ML20082J107

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Amends 103 & 88 to Licenses NPF-11 & NPF-18,respectively, Implement Partial Application of GE ARTS (Aprm/Rbm/Ts) Improvement Program
ML20082J107
Person / Time
Site: LaSalle  Constellation icon.png
Issue date: 04/13/1995
From: William Reckley
NRC (Affiliation Not Assigned)
To:
Shared Package
ML20082J110 List:
References
NUDOCS 9504180108
Download: ML20082J107 (33)
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{{#Wiki_filter:. if>*0% g -t UNITED STATES j j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20556 4001 '+9.....,o COMMONWEALTH EDIS0N COMPANY DOCKET NO. 50-373 LASALLE COUNTY STATION. UNIT _1 AMENDMENT TO FACILITY OPERATING LICEfGE Amendment No. 103 License No. NPF-11 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment filed by the Comonwealth Edison Company (the licensee) dated June 9, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's 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 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 set forth in 10 CFR Chapter I; D. The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission'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 gnclosure to this license amendment and paragraphs 2.C.(2) and 2.C.(34) of Facility Operating License No. NPF-11 are hereby amended ta read as follows:

  • Page 16a is attached, for convenience, for the composite license to reflect this change.

9504180108 950413 PDR ADOCK 05000373 .P PDR

= ~ i (2) Technical Soecifications and Environmental Protection Plan j The Technical-Specifications contained in Appendix A, as revised through Amendment No.103, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. l The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. i (34) Deleted. \\ j 3. This amendment is effective upon date of issuance and shall be i J implemented within 60 days of issuance. ? FOR THE NUCLEAR REGULATORY COMMISSION. ] I William D. Reckley, Proj anager l Project Directorate III-2 i Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation I Attachments:

1. License page 16a
2. Changes to the Technical Specification:

Date of Issuance: April 13. 1995 1 i l i 4 I 4 j: j 1 1 i i i .( l

o l - 16a - l 2.C.(34) Deleted. I i l i i 1 i 1 l i l l i l l i i 1 l i i l Amendment No.103 i (

ATTACHMENT TO LICENSE AMENDMENT NO. 103 FACILITY OPERATING LICENSE NO. NPF-ll DOCKET NO. 50-373 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the area of change. REMOVE INSERT IV IV XII XII B 2-10 B 2-10 3/4 2-2 3/4 2-2 3/4 3-8 3/4 3-8 3/4 3-53a 3/4 3-53a 3/4 4-1 3/4 4-1 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-5 B 3/4 2-5 B 3/4 2-6 B 3/4 2-6 l B 3/4 3-4 B 3/4 3-4 i 6-25 6-25 l

J INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS SECTION PAGE 3/4.0 APPLICABILITY................................................. 3/4 0-1 3/4.1 REACTIVITY CONTRDL SYSTEMS 1 3/4.1.1 SHUTDOWN MARGIN.......................................... 3/4 1-1 i 3/4.1.2 REACTIVITY AN0MALIES..................................... 3/4 1-2 3/4.1.3 CONTROL RODS C o n t rol Rod Op e rab i l i ty.................................. 3/4 1-3 Control Rod Maximum Scram Insertion Times................ 3/4 1-6 ' Control Rod Average Scram Insertion Times................ 3/4 1-7 l Four Control Rod Group Scram Insertion Times............. 3/4 1-8 l Control Rod Scram Accumul ators........................... 3/4 1-9 ) Control Rod Drive Coupl i ng............................... 3/4 1-11 Control Rod Position Indication.......................... 3/4 1-13 Control Rod Drive Housing Support......................... 3/4 1-15 3/4.1.4 CONTROL R00 PROGRAM CONTROLS Rod Worth Minimizer...................................... 3/4 1-16 Rod Block Monitor........................................ 3/4 1-18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM............................ 3/4 1-19 3/4.1.6 ECONOMIC OENERATION CONTROL SYSTEM....................... 3/4 1-23 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLARAR LINEAR HEAT GENERATION RATE............... 3/4 2-1 3/4.2.2 DELETED.................................................. 3/4 2-2 l i 3/4.2.3 MINIMUM CRITICAL POWER RAT 10............................. 3/4 2-3 4 3/4.2.4 LINEAR HEAT GENERATION RATE.............................. 3/4 2-5 i 4 LA SALLE - UNIT 1 IV Amendment No.103

la INDEX t BASES-SECTION. Pffd 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 i 3/4.1.3 CONTROL R0DS...................................... B 3/4-1-2 l -. 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS...................... B 3/4 1-3 3/4.1.5. STANDBY LIQUID CONTROL SYSTEM..................... B 3/4 1-4 i 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM................ B 3/4'l-5 3/4.2 POWER DISTRIBUTION LIMITS l 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 3/4.2.3 MINIMUM CRITICAL POWER RATI0...................... B 3/4 2-2 3/4.2.4 LINEAR HEAT GENERATION RATE....................... B 3/4'2-6 l 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION......... B 3/4 3-1 I 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION............... B 3/4 3-2 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION................................... B 3/4 3-2 l 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION. B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION l INSTRUMENTATION................................... B 3/4 3-4 l 1 l 3/4.3.6 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION....... B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.............. B 3/4 3-4 Selsmic Monitoring Instrumentation................ B 3/4 3-4 LA SALLE - UNIT 1 XII Amendment No. 103 ,y-3

