ML20237G726

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Amend 7 to License NPF-39,incorporating Operating Limits for All Fuel Types for Cycle 2 Operation of Facility,Initiating Change in Slope of flow-biased APRM Scram & Rod Block Setpoints & Modifying Bases Associated W/Core Reloads
ML20237G726
Person / Time
Site: Limerick 
Issue date: 08/14/1987
From: Butler W
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20237G729 List:
References
NUDOCS 8708240204
Download: ML20237G726 (36)
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E WASHINGTON, D. C. 20655 l/ o PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION, UNIT I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 7 License No. NPF-39 ~ 1. The Nuclear Regulatory Corrnission (the Comission) has found that i A. The application for amendment by Philadelphia Electric Company (the licensee) dated April 3, 1987, 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 l provisions of the Act, and the rules and regulations of the l Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and I safety of the public, and (ii) that such activities will be conducted in compliarce 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 1 of the Comission's regulations and all applicable requirements have j been satisfied. j 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 The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 7 , are hereby incorporated into this license. Philadelphia Electric Company shall operate the facility in accordance with the l Technical Specifications and the Environmental Protection Plan. ) ) 1 i 8708240204 G70014 i PDR ADOCK 05000352 l P PDR l 1 .1

1 3. This license amendment is effective prior to startup cf Limerick Unit 1 l in Cycle 2. i l FOR THE NUCLEAR REGULATORY COMMISSION l /S/ l Walter R. Butler, Director Project Directorate I-2 Division of Reactor Projects I/II

Attachment:

Changes to the Technical j. Specifications l i Date of Issuance: August 14, 1987 il l l l 1 l l l ? .f) PDI-2/PM 0C PDI-2/D P j M B e RClark: WButler (/87 of/of/87 7/(0/87 / 87

1 1 \\ l-l. 2 3. This license amendment is effective prior to startup of Limerick ' Unit 1 in Cycle 2. FOR THE NUCLEAR REGULATORY COMMISSION Walter R. Butler, Director 4 Project Directorate I-2 Division of Reactor Projects I/II

Attachment:

Changes to the Technical Specifications l l Date of Issuance: August-14, 1987 f 1 I f l 1 o l 1 j

I i i ) ATTACHMENT TO LICENSE ~ AMENDMENT NO, 7 FACILITY' OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 Replace.the following pages of the Appendix A Technical Specifications with-the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Overleaf pages are provided to maintain document completeness.* Remove Insert iii* iii* iv iv v* v* vi .vi xvii

  • xvii
  • xviii xviii 2-1 2-1 2-2*

2-2* 2-3* 2-3* 2-4 2-4 B 2-1 8 2-1 B 2-2 B 2-2 B 2-3 B 2-3 I B 2-4 B 2-4 3/4 2-1 3/4 2-1 3/4 2-2* 3/4 2-2* l 3/4 2-6a 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 3/4 2-10a 3/4 2-10a 3/4 2-11*. 3/4 2-11* 3/4 2-12 3/4 2-12

! 3/4 3-59* 3/4 3-59* 3/4 3-60 3/4 3-60 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 l B 3/4 2-3 B 3/4 2-3 B 3/4 2-4 B 3/4 2-4 B 3/4 2-5 B 3/4 2-5 1 4 4 l i l l l l l l l

1 i INDEX ( DEFINITIONS l SECTION j PAGE DEFINITIONS (Continued) 1-8 1.45 UNRESTRICTED AREA. 1.46 VENTILATION EXHAUST TREATMENT SYSTEM........................ 1-8 1-8 f 1.47 VENTING. Table 1.1, Surveillance Frequency Notation.. 1-9 J Table 1.2, Operational Conditions.. 1-10 i l L1MERICK - UNIT 1 iii

4 i i l { SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i PAGE { SECTION 2.1 SAFETY LIMITS 2-1 THERMAL POWER, low Pressure or Low Flow..................... 2-1 THERMAL POWER, High Pressure and High Flow.................. 2-1 Reactor Coolant System Pressure............................. 2-2 Reactor Vessel Water Leve1.................................. LIMITING SAFETY SYSTEM SETTINGS 2.2 2-3 Reactor Protection System Instrumental, ion Setpoints......... . Tabis 2.2.1-1 Reactor Protection System 2-4 Instrumentation Setpoints.............. BASES 2.1 SAFETY LIMITS B 2-1 THERMAL POWER, Low Pressure or Low Flow..................... B 2-2 THERMAL POWER, Hign Pressure and High Flow.................. B 2-3 Left Intentionally B1ank.................................... B 2-4 Left Intentionally B1ank.................................... B 2-5 Reactor Coolant System Pressure............................. B 2-5 Reactor Vessel Water Leve1.................................. t 2.2 LIMITING SAFETY SYSTEM SETTINGS B 2-6 Reactor Protection System Instrumentation Setpoints.......... Amendment No. 7 iv LIMERICK - UNIT 1 I

