ML20214P354
| ML20214P354 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 11/24/1986 |
| From: | Muller D Office of Nuclear Reactor Regulation |
| To: | City of Dalton, GA, Georgia Power Co, Municipal Electric Authority of Georgia, Oglethorpe Power Corp |
| Shared Package | |
| ML20214P356 | List: |
| References | |
| DPR-57-A-133, GL-86-010, NPF-05-A-070, TAC 60727, TAC 60730, TAC 60731, TAC 62123, TAC 62124 NUDOCS 8612040169 | |
| Download: ML20214P354 (30) | |
Text
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UNITED STATES 8
NUCLEAR REGULATORY COMMISSION o
. B I~
wasHWGTON, D. C. 20885 k.....
GEORGIA POWER COMPANY.
OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-321 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No. 133 License No. DPR-57 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Georgia Power Company, et a1.,
(thelicensee) dated July 25, 1986, 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; e
i B.
The facility will operate in conformity with the application, the provisions of the Act,-and the rules and regulations of the
, p Comission; t'
4 3,
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 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 l
public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 j
of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical i
Specifications as indicated in the attachment to this 1tcense amendment, i
and paragraph 2.C.(2) of Facility Operating License No. DPR-57 is hereby j
amended to read as follows:
1 I
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I (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.133, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The licensee is further amended by replacing page 4 to revise paragraph 2.C.(3).
4.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMI SION k//
Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing Attachments:
1.
Changes to the Technical Specifications 2.
Page 4 of license l
Date of Issuance:
November 24, 1986 6
4 i
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ATTACHMENT TO LICENSE AMENDMENT NO. 133
. FACILITY OPERATING LICENSE NO. DPR-57 DOCKET N0. 50-321 i
Replace the following pages of the Appendix A Technical Specifications with the enclosed pages..The revised pages are indicated by marginal, lines.
1 Pages V
1.0-7 3.13-1 thru 3.13-12 6-8 6-18 o
c>
J T
g i
j f
b
t LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REQUIREMENTS
.3.g.
AUXILIARY ELECTRICAL SYSTEMS (CONT'0) 4.9.
AUXILIARY ELECTRICAL f.YSTEMS (CONT'0) 8.
Requirements for Continued 8.
Requirements for Continued 3.9 -4 Operation With Inoperable Components Operation Without Inoperable Components C.
Diesel Generator Requirements 3.9-6 (Reactor in the Shutdown or Refuel Mode) 3.10.
REFUELING 4.10.
REFUELING 3.10-1 A.
Refueling Interlocks A.
Refueling Interlocks 3.10-1 8.
Fuel Loading 3.10-1 C.
Core Monitoring During Core C.
Core Monitoring During Core 3.10-2 Alternations Alternations D.
Spent Fuel Pool Water Level D.
Spent Fuel Pool Water Level 3.10-2 E.
Control Rod Drive Maintenance E.
Control Rod Crive Maintenance 3.10-2 F.
Reactor Building Cranes F.
Reactor Building Cranes 3.10-4 6.
Spent Fuel Cask Lifting 6.
Spent Fuel Cask Lifting 3.10-5 Trunnions and Yoke Trunnions and Yoke H.
Time Limitation 3.10-5 3.11.
FUEL RODS 4.11.
FUEL RODS 3.11-1 A.
Average Planar Linear Heat A.
Average Planar Linear Heat 3.11-1 Generation Rate (APLH6R)
Generation Rate (APLH6R) 8.
Linear Heat Generation Rate 8.
Linear Heat Generation Rate 3.11-1 (LH6R)
(LH6R)
C.
Minimum Critical Power C.
Minimum Critical Power 3.11-1 Ratio (MCPR)
Ratio (MCPR) 3.12.
MAIN CONTROL ROOM ENVIRONMENTAL 4.12.
MAIN CONTROL ROOM ENVIRONMENTAL 3.12-1 SYSTEM SYSTEM A.
Ventilation System Operability A.
Ventilation System Tests 3.12-1 8.
Isolation Valve Operability 8.
Isolation Valve 3.12-2 and Closing Time Testing C.