LIMI. TING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) Averaae Power Ranoe Monitor (Continued) the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant 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 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 flux trip remains active until the mode switch is placed in the Run position. The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Fixed 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-Upscale setpoint, a time constant of 6 i 1 seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1. The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown. l 3. Reactor Vessel Steam Dome Pressure-Hiah High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum 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 conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit. LA SALLE - UNIT 1 B 2-10 Amendment No. 103 m

. - =... -....., -. =. =. 1 = i i k I !i t t i i 3/4.2.2 INTENTIONALLY-LEFT BLANK. I i i I a i t 4 I t I t i .t, I i 4 l j. I 'l I LA SALLE - UNIT 1 3/4 2-2 Amendment No. 103

TABLE 4.3.1.1-1 (Continued) REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REOUIREMENTS r.R ?: CHANNEL OPERATIONAL p; CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION SURVEILLANCE RE0VIRED 8. Scram Discharge Volume Water -d Level - High NA M R 1, 2, 5 9. Turbine Stop Valve - Closure NA Q R 1 10. Turbine Control Valve Fast Closure Valve Trip System Oil Pressure - Low NA Q R 1 11. Reactor Mode Switch Shutdown Position NA R NA 1, 2, 3, 4. 5

12. Manual Scram NA W

NA 1, 2, 3, 4, 5 13. Control Rod Drive a. Charging Water Header Pressure - Low NA M R 2, 5 ae b. Delay Timer NA M R 2, 5 + us lx> (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. (b) The IRM, and SRM channels shall be determined to overlap for at least 1/2 decades during each startup and the IRM and APRM channels shall be determined to overlap for at least 1/2 decades during each controlled shutdown, if not performed within the previous 7 days. (c) Within 24 hours prior to startup, if not performed within the previous 7 days. (d) This calibration shall consist of the adjustment of the APRM channel to conform to the power levels calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER 2 25% of RATED THERMAL POWER. The APRM Gain Adjustment Factor (GAF) for any channel shall be equal to the power value deter-37 mined by the heat balance divided by the APRM reading for that channel. E Et Within 2 hours, adjust any APRM channel with a GAF > 1.02. In addition, adjust any APRM channel within ^ 12 hours, if power is greater than or equal to 90% of RATED THERMAL POWER and the APRM channel GAF is < 0.98. Until any required APRM adjustment has been accomplished, notification shall-be posted on the EF reactor control panel. (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a Es calibrated flow signal. (f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH). (g) Measure and compare core flow to rated core flow. (h) This calibration shall consist of verifying the 6 I second simulated thermal power time constant.

TABLE 3.3.6-2 (Centinu d) CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS i !g TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE g m 5. SCRAM DISCHARGE VOLUME E q a. Water Level-High s 765' Sk" s 765' 54" b. Scram Discharge Volume Switch in Bypass N.A. N.A. 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW a. Upscale-5 108/125 of full scale s 111/125 of full scale b. Inoperative N.A. N.A. c. Comparator s 10% flow deviation s 11% flow deviation 4 . A Y' u, i; i E a o. 2 5 i w

  • The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow (W).

_________._______.__.__.________.___._j

e 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM l RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION -l t 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation. APPLICABILITY: OPERATIONAL CONDITIONS I and 2 i ACTION l a. With only one (1) reactor coolant system recirculation loop in l-operation, comply with Specification 3.4.1.5 and: i l 1. ~Within four (4) hours: l a) Place the recirculation flow control-system in the Master Manual mode or lower, and l b) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Specification 2.1'.2, and. l c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting l Condition for Operation by 0.01 per Specification 3.2.3,.

and, d)

Reduce the Average Power Range Monitor (APRM) Scram'and-Rod Block and Rod' Block Monitor. Trip Setpoints and - i Allowable Values to those applicable to single recirculation loop ' operation per Specifications 2.2.1 and l [ 3.3.6. 2. Otherwise, be in at least HOT SHUTDOWN within the next twelve (12) hours. p b. With no reactor coolant recirculation loops in operation: l l 1. Take the ACTION required by Specification 3.4.1.5, and-l 2. Be in at least HOT SHUTDOWN within the next six (6) hours. LA SALLE.- UNIT 1 3/4 4-1 Amendment No. 103 l

