ML20214K627
| ML20214K627 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  |
| Issue date: | 08/15/1986 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20214K632 | List: |
| References | |
| TAC-61083, TAC-61357, NUDOCS 8608210361 | |
| Download: ML20214K627 (72) | |
Text
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O UNITED STATES
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NUCLEAR REGUL TORY COMMISSION r,;
J WASHINGTON. D. C. 20555
...../
MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.
SOUTH MISSISSIPPI ELEGINIC POWER ASSOCIATION DOCKET NO. 50-416 GRAND GULT NUtEEAIF3TATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.16 License No. NPF-29 1.
The Nuclear Regulatory Comission (the Comission) has foun'd that A.
The application for amendment by Mississippi Power & Light Company, Middle South Energy, Inc., and South Mississippi Electric Power Association, (the licensees) dated March 31, 1986 as amended by letters dated May 2, June 9, and June 20, 1986 and supplemented i
July 11, 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 confomity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 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 Connission's regulations and all applicable requirements have been satisfied.
l 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby amended to read as follows:
l Technical Specifications l
The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised i
through Amendment No.16, are hereby incorporated into this license.
l Mississippi Power & Light Company shall operate the facility in j
accordance with the Technical Specifications and the Environmental Protection Plan.
i 8609210361 860812 PDR ADOCK 05000416 P
A 3.-
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION J21L-Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: August 15, 1986 h-
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ATTACHMENT TO LICENSE AMENDMENT NO. 16 FACILITYOPERATINGLICENSENO.NPF-29 DOCKET NO. 50-416 l
Replace the following pages of the. Appendix "A" Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf page(s) provided to maintain document completeness.*
Remove Insert 111 iii*
iv iy v
v vi vi*
xi xi*
xii xii xiii xiii xiy xiv*
2-1 2-1 2-2 2-2*
2-3 2-3*
2-4 2-4 2-4a B 2-1 8 2-1 B 2-2 B 2-2 8 2-3 B 2-3 B 2-4 8 2-4*
B 2-7 B 2-7 B 2-8 8 2-8*
B 2-9 B 2-9 B 2-10 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-2a 3/4 2-2b 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4*
3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 3-5 3/4 3-5 3/4 3-6 3/4 3-6*
3/4 3-7 3/4 3-7*
3/4 3-8 3/4 3-8 3/4 3-43 3/4 3-43 3/4 3-44 3/4 3-44*
3/4 3-55 3/4 3-55 3/4 3-55a 3/4 3-56 3/4 3-56*
3/4 3-57 3/4 3-57 3/4 3-58 3/4 3-58*
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. s Remove Insert 3/4 3-111 3/4 3-112 3/4 4-1 3/4 4-1 3/4 4-la 3/4 4-lb
'=
3/4 4-Ic 3/4 4-2 3/4 4-2 3/4 4-3 3/4 4-3 3/4 4-4 3/4 4-4*
B 3/4 1-1 B 3/4 1-1*
B 3/4 1-2 B 3/4 1-2 B 3/4 2-1 B 3/4 2-1 B 3/4 2-la'-
B 3/4 2-2 B 3/4 2-2 B 3/4 2-3 B 3/4.2-3 B 3/4 2-4 B 3/4 2-4 B 3/4 2-5 B 3/4 2-5 B 3/4 2-6
.B 3/4 2-6 B 3/4 2-6a B 3/4 2-7 B 3/4 2-7 B 3/4 3-3 8 3/4 3-3 B 3/4 3-3a B 3/4 3-4 8 3/4 3-4*
B 3/4 3-7 8 3/4 3 B 3/4 3-8 B 3/4 3-8*
B 3/4 4-1 3 3/4 4-1 B 3/4 4-la B 3/4 4-2 B 3/4 4-2*
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r-s INDEX SAFETYLIMITSANDLIMITINGSAFETYSYSTEMSETTINdS SECTION PAGE 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low F1ow.............................
2-1 THERMAL POWER, High Pressure and High F1ow..........................
2-1 Reactor Coolant System Pressure.....................................
2-1 Reactor Vessel Water Leve1..........................................
2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.................
2-3 BASES 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low F1ow.............................
B 2-1 THERMAL POWER, High Pressure and High F1ow..........................
B 2-2 Reactor Coolant System Pressure.....................................
B 2-5 Reactor Vessel Water Leve1..........................................
B 2-5 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.................
B 2-6 GRAND GULF-UNIT 1 iii
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4 4
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY...................................................
3/4 0-1
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3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN..............................................
3/4 1-1 3/4.1.2 REACTIVITY ANOMALIE3.........................................
3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability......................................
3/4 1-3 Control Rod Maximum Scram Insertion Times....................
3/4 1-6 Control Red Scram Accumulators...............................
3/4 1-8
~
Control Rod Drive Coupling...................................
3/4 1-10 Control Rod Position Indication..............................
3/4 1-12 Control Rod Drive Housing Support............................
3/4 1-14 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control Rod Withdrawa1.......................................
3/4 1-15 l
Rod Pattern Control System...................................
3/4 1-16 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM................................
3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE...................
3/4 2-1 3/4.2.2 DELETED......................................................
3/4 2-3 l
3/4.2.3 MINIMUM CRITICAL POWER RATI0.................................
3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE..................................
3/4 2-7 l
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l GRAND GULF-UNIT 1 iv Amendment No.16 l
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INDEX LIMITING CONOITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 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 3-27 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation.........
3/4 3-37 End-of-Cycle Recirculation Pump Trip System Instrumentation.............................................
3/4 3-41 3/4.3.5 REALTOR CORE ISO.LATION COOLING SYSTEM ACTUATION 4
INSTRUMENTATION.............................................
3/4 3-47 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION...........................
3/4 3-52 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation........................
3/4 3-58 4
i Seismic Monitoring Instrumentation..........................
3/4 3-63 Meteorological Monitoring Instrumentation...................
3/4 3-66 f
Remote Shutdown System Instrumentation and Controls.........
3/4 3-69
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Accident Monitoring Instrumentation.........................
3/4 3-73 Source Range Monitors...........................i............
3/4 3-77 Traversing In-Core Probe System.............................
3/4 3-78 Chlorine Detection System...................................
3/4 3-79 Fire Detection Instrumentation..............................
3/4 3-80 Loose-Part Detection System.................................
3/4 3-90 Radioactive Liquid Effluent Monitoring Instrumentation......
3/4 3-91 Radioactive Gaseous Effluent Monitoring Instrumentation.....
3/4 3-96 3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION....................... 3/4 3-105 3/4.3.9 TURBINE OVERSPEED PROTECTION SYSTEM..........................
3/4 3-110 3/4.3.10 NEUTRON FLUX MONITORING INSTRUMENTATION......................
3/4 3-111 l l
GRAND GULF-UNIT 1 v
Amendment No. 16 l
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0 INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops........................................
3/4 4-1 Jet Pumps..................................................
3/4 4-2 Recirculation Loop F1ow....................................
3/4 4-3 Idle Recirculation Loop Startup............................
3/4 4-4 3/4.4.2 SAFETY VALVES Safety / Relief Va1ves.......................................
3/4 4-5 Safety / Relief Valves Low-Low Set Function..................
3/4 4-7 3/4 4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems..................................
3/4 4-8 Operational Leakage........................................
3/4 4-9 3/4.4.4 CHEMISTRY..................................................
3/4 4-13 3/4.4.5 SPECIFIC ACTIVITY..........................................
3/4 4-16 3/4.4.6 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.....................................
3/4 4-19 Reactor Steam 00me.........................................
3/4 4-23 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES...........................
3/4 4-24 3/4.4.8 STRUCTURA L INTEGRITY....................................... 3/4 4-25 3/4.4.9 RESIDUAL HEAT REMOVAL Hot Shutdown...............................................
3/4 4-26 Cold Shutdown..............................................
3/4 4-27 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ECCS - 0PERATING...........................................
3/4 5-1 3/4.5.2 ECCS - SHUTD0WN............................................
3/4 5-6 3/4.5.3 SUPPRESS ON P00L...........................................
3/4 5-8 GRAND GULF-UNIT 1 vi
7-INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration................................................
3/4 11-1 Dose.........................................................
3/4 11-5 Liquid Waste Treatment.......................................
3/4 11-6 Liquid Holdup Tanks..........................................
3/4 11-7 3/4.11.2 GASEOUS EFFLUENTS Dose Rate....................................................
3/4 11-8 Dose - Noble Gases...........................................
3/4 11-12 Dose - Iodine-131, Iodine-133, Tritium and Radionuclides in Particulate Form........................................
3/4 11-13 Gaseous Radwaste Treatment...................................
3/4 11-14 Ventilation Exhaust Treatment................................
3/4 11-15 Explosive Gas Mixture........................................
3/4 11-16 Main Condenser...............................................
3/4 11-17 3/4.11.3 SOLID RADIOACTIVE WASTE......................................
3/4 11-18 3/4.11.4 TOTAL D0SE...................................................
3/4 11-19 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM...........................................
3/4 12-1 3/4.12.2 LAND USE CENSUS..............................................
3/4 12-11 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...........................
3/4 12-12 GRAND GULF-UNIT 1 xi
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0 INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY...............................................
B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.....................................
B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES................................
B 3/4 1-1 3/4.1.3 CONTROL R0DS........................................
B 3/4 1-2 3/4.1.4 CONTROL R0D PROGRAM CONTR0LS........................
B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM.......................
B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE..........
B 3/4 2-1 3/4.2.2 DELETED.............................................
B 3/4 2-2 l
3/4.2.3 MINIMUM CRITICAL POWER RATI0........................
B 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE.........................
B 3/4 2-7 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...........
B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.................
B 3/4 3-1 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION.....................................
B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION.....................................
B 3/4 3-2 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION...........................
B 3/4 3-3 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION...................
B 3/4 3-3 GRAND GULF-UNIT 1 xii Amendment No.16
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INDEX BASES SECTION PAGE INSTRUMENTATION (Continued) 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation...............
B 3/4 3-4 Seismic Monitoring Instrumentation.................
B 3/4 3-4 Meteorological Monitoring Instrumentation..........
B 3/4 3-4 Remote Shutdown System Instrumentation and Controls.........................................
B 3/4 3-4 Accident Monitoring Instrumentation................
B 3/4 3-4 Source Range Monitors..............................
B 3/4 3-5 Traversing In-Core Probe System....................
B 3/4 3-5 Chlorine Detection System..........................
B 3/4 3-5 Fire Detection Instrumentation.....................
B 3/4 3-5 Loose-Part Detection System........................
B 3/4 3-6 Radioactive Liquid Effluent Monitoring Instrumentation..................................
