ML18059A943

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Proposed Tech Specs Relocating Cycle Specific Operating Limits to COLR
ML18059A943
Person / Time
Site: Palisades Entergy icon.png
Issue date: 04/07/1994
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
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ML18059A942 List:
References
NUDOCS 9404190244
Download: ML18059A943 (50)
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{{#Wiki_filter:ATTACHMENT 1 Consumers Power Company Pali sades Pl ant Docket 50-255 PROPOSED CORE OPERATING LIMITS REPORT TECHNICAL SPECIFICATIONS ,--- -~9404190244 940407 PDR ADOCK 05000255 p PDR PROPOSED PAGES April 7, 1994 20 Pages

PALISADES PLANT FACILITY OPERATING LICENSE DPR-20 APPENDIX A TECHNICAL SPECIFICATIONS As Amended Through Amendment No.

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION DESCRIPTION 1.0 DEFINITIONS 1.1 REACTOR OPERATING CONDITIONS 1.2 PROTECTIVE SYSTEMS 1.3 INSTRUMENTATION SURVEILLANCE 1.4 MISCELLANEOUS DEFINITIONS 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS PAGE NO 1-1 1-1 1-3 1-3 1-4 2-1 2.1 SAFETY LIMITS - REACTOR CORE 2-1 2.2 SAFETY LIMITS - PRIMARY COOLANT SYSTEM PRESSURE 2-1 2.3 LIMITING SAFETY SYSTEM SETTINGS - RPS 2-1 Table 2.3.1 Reactor Protective System Trip Setting Limits 2-2 B2.l Basis - Reactor Core Safety Limit B2.2 Basis - Primary Coolant System Safety Limit B2.3 Basis - Limiting Safety System Settings 3.0 LIMITING CONDITIONS FOR OPERATION 3.0 APPLICABILITY 3.1 3.2 3.3 3.4 3.5

3. I. I 3.1.2 Figure 3-1 Figure 3-2 Figure 3-3 3.1.3 3.1.4
3. I. 5 3.1.6 3.1. 7 3.1.8 PRIMARY COOLANT SYSTEM Operable Components Heatup and Cooldown Rates Pressure - Temperature Limits for Heatup Pressure - Temperature Limits for Cooldown Pressure - Temperature Limits for Hydro Minimum Conditions for Criticality Maximum Primary Coolant Radioactivity Primary Coolant System Leakage Limits Maximum PCS Oxygen and Halogen Concentration Primary and Secondary Safety Valves Overpressure Protection Systems CHEMICAL AND VOLUME CONTROL SYSTEM EMERGENCY CORE COOLING SYSTEM CONTAINMENT COOLING STEAM AND FEEDWATER SYSTEMS 3.6 CONTAINMENT SYSTEM Table 3.6.l Containment Penetrations and Valves 3.7 ELECTRICAL SYSTEMS B2-1 B2-2 B2-3 3-1 3-1 3-lb 3-lb 3-4 3-9 3-10 3-11 3-12 3-17 3-20 3-23 3-25 3-25a 3-26 3-29 3-34 3-38 3-40 3-40b 3-41 Amendment No.