~ - = (- - 3/4.2 POWER' DISTRIBUTION LIMITS 1 BASES The specifications of this ~section assure that the peak cladding temperature following.the postulated design basis loss-of-coolant accident will-not exceed the 2200*F limit specified in 10 CFR 50.46. 'l '3/4.2.1 AVERAGE PLANAR LINEAR HEAT GEN'ERATION RATE This specification assures that the' peak cladding temperature'following the postulated design basis loss-of-coolant accident will not exceed the'11mit specified in 10 CFR 50.46. The specification also assures that fuel rod mechanical integrity is maintained during normal and transient operations., The peak cladding temperature (PCT). following a postulated loss-of-coola'nt 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 secondarily. on the rod-to-rod power distribution within an assembly. The peak clad temperature 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 times 1.02 is used in the.heatup code along with the~ exposure dependent steady ' state gap conductance-and rod-to-rod local. peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE. (AOLHGR) _is-this ^LHGR of the highest powered rod divided by -its local peaking-factor.. However, the current General Electric-(GE) calculational models (SAFER /GESTR described in Reference 3), which are consistent with the requirements of Appendix K to 10 CFR 50,'have established that APLHGR values - l are not expected to be limited by LOCA/ECCS considerations. APLHGR limits are-still required, however, to assure that fuel rod mechanical integrity is - maintained. - They are specified for all resident fuel types'in the-Core Operating Limit Report based on the fuel thermal-mechanical design' analysis. _ -The purpose of the power-and flow-dependent MAPLHGR factors specified in the CORE OPERATING LIMITS REPORT is to define operating-limits at other than rated core flow and core power conditions. At less than 100% of rated flow or rated power, the required MAPLHGR is the minimum of either (a) the product of 1 the rated MAPLHGR limit and the power-dependent MAPLHGR factor or (b) the product of-the rated MAPLHGR limit and the flow-dependent MAPLHGR factor. The .j power-and flow-dependent MAPLHGR factors assure that the fuel remains within the fuel design basis during transients at off-rated' conditions. Methodology _ for establishing these factors is described in Reference 6. i LA SALLE - UNIT 1 B 3/4 2-1 Amendment No. 103 e l

t l ~

  • l POWER DISTRIBUTION SYSTEMS t

BASES 111 2.2 DELETED 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as l specified in Specification 3.2.3 are derived from the established fuel cladding . integrity Safety Limit-MCPR, and an analysis of abnormal operational transients. For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2. To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduc-tion in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in the CORE l OPERATING LIMITS REPORT. I Analyses have been performed to determine the effects on CRITICAL POWER l RATIO (CPR) during a transient assuming that certain equipment is out of service. A detailed description of the analyses is provided in Reference 5. The analyses performed assumed a single failure only and established the licensing bases to allow continuous plant operation with the analyzed equipment out of service. The following single equipment failures are included as part of the transient analyses input assumptions:

1) main turbine bypass system out of service,
2) recirculation pump trip system out of service, l

1 LA SALLE - UNIT 1 B 3/4 2-2 Amendment No. 103 i I

~- I l EQyER DISTRIBUTION SYSIffiS SASES-MINIMUM CRITICAL POWER RATIO Jontinued) The value for r used in Specification 3.2.3 is 0.687. seconds which is conservative for the,following reason: For simplicity in formulating and implementing the LCO, a conservative' f n 1-1, of 598 was used. This represents one full core. data set'- value for IN at BOC plus one full core data set following.a 120 day outage plus twelve 10% of core,19 rods, data sets.. The.12 data' sets 'are equivalent to 24 operating months of surveillance at the increased surveillance frequency of one set per 60 days required by the action statements of l Specifications 3.1.3.2 and 3.1.3.4.. ' That is, a cycle length was. assumed which is longer than any. past or j contemplated refuelingsinterval and,the number of rods tested was maximized in i order to simplify and conservatively reduce-the criteria for the scram time at which MCPR penalization is necessary. ] The purp of the power-and flow-dependent. MCPR limits specified in the CORE OPERATIk.IMITS REPORT is to define operating limits at other than rated core flow and core power conditions. At a given power and flow operating = l condition, the required MCPR is the maximum.of either the power-dependent MCPR limit or the flow-dependent MCPR limit. The required MCPR assures that the Safety Limit MCPR will not be violated.. Tethodol.ogy for establishing the power-and flow-dependent MCPR limits is described in Reference 6. i 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 moderator void content will be very scall..For all-designated control rod patterns which may be employed at this point, operating plant experience indicates that the.resulting MCPR valta.is in excess of requirements by a considerable margin. During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level tvith minimum recirculation pump speed. The MCPR margin will thus'be demonstrated such that future MCPR evaluation below 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 for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in i THERMAL POWER or power shape, regardless of magnitude, that'could place ] . operation at a thermal limit. i L

LA SALLE - UNIT I B 3/4 2-5 Amendment No.- 103 i

- ~

POWERDISTRIBUTIONSYSTEMS BASES 3/4.2.4 LINEAR HEAT GENERATION RATE The specification assures that the LINEAR HEAT GENERATION RATE (LHGR) in any rod is less than. the design linear heat generation even if fuel pellet densification is postulated. The power spike penalty specified is based on the analysis presented in Section 3.2.1 of the GE topical report NEDM-10735 Supplement 6, and assumes a linearly increasing variation in axial gaps between i

core bottom and top and assures with a 95% confidence that no.more than one fuel rod exceeds the design LINEAR HEAT GENERATION RATE due to power spiking.