1 j INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECT 10N PAGE j 3/4.0 APPLICABIL]TY., 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.......................................... 3/4 1-1 l 3/4.1.2 REACTIVITY ANOMALIES.................................... 3/4 1-2 j 3/4.1.3 CdNTROLRODS 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 l 1 Four Control Rod Group Scram Insertion Times............. 3/4 1-8 ' l Control Rod Scram Accumulators........................... 3/4 1-9 Control Rod Drive Coupling............................... 3/4 1-11 I I Control Rod Position Indication.......................... 3/4 1-13 Control Rod Drive Housing Support........................ 3/4 1-15 1 l 1 l ~3/4.1.4 CONTROL ROD PROGRAM CONTROLS l l Rod Worth Minimi2er...................................... 3/4 1-16 1 i Rod Sequence Control System.........................,..... 3/4 1-17 1 Rod Block Monitor........................................ 3/4 1-13 l 3/4.1.5 ST ANDBY LIQUID CONTROL SYSTEM............................ 3/4 1-19 l l Figure 3.1.5-1 Sodium Pentaborate Solution Temperature / Concentration Requirements........................ 3/4 1-21 i Figure 3.1.5-2. Sodium Pentaborate Solution l Vol ume/ Conc e ntrati on Requi reme nts... 3/4 1-22 1 1 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

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1 SECTION ' PAGE POWER DISTRIBUTION LIMITS (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 \\ Figure 3.2.1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial: q Core Fuel Types P8CIB094........... 3/4 2-5 Figure 3.2.1-5. Maximum Average Planar Linear Heat l l i l Generation Rate (MAPLHGR) Versus Average Planar Exposure Initial' Core Fuel Types P8CIB071........... 3/4 2 j 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 t 3/4 2.2 APRM SETP01NTS.......................................... 3/4 2-7 3/4'2.3 MINIMUM CRITICAL POWER RATI0............................ 3/4 2-8 Table 3.2.3'-1 Deleted l j Figure 3.2.3-la Minimum Critical Power Ratio (MCPR) I Versus (P8X8R/BP8X8R Fuel)............. 3/42-10l i i ~ Figure 3.2.3-lb Minimum Critical Power Ratio (MCPR) Versus~(GE8X8EB Fuel).................. 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 1 Table 3.3.1-2 Reactor.' Protection System 3/4 3-6 j Response Times...................... Table 4.3.1.1-1 Reactor Protection System Instrumentation Surveillance Requirements...................... 3/4 3-7 1 i LIMERICK UNIT - 1 vi Amendment No. 7 1 1

l. INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMEN S PAGE SECTION RADI0 ACTIVE EFFLUENTS (Continued) Table 4.11.2.1.2-1 Radioactive Gaseous Waste Sampling and Analysis 3/4 11-9 Program............'. 3/4 11-12. 1 Dose - Loble Gases....................................... Do'se.- Iodine-131,' Iodine-133, Tritium, and 3/4 11-13 Radionuclides in Particulate Form..................... 3/4 11-14 Ventilation Exhaust Treatment System.................... 3/4 11-15 j Explosive Gas Mixture................................... 3/4 11-16 Main Condenser.............. 3/4 11-17 Venting or Purging...................................... 3/4 11-18 1 3/4.11.3 SOLID RADWASTE TREATMENT................................ 3/4 11-20 3/4.11.4 TOTAL D0SE.............................................. i 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4 12-1 { 3/,.12.1 MONITORING PR0 GRAM....................................... 4 Table 3.12.1-1 Radiological Environmental Monitoring Program................. 3/4 12-3 l 1 Table 3.12.1-2 Reporting Levels For Radio-activity Concentrations In Envi ronmental Sampl es.............. 3/4 12-9 l Table 4.12.1-1 Detection Capabilities For Environmental Sample Analysis...... 3/4 12-10 3/4 12-13 l 3/4.12.2 LAND USE CENSUS......................................... 3/4.12.3 INTERLABORATORY COMPARISON PR0 GRAM...................... 3/4 12-14 i ' e* A LIMER7CK - UNIT I xvii f

