ML20247D347
| ML20247D347 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 03/22/1989 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20247D353 | List: |
| References | |
| NUDOCS 8903310052 | |
| Download: ML20247D347 (26) | |
Text
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, UNITED STATES
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- NUCLEAR REGULATORY COMMISSION 1
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WASHINGTON, D. C. 20655
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PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION, UNIT I AM_ENDMENT TO FACILITY OPERATING LICENSE Amendment No. 17 License No. NPF-39 1.
The Nuclear Regulatory Comission. (the Comission) has found that A.
The application for amendment by Philadelphia Electric Company (thelicensee)datedDecember 14, 1988, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (1) that the activities authorired 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 Comission's regulations and all applicable requirements have been satisfied.
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-39 is hereby amended to read as follows:
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. 17, are hereby incorporated into this license. Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
?D0 l0@CK 05000352 52 890322 P
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This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION Q
Walter R. Butler, Director-Project Directorate I Division of Reactor Projects I/II Attactment:
Changes to the~ Technical Specifications Date of Issuance: March 22, 1989 k
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ATTACHMENT TO LICENSE AMENDMENT NO. 17 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 P,eplace the fo'ilowing 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 pages are provided to maintain document completeness.*
Remove Insert y
v vi vi*
xv xv*
xvi xvi xxi xxi*
xxii xxii B 2-5 B 2-5*
B 2-6 B 2-6 3/4 1-3 3/4 1-3*
3/4 1-4 3/4 1-4 3/4 1-11 3/4 1-11 3/4 1-12 3/4 1-12*
3/4 1-13 3/4 1-13 3/4 1-14 3/4 1-14*
3/4 1-15 3/4 1-15*
3/4 1-16 3/4 1-16 3/4 1-17 3/4 1-17 3/4 1-18 3/4 1-18*
i 3/4 10-1 3/4 10-1*
3/4 10-2 3/4 10-2 B 3/4 1-3 B 3/4 1-3 B 3/4 1-4 B 3/4 1-4 l
B 3/4 10-1 B 3/4 10-1 1
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F INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS' SECTION' PAGE
. 3_/4. 0 APPLICABILITY.......................
3/4 0-l' E
~3/4.1-REACTIVITY CONTROL SYSTEMS
. 3/4.1.1~
SHUTDOWN MARGIN..........................................
3/4 1-1 l'
3/4.1.2 REACTIVITY AN0MALIES....................................
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 Rod Average Scram Insertion Times.............
3/4 1-7
-Four Control Rod Group Scram Insertion Times.............
3/4 1-8 Control Rod Scram Accumulators...........................
3/4 1-9 Control Rod Dri ve Coupli ng..............................
3/4'l-11 Control Rod Position Indication...........................
3/4 1-13 Control Rod Drive Housing Support......................,.
3/4 1-15 3/4.1.4
-CONTROL ROD PROGRAM CONTROLS
-Rod Worth Minimizer......................................
3/4 1-16 l
Rod Block Monitor.........................................
3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................
3/4 1-19 Figure 3.1.5-1 Sodium Pentaborate Solution Temperature / Concentration Requirements.........................
3/4 1-21 Figure 3.1.5-2 Sodium Pentaborate Solution Volume / Concentration Requirements...
3/4 1-22 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE...............
3/4 2-1 Figure 3.2.1-1 Maximum Average Planar Linear Heat Generation' Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB278............
3/4 2-2 LIMERICK - UNIT 1 v
Amendment No. 17
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o LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION
' PAGE pnWER DisTRTRUTinN (IMITS (Continued)
Figure 3.2.1-2 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB248...........
3/4 2-3 Figure 3.2.1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB163...........
3/4 2-4 l Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB094...........
3/4 2-5 Figure 3.2.1-5 Maximum Average Planar Linear Heat l
Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial Core Fuel Types P8CIB071...........
3/42 Figure 3.2.1-6 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure For Fuel Type BC320A (GE8X8EB)..............
3/4 2-6a 3/4 2.2 APRM SETP0lNTS..........................................
3/4 2-7 3/4 2.3 MINIMUM CRITICAL POWER RATI0............................
3/4 2-8 Table 3.2.3-1 Deleted l
Figure 3.2.3-la Minimum Critical Power Ratio (MCPR)
Versus (P8X8R/BP8X8R Fuel).............
3/42-10l Figure 3.2.3-lb Minimum Critical Power Ratio (MCPR)
Versus (GE8X8E8 Fue1)..................
3/42-10al Figure 3.2.3-2 K
Factor..............................