References:

1.- General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDO-20566A, i September 1986. 2. " Qualification of the One-Dimensional Core Transient Model for i Boiling ~ Water Reactors," General Electric Company Licensing Topical Report NED0 24154 Vols. I and'Il and NEDE-24154-Vol. III as sup-piemented by letter dated September 5, 1980, from R. H. Buchholz (GE) to P. S. Check (NRC). 3. "LaSalle County Station Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," General Electric Company Report NEDC-32258P,-October 1993. 4. " General Electric Standard Application for Reactor Fuel," NEDE-240ll-P-A (latest approved revision). 5. " Extended Operating Domain and Equipment Out-of-Service for LaSalle County Nuclear Station Units 1 and 2," NEDC-31455, November 1987. 6. " ARTS Improvement Program Analysis for LaSalle County Units 1 and 2," General Electric Company Report NEDC-31531P, December 1993. l LA SALLE - UNIT l-B 3/4 2-6 _ _ Amendment No. 103_ __,

1 i-(' i i INSTRUMENTATION i i l BASES 3 f i~ 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION -{ q The reactor core isolation cooling system actuation instrumentation is t provided to initiate-actions to assure adequate core cooling in the event of 3 reactor isolation from its primary heat sink and the loss of feedwater flow to j the reactor vessel without providing actuation of ar.y of the emergency core. cooling equipment. 3/4.3.6 ' CONTROL ROD WITHDRAWAL BLOCK INSTRUMG4TATION. 7 l~ The control rod block function, are provided consistent with the require-ments of the specification in Section 3/4.1.4, Control Rod Program Controls. l The trip logic is arranged so that a trip in any one of the inputs will result e in a control rod block. 1 3/4.3.7 MONITORING INSTRUMENTATION '3/4.3.7.1 RADIATION MONITORING INSTRUMENTATION i j-The OPERABILITY of the radiation monitoring' instrumentation ensures that: l (1) the radiation levels are continually measured in the areas served by the. individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded. l 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 i i seisn.ic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to j that used in the design basis for the unit. This instrumentation is-j i consistent with the recommendations of Regulatory Guide 1.12 " Instrumentation i l for Earthquakes," April 1974. 1 3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION i i i The OPERABILITY of the meteorological monitoring instrumentation ensures that _ sufficient meteorological data 1s available for estimating potential-1 radiation doses to the public as a result of routine or accidental release of j radioactive materials to the atmosphere. This capability is required to evaluate the need for initiating protective measures to protect the health and j safety of the public. This instrumentation is consistent with the recommenda-i tions of Regulatory Guide 1.23, "Onsite Meteorological Programs," February j 1972. 1 3/4.3.7.4 REMOTE SHUTDOWN MONITORING INSTRUMENTATION i The OPERABILITY of the remote shutdown monitoring instrumentation ensures j that sufficient capability is available to permit shutdown and maintenance of 4 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. i LA SALLE - UNIT 1 B 3/4 3-4 AMENDMENT N0. 103 a ___..____.,_m ,_.,._.r -~,

ADMINISTRATIVE CONTROLS Semiannual Radioactive Effluent Release Report (Continued) Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective. In addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the evaluation was reviewed and accepted by Onsite Review and Investigative Function. 6. Core Operatina limits Reoort a. Core operating limits shall be established and documented in the CORE OPERATING LIMITS REPORT before each reload cycle or any remaining part of a reload cycle for the following: (1) The Average Planar Linear Heat Generation Rate (APLHGR) for Technical Specification 3.2.1. l (2) The minimum Critical Power Ratio (MCPR)its, and power an dependent MCPR lim (including 20% scram time, tau (7), limits) for Technical Specification flow dependent MCPR 3.2.3. (3) The Linear Heat Generation Rate (LHGR) for Technical Specification 3.2.4. (4) The Rod Block Monitor Upscale Instrumentation Setpoints for Technical Specification Table 3.3.6-2. b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or su)plement of the topical reports describing the methodology. For aSalle County Station Unit 1, the topical reports are: i (1) NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel," (latest approved revision). (2) Commonwealth Edison Topical Report NFSR-0085, " Benchmark of BWR Nuclear Design Methods," (latest approved revision). (3) Commonwealth Edison Topical Report NFSR-0085, Supplement 1, " Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma Scan Comparisons," (latest approved revision). (4) Commonwealth Edison Topical Report NFSR-0085, Supplement 2, " Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision). j LA SALLE - UNIT 1 6-25 Amendment No.103 1