i BASES l ~ SECTION PAGE 3,/4.0 APPLICABILITY............................................. B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN....................................... B 3/4 1-1 3/4.1.2-REACTIVITY AN0MALIES.................................. B 3/4 1-1 3/4.1.3 CO NT RO L R0D S.......................................... B 3/4 1-2 3/4.1.4 CONTROL R0D PROGRAM CONTR0LS.......................... B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM......................... B 3/4 1-4 1 3/4.2 POWER DISTRIBUTION LIMITS i 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION i RATE.................................................. B 3/4 2-1 1 LEFT INTENTIONALLY BLANK........................................ B 3/4 2-3 1 3/4.2.2 APRM SETP0lNTS........................................ B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATIO......... ............... B 3/4 2-4 3/4.2.4 LINEAR HEAT GENE RATION RATE........................... B 3/4 2-5 l ~ 3/4". 3 INSTRUMENTATION 1 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION............. B 3/4 3-1 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 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION..... B 3/4 3-3 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION....................................... B 3/4 3-4 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION..................... B 3/4 3-4 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation.................. B 3/4 3-4 LIMERICK - UNIT 1 xviii Amendment No.7

1 I I l 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow. ] 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow. APPLICAB_ILITY: OPERATIONAL CONDITIONS 1 and 2. ACTION: With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel . steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of Specification 6.7.1. i THEAMAL NOWER,-High Pressure and High Flow ~ 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 with l the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow. APPLICABILITY: OPERATIONAL CONDITIONS I and 2. ACTION: With MCPR less than 1.07 and the reactor vessel steam dome pressure greater l I than-785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours and comply with the requirements of Specification 6.7.1. ] l REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig. APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3, and 4. ACTION: With the reactor coolant system pressure, as measured in tha reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN witL reactor coolant system pressure less than or equal to 1325 psig within 2 L:urs and comply with the requirements of Specification 6.7.1. LIERICK - 19(IT 1 2-1 Amendment No. 7

-~ SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SAFETY LIMITS (Continued) ~ REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel. APPLICABILITY: OPERATIONAL CONDITIONS 3, 4, and 5 l ACTION: l With the reactor vessel water level at or below the top of the active irradiated fuel, manually initiate the ECCS to' restore the water level, after depressurizing the. reactor vessel, if required. Comply with the requirements of Specification 6.7.1. j i l-l LIMERICK - UNI' 1 2-2

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS ~ REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The reactor protection system instrumentation setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2.1-1. APPLICABILITY: As shown in Table 3.3.1-1. ACTION: With a reactor protection system instrumentation setpoint less conservative than the value shown in the Allowable Values column of Table 2.2.1-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERABLE status with its setpoint adjusted consistent with the Trip Setpoint value. ) ] l I i I l s I l 1 l a 4s LIMERICK - UNIT 1 2-3 l

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I l s i 2.1 SAFETY LIMITS I BASES l

2.0 INTRODUCTION

i 1 The fuel cladding, reactor pressure vessei and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these l barriers during normal plant operations and anticipated transients. The fuel j cladding

  • integrity Safety Limit is set such that no fuel damage is calculated

) to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the MCPR is not less than 1.07. MCPR greater than 1.07 represents a con-l servative margin relative to the conditions required to maintain fuel cledding I integrity. The fuel cladding is one of the physical barriers which separate 1 the radioactive materials from the environs. The integrity of this cladding i 2 barrier is related to its relative freedom from perforations or cracking. Although,some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, ~, can result from thermal stresses which occur from reactor operation signifi-cantly above design conditions and the Limiting Safety System Settings. While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thernally caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a signi-ficant departure from the condition intended by design for planned operation. 2.1.1 THERMAL POWER, low Pressure or low Flow 1 The use of the (GEXL) correlation is not valid for all critical power 1 calculations at pressures below 785 psig or core flows less than 10% of rated l flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL l POWER with the following basis. Since the pressure drop in the bypass region l is essentially all elevation head, the core pressure drop at low power and flows will alwa flow of 28 x 10gs be greater than 4.5 psi. Analyses show that with a bundle lb/h, bundle pressure drop is necrly independent of bundle i power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving i head will be greater than 28 x 103 lb/h. Full scale ATLAS test data taken l at pressures from 14.7 psia to 800 psia indicate that the fuel assembly criti-l cal power at this flow is approximately 3.35 MWt. With the design peaking j factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL l POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. I i LIERICK - (MIT 1 B 2-1 Amendment No. 7 j i l I i