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION DESCRIPTION 3.0 LIMITING CONDITIONS FOR OPERATION (continued} 3.8 REFUELING OPERATIONS 3.9 Deleted 3.10 3.10.l 3.10.2 3.10.3 3.10.4 3.10.5 3.10.6 3.10.7 CONTROL ROD LIMITS Shutdown Margin Requirements Deleted Part-Length Control Rods Misaligned or Inoperable Rod Regulating Group Insertion Limits Shutdown Rod Limits Low Power Physics Testing POWER DISTRIBUTION INSTRUMENTATION Incore Detectors PAGE NO 3-1 3-46 3-50 3-58 3-58 3-59 3-59 3-60 3-60 3-61 3-61 3.11 3.11.1 3.11.2 Excore Power Distribution Monitoring System 3-65 3-65 3-67 3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY 3-69 3.13 Deleted 3-69 3.14 CONTROL ROOM VENTILATION 3-70 3.15 REACTOR PRIMARY SHIELD COOLING SYSTEM 3-70a 3.16 ESF SYSTEM INITIATION INSTRUMENTATION SETTINGS 3-71 Table 3.16.1 ESF System Initiation Instrument Setting Limits 3-75 3.17 INSTRUMENTATION AND CONTROL SYSTEMS 3-76 Table 3.17.1 Instrument Requirements for RPS 3-78 Table 3.17.2 Instrument Requirements for ESF Systems 3-79 Table 3.17.3 Instrument Conditions for Isolation Functions 3-80 Table 3.17.4 Instrument Requirements for Other Safety Features 3-81 3.18 Deleted 3-82 3.19 IODINE REMOVAL SYSTEM 3-84 3.20 SHOCK SUPPRESSORS (Snubbers) 3-88 3.21 MOVEMENT HEAVY LOADS 3-92 3.22 Deleted 3-96 3.23 3.23.1 3.23.2 3.23.3 POWER DISTRIBUTION LIMITS Linear Heat Rate Radial Peaking Factors Quadrant Power Tilt - Tq 3.24 Deleted 3.25 ALTERNATE SHUTDOWN SYSTEM Table 3.25.1 Alternate Shutdown Minimum Equipment ii 3-103 3-103 3-111 3-112 3-113 3-134 3-135 Amendment No.

PALISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS SECTION DESCRIPTION PAGE NO 6.0 ADMINISTRATIVE CONTROLS (Continued} 6.6 Deleted 6-10 6.7 SAFETY LIMIT VIOLATION 6-10 6.8 PROCEDURES AND PROGRAMS 6-11 6.9 REPORTING REQUIREMENTS 6-14 6.9.1 Routine Reports 6-14 6.9.1.a Start-up Report 6-14 6.9.1.b Annual Report 6-14 6.9.1.c Monthly Operating Report 6-15 6.9.1.d Radioactive Effluent Release Report 6-15 6.9.1.e Radiological Environmental Operating Report 6-15 6.9.1.f Core Operating Limits Report 6-15 6.9.2 Reportable Events 6-17 6.9.3 Nonroutine Reports 6-17 6.9.4 Special Reports 6-26 6.10 RECORD RETENTION 6-26 6.11 RADIATION PROTECTION PROGRAM 6-28 6.12 HIGH RADIATION AREA 6-28 6.13 Deleted 6-33 6.14 Deleted 6-33 6.15 SYSTEMS INTEGRITY 6-33 6.16 IODINE MONITORING 6-33 6.17 POST ACCIDENT SAMPLING 6-34 6.18 OFFSITE DOSE CALCULATION MANUAL 6-35 6.19 PROCESS CONTROL PROGRAM 6-35 6.20 Deleted 6-36 6.21 SEALED SOURCE CONTAMINATION 6-37 6.22 SECONDARY WATER CHEMISTRY 6-38 v Amendment No.

1.4 MISCELLANEOUS DEFINITIONS {Continued) Safety Safety as used in these Technical Specifications refers to those safety issues related to the nuclear process and, for example, does not encompass OSHA considerations. Reportable Event. A reportable event shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. Core Operating Limits Report {COLR) The COLR is the document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 6.9.1.f. Plant operation within these limits is addressed in individual Specifications. 1-5 Amendment No. BS, t-28,

3.I PRIMARY COOLANT SYSTEM 3.I.I Operable Components (Continued)

f.

Nominal primary system operation pressure shall not exceed 2IOO psia.

g.

The reactor inlet temperature and ASI shall be maintained within the limits specified in the COLR. (I) When reactor inlet temperature exceeds the limits specified in the COLR, restore reactor inlet temperature to within limits within 30 minutes. (2) When the ASI exceeds the limits specified in the COLR, within I5 minutes initiate corrective actions to restore the ASI to the acceptable region. Restore the ASI to acceptable values within one hour or be at less than 70% of rated power within the following two hours.

h.

Forced circulation (starting the first primary coolant pump) shall not be initiated unless one of the following conditions is met: (I) Primary coolant cold leg temperature is> 430°F. (2) PCS cold leg temperature is ~ 430°F and S/G secondary temperature is less than PCS cold leg temperature. (3) Shutdown cooling is isolated from the PCS AND PCS cold leg temperature is> 2I0°F and S/G secondary temperature is less than I00°F higher than PCS temperature. (4) Shutdown cooling is isolated from the PCS AND PCS cold leg temperature is ~ I70°F and ~ 2I0°F AND S/G secondary temperature is less than 20°F higher than PCS cold leg temperature. (5) Shutdown cooling is isolated from the PCS AND PCS cold leg temperature is ~ I20°F and< I70°F AND S/G secondary temperature is less than I00°F higher than PCS cold leg temperature.

i.