meog4 ., m EQNiS UNITED STATES i $[q, /- NUCLEAR REGULATORY COMMISSION \\.. WASHINGTON, D.C. 2055MX)01 COMMONWEALTH EDISON COMPANY DOCKET NO. 50-374 LASALLE COUNTY STATION. UNIT 2 l AMENDMENT TO~ FACILITY OPERATING LICENSE Amendment No. 88 License No. NPF-18 L 1. The Nuclear Regulatory Commission (the Commission)_has found that: A. The application for amendment filed by the Commonwealth Edison Company (the licensee) dated June 9, 1994, complies with the standards and requirements of the Atomic Energy Act of 1954,'as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; j B. The facility will operate in conformity with the application, the-l provisions of the Act, and the regulations of the Commission; l l 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 l conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; l D. The issuance of this amendment will not be inimical to the conmon defense and security or to the health and safety of the public; and i E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. ? 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the enclosure to this license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-18 is hereby amended to read as follows: I

i (2) Technical Soecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 88, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. This amendment is effective upon date of issuance and shall be implemented within 60 days of issuance. FOR THE NUCLEAR REGULATORY COMMISSION William D. Reckley, Project nager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance: April 13. 1995 i

l ATTACMENT TO LICENSE AMENDMENT NO. 88 FACILITY OPERATING LICENSE NO. NPF-18 DOCKET NO. 50-374 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified by amendment number and contain a vertical line indicating the area of change. j REMOVE INSERT IV ~IV XII XII B 2-10 B 2-10 3/4 2-2 3/4 2-2 3/4 3-8 3/4 3-B 3/4 3-54 3/4 3 ! 3/4 4-1 3/4 4-1 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-5 8 3/4 2-5 B 3/4 2-6 B 3/4 2-6 B 3/4 3-4 8 3/4 3-4 6-25 6-25 l i

l l 1 l INDEX I I LIMITING CONDITIONS FOR OPERATION AND' SURVEILLANCE RE0VIREMENTS E 'SECTION EAji[ l 3/4.0 APPLICABILITY................................................... 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS t 3/4.1.1 SHUTDOWN MARGIN............................................... 3/4:1-l 3/4.1.2 REACTIVITY ANOMALIES........................................... 3/4 1-2 3/4.1.3. CONTROL RODS Control Rod Operabil i ty....................................... 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 Accumul ators................................. < 3/4 1-9 ' Control Rod Dri ve Coupl i ng..................................... ~ 3/4 1-11 Control Rod Position Indication............................... 3/4 1-13~ Control Rod Drive Housing Support............................. 3/4 1-15 ) t l 3/4.1.4 CONTROL R00 PROGRAM CONTROLS o j Rod Worth Mi n i mi z e r........................................... 3/4 1 -16 Rod B l oc k Mo n i t o r............................................. 3 / 4 1 - 18 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM................................. 3/4 1-19 ) o 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM............................ 3/4 1-23 ) 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.................... 3/4 2-1 l 3/4.2.2 DELETED....................................................... 3/4 2-2 l 3/4. 2.3 MINIMUM CRITICAL POWER RATI0................................... 3/4 2-3 3/4.2.4 LINEAR HEAT GENERATION RATE................................... 3/4 2-5 l l LA SALLE - UNIT 2-IV Amendment No. 88 l

a i a 1 2 1 1 r INDEX BASES 1. i i SECTION PAGE i 3/4.0 j APPL I C AB I L ITY.................................................. B 3/4 0-1 j 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 _ SHUTDOWN MARGIN............................................. B 3/4 1-l' 3/4.1.2 REACTIVITY-AN0MALIES........................................ B 3/4 1-1 ? j 3/4.1.3 C ONT RO L R0D S................................................ B 3 /4 1 -2 j 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS................................ B 3/4 1-3 3/4.1.5 ~ STANDBY LIQUID CONTROL SYSTEM............................... B 3/4 1-4 ) 3/4.1.6 ECONOMIC GENERATION CONTROL SYSTEM.......................... B 3/4 1-5 i i 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.................. B 3/4 2-1 4 3/4.2.2 DELETED..................................................... B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RAT 10................................ B 3/4 2-2. 3/4.2.4 LINEAR HEAT GENERATION RATE................................. B 3/4 2-6 j 3/4.3 INSTRUMENTATION 4 l 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................... B 3/4 3-i 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION......................... B 3/4 3-2 1 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION j I NS T RUM ENTAT I ON............................................. B 3 /4 3 j 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION........... B 3/4 3-3 1 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION j I NSTRUMENTAT ION............................................. B 3/4 3-4 } 3/4.3.6 CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION................ B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION l Radiation Monitoring Instrumentation........................ B'3/4 3i4 j Seismic Monitoring Instrumentation.......................... B 3/4 3-4 I LA SALLE - UNIT 2 XII Amendment No. 88 i. , _ _ _ _ _ _. ~