c n l l l ) 3 SAFETY LIMITS j l BASES 2.1.2 THERMAL POWER, High Pressure and High Flow The fuel ciaddias integrity Safety Limit is set such that no fuel damage is calculated to occLr if the limit is not violated. Since the parameters which result in fuel damage are not'directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage j 1 could occur. Although it is recognized that a departure from nucleate boiling l would not necessarily result in damage to BWR fuel rods, the critical power at l which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty .in the value of-the critical power. Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting' fuel assembly for which more than 99.9% of the ' eel rods in tha core are* expected to avoid boiling transition I consider'ing the power distribution within the core and all uncertainties. l The Safety Limit MCPR is determined using a statistical model that combines l i all of the uncertainties in operating parameters and the procedures used to l calculate critical power. Calculation of the Safety Limit MCPR is described in Reference 1. l l l l l l

Reference:

1. " General Electric Standard Application for Reactor Fuel," HEDE-24011-P-A (latest approved revision). LDERICK - IMIT 1 B 2-2 Amen w nt No. 7

1 l .I 1 l l \\ l LEFT INTENTIONALLY BLANK l ) i e l l j 1 ,i l, LIMERICK - UNIT I B 2-3 Amendment No. 7

.I 1 l l l l 1 l LEFT INTENTIONALLY BLANK 1 \\ t t I LIMERICK - tMIT I E 2-4 Amenchnent No. 7

3/4.2 ' POWER DISTRIBUTION LIMITS ~3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE 4 LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type 3 of fuel as a function of axial location and AVERAGE PLANAR EXPOSURE shall be 1 within limits br. sed on-applicable APLHGR limit values which have been approved for'the respective fuel and lattice types. When hand calculations are required, { the APLHGR for each type of fuel as 'a function of AVERAGE PLANAR EXPOSURE shall q not exceau the limiting value for the most limiting lattice (excluding natural 1 uranium) as shown in the applicable figures for BP/P8X8R and GE8X8EB fuel types. d APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POV O is greater than or val to 25% of RATED THERMAL POWER. ACTION: With an iPLHGR exceeding the li iting value, initiate corrective action within l 15 minutes and restore APLHGR to within the required limits within 2 hours or, reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall t,e verified to be equal to or less than the limiting value a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR. d. The provisions of Specification 4.0.4 are not applicable. r LIMERICK - UNIT 1 3/4 2-7 Amendment No. 7 __=

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POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The APRM flow biased neutron flux-upscale scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be esteb'ished according to the following relationships: TRIP SETPOINT ALLOWABLE VALUE S < (0.58W + 59%)T S < (0.58W + 62%)T l S $ (0.58W + 50%)T Spg 1 (0.58W + 53%)T l RB where: S and So8 are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 100 million lbs/hr, T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER l ~ divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY. T is applied only if le:s than or equal to 1.0. I APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: With the APRM flow biased neutron flux-upscale scram trip setpoint and/or the flow Diased neutron flux-upscale cor.iral rod block trip setpoint less l conservative than the value shown in the Allowable Value column for S or S as above determined, initiate corrective action within 15 minutes ak$,adjustSand/or5 to be consistent with the Trip Setpoint values

  • l i

within6hoursorredb$eTHERMALPOWERtolessthan25%ofRATEDTHERMAL l POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.2.2 The FRTP and the MFLPD shall be determined, the value of T calculated, and the most recent actual APRM flow biased neutron flux-upscale scram and flow biased neutron flux-upscale control rod block trip setpoints verified to be within the above limits or adjusted, as required: a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and Initially and at least once per 12 hours when the reactor is operating c. with MFLPD greater than or equal to FRTP. d. The provisions of Specification 4.0.4 are not applicable.