' B 3/4 3-6 Radioactive Gaseous Effluent Monitoring Instrumentation..................................
B,3/4 3-6 3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION............
B 3/4 3-6 3/4.3.9 TURBINE OVERSPEED PROTECTION.......................
B 3/4 3-7 3/4.3.10 NEUTRON FLUX MONITORING INSTRUMENTATION............
B 3/4 3-7 l
3/4.4 REACTOR COOLANT SYSTEM i
3/4.4.1 RECI RCU LATION SYSTEM...............................
B 3/4 4-1 3/4.4.2 SAFETY /RE' LIEF VALVES...............................
B 3/4 4-2 i
3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems..........................
B 3/4 4-2 l
l Operational Leakage................................
B 3/4 4-2 3/4.4.4 CHEMISTRY..........................................
B 3/4 4-3 3/4.4.5 SPECIFIC ACTIVITY..................................
B 3/4 4-3 i
i 3/4.4.6 PRESSURE / TEMPERATURE LIMITS........................
B 3/4 4-4 l
l 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES...................
B.3/4 4-5 l
3/4.4.8 STRUCTURAL INTEGRITY...............................
B 3/4 4-5 3/4.4.9 RESIDUAL HEAT REM 0 VAL..............................
B 3/4 4-5 GRAND GULF-UNIT 1 xiii Amendment No.16 l
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INDEX BASES SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEM 3/4.5.1/2 ECCS - OPERATING and SHUT 00WN......................
,8 3/4 5-1 3/4.5.3 SUPPRESSION P00L...................................
B 3/4'5-2 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT Primary Containment Integrity......................
B 3/4 6-1 Containment Leakage................................
B 3/4 6-1 Containment Air Locks..............................
B 3/4 6-1 MSIV Leakage Control System........................
B 3/4 6-1 Feedwater Leakage Control System...................
B 3/4 6-2 Containment Structural Integrity...................
B 3/4 6-2 Containment Internal Pressure......................
B 3/4 6-2 Containment Average Air Temperature................
B 3/4 6-2 Containment Purge System...........................
B 3/4 6-2 3/4.6.2 ORYWELL Drywell Integrity..................................
B 3/4 6-3 Orywell Bypass Leakage.................._...........
B 3/4 6-3 Orywell Air Locks..................................
B 3/4 6-3 Drywell Structural Integrity.......................
B 3/4 6-4
'Orywell Internal Pressure..........................
B 3/4 6-4 Orywell Average Air Temperature....................
B 3/4 6-4 Orywell Vent and Purge.............................
B 3/4 6-4 3/4.6.3 DEPRESSURIZATION SYSTEMS...........................
B 3/4 6-4 3/4.6.4 CONTAINMENT AND ORYWELL ISOLATION VALVES...........
B 3/4 6-7 3/4.6.5 DRYWELL POST-LOCA VACUUM BREAKERS..................
B 3/4 6-7 3/4.6.6 SECONDARY CONTAINMENT..............................
B 3/4 6-8 3/4.6.7 ATMOSPHERE CONTR0L.................'................
B 3/4 6-9 GRAND GULF-UNIT 1 xiv
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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or low Flow 2.1.1 THERMAL POWER shall'not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 during two loop operation with the reactor vessel steam dome pressure greater than 1
785 psig and core flow greater than 10% of rated flow.
During single loop operation with the reactor vessel steam done pressure greater than 785 psig and core flow greater than 10% of rated flow, the MCPR shall not be less than 1.07.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With MCPR less than the above limits and the reactor vessel steam dome pressure l
greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specifi-i cation 6.7.1.
4 REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, 3 and 4.
ACTION:
With the reactor coolant system pressure, as measured in the reactor vessel j
steam done, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant i
system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.
l 1
i GRAND GULF-UNIT 1 2-1 Amendment No.16 i
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SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS
~
SAFETY LIMITS (Continued)
REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.
1 APPLICABILITY:
OPERATIONAL CONDITIONS 3, 4 and 5 ACTION:
Wit'h the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the ECCS to restore the water level.
D: pressurize the reactor vessel as necessary for ECCS operation.
Comply with the requirements of Specification 6.7.1.
O s
l GRAND GULF-UNIT 1 2-2 l
s SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS l
- 2. 2 LIMITING SAFETY SYSTEM SETTINGS REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1.
APPLICABILITY:
As shown in Table 3.3.1-1.
ACTION:
With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value.
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GRAND GULF-UNIT 1 2-3
TABLE 2.2.1-1 o
REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS ALLOWA8LE FUNCTIONAL UNIT TRIP SETPOINT VALUES
[
1.
Intermediate Range Monitor, Neutron Flux-High 5 120/125 divisions 5 122/125 divisions 3
of full scale of full scale
-4 9
2.
Average Power Range Monitor:
a.
Neutron Flux-High, Setdown 5 15% of RATED
$ 20% of RATED THERMAL POWER THERMAL POWER b.
Flow Blased Simulated Thermal Power-High
- 1) During two recirculation loop operation:
a) Flow Biased 5 0.66 W+64%, with 5 0.66 W+67%, with a maximum of a maximum of b) High Flow Clamped 5 111.0% of RATED 5 113.0% of RATED THERMAL POWER THERMAL POWER
- 2) During single recirculation loop cperation:
a) Flow Biased 5 0.66 W+40%
$ 0.66 W+43%
b) High Flow Clamped Not Required Not Required OPERA 8LE OPERABLE c.
Neutron Flux-High 5 118% of RATED
$ 120% of RATED THERMAL POWER THERMAL POWER d.
Inoperative NA NA i
3.
Reactor Vessel Steam Dome Pressure - High 5 1064.7 psig 5 1079.7 psig 4.
Reactor Vessel Water Level - Low, Level 3
-> 11.4 inches above
-> 10.8 inches above
{
instrument zero*
instrument zero*
- s
{
5.
Reactor Vessel Water Level-High, Level 8 5 53.5 inches above 5 54.1 inches above instrument zero*
instrument zero*
g N
L
TABLE 2.2.1-1 (Continued)
REACTOR PROTECTION SYSTEN INSTRUENTATION SETPOINTS o
ALLOW 8LE FUNCTIONAL LRIIT TRIP SETPOINT VALUES g
h 6.
Main Steam Line Isolation Valve - Closure 1 65 closed 1 75 closed EQ 7.
Main Steam Line Radiation - High 5 3.0 x full power i 3.6 x full power background background g
8.
Drywell Pressure - High 5 1.23 psig 5 1.43 psig 9.
Scram Discharge Volume Water Level - High 5 60% of full scale 1 63% of full scale
- 10. Turbine Stop Valve - Closure 1 40 psig**
1 37 psig
- 11. Turbine Control Valve Fast Closure, Trip Dil Pressure - Low 1 44.3 psig**
1 42 psig 12.
Reactor Mode Switch Shutdown Position MA MA 9
A
- 13. Manual Scram NA NA "See Bases Figure 8 3/4 3-1.
- Initial setpoint.
Final setpoint to be determined during startup test program. Any required change to this setpoint shall be submitted to the Commission within 90 days of test completion.
3 t
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2.1 SAFETY LIMITS BASES
2.0 INTRODUCTION
The fuel cladding, reactor pressure vessel and primary. system piping are the principal barriers to the release of radioactive materials to the environs.
Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur i
if the limit is not violated.
Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit for the MCPR.
MCPR greater than the applicable Safety Limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.
The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations, however, can result from thermal i
stresses which occur from reactor operation significantly above design condi-tions and the Limiting Safety System Settings.
While fission product migration from cladding perforation is just as measurable as that from use related cracking, 4
the thermally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding i
4 deterioration.
Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR
)
of 1.0.
These conditions represent a significant departure from the condition l
intended by design for planned operation.
1 2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calcula-I tions at pressures below 785 psig or core flows less than.10% of rated flow.
i Therefore, the fuel cladding integrity Safety Limit is established by other i
means.
This is done by establishing a limiting condition on core THERMAL POWER
}
j with the following basis.
Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows j
will always be greater than 4.5 psi.
Analyses show that with a bundle flow of 28 x 108 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr.
Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical i
power at this flow is approximately 3.35 MWt.
With the design peaking factors, j
this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.
Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
4 GRAND GULF-UNIT 1 B 2-1 Amendment No.16 1
iL- - - -.
SAFETY LIMITS BASES 2.1.2 THERMAL' POWER. High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel' damage is calculated to occur if the limit is not violated.
Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal und hydraulic conditions resulting in a departure from nucleate boiling have been use.d to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boil ng would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit.
However, the uncertainties in monitoring the core operating state and in the procedures used to calculate tLe critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for
,which more than 99.9% of the fuel rods in the cor. are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.
The Safety Limit,MCPR is determined using the General Electric. Thermal Analysis Basis GETAB, which is a statistical model that combines all of the uncertainties In operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L),
GEXL, correlation. The GEXL correlation is valid over the range of conditions us:d in the tests of the data used to develop the correlation.
The required input to the statistical model are the uncertainties listed in Bases Table B2.1.2-1 and the nominal values of the core parameters listed in Bases Table B2.1.2-2.
The bases for the uncertainties in the core parametern are given in b
in NEDO-10958-A,the basis for the uncertainty in the GEXL correlation is given NE00-20340 and The bases for the changes in uncertainties m ulting from a single loop operation are given in the GGNS Single Loop Operatio1 Analysis dated February 1986. The power distribution is based on a typical 764 assembly
. core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution during any fuel cycle would not be as severe as the distribution used in the analysis.
a.
" General Electric BWR Thermal Analysis Bases (GETAB) Data, Correlation and Design Application," NED0-10958-A.
b.
General Electric " Process Com) uter Performance Evaluation Accuracy
NED0-20340 and Amendment 1, NEDO-20340-1 dated June 1974 and December 1974, respactively.
GRAND GULF-UNIT 1 B 2-2 Amendment No. 16
Bases Table' B2.1'.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFE 1Y LIMIT
- Standard Deviation Quantity
(% of Point)
Feedwater Flow 1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow 2.5(a)
Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5.0 TIP Readings 6.3(b)
R Factor
- 1. 5 Critical Power 3.6
- The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.
(a) This value increases to 6.0 for single recirculation loop operation.
(b) This value increases to 6.8 for single recirculation loop operation.