The PCS shall not be heated or maintained above 325°F unless a minimum of 375 kW of pressurizer heater capacity is available from both buses ID and IE. Should heater capacity from either bus ID and IE fall below 375 kW, either restore the inoperable heaters to provide at least 375 kW of heater capacity from both buses ID and IE within 72 hours or be in hot shutdown within the next I2 hours. 3-lc Amendment No. l, .§.l, 8§., -+/--1-7-, H-8' 34' a+'

3.1 PRIMARY COOLANT SYSTEM 3.1.1 Operable Components {continued) Basis When primary coolant boron concentration is being changed, the process must be uniform throughout the primary coolant system volume to prevent stratification of primary coolant at lower boron concentration which could result in a reactivity insertion. Sufficient mixing of the primary coolant is assured if one shutdown cooling or one primary coolant pump is in operation.<1> The shutdown cooling pump will circulate the primary system volume in less than 60 minutes when operated at rated capacity. By imposing a minimum shutdown cooling pump flow rate, sufficient time is provided for the operator to terminate the boron dilution under asymmetric flow conditions.<5> The pressurizer volume is relatively inactive, therefore will tend to have a boron concentration higher than rest of the primary coolant system during a dilution operation. Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the pressurizer and the primary system during the addition of boron.<2> The FSAR safety analysis was performed assuming four primary coolant pumps were operating for accidents that occur during reactor operation. Therefore, reactor startup above hot shutdown is not permitted unless all four primary coolant pumps are operating. Operation with three primary coolant pumps is permitted for a limited time to allow the restart of a stopped pump or for reactor internals vibration monitoring and testing. Requiring the plant to be in hot shutdown with the reactor tripped from the C-06 panel, opening the 42-01 and 42-02 circuit breakers, assures an inadvertent rod bank withdrawal will not be initiated by the control room operator. Both steam generators are required to be operable whenever the temperature of the primary coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor. The transient analyses were performed assuming a vessel flow at hot zero power {532°F) of 140.7 x 106 lb/hr minus 63 to account for flow measurement uncertainty and core flow bypass. A DNB analysis was performed in a parametric fashion to determine the core inlet temperature as a function of pressure and flow for which the minimum DNBR is equal to the DNB correlation safety limit. This analysis includes the following uncertainties and allowances: 2% of rated power for power measurement; +/-0.06 for ASI measurement; +/-50 psi for pressurizer pressure; +/-7°F for inlet temperature; and 3% measurement and 3% bypass for core flow< 4>, In addition, transient biases were included in the determination of the allowable reactor inlet temperature specified in the COLR. 3-2 Amendment No. fH-, 8§., -l-1-7, ~' 3-1-,

3.1 PRIMARY COOLANT SYSTEM Basis {continued) The Axial Shape Index alarm channel is being used to monitor the ASI to ensure that the assumed axial power profiles used in the development of the inlet temperature LCO bound measured axial power profiles. The signal representing core power (Q) is the auctioneered higher of the neutron flux power and the Delta-T power. The measured ASI calculated from the excore detector signals and adjusted for shape annealing (Y1) and the core power constitute an ordered pair (Q,Y1). An alarm signal is activated before the ordered pair exceed the boundaries specified in the COLR. The requirement that the steam *generator temperature be ~ the PCS temperature when forced circulation is initiated in the PCS ensures that an energy addition caused by heat transferred from the secondary system to the PCS will not occur. This requirement applies only to the initiation of forced circulation {the start of the first primary coolant pump) when the PCS cold leg temperature is< 430°F. However, analysis (Reference 6) shows that under limited conditions when the Shutdown Cooling System is isolated from the PCS, forced circulation may be initiated when the steam generator temperature is higher than the PCS cold leg temperature. References (1) Updated FSAR, Section 14.3.2. (2) Updated FSAR, Section 4.3.7. (*3) De 1 eted (4) EMF-92-178, Revision 3, Section 15.0.7.1 (5) ANF-90-078 (6) Consumers Power Company Engineering Analysis EA-A-NL-89-14-1 3-3 Amendment No. 3-1-, .§.l., -H-7, -H-8, !3-l, 34' 7' -l4a-' -1.§.6' §9'

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS 3.10.4 3.10.5 Misaligned or Inoperable Control Rod or Part-Length Rod

a.