3/4 2-11 f
3/4.2.4 LINEAR HEAT GENERATION RATE.............................
3/4 2-12 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION...............
3/4 3-1 Table 3.3.1-1 Reactor Protection System Instrumentation.....................
3/4 3-2 Tabic 3.3.1-2 Reactor Protection System 3/4 3-6 Response Times......................
Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements......................
3/4 3-7 LIMERICK UNIT - 1 vi Amendment No. 7
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'INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS-SECTION-
-PAGE ELECTRICAL POWER SYSTEMS (Continued)'
Table 4.8.2.1-1 Battery Surveillance Requirements......................
3/4 8-13 0.C.-Sources -
Shutdown.................
3/4 8 3/4.8.3 ONSITE POWER DISTRIBUTION SYSTEMS Distribution'-
Operating................................
3/4 8-15 Distribution -
Shutdown.................................
3/4'8-18 3/4.8.4.
ELECTRICAL EQUIPMENT PROTECTIVE DEVICES Primary Containment Penetration Conductor Overcurrent Protective Devices....................................
3/4 8-21 Table 3.8.4.1-1 Primary Containment Penetration Conductor Overcurrent Protective Devices...........................
3/4 8-23 Motor-Operated Valves Thermal Overload Protection.......
3/4'8-27 L
Reactor Protection System Electric Power Monitoring......
3/4 8-28 3/4.9 REFUELING OPERATIONS 3/4.9.1 REACTOR MODE SWITCH.......................................
3/4 9-1 3/4.9.2 INSTRUMENTATION.........................................
3/4 9-3 3/4.9.3 CONTROL R0D P0SITION....................................
3/4 9-5 3/4.9.4 DECAY TIME..............................................
3/4 9-6 3/4.9.5 COMMUNICATIONS..........................................
3/4 9-7 3/4.9.6 REFUELING PLATF0RM......................................
3/4 9-8 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L..................
3/4 9-10' 3/4.9.8 WATER LEVEL - REACTOR VESSEL............................
3/4 9-11 4
3/4.9.9 WATER LEVEL - SPEkT FUEL STORAGE P00L.................
3/4 9-12 LIMERICK - UNIT 1 xv
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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE-3/4.9.10 CONTROL R0D REMOVAL Single Control Rod Removal..............................
3/4 9-13' Multiple Control Rod Remova1............................
3/4 9-15 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION High Water Leve1........................................
3/4 9-17 Low Water.Leve1.........................................
3/4 9-18 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT. INTEGRITY...........................
3/4 10-1 3/4.10.2 R0D WORTH MINIMIZER.....................................
3/4 10-2:
3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS.........................
3/4 10 3 3/4.10.4 RECIRCULATION L00PS.....................................
3/4 10-4 3/4.10.5 0XYGEN CONCENTRATION....................................
3/4 10-5 3/4.10.6 TRAINING STARTUPS.......................................
3/4 10-6 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS Concentration...........................................
3/4 11-1 Table 4.11.1.1.1-1 Radioactive Liquid Waste Sampling and Analysis Program........................
3/4 11-2 Dose....................................................
3/4 11-5 Liquid Radwaste Treatment System........................
3/4 11-6 l
Liquid Holdup Tanks.....................................
3/4 11-7 3/4.11.2 GASE0US EFFLUENTS Dose Rate.....................
3/4 11-8 1
LIMERICK - UNIT 1 xvi Amendment No. 17
s INDEX BASES SECTION PAGE CONTAINMENT SYSTEMS (Continued) 3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES...............
B 3/4 6-4 3/4.6.4 VACUUM RELIEF......................................
B 3/4 6-4 3/4.6.5 SECONDARY CONTAINMENT..............................
B 3/4 6-5 t
3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL.............
B 3/4 6-6 l 3/4.7 PLANT SYSTEMS 3/4.7.1 SERVICE WATER SYSTEMS.............................
B 3/4 7-1 3/4.7.2 CONTROL ROOM EMERGENCY FRESH AIR SUPPLY SYSTEM.....
B 3/4 7-1 3/4.7.3 REACTOR CORE ISOLATION COOLING SYSTEM..............
B 3/4 7-1 3/4.7,4 SNUBBERS...........................................
B 3/4 7-2 3/4.7.5 SEALED SOURCE CONTAMINATION.........................
B 3/4 7-3 3/4.7.6 FIRE SUPPRESSION SYSTEMS........................
B 3/4 7-4 3/4.7.7 FIRE RATED ASSEMBLIES..............................
B 3/4 7-4 3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2, and 3/4.8.3 A.C. SOURCES, D.C. SOURCES, and DNSITE POWER DISTRIBUTION SYSTEMS...............................