Radiation Monitors C.
Radiation Monitors 3.12-3 0.
Shutdown Requirements 3.12-3 E.
Chlorine Monitors E.
Chlorine Monitors 3.12-3 HATCH - UNIT 1 v
Amendment No. $0, 133
s MM. Minimum Critical Power Ratio (MCPR) - Minimum Critical Power Ratio (MCPR) is the value of the critical power ratio associated with the most limiting assembly in the reactor core. Critical Power Ratio (CPR) is the ratio of that power in a fuel assembly, which is calculated to l
cause some point in the assembly to experience boiling transition, to the actual assembly operation power.
NN.
Trip System - A trip system means an arrangement of instrument channel trip signals and auxiliary equipment required to initiate action to accomplish a protective function. A trip system may require one or more instrument channel trip signals related to one or more plant parameters in order to initiate trip system action.
Initiation of protective action may require the tripping of a single trip system or the coincident tripping of two trip systems.
00.
(Deleted) l PP.
(Deleted) r QQ. Channel Calibration - A Channel Calibration is the adjustment, as l
necessary, of the channel output such that it responds with the i
necessary range and accuracy to known values of the parameter which the I
channel monitors. The Channel Calibration shall encompass the entire channel including the sensor and alarm and/or trip functions, and shall include the Channel Functional Test. The Channel Calibration may be performed by any series of sequential, overlapping or total channel steps such that the entire channel is calibrated.
2 RR.
Channel Functional Test - A Channel Functional Test shall be:
Analog channels - the injection of a simulated signal into the a.
channel as close to the primary sensor as practicable to verify operability including alarm and/or trip functions.
b.
81 stable Channels - the injection of a simulated signal into the channel sensor to verify operability including alarm and/or trip functions.
SS. Fraction of Limiting Power Density (FLPD) - the ratio of the linear heat generation rate (LH6R) existing at a given location to the design LHGR for the bundle type. Design LNGRs are 18.5 KW/ft for 7x7 bundles and 13.4 KW/ft for 8x8 bundles.
TT. Core Maximum Fraction of Limiting Power Density (CMFLPD) - the CMFLPD is the highest value existing in the core of the FLPD.
Amendment No. 59, 73, 123, 133 HATCH - UNIT 1 1.0-7
+
l (These pages are intentionally-left blank.)
HATCH - UNIT I 3.13-1 through 3.13-12 Amsndment No. $@, 133
ADMINISTRATIVE CONTROLS RESPONSIBILITIES (Continued) 1.
Review of the Security Plan ar.d implementing procedures and shall submit recomended changes to the SRB.
- j. Review of the Emergency Plan and implementing procedures and shall submit recomended changes to the SRB.
k.
Review of any unplanned onsite release of radioactive material to the environs when such release is in excess of I C1, excluding dissolved and entrained gases and tritium for liquid effluents, and in excess of 150 Ci of noble gases or 0.02 C1 of radiciodines for gaseous effluents. Also included is the preparing and forwarding to the General Manager-Plant Hatch and the SR8 reports covering i
evaluation, recommendations and disposition of the corrective action to prevent recurrence.
1.
Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL (ODCM).
m.
Review of proposed change (s) to plant systems and equipment to determine whether the proposed change has a potential radiological environmental impact. Such change (s) will be reported to the Manager-Nuclear Engineering and Chief Nuclear Engineer.
n.
Review of the Fire Protection Program and implementing procedures and shall submit recommended changes to the SR8.
AUTHORITY 6.5.1.7.
The PR8 shall:
a.
Recomend in writing to the General Manager-Plant Hatch approval or
],
disapproval of items considered under 6.5.1.6(a) through (d) above.
~
b.
Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
l c.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Manager of Nuclear Generation or the Vice President and General Manager l
Nuclear Operations and the Safety Review 8oard of disagreement j
between the PRB and the General Manager-Plant Match; however, the General Manager-Plant Hatch shall have responsibility for resolution of such disagreements pursuant to 6.1.1. above.
RECORDS 6.5.1.8.