  • With MFLPD greater than the FRTP during power ascension up to 90% of RATED l

THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times MFLPD, provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER and a notice of adjustment is posted on the reactor control panel. LIMERICK - UNIT 1 3/4 2-7 Amendment No. 7 k

l I l i i POWER DISTRIBUTION LIMITS l ) 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 1 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR) shall be equal to or greater than the MCPR limit shown in Figure 3.2.3-la (BP/P8X8R fuel) and Figure 3.2.3-lb (GE8X8EB fuel), times the K, shown in Figure 3.2.3-2, provided that the end-of-cycle recirculation pump trip (EOC-RPT) system is OPERABLE per Specification 3.3.4.2, with: (Tave I ) B T ~I A B where: A = 0.86 seconds, control rod average scram insertion T time limit to notch 39 per Specification 3.1.3.3, B = 0. 672 + 1. 65[- ]b(0.016), ,,, l 1 T n N I i=1 n I Nt I t,y,, 4,7 n N i i=1 n = number of surveillance tests performed to date in cycle, th N$ = number of active control rods measured in the i surveillance test, $ = average scram time to notch 39 of all rods measured t th in the i surveillance test, and total number of active rods measured in Specification Ny = 4.1.3.2.a. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to l 25% of RATED THERMAL POWER. l l LIMERICK - UNIT 1 3/4 2-8 Amendment No. 7

POWER DISTRIBUTION LIMI15 LIMITING CONDITION FOR OPERATION (Continued) ACTION With the end-of-cycle recirculation pump trip system inoperable per. a. Specification 3.3.4.2, operation may continue and the provisions of Specification 3.0.4 are not applicable provided that, within I hour, MCPR is determined to be greater than or equal to the MCPR limit as a function of the average scram time shown in Figure 3.2.3-la (BP/P8X8R fuel) and Figure 3.2.3-lb (GE8X8EB fuel), EOC-RPT inoper-able curve, times the K shown in Figure 3.2.3-2. f b. With MCPR less than the applicable MCPR' limit shown in Figures 3.2.3-la, 3.2.3-Ib and 3.2.3-2, initiate corrective action within 15 minutes and restore MCPR to within the required limit within 2 hours or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next , 4 hours. SURVEILLANCE REQUIREMENTS 4.2.3 MCPR, with: a. t = 1.0 prior to performance of the initial scram time measurements for the cycle in accordance with Specification 4.1.3.2, or b. t as defined in Specification 3.2.3 used to determine the limit-within 72 hours of the conclusion of each scram time surveillance test required by Specification 4.1.3.2, shall be determined to be equal to or greater than the applicable MCPR limit determined from Figures 3.2.3-la, 3.2.3-lb and 3.2.3-2. l a. At least once per 24 hours, b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is f operating with a LIMITING CONTROL ROD PATTERN for MCPR. I d. The provisions of Specification 4.0.4 are not applicable. ) l l 1 i I i LIMERICK - UNIT 1 3/4 2-9 Amendment No. 7 .I 1

i- ] _j 1 l 1 1 t \\ l 1.44 1.44 l 4 t,42 _ ..:....:....:.... :..~. :... -1,42 y i i I 1.40- -1.40 I

F lCI + FFWTR.

1.38- !A-l + 1.38 IC WITH0g1 EOC-Rpq'wtTH,,,;,,,,,, ?,,,,,, ;,,,,,,j,,,,, - -1.36 1.36-j i i -1.34 1.34-t s- <t' l i l D 1.32- +4++iisih+:- -1.32 ,$... wiTH E'OC-RPT.)CF ICF + FFWTR l y i4- -1.30 1.30-

t. wtTH. EOC-RPT 100 ytow :! -l i

- -1.28 l 1.28- ~ f,. l 1.26- +a+:+++++- -1.26 l i 1.24 - .i a. 4.- o. t.- 5 t.* - -1.24 i. i .S FH..OOS.i l.' INCLUDE -1.22 1.22-1 l i 1.20 l l l l l l l 1.20 0 0.10.20.30.40.50.60.70.80.9 2 T l j l D UIN fTIO NS-LCI-INSREASED CORE FLOW (UP TO 105% RATED) E 1Q2.5 TEMP. REDU; TION; ACHIEVED BY REMOVAL OF FEEDWATER HEATER (5))FEEDWATER EEEIB - rawAL rEEDwATER TEMPERATURE RtDuCTioN AT END-OF-CYCLE (UP TO 60 DEG.F TEWP REDUCTION: ACHIEVED BY REWOVAL OF ALL $TM STAGE HEATERS) l t. I MINIMUM CRITICAL POWER RATIO (MCPR) VERSUS t (P8X8R/BP8X8R FUEL) FIGURE 3.2.3-la LIMERICK - UNIT 1 3/4 2-10 Amendment No. 7 1