B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES............
B 3/4 8-3 3/4.9 REFUELING CPERATIONS 3/4.9.1 REACTOR MODE SWITCH................................
E 3/4 9-1 3/4.9.2 INSTRUMENTATION....................................
B 3/4 9-1 3/4.9.3 CONTROL RCD P0SITION...............................
B 3/4 9-1 3/4.9.4 DECAY TIME.........................................
B 3/4 9-1 3/4.9.5 COMMUNICATIONS.....................................
B 3/4 9-1 LIMERICK - UNIT 1 xxi Amendment No. 8
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INDEX BASES SECTION PAGE REFUELING OPERATIONS (Continued)-
3/4.9.6 REFUELING PLATF0RM................................
B 3/4 9-2 3/4.9.7 CRANE TRAVEL - SPENT FUEL STORAGE P00L............
B 3/4 9-2 3/4.9.8 and 3/4.9.9 WATER LEVEL - REACTOR VESSEL and WATER LEVEL - SPENT FUEL STORAGE P00L.........
B 3/4 9-2 3/4.9.10 CONTROL ROD REM 0 VAL...............................
B 3/4 9-2 3/4.9.11 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION.....
B 3/4 9-2 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY.....................
B 3/4 10-1 3/4.10.2 R0D W)RTH MINIMIZER...............................
B 3/4 10-1 3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS....................
B 3/4 10-1 3/4.10.4 RECIRCULATION L00PS...............................
B 3/4 10-1 3/4.10.5 0XYGEN CONCENTRATION..............................
B 3/4 10-1 3/4.10.6 TRAINING STARTUPS.................................
B 3/4 10-1 3/4.11 RADI0 ACTIVE EFFLUENTS 3/4.11.1 LIQUID EFFLUENTS l
Concentration.....................................
B 3/4 11-1 l
Dose..............................................
B 3/4 11-1 Liquid Radwaste Treatment System..................
B 3/4 11-2 j
Liquid Holdup Tanks...............................
B 3/4 11-2 3/4.11.2 GASE0US EFFLUENTS Dose Rate.........................................
B 3/4 11-2 i
Dose - Noble Gases................................
B 3/4 11-3 l
l Dose - Iodine-131, Iodine-133, Tritium, and Radionuclides in Particulate Form.................
B 3/4 11-3 Ventilation Exhaust Treatment System..............
B 3/4 11-4 1
1 LIMERICK - UNIT 1 xxii Amendment No. 17
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SAFETY LIMITS BASES 1
'2.1.3 REACTOR COOLANT SYSTEM PRESSURE
- The Safety Limit for the reactor coolant _ system pressure has been selected such that it is at a pressure below which it can be shown that the-integrity of the system is not endangered.
The reactor pressure vessel is designed to Section III of'the A3ME Boiler and Pressure Vessel Code 1968 Edition, including Addenda through Summer 1969, which permits a maximum pres-sure transient of 110%, 1375 psig,.of design pressure 1250 psig.
The Safety, Limit of 1325 psig, as measured by the reactor vessel steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system.
The reactor andPressure.VesselCode,77tgoolantsystemisdesignedtotheASMEBoiler Edition, including Addenda through Summer 1977 for the reactor recirculation piping, which permits a maximum pressure transient
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.of 110%, 1375 psig of design pressure, 1250 psig for suction piping and 1500 psig for discharge piping.
The pressure Safety Limit is selected to. be the lowest transient overpressure allowed by the ASME Boiler and Pressure Vessel Code Section III, Class I.
~2.1.4 REACTOR VESSEL WATER LEVEL With fuel-in the reactor vessel during periods when the reactor is shutdown, consideration must be given to water level requirements due to the effect of decay heat.
If the water level should drop below the top of the active irradiated fuel during this period,- the ability;to remove decay heat is reduced.
This reduction in cooling capability could lead to elevated cladding temperatures and clad perforation in the event that the water level became less than two-thirds of the core height.
The Safety Limit has'been established at the top of the active irradiated fuel to provide a point which can be monitored and also provide adequate margin for effective action, l
I LIMERICK - UNIT 1 B 2-5 i
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- 2. 2 LIMITING SAFETY SYSTEM SETTINGS BASES
-k 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System instrumentation setpoints specified in i
Table 2.2.1-1 are the values at which the reactor trips are set for each para-meter. The Trip Setpoints have been selected to ensure that the reactor core-and reactor coolant system are prevented from exceeding their Safety Umits during normal operation and design basis anticipated operational occur"ences and to assist in mitigating the consequences of accidents.