The Plant Review Board shall maintain written minutes of each PR8 meeting that, at a minimum, document the results of all PR8 activities performed under the responsibility and authority provisions of these Technical Specifications. Copies shall be provided to the Manager of 4
Nuclear Generation or Vice President and General Manager Nuclear Operations and the Safety Review Board.
I HATCH - UNIT 1 6-8 Am:ndment No. O, 109, 133
ADMINISTRATIVE CONTROLS d.
Abnormal degradation of systems other than those specified in 6.9.1.8.c. above designed to contain radioactive material resulting f rom the fission process.
SPECIAL REPORTS 6.9.2.
Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified and for each activity shown in Table 6.9.2-1.
Special reports for fire protection equipment operating and surveillance requirements shall be submitted, as required, by the Fire Hazards Analysis and its Appendix B requirements.
6.10.
RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
6.10.1.
The following records shall be retained for at least five years:
a.
Records and logs of unit operation covering time interval at each power level, b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c.
ALL REPORTABLE OCCURRENCES submitted to the Commission.
d.
Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
1 e.
Records of changes made to the procedures required by Specification 6.8.1.
f.
Records of radioactive shipments.
g.
Records of sealed source and fission detector leak tests and results.
h.
Records of annual physical inventory of all sealed source material of record.
6.10.2.
The following records shall be retained for the duration of the unit Operating License:
a.
Records and drawing changes reflecting unit design modifi-cations made to systems and equipment described in the Final Safety Analysis Report, b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
HATCH - UNIT 1 6-18 Amendment Nm (& R32
> (3) Georgia Power Company shall implement and maintain in effect all provisions of the fire protection program, which is referenced in the Final Safety Analysis Report for the facility, as contained in the
. updated Edwin I. Ilatch Nuclear Plant Units 1 and 2 Fire Hazards Analysis and Fire Protection Program, originally submitted by a letter dated July 22, 1986. The licensee may make changes to the fire protection program without prior approval of the Commission only if the changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
1 AmIndment No. 133
[ e at:t o
UNITED STATES g
I NUCLEAR REGULATORY COMMISSION o
B
,i wAsmNGTON, D. C. 20555 GEORGIA POWER COMPANY OGLETHORPE POWER CORPORATION MUNICIPAL ELECTRIC AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 70 License No. NPF-5 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Georgia Power Company, et al.,
(thelicensee)datedJuly 25, 1986 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 confonnity with the application, the provisions of the Act, and the rules and regulations of the Comission; I
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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied.
2.
Accordingly, the license is amended by changes to the Technical i
Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-5 is hereby amended to read as follows:
. i (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 70, are hereby iacorporated in the license.
The licensee shall operate the facility in accordance with the Technical Specifications.
3.
The license is further amended by replacing page 5 to revise paragraph 2.C.(3)(b).
4.
This license amendment is effective as of its date of issuance and shall be implemented within 60 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION t
A$
Daniel R. Muller, Director BWR Project Directorate #2 Division of BWR Licensing Attachments:
1.
Changes to the Technical Specifications 2.
Page 5 of license r
Date of Issuance: November 24, 1986-l 1
ATTACHMENT TO LICENSE AMENDMENT NO. 70 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the Appendix A Techr.ical Specifications with the enclosed pages. The revised areas are indicated by marginal lines. The overleaf pages are provided for convenience.
Pages V
VIII XI XIII 3/4 3-59 3/4 3-60 3/4 7-21 thru 3/4 7-30 B 3/4 3-4 i
B 3/4 7-3 B 3/4 7-4 l
6-7 6-17 t
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREl%N4 SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.