i 1.44 1.44 c 1.42-4- 4-4- -1.42 .-".-"t"">- ." "i"- ~~< - -1.40 1.40- "t"" i i i i j "" j- "l- "i jf,\\C + IU' "l 38 1.38- ~~ WITHOW EOC-RPI'.j.. g +.-j-..!_.-{... 21.36 1.36- - j,34 j j,34 .g. .;....<....g......y....., oc D 1.32 - " " l ~ ~ k " "." " " $ " " "b ~ ~ k" " " l." " ; " I " " - -1.32 " "l " W!TH EOC k' PT,ICF.ICF + prWTR j 2 -"3 -1.30 l 1 1.30 - 0 0 7* 7 to W : 1.28- " "i"" >" r""<-- -1.28

wgTHEOC-gpq'1 :

1.26- - i "":-~~ -1.26 1.24 - .5 4. 3. 4. 4. - -1.24 i. .i !.*lNCLUDE.S FH005 i .i i. 1.22- ><- -1.22 1.20 l l l i l l l 1.20 l 0 0.10.20.30.40.50.60.70.80.9' 1 T l ntrmmons-l LCf - lNCREASED CORE FLOW (UP TO ID5% RATED) Etic.D.s - rEEDWATER HEATING OUT OF SERVICE THROUCHOUT CYCLE (UP TO 60 DEG.F TEMP. REDUCTION: ACHIEVED BY REMOVAL OF FEEDWATER HEATER (S)) EEg.T,B - FINAL FEEDWATER TEMPERATURE REDUCTION AT END-OF-CYCLE (UP 70 60 DEC.F TEMP REDUCTION: ACHIEVED BY REWOVAL OF ALL STH FTAGE HEATERS) l MINIMUM CRITICAL POWER P.ATIO (MCPR) VERSUS t (GE8X8EB FUEL) l L FIGURE 3.2.3-lb LIMERICK - UNIT 1 3/4 2-10a Amendment No. 7 a

'e 00 1 = s 09 LOR TN 09 O C hA W O L F C w IT 0 o A l 7 F M e O ro TU C 2 A d R e O 3 s T e N C 2 R A f F 3 0 O 3 \\% 7 E K R U w G I o F l // e F N r O o IT 0 C A 5A00 5 R 1717 B 9911 M 1111 LA a=== L C OR TT MMMM T M A UUUU fH MMMM N OT IItt 0 O P C TO XXXX 4 N AAAA ES W SD MMMM O EE WWWW L BN UO OOOO F LLLL L Tl FFFF A i U ft S 0 N 0O 03 A C P M S 4 3 3 W 1 1 9 C4sO' $ 7U i

i i POWER DISTRIBUTION LIMITS i 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATICN 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13.4 kW/ft for BP/P8X8R fuel and 14.4 kw/f t for GE8X8EB fuel. APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER. ACTION: With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and restore the LHGR to within the limit within 2 hours or ~- reduce THERMAL POWER to less than 25% of RATED 7 DERMAL POWER within the next 4 hours. 1 s SURVEILLANCE REQUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit: a. At least once per 24 hours, i f b. Within 12 hours after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and c. Initially and at least once per 12 hours when the reactor is operating on a LIMITING CONTROL ROD PATTERN for LHGR. d. The provisions of Specification 4.0.4 are not applicable. 1 i l l LIMERICK - UNIT 1 3/4 2-12 Amendment No. 7 ) i

I TABLE 3.3.6-1 (Continued) l CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION ACTION STATEMENTS ACTION 60 Declare the RBM inoperable and take the ACTION required by l Specification 3.1.4.3. ACTION 61 With the number of.0PERABLE channels one or more less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one inoperable channel in the ) tripped condition within one hour, j t ACTION 62 With the number of OPERABL- 'hannels less than required by the Minimum OPERABLE Channels p.c Trip Function requirement, place the inoperable channel in the tripped condition within one hour. ACTION 63 With the number of OPERABLE channels'less than required by the Minimum OPERABLE Channels per Trip Function requirement, initiate a rod block. NOTES With THERMAL POWER > 30% of RATED THERMAL POWER. i With more than one control rod withdrawn. Not applicable to conte:1 -95 removed per Specification 3.9.10.1 or 3.9.10.2. ( These channels are not required when siyt.een or fewer fuel assecblies, .edjacent to the SRMs, are in the core. (a) The RBM shall be automatically bypassed when a peripheral control rod is selected or the reference APRM channel indicates less than 30% of ~ RATED THERMAL POWER. (b) This function shall be automatically bypassed if detector count rate is > 100 cps or the IRM channels are on range 3 or higher. j i (c) This function is automatically bypassed when the associated IRM channels are on range 8 or higher. (d) This function is autoestically bypassed when the IRM channels are on range 3 or higher. (e) This function is automatically bypassed when the IRM channels are on range 1. ij 'v LIMERICK - UNIT 1 3/4 3-59 Amendment h 4 MAY ) 1 1987