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 bctween each Trip Setpoint and the Allowable Value is equal to or less than the drift allowance assumed for~each trip in the safety analyses.
1.
Intermediate Range Monitor, Neutron Flux - High I
i The IRM system consists of 8 chambers, 4 in each of tu reactor trip systems.
The IRM is a 5 decade 10 range instrument.
The trip setpoint of 120 divisions of scale is active in each of the 10 ranges.
Thes as the IRM is ranged up to accommodate the increase in power level, the trip setpoint is also ranged up.
The IRM instruments provide for overlap with both the APRM and SRM systems.
The most significant source of reactivity changes during the power increase is due to control rod withdrawal.
In order to ensure that the IRM provides the required protection, a rar.ge of rod withdrawal accidents nave been analyzed.
The results of these analyses are in Section 15.4 of the FSAR.
The most severe case involves an initial condition in which THERMAL POWER is at approximately 1% of RATED THERMAL POWER.
Additional conservatism was taken in this analysis by assuming the IRM channel closest to the control rod being withdrawn is bypassed.
The results of this analysis show that the reactor is shutdown and peak power is limited to 21% of RATED THERMAL POWER with the peak fuel enthalpy well below the fuel failure threshold of 170 cal /gm.
Based on this analysis, the 1RM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.
2.
Average Power Range Monitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15% of RATED THERMAL POWER provides adequate thermal margin between the setpoint and the Safety Limits.
The margin accommodates the anticipated maneuvers associated with power plant startup.
Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup is not much colder than that already in the system.
Tempera-ture coefficients are small and control rod patterns are constrained by the RWM.
Of all the possible sources of reactivity input, uniform control rod
{
withdrawal is the most probable cause of significant power increase.
LIMERICK - UNIT 1 B 2-6 Amendment No. 17
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RExCTIVITY CONTROL SYSTEMS
-3/4.1.3 CONTROL RODS CONTROL R00 OPERABILITY
' LIMITING CONDITinN FOR OPERATION 3.1.3.1 All control rods shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
a.
With one control rod inoperable due to being immovable, as a result of excessive friction or mechanical interference, or known to be untrippable:
1.
Within 1 hour:
a)
Verify that the inoperable control rod, if withdrawn, is separated from all other inoperable control rods by at least
.two control cells-in all directions.
b)
Disarm the associated directional control valves ** either:
- 1) Electrically, or
- 2) Hydraulically by closing the dfive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHUTDOWN within tne next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
Restore the inoperable control rod to OPERABLE' status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in at least HOT SHUTD0WN within the next.12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, b.
With one or more control rods trippable but inoperable for causes other than addressed in ACTION a, above:
1.
If the' inoperable control rod (s) is withdrawn, within 1 hour:
a)
Verify that the inoperable withdrawn control rod (s) is separated from all other inoperable withdrawn control rods by at least two control cells in all directions, and b)
Demonstrate the insertion capability of the inoperable with-drawn control rod (s) by inserting the control rod (s) at least one notch by drive water pressure within the normal operating range *.
Otherwise, insert the inoperable withdrawn control rod (s) and disarm the associated directional control valves ** either:
a)
Electrically, or b)
Hydraulically by closing the drive water and exhaust water isolation valves.
- The inoperable control rod may then be withdrawn to a position no further withdrawn than its position when found to be inoperable.
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- May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
I LIMERICK - UNIT 1 3/4 1-3
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REACTIVITY CONTROL SYSTEMS LIMITING CONDITION FOR OPERATION (Continued)
ACTION: (Continued) 2.
If the inoperable control rod (s) is inserted, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> disarm the associated directional control valves ** either:
a)
Electrically, or b)
Hydraulically by closing the drive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
3.
The provisions of Specification 3.0.4 are not applicable.
c.
With more than 8 control rods 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 4.1.3.1.1 The scram discharge volume drain and vent valves shall be demonstrated OPERABLE by:
At least once per 31 days verifying each valve to be open,* and a.
b.
At'least once per 92 days cycling each valve through at least one complete cycle of full travel.
4.1.3.1.2 When above the preset power level of the RWM, all withdrawn control l
rods not required to have their directional control valves disarmed electrically or hydraulically shall be demonstrated OPERABLE by moving each control rod at least one notch:
a.
At least once per 7 days, and b.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when any control rod is immovable as a result of excessive friction or mechanical interference.
4.1.3.1.3 All control rods shall be demonstrated OPERABLE by performance of Surveillance Requirements 4.1.3.2, 4.1.3.4, 4.1.3.5, 4.1.3.6, and 4.1.3.7.