3/4 3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION 3/4 3-24 3/4.3.4 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION 3/4 3-33 3/4.3.5 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION 3/4 3-37 3/4.3.6 MONITORING INSTRUMENTATION l
Radiation Monitoring Instrumentation
,3/4 3-43 4
Seismic Monitoring Instrumentation 3/4 3-47 Remote Shutdown Monitoring Instrumentation 3/4 3-50 i
Post-Accident Monitoring Instrumentation 3/4 3-53 Source Range Monitors 3/4 3-56 Traversing Incore Probe System 3/4 3-57 Chlorine,Dettetors 3/4 3-58 i
i Radioactive Liquid Effluent Instrumentation 3/4 3-60a Radioactive Gaseous Effluent Instrumentation 3/4 3-60f 3/4.3.7 TURBINE OVERSPEED PROTECTION SYSTEM 3/4 3-61 3/4.3.8 DEGRADED STATION VOLTAGE PROTECTION INSTRUMENTATION 3/4 3-63 3/43-66 3/4.3.9 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops 3/4 4-1 1
i HATCH-UNIT 2 V
Amendment No. 27.48.//;70
--n
_. - _, - -.. - ~,., - -. -.,. - -.
INDEX LIMITING CONDITIONS FOR OPERATION AND $URVEILLANCE REQUIREMENTS I
SECTION PAGE REACTOR COOLANT SYSTEM (Continued)
Jet Pumps 3/4 4-2 Idle Recirculation Loop Startup 3/4 4-3 3/4.4.2 SAFETY / RELIEF VALVES 3/4 4-4 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems 3/4 4-5 j
Operational. Leakage 3/4 4-6 3/4.4.4 CHEMISTRY 3/4 4-7 3/4.4.5 SPECIFIC ACTIVITY 3/4 4-10 3/4.4.6 PRESSURE / TEMPERATURE LIMITS i
Reactor Coolant System 3/4 4-13 Reactor Steam Dome 3/4 4-18 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES 3/4 4-19 l
3/4.4.8 STRUCTURAL INTEGRITY 3/4 4-20 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 HIGH PRESSURE COOLANT INJECTION SYSTEM 3/4 5-1 3/4.5.2 AUTOMATIC DEPRES$URIZATION SYSTEM 3/4 5-3 i
3/4.5.3 LOW PRES $URE CORE COOLING SYSTEMS Core Spray System 3/4 5-4 i
3/4 5-7 Low Pressure Coolant Injection System 3/4.5.4
$UPPRESSION CHAMBER 3/4 5-9 i
HATCH-UNI" 2 VI Amendmentl(o A8, 69 i
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J' INDEX LIMITING CONDITIONS FOR OPERATION kND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity........................
3/4 6-1 Primary Containment Leakage..........................
3/4 6-3 Primary Containment Air Loc k.........................
3/4 6-6 MSI V Lea ka ge Control System..........................
3/4 6-7 Primary Containment Structural Integrity.............
3/4 6-8 Primary Containment Internal Pressure................
3/4 6-9 Drywell Avera ge Air Tempera ture......................
3/4 6-10 3/4.6.2 DEPRESSURIZATION SYSTEMS r.
Suppression Chamber..................................
3/4 6-11 Suppression Pool Coo 11ng.............................
3/4 6-14 Y
3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES.................
3/4 6-15 3/4.6.4 VACUUN RELIEF Suppression Chamber - Drywell Vacuum Breakers........
3/4 6-33 Reactor Building - Suppression Chamber Vacuum Breakers...........................................
3/4 6-35
~
3/4.6.5 SECONDARY CONTAIhMENT i
S econda ry Containment Integrity.......................
3/4 6-36 Secondary Containment Automatic Isolation Dampers.....
3/4 6-37 3/4.6.6 CONTAINENT ATMOSPHERE CONTROL S ta nd by Ga s Treatment System.........................
3/4 6-40 Primary Containment Hydrogen Recombiner Systems......
3/4 6-43 Primary Containment ' Hydrogen Mixing System...........
3/4 6-44 HATCH-UNIT 2 VII h.
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS Residual Heat Removal Service Water System............... 3/4 7-1 Service Water Systems....................................
3/4 7-3 3/4.7.2 MAIN CONTROL ROOM ENVIRONMENTAL CONTROL SYSTEM........... 3/4 7-6 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM.................... 3/4 7-9 3/4.7.4 SNUBBERS.................................................
3/4 7-11 3/4.7.5 SEALED SOURCE CONTAMINATION..............................