A control rod or a part-length rod is considered misaligned if it is out of position from the remainder of the bank by more than 8 inches.

b.

A control rod is considered inoperable if it cannot be moved by its operator or if it cannot be tripped. A part-length rod is considered inoperable if it is not fully withdrawn from the core and cannot be moved by its operator. If more than one control rod or part-length rod becomes misaligned or inoperable, the reactor shall be placed in the hot shutdown condition within 12 hours.

c.

If a control rod or a part-length rod is misaligned, hot channel factors must promptly be shown to be within design limits or reactor power shall be reduced to 75% or less of rated power within two hours. In addition, shutdown margin and individual rod worth limits must be met. Individual rod worth calculations will consider the effects of xenon redistribution and reduced fuel burnup in the region of the misaligned control rod or part-length rod. Regulating Group Insertion Limits

a.

The regulating groups shall be limited to the withdrawal sequence, overlap, and insertion limits specified in the COLR.

b.

With any regulating group inserted beyond its limit,

1.

Restore all regulating groups to within insertion limit within 2 hours. 3-60 Amendment No. 3-1-,

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS 3.10.6 3.10.7 Shutdown Rod Limits

a.

All shutdown rods shall be withdrawn before any regulating rods are withdrawn.

b.

The shutdown rods shall not be withdrawn until normal water level is established in the pressurizer.

c.

The shutdown rods shall not be inserted below their exercise limit until all regulating rods are inserted. Low Power Physics Testing Sections 3.10.1.a, 3.10.1.b, 3.10.3, 3.10.4.b, 3.10.5 and 3.10.6 may be deviated from during low power physics testing and CROM exercises if necessary to perform a test but only for the time necessary to perform the test. Sufficient control rods shall be withdrawn at all times to assure that the reactivity decrease from a reactor trip provides adequate shutdown margin. The available worth of withdrawn rods must include the reactivity defect of power and the failure of the withdrawn rod of highest worth to insert. The requirement for a shutdown margin of 2.0% in reactivity with 4-pump operation, and of 3.75% in reactivity with less than 4-pump operation, is consistent with the assumptions used in the analysis of accident conditions (including steam line break) as reported in Reference 1 and additional analysis. Requiring the boron concentration to be at cold shutdown boron concentration at less than hot shutdown assures adequate shutdown margin exists to ensure a return to power does not occur if an unanticipated cooldown accident occurs. This requirement applies to normal operating situations and not during emergency conditions where it is necessary to perform operations to mitigate the consequences of an accident. By imposing a minimum shutdown cooling pump flow rate of 2810 gpm, sufficient time is provided for the operator to terminate a boron dilution under asymmetric conditions. For operation with no primary coolant pumps operating and a recirculating flow rate less than 2810 gpm the increased shutdown margin and controls on charging pump operability or alternately the surveillance of the charging pumps will ensure that the acceptance criteria, for an inadvertent boron dilution event will not be violated.< 1> The change in insertion limit with reactor power insures that the shutdown 3-61 Amendment No. 3-l, M,.§.7., 68, H-8, 3-7,

3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS 3-62 Amendment No 3+, -l-l-8,

3.11.2 POWER DISTRIBUTION INSTRUMENTATION EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION The excore monitoring system shall be operable with:

a.

The target Axial Offset (AO) and the Excore Monitoring Allowable Power Level (APL) determined within the previous 31 days using the incore detectors, and the measured AO not deviated from the target AO by more than 0.05 in the previous 24 hours. The APL shall be determined as specified in the COLR.

b.

The AO measured by the excore detectors calibrated with the AO measured by the incore detectors.

c.