- These valves may be closed intermittently for testing under administrative controls.
- May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
LIMERICK - UNIT 1 3/4 1-4 Amendment No. 17
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REACTIVITY CONTROL SYSTEMS CONTROL R0D DRIVE COUPLING LIMITING CONDITION FOR OPERATION-3.1.3.6. All control rods shall be coupled to their drive mechanisms.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 5*.
ACTION:
In OPERATIONAL CONDITIONS 1 and 2 with one control rod not coupled a.
to its associated drive mechanism, within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 5:
1.
If permitted by the RWM, insert the control rod drive mechanism l
to accomplish recoup' ling and verify recoupling by withdrawing the control rod, and:
a)
Observing any indicated response of the nuclear instruments-tion, and b)
Demonstrating-that the control rod will not go to the over-travel position.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
2.
If recoupling is not accomplished on the first attempt or, if not' permitted by the RWM, then until permitted by the RWM, declare the control rod inoperable, insert the control rod and
. disarm the associated directional control valves ** either:
a)
Electrically, or b)
Hydraulically by closing the drive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In OPERATIONAL CONDITION 5* with a withdrawn control rod not coupled to its associated drive mechanism, within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:
1.
Insert the control rod to accomplish recoupling and verify recoup-ling by withdrawing the control rod and demonstrating that the control rc,d will not go to the overtravel position, or 2.
If recoupling is not accomplished, insert the control rod and disarm the associated directional control valves ** either:
a)
Electrically, or b)
Hydraulically by closing the drive water and exhaust water l
isolation valves.
c.
The provisions of Specification 3.0.4 are not applicable.
i
- At least each withdrawn control rod.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
- May be rearmed intermittently, under administrative control, to permit l
testing associated with restoring the control rod to OPERABLE status.
LIMERICK - UNIT 1 3/4 1-11 Amendment No. 17
c REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.l.3.6 Each affected control rod shall be demonstrated to be coupled to its drive mechanism by observing any indicated response of the nuclear.instrumen-
.tation while withdrawing the control rod to the fully withdrawn position and then verifying that the control rod drive does not go to the overtravel position:
a.
Prior to reactor criticality after completing CORE ALTERATIONS that could have affected the control rod drive coupling integrity, l
b.
Anytime.the control rod is withdrawn to the " Full out" position in subsequent operation, and c.
Following maintenance on or modification to the control rod or control rod drive system which could have affected the control rod drive coupling integrity.
l l
l LIMERICK - UNIT 1 3/4 1-12
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'1 REACTIVITY CONTROL SYSTEMS CONTROL R00 POSITION INDICATION LIMITING C0'NDITION FOR OPERATION 3.1.3.7 The control' rod position indication system shall be OPERABLE.
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 5*.
ACTION:
In OPERATIONAL CONDITION 1 or 2 with one or more control rod position a.
indicators inoperable, within 1 hour:
1.
Determine the position of the control rod by using an alternate method, or:
a)
Moving the control rod, by single notch movement,. +o a position with an OPERABLE position: indicator, b)'
Returning the control rod, by single notch movement, to its original position, and c)
Verifying no control rod drift alarm at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, or 2.
Move the control rod to a position with an OPERABLE position indicator, or 3.
When THERMAL POWER is:
a)
Within the preset power level of the RWM, declare the-1-
Greater than the preset power level of the RWM, declare l
the control rod inoperable, insert the control rod and disarm the associated directional control valves ** either:
1)
Electrically, or 2)
Hydraulically by closing the drive water and exhaust water isolation valves.
Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
b.
In OPERATIONAL CONDITION 5* with a withdrawn control rod position indicator inoperable, move the control rod to a position with an OPERABLE position indicator or insert the control rod, c.
The provisions of Specification 3.0.4 are not applicable.
- At least each withdrawn control rod.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
- May be rearmed intermittently, under administrative control, to permit testing associated with restoring the control rod to OPERABLE status.
LIMERICK - UNIT 1 3/4 1-13 Amendment No. 17
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l REACTIVITY CONTROLLSYSTEMS SURVEILLANCE REQUIREMENTS 4.1.3.~7
~ The control rod position 'iridication system shall be determined'
-OPERABLE by verifying:
At least once per.-24 hours:that the-' position of each control rodEis.
a.
indicated, b.
That'the> indicated control rod position changes during the movement of the control rod. drive when performing Surveillance Requirement 4.1.3.1.2,.and
- That the control rod position indicator corresponds to the control-c.
rod position indicated by the;" Full out". position' indicator when -
. performing Surveillance Requirement 4.1.3.6b.