3/4 7-19 3/4.7.6 (Deleted) 3/4.7.7 (Deleted) 3/4.7.8 SETTLEMENT OF CLASS I STRUCTURES......................... 3/4 7-31 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1 A.C. SOURCES A.C. Sources-Operating...................................
3/4 8-1 g
A.C. Sources-Shutdown....................................
3/4 8-9 3/4.8.2 ONSITE POWER DISTRIBUTION SYSTEMS A.C. Distribution - Operating..........................
3/4 8-10 A.C. Distribution - Shutdewn.............................
3/4 8-12 D.C. Distribution - Operating...........................
3/4 8-13 D.C. Distribution - Shutdown............................
3/4 8-16 A.C. Ci rcuits Inside Prima ry Containment................. 3/4 8-17 Primary Containment Penetration Conductor Overcurrent Protective Devices...........................
3/4 8-18 HATCH-UNIT 2 VIII Amendment No. 31, 70
]
f INDEX-BASES SECTION M
3/4.0 APPLICABILITY
... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SRUTDOWN MARGIN 3 3/4 1-1 3/4.1.2 REACTIVITY ANOMALIES B 3/4 1-1 3'/4.1. 3 CONTROL RC'DS B 3/4 1-2 3/4.1. 4 CONTROL RCD PROGRAM CONTROLS B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM B 3/4 1-4 3/4.2 70WER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR BEAT GENERATION RATE B 3/4 2-1 3/4.2.2 APRM SETPOINTS S 3/4 2-3 3/4.2.3 MINIMUM CRITICAL 70WER RATIO 3 3/4 2-3 i
3/4.2.4 LINEAR EEAT GENERATION RATE B 3/4 2-5 3/4.3 IN u avrstfTATION 3/4.3.1 REACTOR PROTECTON SYSTEM INSTRUMENTATION B 3/4 3-1 3/4.3.2 ISCIATION ACTUATION INSTRUMENTATION B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION 3 3/4 3-2 i
l 3/4.3.4 REACTOR CORE ISCLATION COOLING SYSTEM ACTUATION INSTRUMENTATION 3 3/4 3-3 3/4.3.5 CONTROL RCD WITEDRAWAL BLOCK INSTRUMENTATION 3 3/4 3-3
-3/4.3.6 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation B 3/4 3-3 p'
Seismic Monitoring Instrumentation B 3/4 3-3 EATCE-UNIT 2 X
Amendment No. 48 g @ g gggg
=
INDEX BASES
~
SECTION PAGE INSTRUMENTATION (Continued)
Remote Shutdown Monitoring B 3/4 3-3 Instrumentation Post-Accident Monitoring B 3/4 3-4 Instrumentation Source Range Monitors B 3/4 3-4 Traversing Incore Probe System B 3/4 3-4 Chlorine Detectors B 3/4 3-4 I
Radioactive Liquid Effluent Instrumentation B 3/4 3-5 Radioactive Gaseous Effluent Instrumentation B 3/4 3-5 3/4.3.7 TURBINE OVERSPEED PROTECTION SYSTEM B 3/4 3-5 3/4.3.8 DEGRADED STATION VOLTAGE PROTECTION INSTRUMENTATION B 3/4 3-Sa 3/4.4 REACTOR COOLANT SYSTEM
~
3/4.4.1 RECIRCULATION SYSTEM B 3/4 4-1 3/4.4.2 SAFETY / RELIEF VALVES B 3/4 4-1 3/4.4.3 REACTOR COOLANT SYSTEM LEAVAGE i
Leakage Detection Systems B 3/4 4-2 Operational Leakage B 3/4 4-2 3/4.4.4 CHEMISTRY B 3/4 4-2 3/4.4.5 SPECIFIC ACTIVITY B 3/4 4-3 3/4.4.6 PRESSURE / TEMPERATURE LIMITS B 3/4 4-4 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES B 3/4 4-6 l
HATCH-UNIT 2 XI Amendment No. 48, 70