l R W WO l R RE E EW O OP l n P P u I o a L f f i L LA o t h A AM f a t M MR s o v i R RE n e w E EH o s l H HT i n e T T s s o E U D p i i "6 L 4 D 3 DE c v s A 4 E 5 ET i i 1 T TA 5 d v / V +f A + AR 0 e i 9 E o R .R 1 s 5l d f p 2 a 7 L W W A.f o x c 1 c 5 B m f A 6u% o 8N o /s 2 m0 0 1 e '7 W 6.i 1 5 6 8 . 1l ./l 5 %4 O 0 x 1 A. % A.1 A.1 A.1 l A. 3 a 2 3 0 31 L u c L a A 5m<R> <~ >3 N<H> N<TN>s 5 S TN I R O R RE P E EW T W WO l E O OP l n S P P u o a L f f i N L LA o t O h A AM f a I t M MR s o v F T i R RE n e A w E EH o s l 2 T H HT i n e N T T s o E D i i 6 "6 M T 1 D 0 DE s v s 3 U N 4 E 5 ET p i i 1 R I T TA c d v / T O +f A + AR e i 9 3 S P o R R s 5l d E N T W W f 0 2a 5 L I E m f f o 1 s 1 c 5 B S 6u% o 8 o p / s 2 A K m7 6.i 0 5 .%2 .x.c 8 1 e 7 T_ C P 0l ./l 5 O I 0 x 1 A. 5 0A.41 A.1 A. 3 A.1 l A.5 a 2 L R a u c B T 5 m 5N > <H>3 M<H> N<TN>s 5 'D O R e l L a O c p R s u T p t N U r O a C t S xu S d l e R e F l O p a n Tn m n c i Ii a o s N E l r p ST Ol M c t U RT Ml Uhh d u Ou u L gc e w e Tf Ef Oit R s o N I G VHi O a l x Nt Nt El S-w T i f e de u O o e Ao t I b v ev l Mn v R n v Ge N hie sieF ie ie R vt O w gtl atl Er tl Er tl A ea Meo i aa l aan Goeaa T oe aa HL o l l hrc Brco Ntl rc Atl rc C l N Kaf es esr Acaes I ca es S rF O C c pn wpnt R ecpn Dec pn I e I Os .ow oowu t sow Et s ow T L p. 1 no l noe Eepno Mep no, Dt a C BUi 1ID FIDN CDUID RDU ID MWa N M R E A U D R U T R N P O F O C R a bc Aabcd Sabcd I ab cd S a P I R T 1 2 3 4 5 Cx0n7 $e w wA wE (h${. M

3/4.2 POWER DISTRIBUTION LIMITS BASES 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature (PCT) following the postulated design basis Loss-of-Coolant Accident (LOCA) will not exceed the limits specified in 10 CFR 50.46 and that the. fuel design analysis limits specified in NEDE-24011-P-A (Reference 2) will not be exceeded. l Mechanical Design Analysis: NRC approved methods (specified in Refer-ence 2) are used to demonstrate that all fuel rods in a lattice operating at the bounding power history, meet the fuel design limits specified in Reference 2. l No single fuel rod follows, or is capable of following, this bounding power history. This bounding power history is used as the basis for the fuel design analysis MAPLHGR limit. LOCA Analysis: A LOCA analysis is performed.in accordance with 10 CFR 50 Appendix K to demonstrate that the permissible planar power (MAPLHGR) limits comply with the ECCS limits specified in 10 CFR 50.46. The analysis is performed for the most limiting break size, break location, and single failure combination for the plant. The Technical Specification MAPLHGR limit is the most limiting composite of the fuel mechanical design analysis MAPLHGR and the ECCS MAPLHGR limit. Only the most and least limiting MAPLHGR values are shown in the Technical l Specifications for multiple lattice fuel. Compliance with the specific lattice MAPLHGR operating limits, which are available in Reference 3, is ensured by use of the process computer. l 1 r LIMERICK - UNIT 1 B 3/4 2-1 teendment No. 7