The quadrant tilt measured by the excore detectors calibrated with the quadrant tilt measured by the incore detectors. APPLICABILITY: (1) Items a, b, and c above are applicable when the excore detectors are used for monitoring LHR. (2) Item c above is applicable when the excore detectors are used for monitoring quadrant tilt. (3) Item b above is applicable for each channel of the TM/LP trip and the Axial Shape Index (ASI) alarm. ACTION 1: With the excore monitoring system inoperable, do not use the system for monitoring LHR. ACTION 2: If the measured quadrant tilt has not been calibrated with the incores, do not use the system for monitoring quadrant tilt. ACTION 3: When the measured AO uncertainty is greater than specified in Specification 4.18.2, the TM/LP trip function and the ASI alarm setpoints shall be conservatively adjusted within twelve (12) hours or that channel shall be declared inoperable. The operability requirements for TM/LP and ASI are given in Table 3.17.1 and 3.17.4, respectively. 3-67 Amendment No. 43,.sG,.§8, 68, 1-8,

3.11.2 POWER DISTRIBUTION INSTRUMENTATION EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION The excore power distribution monitoring system consists of Power Range Detector Channels 5 through 8. The operability of the excore monitoring system ensures that the assumptions employed in the PDC-11 analysis< 1> for determining AO limits that ensure operation within allowable LHR limits are valid. Surveillance requirements ensure that the instruments are calibrated to agree with the incore measurements and that the target AO is based on the current operating conditions. Updating the Excore Monitoring APL ensures that the core LHR limits are protected within the +/-0.05 band on AO. The APL considers LOCA based LHR limits, and factors are included to account for changes in radial power shape and LHR limits over the calibration interval. References (1) XN-NF-80-47 3-68 Amendment No..§8, 68, -l-l-8, !43-,

3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY Applicability Applies to the moderator temperature coefficient of reactivity for the core. Objective To specify a limit for the positive moderator coefficient. Specifications The moderator temperature coefficient (MTC) shall be less positive than +0.5 x 10-4 Ap/° F at ~ 2% of rated power. The limitations on moderator temperature coefficient (MTC) are provided to ensure that the assumptions used in the safety analysis< 1> remain valid. Reference (1) EMF-92-178, Revision 3, Section 15.0.5 3.13 Deleted 3-69 Amendment No..f-18., 3-7-, -143-, ~' &9,

3.23 POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE CLHR} LIMITING CONDITION FOR OPERATION The LHR in the peak power fuel rod at the peak power elevation Z shall be maintained within the limits specified in the COLR. APPLICABILITY: Power operation above 50% of rated power. ACTION 1: When using the incore alarm system to monitor LHR, and with four or more coincident incore alarms, initiate within 15 minutes corrective action to reduce the LHR to within the limits and restore the incore readings to less than the alarm setpoints within 1 hour or failing this, be at less than 50% of rated power within the following 2 hours. ACTION 2: When using the excore monitoring system to monitor LHR and with the AO deviating from the target AO by more than 0.05, discontinue using the excore monitoring system for monitoring LHR. If the incore alarm system is inoperable, within 2 hours be at 85% (or less) of rated thermal power and follow the procedure in ACTION 3 below. ACTION 3: If the incore alarm system is inoperable and the excore monitoring system is not being used to monitor LHR, operation at less than or equal to 85% of rated power may continue provided that incore readings are recorded manually. Readings shall be taken on a minimum of 10 individual detectors per quadrant (to include a total number of 160 detectors in a IO-hour period) within 4 hours and at least every 2 hours thereafter. If readings indicate a local power level equal to or greater than the alarm setpoints, the action specified in ACTION I above shall be taken. 3-103 Amendment No. 6-8, -l-18,

I

  • POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE (LHR)