GRAND GULF-UNIT 1 B 2-3 Amendment No.16 I
O
+
Bases Table B2.1.2-2 NOMINAL'VALUISOFPARAMETERSUSEDIN THE STATISTICAL ANALYSIS OF FUEL CLADDING INTEGRITY SAFETY LIMIT THERMAL 00WER 3323 MW Core Flow 108.5 M1b/hr Dome Pressure 1010.4 psig 2
Channel Flow Area 0.1089 ft R-Factor High enrichment - 1.043 Medium enrichment - 1.039 Low enrichment - 1.030 m
I 1
I i
I i
l GRAND GULF-UNIT 1 B 2-4
[
i I
d 8
LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)
Average Power Range Monitor (Continued) l amount, the rate of power rise is very slow.
Generally the heat flux is in near equilibrium with the fission rate.
In an assumed uniform rod withdrawal j
approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.
The 15% neutron 4
flux trip remains active until the mode switch is placed in the Run position.
t j
The APRM trip system is calibrated using heat balance data taken during i
steady state conditions.
Fission chambers provide the basic input to the sys-tem and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux-High 118% setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer
).
associated with the fuel.
For the Flow Biased Simulated Thermal Power-High setpoint, a time constant of 6 i 1 seconds is introduced into the fl.ow biased j
APRM in order to simulate the fuel thermal transient characteristics.
A more l
conservative maximum value is used for the flow biased setpoint as shown in 1
Table 2.2.1-1.
In these flow biased equations, the variable w, is the loop i
recirculation flow as a percentage of the total loop recirculation flow that i
produces a rated core flow of 112.5 million lb/hr.
)
i The APRM setpoints were selected to provide adequate margin for the Safety l
Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.
1 3.
Reactor Vessel Steam Dome Pressure-High r
High pressure in the nuclear system could cause a rupture to the nuclear i
system process barrier resulting in the release of fission products.
A pres-sure increase while operating will also tend to increase the power of the 4
reactor by compressing voids thus adding reactivity.
The trip will quickly i
reduce the neutron flux, counteracting the pressure increase. The trip set-ting is slightly higher than the operating pressure to permit normal operation without spurious trips.
The setting provides for a wide margin to the maximum 1
allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during a transient.
This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure trip is bypassed.
For a turbine trip under these 4
conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.
~
]
l i
l GRAND GULF-UNIT 1 B 2-7 Amendment No. 16 I
i
LIMITING SAFETY SYSTEM SETTINGS
^
BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) i 4.
Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint was chosen far enough below the normal operating level to avoid spurious trips but high enough above the fuel to 1
assure that there is adequate protection for the fuel and pressure limits.
5.
Reactor Vessel Water Level-Hi.5 A reactor scram from high reactor water level, approximately two feet above i
normal operating level, is intended to offset the addition of reactivity effect associated with the introduction of a significant amount of relatively cold feedwater.
An excess of feedwater entering the vessel would be detected by i
the level increase in a timely manner.
This scram feature is only effective when the reactor mode switch is in the Run position because at THERMAL POWER levels below 10% to 15% of RATED THERMAL POWER, the approximate range of power i
Isvel for changing to the Run position, the safety margins are more than i
adequate without a reactor scram.
6.
Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit 4
i the amount of fission product release for certain postulated events.
The MSIV's are closed automatically from measured parameters such as high steam flow, high steam line radiation, low reactor water level, high steam tunnel temperature and low steam line pressure.
The MSIV's closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.
j 7.
Main Steam Line Radiation-High The main steam line radiation detectors are provided to detect a gross failure of the fuel cladding.
i When the high radiation is detected, a trip is i
initiated to reduce the continued failure of fuel cladding.
At the same time j
the main steam line isolation valves are closed to limit the release of fission i
products.
Tha trip setting is high enough above background radiation levels to prevent spuaious trips yet low enough to promptly detect gross failures in
~
tha fuel claddiag.
. 8.
Drywell Pressure-High High pressure in the sfrywell could indicate a break in the primary pressure b:undary systems or a loss of drywell cooling.
The reactor is tripped in order l
to minimize the possibility of fuel damage and to reduce the amount of energy 1
b3ing added to the coolant and to the primary containment.
i The trip setting was solected to be as low as possible to minimize heat loads of equipment located i
in the primary containment and to avoid spurious trips.
Negative barometric pr:ssure fluctuations are accounted for in the trip setpoints and allowable values specified for drywell pressure-high.
GRAND GULF-UNIT 1 B 2-8 I
m.
.~.
r LIMITING SAFETY SYSTEM SETTING BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 9.
Scram Discharoe Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram.
Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered.
The reactor is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig when they are tripped.
The trip setpoint for each scram discharge volume is equivalent to a contained volume of 26 gallons of water.
.10.
Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves.
With a trip setting of 40 psig, the resultant increase in heat flux.is such that adequate thermal margins are maintained during the worst case transient assuming the turbine bypass valves fail to operate.
As indicated in Table 3.3.1-1, this function is automatically bypassed below the tJrbine first-stage pressure value equivalent to THERMAL POWER less than 40% of RATED THERMAL POWER.
The automatic bypass setpoint is feedwater temperature dependent as a result of the subcooling changes that affect the turbine first-stage pressure-reactor power relationship.
For RATED THERMAL POWER operation with feedwater tempera-ture greater than or equal to 420*F, an allowable setpoint of 126.9% of control valve wide-open turbine first-stage pressure is provided for the bypass function.
This setpoint is also applicable to operation at less than RATED THERMAL POWER i
with the correspondingly lower feedwater temperature.
The allowable setpoint is reduced to < 22.5% of control valve wide-open turbine first-stage pressure for RATED THERMAL POWER operation with a feedwater temperature between 370*F and 420*F.
Similarly, the reduced setpoint is applicable to operation at less than RATED THERMAL POWER with the correspondingly lower feedwater temperature.
11.
Turbine Control Valve Fast Closure. Trip 011 Pressure-Low The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load rejection coincident with failure of the turbine bypass valves. The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by a low EHC fluid pressure in the contro1* valve and in less than 100 milliseconds after the start of control valve fast closure. This loss of pressure is sensed by pressure transmitters which output to trip units whose contacts form the one-out-of-two twice logic GRAND GULF-UNIT 1 8 2-9 Amendment No. 16 l
~ " - - - _, _ _.
LIMITING SAFETY SYSTEM SETTING BASES (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION S This The trip setpoint is 43.3 psi ~g.i tic from tha input to the Reactor Protection System.
similar to that for the stop15.2.2 of the Fin trip setting and a different valve character s valve combine to produce transients which are very Relevant transient analyses are discussed in SectionAs wit i
valve.
is also bypassed below 40% of RAIED THERMAL PO Safety Analysis Report.
Valve Closure.
Reactor Mode Switch Shutdown Position i
ignals into system trip channels which are redundant to th 12.
t l reactor trip capability.
mentation channels and provides additional manua i
ls into system Manual Scram The Manual Scram pushbutton switches introduce trip s gnatic pro 13.
trip channels which are redundant to the automachanne lity.
Amendment No.16 B 2-10 GRAND GULF-UNIT 1
3/4.2 ' POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for eich type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the lin:its shown in Figure 3.2.1-1 as multiplied by the smaller of either the flow-dependent MAPLHGR factor (MAPFAC ) of Figure 3.2.1-2, or the power-dependent f
MAPLHGR factor (MAPFAC ) of Figure 3.2.1-3.
p APPLICABILITY:
OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
ACTION:
With an APLHGR exceeding the applicable limits, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
i SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the applicable limits:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c.
Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.
d.
The provisions of Specification 4.0.4 are not applicable.
GRAND GULF-UNIT 1 3/4 2-1 Amendment No.16 l
o5 N
5 I
I I
l l
l l
CURVE REL TYPE n%
A SCR2iO
't B
SCRl60 h
13 c
acnori U
7 A
12A 12.6 12.6
(
12.
^
512.4 A
_E4 g
D I2.8 5
12.0 12.I 12 w
)E sin
- l a.
IL7 11.5
- 11. 5 V L4 IL3 f
^
zg ll _11. 5 II.2 II.I II.O NY 5:*
20 10.2 W
10.4 10.4 3
7 Ek g
NE gy 2,
9.s 4
93
$3 9
9.0 9.0 1
I I
I I
I I
eO 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000 AWERAGE PLANAR EXPOSURE (mwd /t)
[
m g
FIGURE 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) 3 l
g VERSUS AVERAGE PLANAR EXPOSURE FOR Tid 0 LOOP OPERATION INITIAL CORE FUEL TYPES 8CR210, 8CR160 and 8CR071 2
3
I l
3 l.1 i
1.0
/( r-CLAMPED DURING I
0*9 J/
/ ONE LOOP OPERATION r
=g
- FOR MAX FLW = 102.55 g' '
i k
1 4
4 0.8 E
FOR MAX FLW
- 107.05
[
A l
lO.7 li l
i 0.5 l
l 1
O.S l
0 to 40 60 80 10 0 120 CORE FLOW (% RATED), F I
f FIGURE 3.2.1-2 MAPFAC 7
GRAND GULF-UNIT 1 3/4 2-2a Amendment No.16
-. ~
c 1.1 1.0
/
0.9 FOR P > 70%;
0.8 r
f0R 25% cps 40%;
DURING ONE LOOP OPERATION CORE FLOW F $50%
l l
e"c*oA5% *i
\\
/
O.7 a
/
O6 FOR 25% $ P S40%;
0.5 coHE FLo# F > So%
i 1
0.4 4
O 20 40 60 80 10 0 120 CORE THERMAL POWER (% RATED)P i
FIGURE 3.2.1-3 MAPFACp GRAND GULF-UNIT 1 3/4 2-2b Amendment No.16 i
. POWER DISTRIBUTION LIMITS r
3/4.2.2 [0ELETE03
's
- ~,
~
s r
e r
i r
/
J f'
1 4
e l
e i
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r t
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2 GRAND GULF-UNIT 1 3/4 2-3 Amendment No.16 1
1
.._,,_m,,
,ym.,..,-___.
._.- - -, - ~ _. _...
_,~.,.y,-.m_-y-
_m
,__,-.-..~.._F,_
._,-.-,-w___
- -.. ~ _.., _. - -,.. _
d h*[
POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 1
[
- t..
fu3:2.3
Than both MCPRThe MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater and MCPR showninFigurls3.2.3-1and3.2.3-2. limits at indicated core flow and THERMAL POWER as P
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWFR.
ACTION:
With MCPR less than the applicable MCPR limits determined from Figures 3.2.3-1 and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 2SV;ofRATEDTHERMALPOWERwithinthenext4 hours.
S'bRVEILLANCEREQUIREMENTS 4.2.3 MCPR shall be determined to be equal to or greater than the applicable MCPR limits determined from Figures 3.2.3-1 and 3.2.3-2:
a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating c.
with a LIMITING CONTROL ROD PATTERN for MCPR.
d.