LIMITING SAFETY SYSTEM SETTINGS l BASES i REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) Averaae Power Ranae Monitor (Continued). the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed unif'.rm rod withdrawal 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 flux trip remains active until the mode switch is placed in the Run position. The APRM trip system is calibrated using heat balance data taken during l steady state conditions. Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Fixed 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-Upscale setpoint, a time constant of 6 i I seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1. The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unnecessary shutdown. 3. Reactor Vessel Steam Dome Pressure-Hiah High pressure in the nuclear system could cause a rupture to the. nuclear system process barrier resulting in the release of fission products. A pressure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip setting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum 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 conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit. LA SALLE - UNIT 2 B 2-10 Amendment No. 88

l i i l l t l l l 3/4.2.2 INTENTIONALLY LEFT BLANK. l l t l l l l 4 l l i I l I l I t I LA SALLE - UNIT 2 3/4 2-2 Amendment No. 88

TARLE 4.3.1.1-1 (Continued) REACTOR PROTECTION' SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5 El CHANNEL OPERATIONAL i F: CHANNEL FUNCTIONAL CHANNEL-CONDITIONS FOR WHICH [' FUNCTIONAL UNIT-CHECK _ TEST CALIBRATION SURVEILLANCE REQUIRED Ei 8. Scram Discharge Volume Water

1 -

Level - High-NA M R 1,-2, 5 9.

Turbine Stop Valve - Closure NA Q

R 1 t c.

10. - Turbine Control Valve Fast Closure Valve Trip System 011 Pressure - Low NA Q

R 1

11. Reactor Mode Switch Shutdown Position NA R

NA-1,2,3,4,5

12. Manual Scram NA W

.NA 1,2,3,4,5 13. Control Rod. Drive a. Charging Water Header Pressure - Low NA M R 2, 5 b. Delay Timer NA M R 2, 5 wls w oo (a) Neutron detectors may be excluded from CHANNEL CALIBRATION. (b) The-IRM and SIU1 channels shall be determined to overlap for at least 1/2 decades ~during each startup and:the IRM and^ APRM channels shall be determined to overlap for at-least 1/2 decades-during each con-trolled shutdown,'if not performed within the previous '7 days. .t (c) Within ;24 hours-prior to startup, if 'not performed within the previous :7 days. (d) This calibration shall' consist of the adjustment of the APRM channel to conform to the power. levels calculated by a heat _ balance during OPERATIONAL CONDITION 1 when THERMAL POWER 2 25% of RATED THERMAL 3 POWER. The APRM Gain Adjustment Factor-(GAF) for any. channel shall be equa! Lto.the power value deter-mined'by the heat balance divided by the APRM reading for that channel. Eg Within 2 hours, adjust ~any'APRM channel with a GAF > 1.02. In addition, adjust any APRM channel within i jt 12 hours, if power is greater than or equal to 90% of RATED' THERMAL POWER'and the APRM channel GAF is g- < 0.98. Until any_ required APRM adjustment has'been accomplished, notification shall be posted on'the l 3l

  • reactor control panel.

EF '(e) _This calibration shall-consist of the' adjustment of the APRM flow biased ~ channel to conform to a 6 calibrated flow signal. (f) The LPRMs shall be calibrated at least once per 1000 effective fullf power hours (EFPH). _ (g)- Measure and compare core flow to rated core flow.. ~ (h) This calibration.shall consist of verifying the 6 i I second' simulated thermal power time constant. l ~ _.________,____________._______._.______m-

f ~ h' TABLE 3.3.6-2 (Continued) ? l;; CONTROL R0D WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS h TRIP FUNCTION TRIP SETPOINT ALLOWABLE VALUE Q N 5. SCRAM DISCHARGE VOLUME a. Water Level-High s 765' 5%" s 765' 5%" b. Scram Discharge Volume Switch in Bypass N.A. N.A. 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW a. Upscale s 108/125 of full scale s 111/125 of full scale b. Inoperative N.A. N.A. c. Comparator s 10% flow deviation 5 iik tiow deviation Y' W 6 E. 2 ,' a .F

  • The Average Power Range Monitor rod block function is varied as a' function of recirculation loop flow (W).

l m w a .t- - - +

3/4.4 REACTOR COOLANT SYSTEM j 3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITION FOR OPERATION l 3.4.1.1 ~ Two reactor coolant system recirculation loops shall be in operation. l APPLICABILITY: OPERATIONAL CONDITIONS I and 2 ACTION a. With only one (1) reactor coolant system recirculation loop in operation, comply with Specification 3.4.1.5 and: 1. Within four (4) hours: a) Place the recirculation flow control system in the Master Manual mode or lower, and b) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Specification 2.1.2, and c) Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Limiting Condition for Operation by 0.01 per Specification 3.2.3,

and, l

d) Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Values to those applicable to single recirculation loop operation per Specifications 2.2.1 and l 3.3.6. l 2. Otherwise, be in at least HOT SHUTDOWN within the next twelve (12) hours. b. With no reactor coolant recirculation loops 'in operation: 1. Take the ACTION required by Specification 3.4.1.5, and 2. Be in at least HOT SHUTDOWN within the next six (6) hours. l l LA SALLE - UNIT 2 3/4 4-1 Amendment No. 88 L-