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based { on a power distribution which would yield the design LHGR at RATED THERMAL i POWER. The flow biased neutron flux-upscale scram trip setpoint and flow ) biased neutron flux-upscale control rod block functions of the APRM instruments I must be adjusted to ensure that the MCPR does not bec0me less than the Safety Limit MCPR specified in Reference 2 or that > 1% plastic strain does not occur i the degraded situation. The scram and ro3 block setpoints are adjusted in accordance with the forcula in Lnis specification when the combination of THERMAL POWER and CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the' degraded condition. I 'f 1 I LIMERICK - UNIT I B 3/4 2-2 Amendment No.7 4 l Io ( t

i t I i l l k a l l l 1 1 LEFT INTENTIONALLY BLANK i 4 l l l I 1 l 1 I LIMERICK - UNIT 1 B 3/4 2,3 Amendment No.7 P

POWER DISTRIBUTION LIMITS BASB .c 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 specified in Reference 2, and an analysis l 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 tran-sients have been analyzed to determine which resul.t in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and i coolant' temperature decrease. The limiting transient yields the largest delta l MCPR. When added to the Safety Limit MCPR, the required minimum operating l limit MCPR of Specification 3.2.3 is obtained and presented in Figure 3,2.3-1: 1 1 The evaluation of a given transient begins with the system initial para- ] meters shown in FSAR Table 15.0-2 that are input to a GE-core dynamic behavior ) transient computer program. The codes used to evaluate transients are discussed in Reference 2. I The purpose of the K factor of Figure 3.2.3-2 is to define operating limits at other than ratec, core flow conditions. At less than 100% of rated j flow the required MCPR is the product of the MCPR and the K factor. The K factorsassurethattheSafetyLimitMCPRwillnotbeviola[edduringaflowf increase transient resulting from a motor generator speed control failure. The K factors may be applied to both manual and automatic flow control modes. f The K, factors values shown in Figure 3.2.3-2 were developed generically and are aphiicable to all BWR/2, BWR/3, and BWR/4 reactors. The K factors were f derived using the flow control line corresponding to RATED THERMAL POWER at rated core flow. For the manual flow control mode, the K factors were calculated such that f for the maximum f E. rr.te, as limited by the pump scoop tube set point and the corresponding TF il POWER along the rated flow control line, the limiting bundle's relat' sower was adjusted until the MCPR changes with different core flows. The rauo of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR, determiner the K. f LIMERICK - UNIT 1 B 3/4 2-4 Amendment No. 7

1 l POWER DISTRIBUTION LIMITS l BASES MINIMUM CRITICAL POWER RATIO (Continued) For operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established such that the MCPR was equal to the operating limit MCPR at RATED THERMAL POWER and rated thermal flow. l The K, factors shown in Figure 3.2.b-2 are conservative for the General Electric B6iling Water Reactor plant operation because the operating limit i MCPRs of Specification 3.2.3 are greater than the original 1.20' operating limit MCPR used for the generic derivation of K. f 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 I moderator void content will be very small. For all designated control rod j patterns.which may be employed at this point, operating plant experience indi-l cates that the-resulting MCPR value is in excess of requirements by a considerable l margin. During initial start-up testing of the plant, a MCPR evaluation will l 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 require-ment for calculating MCPR when a limiting control rod pattern is approached ensures that MCPR will be known following a change in THERMAL POWER or power j shape, regardless of magnitude, that could place operation at a thermal limit. 4 3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in l any rod is less than the design linear heat generation even if fuel pellet densification is postulated. ] l

References:

) 1. General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K. HEDE-20566, November 1975. l 2. " General Electric Standard Application for Reattor Fuel," NEDE-24011-P-A (latest approved revision). ) 3. " Basis of MAPLHGR Technical Specifications for Limerick Unit 1," J l P!ED0-31401, February 1987. j l 4.

Deleted, j

5. Increased Core Flow and Pe.rtial Feedwater Heating Analysis for Limerick Generating Station Unit 1 Cycle 1, NEDC-31323, October 1986 l including Errata and Addenda Shtet No. I dated November f>,1986. i 1 l I J LIMERICK - UNIT 1 B 3/4 2-5 Am&dment No. 7 j l 1}}

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