INDEX BASES SECTION PAGE 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS B 3/4 7-1 3/4.7.2 MAIN CONTROL ROOM ENVIRONMENTAL CONTROL SYSTEM B 3/4 7-1 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM B 3/4 7-1 3/4.7.4 SNUBBERS B 3/4 7-2 3/4.7.5 SEALED SOURCE CONTAMINATION B 3/4 7-3 3/4.7.6 (Deleted) 3/4.7.7 (Deleted) 3/4.7.8 SETTLEMENT OF CLASS I STRUCTURES B 3/4 7-4 3/4.8 ELECTRICAL POWER SYSTEMS B 3/4 8-1 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH B 3/4 9-1 3/4.9.2 INSTRUMENTATION B 3/4 9-1 3/4.9.3 CONTROL ROD POSITION B 3/4 9-1 3/4.9.4 DECAY TIME B 3/4 9-1 3/4.9.5 SECONDARY CONTAINMENT B 3/4 9-1 3/4.9.6 COMMUNICATIONS B 3/4 9-2 3/4.9.7 CRANE AND HOIST OPERABILITY B 3/4 9-2 3/4.9.8 CRANE TRAVEL-SPENT FUEL STORAGE POOL B 3/4 9-2 3/4.9.9 WATER LEVEL - REACTOR VESSEL AND and WATER LEVEL - SPENT FUEL STORAGE 3/4.9.10 POOL B 3/4 9-2 3/4.9.11 CONTROL ROD REMOVAL B 3/4 9-2 3/4.9.12 REACTOR COOLANT CIRCULATION B 3/4 9-3 HATCH-UNIT 2 XIII Amendment No. #$, 51, 70
(
INDEX BASES
~
SECTION g
REACTOR COOLANT SYSTEM (Continued) 3/4.4.8 STRUCTURAL INTEGRITY B 3/4 4-6 3 /4. 5 EMERGENCY CORE COOLING SYSTDIS 3/4.5.1 BICH PRESSURE COOLAN't INJECTION SYSTEM B 3/4 5-1 3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM B 3/4 5-1 3/4. 5.3 LOW P.1 ESSURE CORE COOLING SYSTEMS Core Spray System B 3/4 5-2 Low Pressure Coolant Injection System B 3/4 5-3 3/4.5.4 SUPPRESSION CHAMBER S 3/4 5-3 3/4.6 CONTAINMENT SYSTEMS
'f 3/4.6.1 PRIMARY CONTAINMENT INTEGRITY Primary Containment Integrity B 3/4 6-1 Primary Containment Leakage B 3/4 6-1 Primary Containment Air Lock B 3/4 6-1
-MISV Leakage Control System B 3/4 6-2 2
I Primary Containment Structural Integrity B 3/4 6-2 l
Primary Containment Internal Pressure B 3/4 6-2 Drywell Average Air Temperature B 3/4 6-2 3/4.6.2 DEPRESSURIZATION SYSTEMS B 3/4 6-3 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES B 3/4 6-4 3/4.6.4 VACUUM RELIEF B 3/4 6-5 3/4.6.5 SECONDARY CONTAINMENT B 3/4 6-5
-' ~l 3/4.6.6 CONTAINMENT ATMOSPEERE CONTROL B 3/4 6-5 c;
EATCH-UNIT 2 XII Amendment No. 48
1 (This page is intentionally left blank.)
I.
HATCH - UNIT 2 3/4 3-59 Amendment No. 19, 70
(This page is intentionally left blank.)
4 I.
HATCH - UNIT 2 3/4 3-60 Amendment No. 19, 70
(These pages are intentionally left blank.)
~t t
L L,
J J
HATCH - UNIT 2 3/4 7-21 through 3/4 7-30 Amendment No. 70
INSTRUMENTATION BASES o
fU 3/4.3.4 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION b'
The reactor core isolation cooling (RCIC) system actuation instru-mentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without requiring actuation of any of the emergency core cooling equipment.
3/4.3.5 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls, and Section 3.4.2. Power Distribution Limits. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.
e 3/4.3.6 MONITORING INSTRUMENTATION 3/4.3.6.1 RADIATION MONITORING INSTRUMENTATION k
The OPERABILITY of the radiation monitoring instrumentation ensures that:
(1) the radiation levels are continually measured in the areas e
served by the individual channels, and (2) the alam or automatic action is initiated when the radiation level trip setpoint is exceeded.