3/4.2 POWER DISTRIBUTION LIMITS BASES { The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46. 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46. This specification also assures that fuel rod mechanical integrity is maintained during normal and transient operations. 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 secondarily on the rod to rod power distribution within an assembly. The peak I clad temperature 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 times 1.02 is used in the heatup code along with the exposure dependent steady-state gap conductance and rod-to-rod local peaking factor. The Technical Specification AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking i factor. However, the current General Electric (GE) calculational models (SAFER /GESTR described in Reference 3), which are consistent with the j requirements of Appendix K to 10 CFR 50, have established that APLHGR values are not expected to be limited by LOCA/ECCS considerations. APLHGR limits are still required, however, to assure that fuel rod mechanical integrity is maintained. They are specified for all resident fuel types in the Core Operating Limit Report based on the fuel thermal-mechanical design analysis. The purpose of the power-and flow-dependent MAPLHGR factors specified in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow and core power conditions. At less than 100% of rated flow or rated power, the required MAPLHGR is the minimum of either (a) the product of the rated MAPLHGR limit and the power-dependent MAPLHGR factor or (b) the product of the rated MAPLHGR limit and the flow-dependent MAPLHGR factor. The power-and flow-dependent MAPLHGR factors assure that the fuel remains within the fuel design basis during transients at off-rated conditions. Methodology l for establishing these factors is described in Reference 6. i LA SALLE - UNIT 2 B 3/4 2-1 Amendment No. 88 i

~

  • POWER DISTRIBUTION SYSTEMS BASES 3/4.2.2 DlLETED 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR and an analysis of abnormal operational transients.

For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2. l To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduc-tion in CRITICAL POWER RATIO (CPR). The type of transients evaluated were. loss of flow, increase in pressure and power, positive reactivity insertion, l and coolant temperature decrease. The limiting transient yields the largest delta MCPR. When added to the Safety Limit MCPR, the required minimum operating limit MCPR of Specification 3.2.3 is obtained and presented in the CORE OPERATING LIMITS REPORT. I Analyses have been performed to determine the effects on CRITICAL POWER RATIO (CPR) during a transient assuming that certain equipment is out of service. A detailed description of the analyses is provided in Reference 5. The analyses performed assumed a single failure only and established the licensing bases to allow continuous plant operation with the analyzed equipment out of service. The following single equipment failures are included are part of the transient analyses input assumptions: 1. main turbine bypass system out of service, 2. recirculation pump trip system out of service, i l l l LA SALLE - UNIT 2 B 3/4 2-2 Amendment No. 88 p+.,

POWER DISTRIBUTION SYSTEMS BASES MINIMUM CRITICAL POWER RATIO (Continued) 1 The value for r used in Specification'3.2.3 is 0.687 seconds which is conservative for the,following reason: For simplicity in formulating and implementing the LCO, a conservative n value for-I N, of 598 was used. This represents one full core data set i=1 at BOC plus one full core data set following a 120 day outage plus twelve 10% of core,19 rods, data sets. The 12 data sets are equivalent to 24 operating months of surveillance at the increased surveillance l frequency of one set per 60 days required by the action statements of i Specifications 3.1.3.2 and 3.1.3.4. i That is, a cycle length was assumed which is longer than any past or j contemplated refueling interval and the number of rods tested was maximized in order to simplify and conservatively reduce the criteria for the scram time at which NCPR penalization is necessary. l The purpose of the power-and flow-dependent MCPR limits specified.in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow and core power conditions. At a given power and flow operating condition, the required MCPR is the maximum of either the power-dependent MCPR l limit or the flow-dependent MCPR limit. The required MCPR assures that the Safety Limit MCPR will not be violated. Methodology for establishing the a power-and flow-dependent MCPR limits is described in Reference 6. 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 moderator 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 considerable margin. During initial start-up testing of the plant, a MCPR i i 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 below 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 for calculating MCPR when a limiting control rod patieni is approached ensures that MCPR will be known following a change in THERMAL POWER or power shape, regardless of magnitude, that could place operation at a thermal limit. LA SALLE - UNIT 2 B 3/4 2-5 Amendment No. 88

  • ^
  • I l

. POWER DISTRIBUTION SYSTEMS BASES 3/4.2.4' LINEAR HEAT GENERATION RATE . The specification assures that the LINEAR HEAT GENERATION RATE.(LHGR) in any rod is'less than the design linear heat generation even if fuel pellet J densification is postulated. _ The power spike penalty specified is based on the analysis presented in Section 3.2.1 of the GE topical report NEDM-10735 l Supplement 6, and assumes a linearly' increasing variation'in axial gaps between core bottom and top and assures with a 95% confidence that no more than one fuel rod exceeds the design LINEAR HEAT GENERATION RATE due to power l spiking.' l

