of l
offsite power sources whenever the voltage'setpoint and time delay limits have been exceeded. This action shall provide voltage protection for the emergency power systems by preventing sustained degraded voltage conditions due to the offsite power source and interaction between the offsite and-onsite emergency power systems. The undervoltage relays have a time delay characteristic that provides protection against both a loss of voltage and degraded voltage condition and thus minimizes the effect of short duration disturbances without exceeding the maximum time delay, including margin, that is assumed in the FSAR accident analyses.
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HATCH - UNIT.1 3.2-68 Amendment No.186 "
-.r RASES FOR LIMIi1NG CONDITIONS FOR OPERATION M. (Deleted)
N.
Instrumentation Which Arms low Low Set System (Table 3.2-141 The bases for these trip f;nctions are found in the bases for Section 3.6.H, page 3.6-21, 3.2.1 References 1.
FSAR Appendir G Plant Nuclear Safety Operational Analysis i
2.
FSAR Section 7.3, Primary Containment and Reactor Vessel Isolation Control System 3.
FSAR Section 14, Plant Safety Analysis 4.
FSAR Section 6, Core Standby Cooling Systems S.
FSAR Section 14.4.4, Refueling Accident 6.
FSAR Section 6.5.3, Integrated Operation of the Core Standby Cooling Systems 7.
FSAR Section 6.5.3.1, Liquid Line Breaks 8.
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HATCH - UNIT 1 3.2-68a Amendment No.186
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BASES FOR LIMITING CONDITIONS FOR OPERATION M. (Deleted)
N.
Instrumentation Which Arms low Low Set System (Table 3.2-14)
The bases for these trip functions are found in the bases for Section 3.6.H.
page 3.6-21.
3.2.1 References 1
1.
FSAR Appendix G, Plant Nuclear Safety Operational Analysis I
2.
FSAR Section 7.3, Primary Containment and Reactor Vessel Isolation.
Control System 3.
FSAR Section 14, Plant Safety Analysis i
4 FSAR Section 6, Core Standby Cooling Systems l
S.
FSAR Section 14.4.4, Refueling Accident 6.
FSAR Section 6.5.3, Integrated Operation of the Core Standby Cooling Systems 7.
FSAR Section 6.5.3.1, Liquid Line Breaks 8.
- w 44 HATCH - UNIT 1 3.2-6sa Amendment No. 186.
BASES FOR SURVEILLANCE REQUIREKNTS 4.9.A.2.e.
Fuel 011 Transfer Ptses Following the mnthly test of the diesels, the fuel oil transfer p(mps shall be operated to refill the day tank and tu check the operation of these ptstps, 3.
125/250 Volt DC Ememency Power System (Plant Batteries IA and 18)
The plant batteries may deteriorate with time, but precipitous j
failure is unlikely. The type of surveillance described in this specification is that which has been demonstrated through experience to provide an indication of a cell becoming irngular or inoperable long before it fails.
j 4.
Ememency 4160 Volt Buses (IE.1F. and IG)
The emergency 4160 volt buses (lE, IF, and IG) are monitored to assure readiness and capability of transmitting power to the emergency load.
These buses distribute AC power to the mquired engineered safety feature equipment. The nomal feeds and backup to the emergency buses (IE,IF,andIG)aretakenfromthestartupauxiliary transfomers. If neither startup auxiliary transfomer is j
available, buses IE, IF, and IG will be energized from the standby diesel generators.
i 5.
Ememency 600 Volt Buses (IC and ID)
The emergency 600 volt buses (IC and ID) are monitored to assure readiness and capability of transmitting the emergency load.
- 6. Ememency 250 Volt DC to 600 Volt AC Inverters The emergency 250 volt DC to 600 volt AC inverters are mnitored to. assure readiness and capability of transmitting power to the emergency loads.
- 7. Loaic Systems The periodic testing of the logic systems will verify the ability of the logic systems to bring the auxiliary electrical systens to running standby readiness with the presence of an accident signal and/or a degraded voltage or LOSP signal.
The periodic testing of the relays which initiate energization of the emergency buses by the diesel generators when voltage is lost on the emergency buses will verify operability of these relays.
l The periodic simlation of accident signals will confinn the ability of the 600 volt load shedding logic system to sequentially shed and restart 600 volt loads if an accident signal were present and diesel generator voltage were the only source of electrical power.
D. RPS M3 Sets The surveillance requirements for the RPS power supply equipment will ensure the timely detection of potential component failures that might be caused by a sustained over-voltage or under-voltage corditions.
E. References 1.
" Proposed IEEE Criteria for Class IE Electric Systems for Nuclear PoyerGeneratingStations"(IEEEStandardNo.308), June,1969.
N
- 2. Anerican Society for Testing and Materials,1970 Annual Book of ASPI Standards. Part 17.
HATCH - UNIT 1 3.9-12 AmendmerTD No.186
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