The limitation of LHR ensures that, in the event of a LOCA, the peak temperature of the cladding will not exceed 2200°F.< 1> Either of the two core power distribution monitoring systems {the incore alarm system or the excore monitoring system) provides adequate monitoring of the core power distribution and is capable of verifying that the LHR does not exceed its limits. The incore alarm system performs this function by continuously monitoring the local power at many points throughout the core and comparing the measurements to predetermined setpoints above which the limit on LHR could be exceeded. The excore monitoring system performs this function by providing comparison of the measured core AO with predetermined AO limits based on incore measurements. An Excore Monitoring Allowable Power Level {APL}, which may be less than rated power, is applied when using the excore monitoring system to ensure that the AO limits adequately restrict the LHR to less than the limiting values.<4> If the incore alarm system and the excore monitoring system are both inoperable, power will be reduced to provide margin between the actual peak LHR and the LHR limits and the incore readings will be manually collected at the terminal blocks in the control room utilizing a suitable signal detector. If this is not feasible with the manpower available, the reactor power will be reduced to a point below which it is improbable that the LHR limits could be exceeded. The time interval of 2 hours and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the core power distribution to detect significant changes until the monitoring systems are returned to service. To ensure that the design margin of safety is maintained, the determination of both the incore alarm setpoints and the APL takes into account the local LHR measurement uncertainty factors< 5> specified in the COLR. References (1) EMF-91-77 (2) (Deleted) (3) (Deleted) (4) XN-NF-80-47 (5) FSAR Section 3.3.2.5 {next page is 3-111) 3-104 Amendment No. 68, 82-, HS, 7, -l-43-, -144,

POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION The radial peaking factors F~, and F~ shall be maintained within the limits specified in the COLR. APPLICABILITY: Power operation above 25% of rated power. ACTION:

1.

For P < 50% of rated with any radial peaking factor exceeding its limit, be in at least hot shutdown within 6 hours.

2.

For P ~ 50% of rated with any radial peaking factor exceeding its limit, reduce thermal power within 6 hours to less than the lowest value of: Basis [1 F - 3.33( r - 1) ] x Rated Power F L Where Fr is the measured value of either F~ or F~, and FL is the corresponding limit specified in the COLR. The limitations on F~, and F~ are provided to ensure that assumptions used in the analysis for establishiny DNB margin, LHR and the thermal margin/low-pressure and variable high-power trip set points remain valid during operation. Data from the incore detectors are used for determining the measured radial peaking factors. The periodic surveillance requirements for determining the measured radial peaking factors provide assurance that they remain within prescribed limits. Determining the measured radial peaking factors after each fuel loading prior to exceeding 50% of rated power provides additional assurance that the core is properly loaded. To ensure that the design margin of safety is maintained, the determination of radial peaking factors takes into account the appropriate measurement uncertainty factors< 1> specified in the COLR. Reference (1) FSAR Section 3.3.2.5 ~ 3-111 Amendment No. 68, -l-18, -l3-7, +/-43-, +/-44, 66,

ADMINISTRATIVE CONTROLS 6.9.1 Routine Reports (continued)

c.

Monthly Operating Report Routine reports of operating statistics and shutdown experience shall be submitted on a monthly basis to the NRC to arrive no later than the fifteenth of each month following the calendar month covered by the report.

d.

Radioactive Effluent Release Report The Radioactive Effluent Release Report shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PROCESS CONTROL PROGRAM and (2) in conformance with 10 CFR 50.36a and Section IV.B.l of Appendix I to 10 CFR 50.

e.

Radiological Environmental Operating Report The Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (1) the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to 10 CFR 50.

f.

Core Operating Limits Report (COLR) Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: 3.1.1 3.10.5 3.11.2 3.23.1 3.23.2 Tinle Function and ASI Limits. Reguiating Group Insertion Limits Excore Monitoring Allowable Power Level Linear Heat Rate (LHR) Limits Radial Peaking Factor Limits 6-15 Amendment No. -+/--&, ~' 36, && ' l-G&' -l-§4'

ADMINISTRATIVE CONTROLS ~ 6.9.1 Routine Reports

f.

COLR (continued) The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the latest approved revision of the following documents:

1.

XN-75-27(A) and Supplements, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company.

2.

ANF-84-73(P)(A), Appendix Band Supplements, "Advanced Nuclear Fuels Corporation Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation.

3.
  • XN-NF-82-2l(P)(A), "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company.
4.

ANF-84-093(P)(A) and Supplement, "Steamline Break Methodology for PWRs," Advanced Nuclear Fuels Corporation.

5.

XN-75-32(P)(A), Supplements, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company.

6.