The provisions of Specification 4.0.4 are not applicable.
j I
a 9,
l y
9 e
- r O
GRAkuGULF-UNIT 1 3/4 2-4
t s
.1.7 t
l.6
~
T
\\
-FOR MAX FLOW = 107.0%
g,3 3\\\\
i N
s 1.4 FOR MAX FLOW = IO2.$%s
\\
- l. 3 g
E
~ 4 U
A
- 1. 2
\\\\
- 8. I RATED MCPR OPERATING LMT = 1.18 1
I I
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I O
20 40 60 80 100 120 CORE FLOW (% RATED), F r
FIGURE 3.2.3-1 MCPR7 GRAND GULF-UNIT 1 3/4 2-5 Amendment No.16
I I
I 8
3 l
THERNAL POWER 25%iPi40%
2.2 FLOW > 50%
l
\\
2.O i
eTHERNA _ POWER 25% $ Pn40%
l CORE FLOW 5 50%
i b
I.8 THERNAL POWER 4Wo < Pi70%
\\
ALL CORE FLOWS E
w U
2 g,4 CORE OWS N
A
- 1. 2 l.O O
20 40 60 60 10 0 120 l
CORE THERMAL POWER (% RATED)P l
FIGURE 3.2.3-2 MCPR P
GRAND GULF-UNIT 1 3/4 2-6 Amendment No.16
TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a) A channel may be placed in an inoperable status 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 required surveillance without placing the trip system in the tripped condi-tion provided at least one OPERABLE channel in the same trip system is monitoring that parameter.
(b) The " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn
- per Specification 3.9.2 and shutdown margin demonstrations performed per Specification 3.10.3.
(c) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.
(d) This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.
(e) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.
(f) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required.
(g) With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(h) This function shall be automatically bypassed when operating below the appropriate turbine first stage pressure setpoint of:
(1) 1 26.9%** of the value of turbine first-stage pressure at valves wide open (VWO) steam flow when operating with rated feedwater l
temperature of greater than or equal to 420*F, or (2) i 22.5%** of the value of turbine first-stage pressure at VWO steam flow when operating with rated feedwater temperature between 370*F and 420*F.
- Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
- Allowable setpoint values of turbine first-stage pressure equivalent to
- THERMAL POWER less than 40% of RATED THERMAL POWER.
I e
GRAND GULF-UNIT 1 3/4 3-5 Amendment No.16 l
TABLE 3.3.1-2 S
g REACTOR PROTECTION SYSTEM RESPONSE TIMES
.e G;
RESPONSE TIME c'-
FUNCTIONAL UNIT 5
(Seconds)
-d 1.
a.
Neutron Flux - High NA b.
Inoperative NA 2.
Average Power Range Monitor *:
a.
Neutron Flux - High, Setdown NA b.
Flow Biased Simulated Thermal Power - High 5 0.09**
c.
Neutron Flux - High 5 0.09 d.
Inoperative NA 3.
Reactor Vessel Steam Dome Pressure - High 5 0.35 u,
):
4.
Reactor Vessel Water Level - Low, Level 3 5 1.05 5.
Reactor Vessel Water Level - liigh, Level 8 u,
cn' 6.
Main Steam Line Isolation Valve - Closure 5 1.05
< 0.06 7.
Main Steam Line Radiation - High NA 8.
Drywell Pressure - High NA 9.
Scram Discharge Volume Water Level - High NA i
10.
Turbine Stop Valve - Closure 11.
Turbine Control Valve Fast Closure, Valve Trip System
-< 0.10 Oil Pressure - Low
< 0.10, 12.
Reactor Mode Switch Shutdown Position HA i
13.
Manual Scram NA
- Neutron detectors are exempt from response time testing.
Response time shall be measured from the detector output or from the input of the first electronic component in the channel.
- Not including simulated thermal power time constant.
- Medsured from start of turbine control valve fast closure.
e
l o
l s
TABLE 4.3.1.1-1 z
REACTOR PROTECTION SYSTEM INSTRUNENTATION SURVEILLANCE REQUIRENENTS 5
CHANNEL OPERATIONAL I
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH E
FUNCTIONAL UNIT CHECK TEST CALIBRATION (a) SURVEILLANCE REQUIRED' M
1.
a.
Neutron Flux - High S/U.S,(b)
S/U, W R
2 H
S W
R 3,4,5 b.
Inoperative NA W
NA 2,3,4,5 II) 2.
Average Power Range Monitor:
a.
Neutron Flux - High, S/U,5,(b)
S/U, W SA 2
Setdown.
S W
SA 3, 5 b.
Flow Blased Simulated Thermal Power - High 5,D(h) y y(d)(e), SA, R 1
III' Y
c.
Neutron Flux - High S
W W(d), SA 1
O d.
Inoperative NA W
NA 1,2,3,5 1
3.
Reactor Yessel Steam Dome I9)
III Pressure - High S
M R
1, 2 4.
I9)
Low, Level 3 S
M R
1, 2 5.
?
I9)
High, Level 8 S
M R
1 6.
Main Steam Line Isolation Valve - Closure NA M
R 1
7.
Main Steam Line Radiation -
III High 5
M R
1, 2 I9)
I 8.
DryweII Pressure - High S
M R
1, 2
TABLE 4.3.1.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTrip4ENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH q;
J.
FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE REQUIRED zy 9.
Scram Discharge Volume Water R'g) 1,2,5(j) f i
Level - High S
M g
I9) 1 10.
Turbine Stop Valve - Closure S
M R
l
- 11. Turbine Control Valve Fast Closure Valve Trip System Oil R(9) 1 Pressure - Low S
M 12.
Reactor Mode Switch Shutdown Position NA R
NA 1,2,3,4,5 13.
Manual Scrap NA M
NA 1,2,3,4,5
~
wk Neutron detectors may be excluded from CHANNEL CALIBRATION.
y (a)
The IRM and SRM channels shall be determined to overlap for at least 1/2 decade during each m
(b) startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be deter-mined to overlap for at least 1/2 decade during each controlled shutdown, if not performed within the previous 7 days.
[ DELETED]
(c)
This calibration shall consist of the adjustment of the APRM channel to conform to the power values (d) calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED l
THERMAL POWER.
Adjust the APRM channel if the absolute difference is greater tiian 2% of RATED l
THERMAL POWER.
This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a (e) calibrated flow signal.
(f) The LPRMs shall be calibrated at least once per 1000 MWD /T using the TIP system.
Calibrate trip unit at least once per 31 days.
(g)
Verify measured drive flow to be less than or equal to-established drive flow at the existing flow con-y (h) trol valve position.
Tliis calibration shall consist of verifying the 6 i 1 second simulated thermal power time constant.
g g-(i)
Not applicable when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.
g (j)
(k) Not applicable when DRYWELL INTEGRITY is not required.
y (1) Applicable with any control rod withdrawn.
Not applicable to control rods removed per Specifica-r+
tion 3.9.10.1 or 3.9.10.2.
m m
O
t TABLE 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUMOPERABLECHANgggSPER TRIP FUNCTION TRIP SYSTEM 1.
Turbine Stop Valve - Closure 2(b) 2.
Turbine Control Valve - Fast Closure 2(b)
(a) A trip system may be placed in an inoperable status 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 required surveillance provided that the other trip system is OPERABLE.
(b) This function shall be automatically bypassed when operating be'10w the appropriate turbine first stage pressure setpoint of (1) < 26.9% of the value of turbine first-stage pressure at valves wide open (VWO) steam flow when operating with rated feedwater temperature of greater than or equal to 420*F; or (2) < 22.5% of the value of turbine first-stage pressure at VWO steam T10w when operating with rated feedwater temperature between 370*F and 420*F.
These represent allowable setpoint values of turbine first-stage pressure i
equivalent to THERMAL POWER less than 40% of RATED THERMAL POWER.
l i
GRAND GULF-UNIT 1 3/4 3-43 Amendment No.16
4 TABLE 3.3.4.2-2 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM SETPOINTS TRIP ALLOWABLE TRIP FUNCTION SETPOINT VALUE 1.
Turbine Stop Valve - Closure
> 40 psig*
>. 37 psig
~
2.
Turbine Control Valve - Fast Closure 1 44.3 psig*
1 42 psig
- Initial setpoint.
Final setpoint to be determined during startup test program.
Any required change to this setpoint shall be submitted to the Commission within 90 days of test completion.
I e
9 GRAND GULF-UNIT 1 3/4 3-44
as TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS h
TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE T
l 1.
ROD PATTERN CONTROL SYSTEM Q
a.
Low Power Setpoint 20 + 15, -0% of RATED THERMAL 20 + 15. -0% of RATED THERMAL POWER g
POWER b.
High Power Setpoint 5 70% of RATED THERMAL POWER
$ 70% of RATED THERMAL POWER 2.
APRM a.
Flow Biased Neutron Flux-Upscale i
1)
During two recirculation loop operation y
a)
Flow Biased 1 0.66 W+58%, with a maximum 1 0.66 W+61%, with'a maximum i
of of l
y b)
High Flow Clamped 5 108.0% of RATED THERMAL
$ 110.0% of RATED THERMAL t
m POWER POWER 2)
During single recirculation loop operation:
a)
Flow Biased 1 0.66 W+34%
1 0.66 W+37%
b)
High Flow. Clamped Not required OPERABLE Not required OPERABLE b.
Inoperative NA NA c.
Downscale 1 4% of RATED THERMAL POWER 2 3% of RATED THERMAL POWER d.
Neutron Flux - Upscale Startup 5 12% of RATED THERMAL POWER
$ 14% of RATED-THERMAL POWER 3.
SOURCE RANGE MONITORS g
a.
Detector not full in NA NA 5
,e b.
Upscale 5 1 x 10 cps 5 1.5 x 10 cp, i;.
g.
- c., Inoperative NA NA T;
e d.
Downscale 2 0.7 cps 1 0.5 cps
,e I
g j.
l_
l
a TABLE 3.3.6-2 (C:ntinued)
CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS h
TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE
?
E 4.
g a.
Detector not full in NA NA b.
Upscale
$ 108/125 of full scale
$ 110/125 of full scale c.
Inoperative NA NA d.
Downscale 1 5/125 of full scale 1 3/125 of full scale 5.
Water Level-High 5 32 inches 5 33.5 inches 6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW R
a.
Upscale
$ 111% of rated flow
$ 114% of rated flow a
Y 7.
REACTOR MODE SWITCH SHUTDOWN 3
P0SITION-NA NA w
t
,t a
f 4
TABLE 4.3.6-1 CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL OPERATIONAL G
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH E
TRIP FUNCTION CHECK.
TEST CALIBRATION (a)
SURVEILLANCE REQUIRED 5
1.
ROD PATTERN CONTROL SYSTEM
{g
), D(c) b a.