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3/4.3.6.2 SEISMIC MONITORING INSTRUMENTATION F-The OPERABILITY of the seismic monitoring instrumentation ensures that suffictent capability is available to promptly detemine the magnitude of a seismic event and evaluate the response of those features important 3
to safety. This capability is required to permit tamparison of the measured response to that used in the design basis for the unit.
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3/4.3.6.3 REMOTE SHUTDOWN MONITORING INSTRUMENTATION The OPERABILITY of the remote shutdown monitoring instrumentation
- ,i ensures that sufficient capability is available to permit shutdown and Q
maintenance of HOT SHUTDOWN of the unit from locations outside of the
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control room. This capability is required in the event control room habitability is lost and is consistent with General Design Criterion 19 of
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1.0 CFR Part 50.
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Q, HATCH - UNIT 2 B 3/4 3-3 J.
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INSTRUMENTATION BASES MONITORING INSTRUMENTATION (Continued) 3/4~.3.6.4 POST-ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the post-accident monitoring instrumentation ensures that sufficient information is available on selected plant param-eters to monitor and assess important variable following an accident.
3/4.3.6.5 SOURCE RANGE MONITORS The source range monitors provide the operator with information on the status of the neutron level in the core at very low power levels
.during startup. At these power levels, reactivity additions should not be made without this flux level information available to the operator. When the intermediate range monitors are on scale adequate information is available without the SRMs and they can be retracted.
3/4.3.6.6 TRAVERSING INCORE PROBE SYSTEM The OPERABILITY of.the traversing incore probe system with the specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating each detector to be used and normalizing their respective outputs.
3/4.3.6.7 CHLORINE DETECTORS The OPERABILITY of the chlorine detectors ensures that an accidental e
chlorine release will be detected promptly and the necessary protective actions will be automatically initiated to provide protection for control room personnel. Upon detection of a high concentration of chlorine the control room emergency ventilation system will automatically be placed in the isolation mode of operation to provide the required protection.
i-3/4.3.6.8 (Deleted)
Amendment No. 70 HATCH - UNIT 2 B 3/4 3-4 l
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PLANT SYSTEMS BASES 3/4.7.4 SNUBBERS (Continued)
The service life of a snubber is evaluated via manufacturer input and information through consideration of the snubber service conditions and associated installation and maintenance records (newly installed snubber, seal replaced, spring replaced, in high radiation area, in high temperature i
area,etc...).
The requirement to monitor the snubber service life is included to ensure that the snubbers periodically undergo a performance evaluation in view of their age and operating conditions. These records will provide statistical bases for future consideration of snubber service life. The requirements for the maintenance of records and the snubber service life review are not intended to affect plant operation.
3/4.7.5 SEALED SOURCE CONTAMINATION The limitations on sealed source removable contamination ensure that the i
total body or individual. organ irradiation does not exceed allowable limits in the event of ingestion or inhalation of the. source material. The limitations on removable contamination for sources requiring leak testing, including alpha emitters, is based on 10 CFR 70.39(c) limits for plutonium.
Quantities of interest to this specification which are exempt from the leakage testing are consistent with the criteria of 10 CFR Part 30.11-20 and 170.19.
Leakage from sources excluded from the requirements of this specification is not likely to represent more than one maximum permissible
-body burden for total body irradiation if the source material is inhaled or
' ingested.
i 3/4.7.6 (Deleted)
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HATCH - UNIT 2 B 3/4 7-3 Amtndment No. 51, 70 l
PLANT SYSTEMS BASES 3/4.7.7 (Deleted) 3/4.7.8 SETTLEMENT OF CLASS 1 STRUCTURES In order to assure that settlement does not exceed predicted and allowable settlement values, a program has been established to conduct a survey at the site.
-The allowable total and differential settlement values are based on original settlement predictions.
In establishing these tabulated values, an assumption is made that pipe and conduit connections have been designed to safely withstand the stresses which would develop due i
to total and differential settlement.