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' LIMERICK - UNIT 1 3/4 1-14
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REACTIVITY CONTROL SYSTEMS
' CONTROL R00 ORIVE HOUSING SUPPORT'
-l LIMITING CONDITION FOR OPERATION 3.1.3.8.The control rod drive housing _ support shall be in place.
APPLICABILITY: OPERATIONAL CONDITIONS 1.-2 and 3.
ACTION:
With the control rod drive housing support n'ot in place, be in at least HOT j
SHUT 00WN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUT 00WN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4 SURVEILLANCE REQUIREMENTS 4.1.3,8 The control rod drive housing support shall be verified to be in place by a visual inspection prior to startup any time it has been disassembled or-when' maintenance has'been performed in the control-rod drive: housing _ support
' area.
LIMERICK - UNIT 1 3/4 1-15 1
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l REACTIVITY CONTROL SYSTEMS 3/4.1.4 CONTROL R0D PROGRAM CONTROLS 1
J R00 WORTH MINIMIZER LIMITING CONDITION FOR OPERATION l
3.1.4.1 The rod worth minimizer (RWM) shall be OPERABLE.
1 l
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2*, **, when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER.
ACTION:
With the RWM inoperable after the first 12 control rods'are fully a.
withdrawn, operation may continue provided that control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or technically qualified member of the unit technical staff.
b.
With the RWM inoperable before the first 12 control rods are fully withdrawn, one startup per calendar year may be performed provided that control rod movement and compliance with the prescribed control rod pattern are verified by a second licensed operator or technically qualified member of the unit technical staff.
c.
Otherwise, with the RWM inoperable, control <
movement shall not be permitted except by full scram.***
- See Special Test Exception 3.10.2.
- Entry into OPERATIONAL CONDITION 2 and withdrawal of selected control rods is permitted for the purpose of determining the OPERABILITY of the RWM prior to i
withdrawal of control rods for the purpose of bringing the reactor to criticality.
- Control rods may be moved, under administrative control, to permit testing associated with demonstrating OPERABILITY of the RWM.
LIMERICK - UNIT 1 3/4 1-16 Amendment No. II, 17
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' REACTIVITY CONTROL SYSTEMS SURVEILLANCE REQUIREMENTS 4.1.4.1-The RWM shall be demonstrated OPERABLE:
i In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of a.
control rods for the' purpose of making the reactor critical, and in OPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after RWM automatic initia-tion when reducing THERMAL POWER, by verifying proper indication of the selection error of at least one out-of-sequence control rod.
b.
In OPERATIONAL CONDITION 2 within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> prior to withdrawal of control rods'for the purpose of making the reactor critical, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod.
c.
In OPERATIONAL CONDITION 1 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after RWM automatic initiation when reducing THERMAL POWER, by verifying the rod block function by demonstrating inability to withdraw an out-of-sequence control rod, d.
By verifying that the control' rod patterns and sequence input to the RWM computer are correctly loaded following any loading of the program into the computer.
3.1.4.2 DELETED 4.1.4.2 DELETED l
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L LIMERICK - UNIT 1 3/4 1-17 Amendment No. 17
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REACTIVITY CONTROL SYSTEMS ROD BLOCK MONITOR LIMITING CONDITION FOR OPERATION 3.1.4.3 Both rod block monitor (RBM) channels shall be OPERABLE.
APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater. than or equal to 30% of RATED THERMAL POWER.
ACTION:
a.
With one RBM channel inoperable:
1.
Verify that the reactor is not operating on a LIMITING CONTROL-R0D PATTERN, and 2.
Restore the inoperable RBM channel to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Otherwise, place the inoperable rod block monitor channel in the tripped condition within the next hour.
b.
With both RBM channels inoperable, place at least one inoperable rod block monitor channel in the tripped condition within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
SURVEILLANCE REQUIREMENTS 4.1.4.3 Each of the above required RBM channels shall be demonstrated OPERABLE by performance of a:
a.
CHANNEL FUNCTIONAL TEST and CHANNEL CALIBRATION at the frequencies and for the OPERATIONAL CONDITIONS specified in Table 4.3.6-1.
b.
CHANNEL FUNCTIONAL TEST prior to control rod withdrawal when the reactor is operating on a LIMITING CONTROL ROD PATTERN.
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LIMERICK - UNIT 1 3/4 1-18 i
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~3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY LIMITING CONDITION FOR OPERATION l
l 3.10.1 The provisions of Specifications 3.6.1.1, 3.6.1.3, and 3.9.1 and Table 1.2 may be suspended to permit the reactor pressure vessel closure head and the drywell head to be removed and the primary containment air lock doors to be open when the reactor mode switch is in the Startup position during low power PHYSICS TESTS with THERMAL POWER less than 1% of RATE 0 THERMAL POWER and reactor coolant temperature less than 200 F.