Low Power Setpoint NA Q
1, 2
{gfb),D(c)
Q 1, 2 b.
High Power Setpoint NA 2.
APRM a.
Flow Biased Neutron Flux-II)(9)
Upscale NA W
W
, SA 1
b.
Inoperative NA S/U,W Wg) 1, 2, 5 N
, SA 1
w c.
Downscale NA W
A d.
Neutron Flux - Upscale, Startup NA S/U(b) M Q
2, 5 y
3.
SOURCE RANGE MONITORS a.
Detector not full in NA S/U,W NA 2, 5 b.
Upscale NA S/U,W Q
2, 5 c.
Inoperative NA S/U,W NA 2, 5 d.
Downscale-NA S/U,W Q
2, 5 4.
INTERMEDIATE RANGE MONITORS a.
Detector not full in NA S/U,W NA 2, 5 b.
Upscale NA S/U,W Q
2, 5 c.
Inoperative NA S/U,W NA 2, 5 d.
Downscale NA S/U,W Q
2, 5 5.
Water Level-High NA M
R 1, 2, 5*
6.
REACTOR COOLANT SYSTEM RECIRCULATION FLOW a.
Upscale NA M
Q 1
7.
REACTOR MODE SWITCH SHUTDOWN POSITION NA R
NA 3, 4
)
INSTRUMENTATION TABLE 4.3,6-1 (Continued)
CONTROL ROD BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTES:
a.
Neutron detectors may be excluded from CHANNEL CALIBRATION.
b.
Within 7 days prior to startup.
c.
Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to control rod movement and as each power range above the RPCS low power setpoint is entered for the first time during any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period during power increase or decrease.
d.
At least once per 31 days while operation continues within a given power range above the RPCS low power setpoint.
~
e.
[ Deleted]
f.
This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.
Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.
l g.
This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.
h.
This calibration shall consist of verifying the trip setpoint only.
With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
i GRAND GULF-UNIT 1 3/4 3-57 Amendment No. 16
0 i.
INSTRUMENTATION l
3/4.3.7 MONITORING INSTRUMENTATIO'N RADIATION MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.1.The radiation monitoring instrumentation channels shown in Table 3.3.7.1-1 shall be OPERABLE with their alarm / trip setpoints within the specified limits.
t APPLICABILITY:
As shown in Table 3.3.7.1-1.
ACTION:
a.
With a radiation monitoring instrumentation channel alarm / trip setpoint exceeding the value shown in Table 3.3.7.1-1, adjust the setpoint to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or declare the channel l
b.
With one or more radiation monitoring channels inoperable, take the i
ACTION required by Table 3.3.7.1-1.
l l
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l SURVEILLANCE REQUIREMENTS 4.3.7.1 Each of the above required radiation monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION operations for the conditions and at the frequencies shown in Table 4.3.7.1-1.
l l
l l
GRAND GULF-UNIT 1 3/4 3-58 l
]
o INSTRUMENTATION 3/4.3.10 NEUTRON FLUX MONITORING INSTRUMENTATION
\\
LIMITING CONDITION FOR OPERATION 3.3.10 The APRM and LPRM* neutron flux noise levels shall not exceed three (3) times their established baseline value.
APPLICABILITY:
OPERATIONAL CONDITION 1 with operation in Region I as speci-fied in Figure 3.4.1.1-1.
ACTION:
With no established baseline flux noise. levels, immediately initiate a.
action to either reduce THERMAL POWER to within Region III as speci-fied in Figure 3.4.1.1-1 or increase flow to within Region II as specified in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
b.
With the flux noise levels greater than three (3) times their estab-lished baseline noise levels, initiate corrective action within 15 minutes to reduce the noise levels to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; if unsuccessful, either reduce THERMAL POWER to within Region III as specified in Figure 3.4.1.1-1 or increase. flow to within Region II as specified in Figure 3.4.1.1-1 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
SURVEILLANCE REQUIREMENTS 4.3.10.1 The APRM and LPRM* neutron flux noise levels shall be determined to be less than or equal to the limit of Specification 3.3.10:
a.
Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after entering the applicable region, and b.
At least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, and Within 30 minutes after completion of a change in THERMAL POWER of c.
at least 5% of RATED THERMAL POWER.
The provisions of Specification 4.0.4 are not applicable.
4.3.10.2 Two loop baseline APRM and LPRM neutron flux noise levels shall be established at a point in Region II less than 60% of rated total core flow prior to operation in Region I of Figure 3.4.1.1-1 provided the baseline has not been established since the last CORE ALTERATION.
- Detector A and C of one LPRM string per core octant plus detector A and C of one LPRM string in the central regicn of the core shall be monitored.
i i
GRAND GULF-UNIT 1 3/4 3-111 Amendment No.16
.--.-----x
,,---,--,-.m.---
v1--,-
y
..wi--
.,-w----..-..w-.-y-.-,.,...--------,,,--,.--,---.,-r----
---r
- - - - - - - - -w--ww
INSTRUMENTATION SURVEILLANCE' REQUIREMENTS (Continued ~)
4.3.10.3 Single loop baseline APRM and LPRM neutron flux noise levels shall be cstablished at a point in Region II less than 60% of rated core flow prior to
}
single loop operation in Region I of Figure 3.4.1.1-1 provided the baseline has nit been established since the last CORE ALTERATION; or in lieu of est'ablishing.
single loop baseline data, the baseline established in 4.3.10.2 may be utilized for single loop operation in Region I of Figure 3.4.1.1-1.
l l
l GkAhD CULF-UNIT 1 3/4 3-112 Amendment No. 16
3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS 4
LIMITING CONDITION FOR OPERATION 3.4.1.1 The reactor coolant recirculation system shall be in operation and not in Region IV as specified in Figure 3.4.1.1-1 with either:
a.
Two recirculation loops operating with limits and setpoints per Specifications 2.1.2, 2.2.1, 3.2.1, and 3.3.6, or b.
A single recirculation loop operating with:
1.
A volumetric loop flow rate less than 44,600 gpm, and 1
i 2.
The loop recirculation flow control in the manual mode, and 3.
Limits and setpoints per Specifications 2.1.2, 2.2.1; 3.2.1, and 3.3.6.
APPLICABILITY:
OPERATIONAL CONDITIONS 1* and 2*.
ACTION:
a.
During single loop operation, with the volumetric loop flow rate greater than the above limit, immediately initiate corrective action to reduce flow to within the above limit within 30 minutes.
j b.
During single loop operation, with the loop flow control nct in the manual mode, place it in the manual mode within115 minutes.
i i
c.
With no reactor coolant system recirculation loops in operation, imme-i diately initiate an orderly reduction of THERMAL POWER to within Region III as specified in Figure 3.4.1.1-1, and initiate measures to place the unit in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
i d.
During single loop operation, with temperature differences exceeding l
the limits of SURVEILLANCE REQUIREMENT 4.4.1.1.5, suspend the THERMAL POWER or recirculation loop flow increase.
e.
With operation in Region IV as specified in Figure 3.4.1.1-1, initiate corrective action within 15 minutes to either reduce power to within Region III of Figure 3.4.1.1-1 or increase flow to w.ithin i
Region I or Region II of Figure 3.4.1.1-1 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
j "See Special Test Exception 3.10.4.
i GRAND GULF-UNIT 1 3/4 4-1 Amendment No.16
~ -
REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION'(Continued) f.
With a change in reactor operating conditions, from two recircula-tion loops operating to single loop operation, or restoration of two loop operation, the limits and setpoints of Specifications 2.1.2, 2.2.1, 3.2.1, and 3.3.6 shall be implemented within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or declare the associated equipment inoperable (or the limits to be "not satisfied"), and take the ACTIONS required by the referenced specifications.
SURVEILLANCE REQUIREMENTS 4.4.1.1.1 The reactor coolant recirculation system shall be verified to be in cperation and not in Region IV of Figure 3.4.1.1-1 at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.1.1.2 Each reactor coolant system recirculation loop flow control valve in an operating loop shall be demonstrated OPERABLE at least once per 18 months by:
Verifying that the control valve fails "as is" on loss of hydraulic a.
pressure at the hydraulic unit, and b.
Verifying that the average rate of control valve movement is:
1.
Less than or equal to 11% of stroke per second opening, and 2.
Less than or equal to 11% of stroke per second closing.
4.4.1.1.3 During single loop operation, verify that the loop recirculation flow control in the operating loop is in the manual mode at least once per 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
I 4.4.1.1.4 During single loop operation, verify that the dolumetric loop flow rate of the loop in operation is within the limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.1.1.5 During single loop operation, and with both THERMAL POWER less than 36% of RATED THERMAL POWER and the operating recirculation pump not on high spied, verify that the following differential temperature requirements are met within 15 minutes prior to beginning either a THERMAL POWER increase or a recirculation loop flow increase and within every hour during the THERMAL POWER or recirculation loop flow increase:
Less than 100*F, between the reactor vessel steam space coolant and a.
the bottom head drain line coolant, and b.
Less than 50*F, between the coolant of the loop not in operation and the coolant in the reactor vessel, and Less than 50*F, between the coolant in the operating loop and the c.
coolant in the loop not in operation.
GRAND GULF-UNIT 1 3/4 4-la Amendment No.16 f
~
SURVEILLANCE REQUIREMENTS (Continued)
The differential temperature requirements 4.4.1.1.5.b and c do not apply when the loop not in operation is isolated from the reactor pressure vessel.
4.4.1.1.6 The limits and setpoints of Specifications 2.2.1, 3.2.1, and 3.3.6 shall be verified to be within the appropriate limits within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of an operational change to either one or two loops operating.
i i
l i
1 GRAND GULF-UNIT 1 3/4 4-lb Amendment No. 16
W e
02 1
0 1
1 0
E 0
N 1
I L
DO 0
R 9
08 P
0 A
8 M
G N
I W
T 0O AR 7 L E
N F
P IO O
Gn w
E N
R W
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OGg O
O R
l I
0 L
F C
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F R
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54 T
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-1 N
T ili I'
ll il O
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4 R 3
l.l' lIIIl;lj' l EP E
RUG 0
I w
3 F
o LF
%e 0*
s N
2 Oa I
G E
R 0
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0 0
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=
0 3
0 4
m
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1 e
7 6
1 0
1 1
gW[ d2 wE o"5 O p $ e 'E
@ r[3tH
- e. iN AS2
j REACTOR COOLANT SYSTEM JET PUMPS
~
LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.
l APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
SURVEILLANCE REQUIREMENTS 4.4.1.2.1 Each of the above required jet pumps in an operating loop shall be l
~
demonstrated OPERABLE with THERMAL POWER in excess of 25% of RATED THERMAL POWER and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by determining recirculation loop flow, total core flow and diffuser-to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur:
l a.