HATCH - UNIT 2 B 3/4 7-4 Amendment No. 70
ADMINISTRATIVE CONTROLS RESP 0NSIBILITIES (Continued)
'i. Review of the Security Plan and implementing procedures and shall
. submit recommended changes to the SRB.
j.
Review of the Emergency Plan and implementing procedures and shall submit recommended changes to the SRB.
k.
Review of any unplanned onsite release of radioactive material to the environs when such release is in excess of I C1, excluding dissolved and entrained gases and tritium for liquid effluents, and in excess of 150 Ci of noble gases or 0.02 C1 of radioiodines for gaseous effluents. Also included is the preparing and forwarding to the General Manager-Plant Hatch and the SRB reports covering evaluation, recommendations and disposition of the corrective action to prevent recurrence.
~1.
Review of changes to the PROCESS CONTROL PROGRAM and the OFFSITE DOSE CALCULATION MANUAL (0DCM).
Review of proposed change (s) to plant systems and equipment to m.
determine whether the proposed change has a potential radiological environmental impact. Such change (s) will be reported to the Manager-Nuclear Engineering and Chief Nuclear Engineer.
Review of the Fire Protection Program and implementing procedures n.
and shall submit recommended changes to the SRB.
AUTHORITY
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6.5.1.7 The PRB shall:
I Recommend in writing to the General Manager-Plant Hatch approval or t
a.
disapproval of items considered under 6.5.1.6(a) through (d) above.
.I b.
Render determinations in writing with regard to whether or not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
Provide written notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to the Manager of c.
Nuclear Generation or the Vice President and General Manager Nuclear Operations and the Safety Review Board of disagreement between the PRB and the General Manager-Plant Hatch; however, the General Manager-Plant Hatch shall have responsibility for resolution of.such disagreements pursuant to 6.1.1 above.
RECORDS 6.5.1.8 The Plant Review Board shall maintain written minutes of each PRB meeting that, at a minimum, document the results of all PRB activities performed under the responsibility and authority provisions of these Technical Specifications. Copies shall be provided to the Manager of Nuclear Generation or the Vice President and General Manager Nuclear Operations and the Safety Review Board.
HATCH - UNIT 2 6-7 Amendment No. f7, 48, 70
" ADMINISTRATIVE CONTROLS d.
Abnormal degradation of systems other than those specified in 6.9.1.8.c above designed to contain radioactive material resulting from the fission process.
SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director.of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. Special reports for fire protection equipment operating and surveillance requirements shall be submitted, as required, by the Fire Hazards Analysis and its Appendix B requirements.
6.10 RECORD RETENTION In addition to the applicable record retention requirements of Title 10, Code of Federal Regulations, the following records shall be retained for at least the minimum period indicated.
6.10.1 The following records shall be retained for at least five years:
a.
Records and logs of unit operation covering time interval at each power level.
b.
Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
c.
ALL REPORTABLE OCCURRENCES submitted to the Commission.
d.
Records of surveillance activities, inspections and calibrations l
required by these Technical Specifications.
p e.
Records of changes made to the procedures required by i,
Specification 6.8.1.
f.
Records of radioactive shipments.
g.
Records of sealed source and fission detector leak tests and results.
h.
Records of annual physical inventory of all sealed source material of record.
6.10.2 The following records shall be retained for the duration of the unit Operating License:
i a.
Records and drawing changes reflecting unit design modifi-cations made to systems and equipment described in the Final Safety Analysis Report.
b.
Records of new and irradiated fuel inventory, fuel transfers and assembly burnup histories.
HATCH - UNIT 2 6-17 Amendment No. 70
(b) Georgia Power Company shall implement and maintain in effect all provisions of the. fire protection program, which is referenced in the Final Safety Analysis Report for the facility, as contained in the updated Edwin I. Hatch Nuclear Plant Units 1 and 2 Fire Hazards Analysis and Fire Protection Program, originally submitted by a letter dated July 22, 1986. The licensee may make changes to the fire protection program without prior approval of the Commission only if the changes would not adversely affect the ability to achieve and maintain safe shutdown in the event of a fire.
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Amendment No. 70