APPLICABILITY: OPERATIONAL CONDITION 2, during low power PHYSICS TESTS.
ACTION:
With THERMAL POWER greater than or equal to 1% of RATED THERMAL POWER or with the reactor coolant temperature greater than or equal to 200 F, immediately place the reactor mode switch in the Shutdown position.
SURVEILLANCE REQUIREMENTS 4.10.1 The THERMAL POWER and reactor coolant temperature shall be verified to be within the limits at least once per hour during low power PHYSICS TESTS.
LIMERICK - UNIT 1 3/4 10-1
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SPECIAL TEST EXCEPTIONS 3/4.10.2 R0D WORTH MINIMIZER l
LIMITING CONDITION FOR OPERATION 3.10.2 The sequence constraints imposed on control rod groups by the rod worth minimizer (RWM) per Specification 3.1.4.1 may be suspended for the following tests provided that control rod movement prescribed for this testing is verified by a second licensed operator or other technically qualified member of the unit technical staff present at the reactor console:
a.
Shutdown margin demonstration, Specification 4.1.1.
b.
Control rod scram, Specification 4.1.3.2.
c.
Control rod friction measurements, d.
Startup Test Program l
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2 when THERMAL POWER is less than or equal to 10% of RATED THERMAL POWER.
ACTION:
With the requirements of the above specifications not satisfied, verify that the RWM is OPERABLE per Specification 3.1.4.1.
SURVEILLANCE REQUIREMENTS 4.10.2 When the sequence constraints imposed by the RWM are bypassed, verify; l
j a.
That movement of control rods is blocked or limited to the approved control rod withdrawal sequence during scram and friction tests.
b.
That movement of control rods during shutdown margin demonstrations is limited to the prescribed sequence per Specification 3.10.3.
c.
Conformance with this specification and test procedures by a second licensed operator or other technically qualified member of the unit technical staff.
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LIMERICK - UNIT 1 3/4 10-2 Amendment No.17
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' REACTIVITY CONTROL SYSTEMS,
BASES CONTROL RODS'(Continued) t Control rod coupling integrity is required to ensure compliance with the analysis of the rod drop accident in the FSAR.
The overtravel position feature
)
provides the only positive means of determining that a rod is properly coupled i
and therefore this check must be performed prior to achieving criticality after j
completing CORE ALTERATIONS that could have affected the control rod coupling 1
integrity.
The subsequent check is performed as a backup to the initial demon-stration.
In order to ensure that the control rod patterns can be followed and there-fore that other parameters are within their limits, the control rod position indication system must be OPERABLE.
The control rod housing support restricts the outward movement of a control rod to less than 3 inches in the event of a housing failure.
The amour.t of rod reactivity which could be added by this small amount of rod withdrawal is less than a normal withdrawal increment and will not contribute to any damage to the primary coolant system.
The support is not required when there is no pressure to act as a driving force to rapidly eject a drive housing.
The required surveillance intervals are adequate to determine that the rods are OPERABLE and not so frequent as to cause excessive wear on the system components.
3/4.1.4 CONTROL R0D PROGRAM CONTROLS Control rod withdrawal and insertion sequences are established to assure that the maximum insequence individual control rod or control rod segments which are withdrawn at any time during the fuel cycle could not be worth enough to result in a peak fuel enthalpy greater than 280 cal /gm in the event of a control rod drop accident The specified sequences are characterized by homogeneous, scattered patterns of control rod withdrawal.
When THERMAL POWER is greater l
than 10% of RATED THERMAL POWER, there is no possible rod worth which, if-I dropped at the design rate of the velocity limiter, could result in a peak enthalpy of 280 cal /gm.
Thus requiring the RWM to be OPERABLE when THERMAL l
POWER is less than or equal to 10% of RATED THERMAL POWER provides adequate l
control.
The RWM provides automatic supervision to assure that out-of-l sequence rods will not be withdrawn or inserted.
The analysis of the rod drop accident is presented in Section 15.4.9 of the FSAR and the techniques of the analysis are presented in a topical report, Reference 1, and two supplements, References 2 and 3.
Additional pertinent r
analysis is also contained in Amendment 17 to the Reference 4 topical report.
The RBM is designed to automatically prevent fuel damage in the event of l
erroneous rod withdrawal from locations of high power density during high power operation.
Two channels are provided.