The indicated recirculation loop flow differs by more than 10% from the established flow control valve position-loop flow characteristics.
b.
The indicated total core flow differs by more than 10% from the established total core flow value derived from recirculation loop flow measurements.
The indicated diffuser-to-lower plenum differential pressure of any c.
4 individual jet pump differs from established patterns by more than 10%.
f 4.4.1.2.2 The provisions of Specification 4.0.4 are not applicable provided the diffuser-to-lower plenum differential pressures of the individual jet pumps are determined to be within 50%* of the loop average within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after entering OPERATIONAL CONDITION 2 and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.
l
- Initial value.
Final value to be determined during startup test program.
Any required changes to the value shall be submitted to the Commission within 90 days of test completion.
I i
GRAND GULF-UNIT 1 3/4 4-2 Amendment No 16 l
REACTOR COOLANT SYSTEM RECIRCULATION LOOP FLOW
~
LIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation loop flow mismatch, when two loops are in operation, l
shall be maintained within:
5% of rated recirculation flow with core flow greater than or equal a.
to 70% of rated core flow.
b.
10% of rated recirculation flow with core flow less than 70% of rated core flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.
ACTION:
l With recirculation loop flows different by more than the specified limits, restore the recirculation loop flows to within the specified limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
If unsuccessful, either:
Shut down one recirculation loop and comply with the requirements of a.
Specification 3.4.1.1, or b.
Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
l l
SURVEILLANCE REQUIREMENTS 1
4.4.1.3 Recirculation loop flow mismatch shall be verified to be within the limits at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
1
- See Special Test Exception 3.10.4.
GRAND GULF-UNIT 1 3/4 4-3 Amendment No.16
REACTOR COOLANT SYSTEM IDLE RECIRCULATION LOOP STdRTUP LIMITING CONDITION FOR OPERATION 1
3.4.1.4 An idle recirculation loop shall not be started unless the temperature differential between the reactor pressure vessel steam space coolant and the bottom head drain line coolant is less than or equal to 100F,* and:
When both loops have been idle, unless the temperature differential
^
a.
between the reactor coolant within the idle loop to be started up and the coolant in the reactor pressure vessel is less than or equal to 50*F, or i
b.
When only dne loop has been idle, unless the temperature differential between the reactor coolant within the idle and operating recirculation loops is less than or equal to 50'F and the operating loop flow rate is less than or equal to 50% of rated loop flow.
l l
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, 3 and 4.
ACTION:
l With temperature differences and/or flow rates exceeding the above limits, suspend startup of any idle recirculation loop.
i SURVEILLANCE REQUIREMENTS 4.4.1.4 The temperature differentials and flow rate shall be determined to be within the limits within 15 minutes prior to startup of an idle recirculation I
loop.
CBelow 25 psig, the temperature differential is not applicable.
GRAND GULF-UNIT 1 3/4 4-4
3/4.1 REACTIVITY CONTROL SYSTEMS BASES 3/4.1.1 SHUTDOWN MARGIN A sufficient SHUTDOWN MARGIN ensures that 1) the reactor can be made sub-critical from all operating conditions, 2) the reactivity transients associated with postulated accident conditions are controllable within acceptable limits, and 3) the reactor will be maintained sufficiently subcritical to preclude inadvertent criticality in the shutdown condition.
Since core reactivity values will vary through core life as a function of fuel depletion and poison burnup, the demonstration of SHUTDOWN MARGIN will be performed in the cold, xenon-free condition and shall show the core to be subcritical by at least R + 0.38% delta k/k or R + 0.28% delta k/k, as appro-priate.
The value of R in units of % delta k/k is the difference between the calculated value of maximum core reactivity during the operating cycle and the
- calculated beginning-of-life core reactivity.
The value of R must be positive
.; or zero and must be determined for each fuel loading cycle.
Two different values are supplied in the Limiting Condition for Operation f to provide for the different methods of demonstration of the SHUTDOWN MARGIN.
I The highest worth rod may be determined analytically or by test.
The SHUTDOWN I MARGIN is demonstrated by an insequence control rod withdrawal at the beginning i of life fuel cycle conditions, and, if necessary, at any future time in the a cycle if the first demonstration indicates that the required margin could be reduced as a function of exposure.
Observation of subtriticality in this condi-tion assures subcriticality with the most reactive control rod fully withdrawn.
This reactivity characteristic has been a basic assumption in the analysis of plant performance and can be best demonstrated at the time of fuel loading, but the margin must also be determined anytime a control rod is incapable of insertion.
3/4.1.2 REACTIVITY ANOMALIES Since the SHUTDOWN MARGIN requirement for the reactor is small, a careful check on actual conditions to the predicted conditions is necessary, and the changes in reactivity can be inferred from these comparisons of rod patterns.
Since the comparisons are easily done, frequent checks are not an imposition on normal operations.
A 1% change is larger than is expected for normal operation so a change of this magnitude should be thoroughly evaluated.
A change as large as 1% would not exceed the design conditions of the reactor l
and is on the safe side of the postulated transients.
GRAND GULF-UNIT 1 B 3/4 1-1 u
-r
--w-
-w-
--,-w,---------
,,n..-
w-
--,-p.-
-.v,-
=
P
_y L
- l REACTIVITY CONTROL SYSTEMS r
g
^
BASES 3/4.1.3 CONTROL RODS
[
The specifications of this section ensure that (1) the minimum SHUTDOWN g
~
MARGIN is maintained, (2) the control rod insertion times are consistent with
=
those used in the accident, non-accident and transient analyses, and (3) the
}
potential effects of the rod drop accident and rod withdrawal error event are i
limited.
The ACTION statements permit variations from the basic requirements B
but at the same time impose more restrictive criteria for continued operation.
A limitation on inoperable rods is set such that the resultant effect on total 1
rod worth and scram shape will be kept to a minimum.
The requirements for the
{
various scram time measurements ensure that any indication of systematic prob-l*
f lems with rod drives will be investigated on a timely basis.
j Damage within the control rod drive mechanism could be a generic problem, i
/
therefore with a control rod immovable because of excessive friction or mechani-cal interference, operation of the reactor is limited to a time period which
~
is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.
I Control rods that are inoperable for other reasons are permitted to be j-taken out of service provided that those in the nonfully-inserted position are q
j-consistent with the SHUTDOWN MARGIN requirements.
The number of control rods permitted to be inoperable but trippable could
[
be more than the eight allowed by the specification, but the occurrence of eight F
inoperable rods could be indicative of a generic problem and the reactor must be shut down for investigation and resolution of the problem.
The control rod system is designed to bring the reactor subcritical at a
-g rate fast enough to prevent the MCPR from becoming less than the Safety Limit l
=
during the limiting power transient analyzed in Section 15.4 of the FSAR.
This
?
analysis shows that the negative reactivity rates resulting from the scram with E
the average response of all the drives as given in the specifications, provide
-]
E the required protection and MCPR remains greater than the Safety Limit.
The l
5 occurrence of scram times longer than those specified should be viewed as an 1
[
indication of a systematic problem with the rod drives and therefore the sur-l J
F veillance interval is reduced in order to prevent operation of the reactor for 2
long periods of time with a potentially serious problem.
-2 I
The scram discharge volume is required to be OPERABLE so that it will be 5
available when needed to accept discharge water from the control rods during a j
c reactor scram and will isolate the reactor coolant system from the containment E
when required, g
E Control rods with inoperable accumulators are declared inoperable and Spec-ification 3.1.3.1 then applies.
This prevents a pattern of inoperable accumu-
]
lators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be c
slowly scrammed via reactor pressure or inserted with normal drive water pres-
}
sure. Operability of the accumulator ensures that there is a means available E
to insert the control rods even under the most unfavorable depressurization of i
the reactor.
m 2
d A
GRAND GULF-UNIT 1 B 3/4 1-2 Amendment No. 16 m
3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not i
exceed the 2200'F limit specified in 10 CFR 50.46.
J 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following I
the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.
The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondar-ily on the rod to rod power distribution within an assembly. The peak clad tem-perature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR i
' times 1.02 is used in the heatup code along with the exposure dependent steady j
state gap conductance and rod-to-rod local peaking factor.
The Technical Speci-fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor.
The MAPLHGR limits of Figure 3.2.1-1 are multiplied by the smaller of either the flow-dependent MAPLHGR factor (MAPFAC ) or the power-dependent MAPLHGR factor (MAPFAC ) corre-f p
sponding to existing core flow and power state to ensure the adherence to fuel mechanical design bases during the most limiting transient.
The maximum factor for single loop operation is 0.86.
MAPFAC 's are determined using the three-dimensional BWR simulator code to f
analyze slow flow runout transients.
Two curves are provided for use based on the existing setting of the core flow limiter in the Recirculation Flow Control System.
The curve representative of a maximum core flow limit of 107.0% is more restrictive due to the larger potential flow runout transient.
f MAPFAC 's are generated using the same data base as the MCPR to protect p
p l
the core from plant transients other than core flow increases.
The daily requirement for calculating APLHGR when THERMAL POWER is greater j
than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-j tion shifts are very slow when there have not been significant power or control rod changes.
The requirement to calculate APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER i
ensures thermal limits are met after power distribution shifts while still l
allotting time for the power distribution to stabilize.
The requirement for calculating APLHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit.,
i GRAND GULF-UNIT 1 B 3/4 2-1 Amendment No.16 l
POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)
The calculational procedure used to establish the APLHGR limits is based
'l cn a loss-of-coolant accident analysis.
The analysis was performed using General Electric (GE) calculational models which are consistent with the rcquirements of Appendix K to 10 CFR 50.
A complete discussion of each code employed in the analysis is presented in References 1 and 6.
Differences in l
this analysis compared to previous analyses can be broken down as follows.
a.
Input Chances 1.
Corrected Vaporization Calculation - Coefficients in the vaporization correlation dsed in the REFLOOD code were corrected.
2.
Incorporated more accurate bypass areas - The bypass ateas in the top guide were recalculated using a more accurate technique.
3.
Corrected guide tube thermal resistance.
4.
Correct heat capacity ^of reactor internals heat nodes.
i i
(
i
- l GRAND GULF-UNIT 1 B 3/4 2-la Amendment No.16
. ;p c
POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued) i
(,
b.
Model Change 1.
Core CCFL pressure differential - 1 psi - Incorporate the a'ssumption
". / '"
that flow from the bypass to lower plenum must overcome a 1 psi pressure drop in core.
2.
Incoporate NRC pressure transfer assumption - The assumption used in I
the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.
A few of the changes affect the accident calculation irrespective of CCFL.