References:

l l 1. General Electric Company Analytical Model for Loss-of-Coolant } Analysis in Accordance with 10 CFR 50, Appendix K, NED0-20566A, i September 1986. 2. " Qualification of the One-Dimensional Core Transient Model for. l Boiling Water Reactors," Ger:eral Electric Co. Licensing: Topical i Report NED0 24154 Vols. I and II and NEDE-24154 Vol. III as sup-plemented by letter dated September 5, 1980, from R. H. Buchholz l (GE) to P. S. Check (NRC). l i y 3. "LaSalle County Station Units 1 and 2 SAFER /GESTR - LOCA Loss-of-j Coolant Accident Analysis," General Electric Co. Report NEDC-32258F, j October 1993. 4. " General Electric Standard Application for Reactor Fuel," NEDE-24011-P-A (latest approved revision). 5. " Extended Operating Domain and Equipment Out-of-Service for LaSalle County Nuclear Station Units I and 2," NEDC-31455,, November 1987. 6. " ARTS Improvement Program Analysts for LaSalle County Station Units 1 and 2," General Electric Co. Report NEDC-31531P, December 1993. ) 1 u 1 ( LA.SALLE - UNIT 2 B 3/4 2-6 Amendment No 88 l !1 l o

r INSTRUMENTATION l -BASES-3/4.3.5 REACTOR CORE ISOLATI'ON COOLING SYSTEM ACTUATION INSTRUMENTATION L L The reactor core isolation' cooling system actuationLinstrumentation-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 : 1 cooling equipment. 3/4.3.6 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION L The control rod block functions are.provided ' consistent with the' require - ments of the: specification in Section 3/4.1.4,, Control' Rod Program Controls.. The trip logic is arranged sol that= a trip in any one of the. inputs will result) in a control rod block;- 3/4.3.7 MONITORING INSTRUMENTATION i 3 /4 '. 3. 7.1 RADIATION MONITORING INSTRUMENTATION. j The OPERABILITY of-the' radiation monitoring instrumentation ensures that: (1) the radiation levels 'are continually measured in the: areas served by the-individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded. 3.4.3.7.2 SEISMIC MONITORING INSTRUMENTATION-The OPERABILITY of the seismic monitoring inshumentation 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 that used in the design basis for the unit. This instrumentation is consistent with the recommendations of Regulatory Guide 1.12 " Instrumentation for Earthquakes," April 1974. 3/4.3.7.3 METEOROLOGICAL MONITORING INSTRUMENTATION The OPERABILITY of. the meteorological monitoring instrumentation ensures that sufficient meteorological data is 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 recommenda-tions of' Regulatory Guide 1.23 "Onsite Meteorological Programs," February 19'i2. 3/4.3.7.4 REMOTE SHUTDOWN MONITORI@ INSTRUMENTATION The OPERABILITY of the remote shutdown monitoring instrumentation ensures-that sufficient capability is available to permit shutdown and maintenance 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. LA SALLE - UNIT 2 B 3/4 3-4 Amendment No. 88'

i I -o ADMINISTRATION CONTROLS i ' Core Operatino Limits Report (Continued) (1) The Average Planar Linear Heat Genemion Rate (APLHGR) for-Technical Specification 3.2.1. (2).The minimum Critical Power Ratio (MCPR)its, and power and scram time, tau (7), dependent MCPR lim (including 205 flow de 3.2.3. pendent MCPR limits) for Technical Specification (3) The Linear Heat-Generation Rate' (LHGR)'for Technical Specification 3.2.4. 1 (4) The Rod Block Monitor Upscale Instrumentation Setpointsifor i Technical Specification Table 3.3.6-2. I b. The analytical methods used to determine the core. operating limits shall be those previously reviewed and approved by the NRC in the latest approved revision or su)plement of the topical reports describing the methodology. For aSalle County Station Unit 2, the topical reports are: (1) NEDE-240ll-P-A, " General Electric Standard Application for Reactor Fuel," (latest approved revision).. l (2) Commonwealth Edison Topical Report NFSR-0085, " Benchmark of-BWR Nuclear Design Methods," (latest approved revision). (3) Commonwealth Edison Topical Report NFSR-0085,' Supplement 1, i " Benchmark of BWR Nuclear Design Methods -- Quad Cities Gamma Scan Comparisons,".(latest approved revision). (4) Commonwealth Edison Topical Report NFSR-0085, Supplement 2, i " Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," (latest approved revision). ] c. The core operating limits shall be determined so that all l applicable limits thermal-hydraulic '(e.g., fuel thermal-mechanical limits, core imits, ECCS Limits, nuclear limits such as shutdown margin, and transient and accident analysis limits).of the safety analysis are met. d. The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon-issuance, for each reload cycle to the U.S. Nuclear Regulatory CommissionDocumentControlDeskwithcopiestothe. Regional-Administrator and Resident Inspector. ) B. Deleted. ~i i 4 i LA SALLE - UNIT 2' 6-25 Amendment No.-88 -}}

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