Tripping one of the channels will block erroneous rod withdrawal to prevent fuel damage.
This system backs up the written sequence used by the operator for withdrawal of control rods.
I LIMERICK - UNIT 1 B 3/4 1-3 Amendment No. 17 t
R REACTIVITY CONTROL SYSTEMS BASES 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM The standby liquid control system provides a backup capability for bringing the reactor from full power to a cold, Xenon-free shutdown, assuming that the withdrawn control rods remain fixed in the rated power pattern.
To meet this objective it is necessary to inject a quantity of boron which produces a concen-tration of 660 ppm in the reactor core and other piping systems connected to the reactor vessel.
To allow for pc,tential leakage and improper mixing, this concentration is increased by 25%.
The required concentration is achieved by having available a minimum quantity of 4,620 gallons of sodium pentaborate solution containing a minimum of 5,500 lbs of sodium pentaborate.
This quan-tity of solution is a net amount which is above the pump suction shutoff level setpoint thus allowing for the portion which cannot be injected.
The pumping rate of 41.2 gpm provides a negative reactivity insertion rate over the permis-sible solution volume range, which adequately compensates for the positive reactivity effects due to elimination of steam voids, increased water density from hot to cold, reduced doppler effect in uranium, reduced neutron leakage from boiling to cold, decreased control rod worth as the moderator cools, and xenon decay.
The temperature requirement ensures that the sodium pentaborate always remains in solution.
With redundant pumps and explosive injection valves and with a highly reliable control rod scram. system, operation of the reactor is permitted to continue for short periods of time with the system inoperable or for longer periods of time with one of the redundant components inoperable.
The SLCS system consists of three separate and independent 100% capacity pumps and explosive valves.
Only two of the separate and independent pumps and explosive valves are required to meet the minimum requirements of this technical specification and satisfy the single failure criterion.
Surveillance requirements are established on a frequency that assures a high reliability of the system.
Once the solution is established, boron con-centration will not vary unless more boron or water is added, thus a check on the temperature and volume once each 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> assures that the solution is available for use.
Replacement of the explosive charges in the valves at regular intervals will assure that these valves will not fail because of deterioration of the charges.
1.
C. J. Paone, R. C. Stirn and J. A. Woolley, " Rod Drop Accident Analysis for Large BWR's," G. E. Topical Report NED0-10527, March 1972.
2.
C. J. Paone, R. C. Stirn, and R. M. Young, Supplement 1 to NED0-10527, July 1972.
3.
J. M. Haun, C. J. Paone, and R. C. Stirn, Addendum 2, " Exposed Cores,"
l Supplement 2 to NED0-10527, January 1973.
4.
Amendment 17 to General Electric Licensing Topical Report NEDE-24011-P-A,
" General Electric Standard Application for Reactor Fuel".
LIMERICK - UNIT 1 B 3/4 1-4 Amendment No.
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3/4.10 SPECIAL TEST EXCEPTI.0NS BASES 3/4.10.1 PRIMARY CONTAINMENT INTEGRITY The requirement for PRIMARY CONTAINMENT INTEGRITY is not applicable during the period when open vessel tests are being performed during the low power PHYSICS TESTS.
3/4.10.2 R0D WORTH MINIMIZER l
In order to perform the tests required in the technical specifications it is necessary to bypass the sequence restraints on control rod movement.
The additional surveillance requirements ensure that the specifications on heat generation rates and shutdown margin requirements are not exceeded during the period when these tests are being performed and that individual rod worths do not exceed the values assumed in the safety analysis.
3/4.10.3 SHUTDOWN MARGIN DEMONSTRATIONS Performance of shutdown margin demonstrations with the vessel head removed requires additional restrictions in order to ensure that criticality does not occur.
These additional restrictions are specified in this LCO.
3/4.10.4 RECIRCULATION LOOPS This special test exception permits reactor criticality under no flow conditions and is required to perform certain startup and PHYSICS TESTS while at low THERMAL POWER levels.
3/4.10.5 OXYGEN CONCENTRATION Relief from the oxygen concentration specifications is necessary in order to provide access to the primary containment during the initial startup and testing phase of operation.
Without this access the startup and test program could be restricted and delayed.
3/4.10.6 TRAINING STARTUPS This special test exception permits training startups to be performed with the reactor vessel depressurized at low THERMAL POWER and temperature while controlling RCS temperature with one RHR subsystem aligned in the shutdown cooling mode in order to minimize contaminated water discharge to the radioactive waste disposal system.
LIMERICK - UNIT 1 8 3/4 10-1 Amendment No. 17 L __-