These changes are listed below.
' I a.
Input Change 1.
Break Areas - The DBA break area was calculated more accurately.
b.
Model Change 1.
Improved Radiation and Conduction Calculation - Incorporaf. ion of CHASTE 05 for heatup calculation.
A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.
3/4.2.2 [ DELETED]
s.
f f
.c J
l i
GRAND GULF-UNIT 1 B 3/4 2-2 Amendment No.16
77, j#
1p Bases Table B 3.2.1-1 n.
SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS l
r Plant Parameters; Core THERMAL POWER....................
3993 MWt* which corresponds y
to 105% of rated steam flow Vessel Steam Output...................
17.3 x 10s 1bs/hr which cor-responds to 105% of rated steam flow J
Vessel Steam Dome Pressure.............
1060 psia Design Basis Recirculation Line Break Area for:
a.
Large Breaks 3.1 ft2, b.
Small Breaks 0.1 ft2, Fuel Parameters:
P r
PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GE0 METRY (kW/ft)
FACTOR RATIO Initial Core 8-x 8 RP 13.4
- 1. 4 MCPR **
f
' A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 6.3.3 of the FSAR.
- This power level meets the Appendix requirement of 102%.
The core heatup
y calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.
- During single loop operation, departure from nucleate boiling is assumed to i
occur 0.1 second following the LOCA regardless of initial MCPR.
l v~ i.f,
"i
,,,/
i/r
/
L' GRAND GULF-UNIT 1 B 3/4 2-3 Amendment No.16
5 POWER DISTRIBUTION LIMITS BASES 3/4.2.3 MINIMUM CRITICAL POWER RATIO x
The required operating limit MCPRs at steady state operating conditions es specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational trahs-l 1ents.
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is re quired that the resulting MCPR does not decrease below the Safety Limit MCPR at cny time during the transient assuming instrument trip setting given in Specification 2.2.
To assure that the fuai c? adding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO'(CPR).
The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and co_olant temperature decrease.
The limiting transient yields the largest delta CPR.h When added to the Safety Limit MCPR, the required cperating limit MCPR of Specification 3.2.3 is obtained.
The power-flow map of Figure B 3/4 2.3-1 defines the analytical basis for generation of the MCPR operating limits.
The evaluation of a given transient begins with the system initial parameters shown in FSAR Table 15.0-2 and in Table 15.C.3-1 of Reference 5 that are input to a GE-core dynamic behavior transient computer program.
The evaluation of transients during operation in the ME00 begins with the system initial parameters shown in Tables 15.D.4.-2 and 3 of Reference 7.
The code used to eval-l uste pressurization events?is described in NEDO-24154(3) and the program used in ncn pressurization events'is described in NEDO-10802(2)
The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient. thermal hydraulic TASC code described in NEDE-25149(4)
The principal result of this evaluation is the reduction in MCPR caused by the transient.
The purpose of the MCPR and MCPR is to define o'perating limits at other f
p than rated core flow and power conditions.
The MCPR s are established to protect the core from inadvertent core flow f
increases such that the 99.9% MCPR limit requirement can be assured.
The ref-crence core flow increase event used to establish the MCPR is a hypothesized f
slow flow runout to maximum, that does not result'in a scram from neutron flux overshoot exceeding the APRM neutron flux-high level (Table 2.2.1-1 item 2).
The maximum runout flow value is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System.
Two flow rates have been censidered:
102.5% core flow and 107.0% core flow (for increased Core Flow oper-ation). With this basis, the MCPR curves are generated from a series of steady f
state core thermal hydraulic calculations performed at several core power and flow conditions along the steepest flow control line.
In the actual calculations a conservative highly steep generic repres'entation of the 105% steam flow rod-line flow control line has been used.
Assumptions used in the original calcula-tions of this generic flow control line were consistent with a slow flow increase transient duration of several minutes:
(a) the plant heat balance was assumed l
GRAND GULF-UNIT 1 8 3/4 2-4 Amendment No. 16 I
4 m
11 0 z
A - NATURAL CIRCULATION D
3 - Low RECIRCUt.ATioN PUMP SPEEo MLVE MINIMUM POSITION m
C - LOW REORCULATION PUMP SPEED WI.VE MAXIMUM POSITION 10 0 C
o - IIATED MECNICULATeoM PUtr SPEED WLVE MINIMUM POSITioM 4
Te o+
m
,e e
's W
3 5
i 30 8
3=
i N
~
04 C
g
=
2 o 'd 1*
i 40 p
Aj 1
/
O
/
D' i
e B"
CW#
o ui 4o
,3
$L 30 CAVITATION REGION ins to 10 I
o g
o 10 to 30 40 so so 70 so so 10 0 110 12 0 g
Percent of RATED CORE FLOW 5
g FIGURE B 3/4 2.3-1 POWER - FLOW OPERATING MAP l
em i
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) to be in equilibrium, and (b) core xenon concentration was assumed to be constant The generic flow control line is used to define several core power / flow states at which to perform steady-state core thermal-hydraulic evaluations.
The first state analyzed corresponded to the maximum core power at maxi-tua core flow (either 102.5% for Rated Core Flow operation or 107% of rated for Increased Core Flow operation) after the flow runout.
Several evaluations were parformed at this state iterating on the normalized core power distribution input until the limiting bundle MCPR just exceeded the safety limit Specifica-tion (2.1.2).
Next, similar calculations of core MCPR performance were deter-mined at other power / flow conditions on the generic flow control line, assuming the same normalized core power distribution.
The result is a definition of the MCPRf performance requirement such that a flow increase event to maximum will not violate the safety limit.
(The assumption of constant power distri-bution during the runout power increase has been shown to be conservative.
Increased negative reactivity feedback in the high power limiting bundle due to doppler and voids would reduce the limiting bundle relative power in an actual runout.)
The MCPR is established to protect the core from plant transients other p
than core flow increase including the localized rod withdrawal error event.
i Core power dependent setpoints are incorporated (incremental control rod with-drawal limits) in the Rod Withdrawal Limiter (RWL) System Specification (3.3.6).
These setpoints allow greater control rod withdrawal at lower core powers where core thermal margins are large.
However, the increased rod withdrawal requires higher initial MCPR's to assure the MCPR safety limit Specification (2.1.2) is i
not violated.
The analyses that establish the power dependent MCPR require-ments that support the RWL system are presented in GESSAR II, Appendix 158.
For core power below 40% of RATED THERMAL POWER, where the EOC-RPT and the reactor scrams on turbine stop valve closure and turbine control valve fast closure are bypassed, separate sets of MCPR limits are provided for high and
)
p i
low core flows to account for the significant sensitivity tc initial core j
flows.
For core power above 40% of RATED THERMAL POWER, bounding power-dependent MCPR limits were developed.
The abnormal operating transients analyzed for single loop operation are discussed in Reference 5.
The current MCPR limits were found to be bounding.
No change to the operating MCPR limit p
is required for single loop operation.
At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the modera-tor void content will be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerabfe margin.
GRAND GULF-UNIT 1 B 3/4 2-6 Amendment No.
16
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued)
During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed.
The,MCPR margin will thus be demonstrated such that future MCPR evaluation be-low this power level will be shown to be unnecessary.
The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.
The requirement to calculate MCPR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER in-crease of at least 15% of RATED THERMAL POWER ensures thermal limits are met l
l l
i GRAND GULF-UNIT 1 B 3/4 2-6a Amendment No. 16
[
POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Continued) cfter power distribution shifts while still a11otting time for the power dis-tribution to stabilize.
The requirement for calculating MCPR after initially datermining a LIMITING CONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, that could place cperation exceeding a thermal limit.
3/4.2.4 LINEAR HEAT GENERATION RATE i
This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet dInsification is postulated.
The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distri-bution shifts are very slow when there have not been significant power or control rod changes.
The requirement to calculate LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts..while still allotting time for the power distribution to stabilize.
The requirement for calculating LHGR after initially determining a LIMITING CONTROL ROD PATTERN cxists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit.
References:
i 1.
. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.
2.
R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NEDO-10802).
3.
Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors, NEDO-24154, October 1978.
4.
TASC 01-A Computer Program for The Transient Analysis of a Single Channel, Technical Description, NEDE-25149, January 1980.
5.
GGNS Reactor Performance Improvement Program, Single Loop Operation Analysis, General Electric Final Report, February 1986.
i 6.
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, Amendment 2, One Recircula-tion Loop Out-of-Service, NEDO-20566-2, Revision 1, July 1978.
7.
General Electric Company, " Maximum Extended Operating Domain Analysis," March 1986.
GRAND GULF-UNIT 1 B 3/4 2-7 Amendment No. 16
o INSTRUMENTATION BASES RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued) feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.
A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the 'other two turbine control valves provides input to the second EOC-RPT system.
Similarly, a closure sensor for each of two turbine stop valves provides input to one EOC-RPT system; a closure sensor from each of the other two stop valves provides input to the other EOC-RPT system.
For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closure of turbine control valves and a 2-out-of-2 logic for the turbine stop valves.
The operation of either logic will actuate the EOC-RPT system and trip both recirculation pumps.
Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administrative 1y controlled.
The manual bypasses and the automatic Operating Bypass at less than 40% of RATED THERMAL POWER are annunciated in the control The automatic bypass setpoint is feedwater temperature dependent due to l
room.
the subcooling changes that affect the turbine first-stage pressure reactor power relationship.
For RATED THERMAL POWER operation with feedwater tempera-ture greater than or equal to 420*F, an allowable setpoint of 1 26.9% of control valve wide open turbine first-stage pressure is provided for the bypass func-tion.
This setpoint is also applic'ble to operation at less than RATED THERMAL a
POWER with the correspondingly lower feedwater temperature.
The allowable setpoint is reduced to 1 22.5% of control valve wide open turbine first-stage pressure for RATED THERMAL POWER operation with a feedwater temperature between 370'F and 420*F.
Similarly, the reduced setpoint is applicable to operation at less than RATED THERMAL POWER with the correspondingly lower feedwater temperature.
The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e., 190 ms, less the time allotted from start of motion of the stop valve or turbine control valve until the sensor relay contact supplying the input to the reactor protection system opens, i.e., 70 ms, and less the time allotted for breaker arc suppression determined by test, as correlated'to manufacturer's test results, i.e., 50 ms, and plant pre-operational test results.
Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater than the drift allowance assumed for each trip in the safety analyses.
3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION The reactor core isolation cooling system actuation instrumentation 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 providing actuation of any of the emergency core cooling equipment.
Operation with a trip set less conservative than its Trip Setpoin't but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater than the drift allowance assumed for each trip in the safety analyses.
GRAND GULF-UNIT 1 B 3/4 3-3 Amendment No.16
INSTRUMENTATION BASES 3/4.3.6 CONTROL R0D BLOCK INSTRUMENTATION The control rod block functions are provided consistent with the require-ments 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 th:t a trip in any one of the inputs will result in a control rod block.
The OPERABILITY of the control rod block instrumentation in OPERATIONAL CONDITION 5 is to provide diversity of rod block protection to the one-rod-out interlock.
i GRAND GULF-UNIT 1 B 3/4 3-3a Amendment No. 16
'l INSTRUMENTATION BASES 3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION (Continued) 1 Operation with a trip set less conservative than its Trip Setpoint but l
within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater than the drift allowance assumed for.each trip in the safety analyses.
3/4.3.7 MONITORING INSTRUMENTATION 3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that:
(1) the radiation levels are continually measured in the areas served by the individual channels; (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded; and (3) sufficient information is available on selected plant parameters to monitor and assess these variables following an accident.
This capability is consistent with the recommendations of NUREG-0737, " Clarification of TMI Action Plan Requirements," November, 1980.
3/4.3.7.2 SEISMIC MONITORING INSTRUMENTATION The OPERABILITY of the seismic monitoring instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety.
This capability is required to permit comparison of the measured response to l
that used in the design basis for the unit.
3/4.3.7.3 ' METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of the meteorological monitoring instrumentation ensures that sufficient meteorological data are available for estimating potential radiation doses to the public as a result of routine or accidental release of radioactive materials to the atmosphere.
This capability is required to evaluate the need for initiating protective measures to protect the health and safety of the public.
This instrumentation is consistent with the recommendations of Regulatory Guide 1.23 "Onsite Meteorological Programs," February, 1972.
3/4.3.7.4 REMOTE SHUTDOWN SYSTEM INSTRUMENTATION AND CONTROLS The OPERABILITY of the remote shutdown system instrumentation and controls ensures that sufficient capability is available to permit shutdown and main-tenance of HOT SHUTDOWN of the unit from locations outside of the control room.
This capability is required in the event control room habitability is lost and is consistent with General Design Criteria 19 of 10 CFR 50.
3/4.3.7.5 ACCIDENT MONITORING INSTRUMENTATION The OPERABILITY of the accident monitoring instrumentation ensures that sufficient information is available on selected plant parameters to monitor and assess important variables following an accident.
This capability is consistent with the recommendations of NUREG-0578, "TMI-2 Lessons Learned Task Force Status Report and Short-Term Recommendations".
GRAND GULF-UNIT 1 8 3/4 3-4
INSTRUMENTATION BASES 3/4.3.9 TURBINE OVERSPEED PROTECTION This specification is provided to ensure that the turbine overspeed
~
protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed.
Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety related components, equipment or structures.
3/4.3.10 NEUTRON FLUX MONITORING INSTRUMENTATION This specification is to ensure that neutron flux limit cycle oscillations are detected and suppressed.
j Stability tests at operating plants were reviewed to identify a region of the operating map where surveillance should be performed.
To account for varia-bility, a conservative decay ratio of 0.6 was chosen as the basis for defining
~
the region of potential instability.
The resulting region corresponds to core flow less than 45% of rated and THERMAL POWER greater than the 80% rod line.
The 80% rod line is illustrated in Figure 3.4.1.1-1.
Neutron flux noise limits are also established to ensure the early detec-tion of limit cycle oscillations.
Typical APRM neutron flux noise levels at up to 12% of rated power have been observed.
These levels are easily bounded by values considered in the thermal / mechanical fuel design.
Stability tests have shown that limit cycle oscillations result in peak-to peak magnitude of 5 to 10 times the typical values.
Therefore, actions taken to suppress flux oscillations exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle oscillations.
The specification includes the surveil-lance requirement to establish the requisite baseline noise data and prohibits operation in the region of potential instability if the appropriate baseline data is unavailable.
i I
GRAND GULF-UNIT 1 B 3/4 3-7 Amendment No.16 l
J
.-,,,,_,_..__._,_._,..x..
.y m.
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- _ - _ _ _ _, _ - ~,.. _.
5007 NCTE: SCALE IN INCHES WATER LEVEL NOMENCLATURE ABOVE VESSEL Z8R0 HEIGHT ABOVE HEIGHT RELATIVE VESSEL ZERO TO INSTRUMENT LEVEL NO.
(INCHESI ZEMO (INCHES)
(8) 586.5 53.5 800 - -
(7) 573.7 40.7 (4) 565.7 32.7 (3) 544.4
. i t.4 750 - -
(2) 491.4
- 41.6 VESSEL-,-
(1) 382.7
-15 0.3 FLANGE w
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I648-MA!N STEAM LINE e00- -
-S66.S(0)
+60-~
515 "O
II d ~~40.7
-Hi ALARM 560~2565.7t4) gg 32.7 (4) _to ALARg Ag RT Sl7.3g--533-IPETR. MENT-
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( 33' Ax SCRAM
~0444(3I ARATCR
^"0 0
- 0. -
ZERO DownsMiPT "II' g RECIRC. PUMPS
- 4944(2)
> -- 41.6 (2)
= 479.3 d pE NTIATE ACIC. AND HPCS. TRIP RECIRC.
450" PUMPS : START Div III z
DIESEL ISCLATE CTMT.
j AND AUX. SLOG. AND QW.
g 4!6.3 400- -
8 -- 150.3 (t )
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C imTIATE R* LPCS AND 350-ut366,3 START OtESEL: Olv I AN02r.
CONTR19UTE TO A.D.S. AND CLOSE MSIV S g
z 300- -
1 ACTIVE FUEL
~
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w 216.3 - -
200-= 216.3 (8) TRIP MAIM AND RCic.
TURSINES : CLCSE HPCS.
/
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CISCHARGE #
$J. PAyH i R X SCRAM I
PECIRC 172.3 o-NOZZLE SUCTION is N0ZZLE
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103- -
50 - -
BASES FIGURE B 3/4 3-1 REACTOR VESSEL WATER LEVEL GRAND GULF-UNIT 1 B 3/4 3-8
3/4.4 REACTOR COOLANT SYiTEM BASES I
3/4.4.1 RECIRCULATION SYSTEM Operation with one reactor core coolant recirculation loop inoperable has been evaluated and found to remain within design limits and safety margins pro-vided certain limits and setpoints are modified.
The "GGNS Single Loop Opera-tion Analysis" identified the fuel cladding integrity Safety Limit, MAPLHGR limit and APRM setpoint modifications necessary to maintain the same margin of
]
safety for single loop operation as is available during two loop operation.
Additionally, loop flow limitations are established to ensure vessel internal i
vibration remains within limits. A flow control mode restrict. ion is also incorporated to reduce valve wear as a result of automatic flow control attempts and to ensure valve swings into the cavitation region do not occur.
An inoperable jet pump is not, in itself, a sufficient reason to declare i -
a recirculation loop inoperable, but it does, in case of a design-basis-accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump 1noperable.
Jet pump failure can be detected by monitoring jet pump per-1 formance on a prescribed schedule for significant degradation.
During two loop operation, recirculation loop flow mismatch limits are in compliance with ECCS LOCA analysis design criteria.
The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA.
In where the mismatch limits cannot be maintained, continued operation,casesis per-mitted with one loop in operation.
In accordance with BWR thermal hydraulic stability recommendations, opera-tion above the 80% rod line with flow less than 39% of rated core flow is restricted.
In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within.50'F of each other prior to startup of an idle loop.
The loop temperature must also be within 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.
Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 100'F.
During single loop operation, the condi-tion may exist in which the coolant in the bottom head of.the vessel is not circulating. These differential temperature criteria are also to be met prior to power or flow increases from this condition.
The recirculation flow control valves provide regulation of individual recirculation loop drive flows; which, in turn, will vary the flow rate of coolant through the reactor core over a range consistent with the rod pattern and recirculation pump speed.
The recirculation flow control system consists of the electronic and hydraulic components necessary for the positioning of the two hydraulically actuated flow control valves.
Solid state control logic will generate a flow control valve " motion inhibit" signal in response to any one of several hydraulic power unit or analog control circuit failure signals.
The " motion inhibit" signal causes hydraulic power unit shutdown and hydraulic isolation such that the flow control valve fails "as is."
This design feature insures that the flow control valves do not respond to potentially erroneous control signals.
GRAND GULF-UNIT 1 B 3/4 4-1 Amendment No.16
REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM (Continued)
Electronic limiters exist in the position control loop of each flow control vnive to limit the flow control valve stroking rate to 1011% per second in the op:ning and closing directions on a control signal failure. The analysis of tha recirculation flow control failures on increasing and decreasing flow are prosented in Sections 15.3 and 15.4 of the FSAR respectively.
The required surveillance interval is adequate to ensure that the flow control valves remain OPERABLE and not so frequent as to cause excessive wear on the system components.
6 i
GRAND GULF-UNIT 1 B 3/4 4-la Amendment No. 16 j
L REACTOR COOLANT SYSTEM BASES i
3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves (SRV) operate to prevent the reactor coolant system from being pressurized above the Safety Limit i
of 1325 psig in accordance with the ASME Code.
A total of 13 OPERABLE safety /
4 relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.
Any combination of 6 SRVs operating-in the relief mode and 7 SRVs operating in the safety mode is acceptable.
Demonstration of the safety / relief valve lift settings will occer only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.
The low-low set system ensures that safety / relief valve discharges are minimized for a second opening of these valves, following any overpressure transient.
This is achieved by automatically lowering the closing setpoint of 6 valves and loweririg the opening setpoint of 2 valves following the initial
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opening.
In this way, the frequency and magnitude of the containment blowdown duty cycle is substantially reduced.
Sufficient redundancy is provided for the low-low set system such that failure of any one valve to open or close at its reduced setpoint does not violate the design basis.
3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.1 LEAKAGE DETECTION SYSTEMS l
TheRCSleakagedetectionsystemsrequiredbiy'thisspecificationare i
provided to monitor and detect leakage from the peactor coolant pressure boundary.. These systems provide the ability to measure leakage from fluid systems in the drywell.
3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of' cracks in pipes.
The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also con-sidered.
The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly.
However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shut down to allow further investigation and corrective action.
Service 1
i sensitive reactor coolant system Type 304 and 316 austenttic stainless steel piping, i.e., those that are subject to high stress or that contain relatively stagnant, intermittent, or low flow fluids, requires additional surveillance l
and leakage limits.
I The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA.
Leakage from the RCS pressure j
isolation valves is IDENTIFIED LEAKAGE and will be considered as a portion of l
the allowed limiit.
GRAND GULF-UNIT 1 B 3/4 4-2 l
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