EXEM PWR Large Break LOCA Model as defined by: a) XN-NF-82-20(A), and Supplements, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company. b) XN-NF-82-07(P)(A), "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company. c) XN-NF-81-SB(A) and Supplements, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company. d) XN-NF-85-16(A), Volume 1 and Supplements, and Volume 2 and Supplement, "PWR 17x17 Fuel Cooling Tests Program," Exxon Nuclear Company. e) XN-NF-85-44(A), and Supplement, "Scaling of FCTF Based Reflood Heat Transfer Correlation for other Bundle Designs," Exxon Nuclear Company. 6-16 Amendment No.

ADMINISTRATIVE CONTROLS 6.9.1 Routine Reports

f.

COLR (continued)

7.

XN-NF-78-44(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company.

8.

ANF-1224(P)(A) and Supplement, "Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Advanced Nuclear Fuels Corporation.

9.

ANF-89-192(P), "Justification of the ANFP DNB Correlation for High Thermal Performance Fuel in the Palisades reactor," Advanced Nuclear Fuels Corporation.

10.

ANF-89-15l(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation.

11.

EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met. The COLR, including any mid cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC. 6.9.2 Reportable Events The Commission shall be notified of Reportable Events and a report submitted pursuant to the requirements of 10 CFR 50.73. 6.9.3 Nonroutine Reports A report shall be submitted in the event that (a) the Radiological Environmental Monitoring Programs are not substantially conducted as described in the ODCM or (b) an unusual or important event occurs from plant operation that causes a significant environmental impact or affects a potential environmental impact. Reports shall be submitted within 30 days. (Next page is 6-26) 6-17 Amendment No.

ATTACHMENT 2 Consumers Power Company Pali sades Pl ant Docket 50-255 PROPOSED CORE OPERATING LIMITS REPORT TECHNICAL SPECIFICATIONS EXISTING PAGES WITH PROPOSED CHANGES MARKED April 7, 1994 28 Pages

. 1.4 MISCELLANEOUS DEFINITIONS (Continued) Safety Safety as used in these Technical Specifications refers to those safety issues related to the nuclear process and, for example, does not encompass OSHA considerations. Reportable Event A reportable event shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50. A\\;) 1/ BE. F' "' I Tl b "' FD rZ. S:.e 6!:- I,./ S If re:r -::#= 1 1-5 Amendment No. $~, tza-eptember 5, 19S9

INSERT #1 Page 1-5 Core Operating Limits Report (COLR) The COLR is the document that provides cycle specific parameter limits for the current reload cycle. These cycle specific parameter limits shall be determined for each reload cycle in accordance with Specification 6.9.1.f. Plant operation within these limits is addressed in individual Specifications.

TC> CC>l-R 3.1 PRIMARY COOLANT SYSTEM (Continued) 3.1.1 Operable Components (Continued)

f.
g.

Nominal primary psi a. shall not exceed 2100 = reactor i e ure = nominal ing p recirculating mass o 06 lb/h corrected to the operating temperatur conditions. 7He. Col..{( tRe (-z) When the AS! exceeds the limits specified in Fi!JtH'e J,Q, within 15 minutes, initiate corrective actions to restore the AS! to the acceptable region. Restore the AS! to acceptable values within one hour or be at less than 70% of rated power within the following two hours. 3-lc Amendment No. pJ, ~J,$~, JJl, JJ~, Jp~,~ Fel:U"tta1 y 20, ~

INSERT #2 Page 3-lc (1) When reactor inlet temperature exceeds the limits specified in the COLR, restore reactor inlet temperature to within limits within 30 minutes.

n"l(.:r MD"/!:.-'!:> Tb f~'llZ'.. "3-lc ~,,~ ~XT Moue.t:> 1() r~~IZ:,. '!-'2.. 3.I PRIMARY COOLANT SYSTEM (Cont'd)

3. I. I Operable Components (cont'd)
h.

Forced circulation (starting the first primary coolant pump) shall not be initiated unless one of the following conditions is met: (I) (2) (3) (4) (5) Primary coolant cold leg temperature is > 430°F. PCS cold leg temperature is ~ 430°F and S/G secondary temperature is less than PCS cold leg temperature. Shutdown cooling is isolated from the PCS AND PCS cold leg temperature is > 2I0°F and S/G secondary temperature is less than I00°F higher than PCS temperature. Shutdown cooling is isolated from the PCS AND PCS cold leg temperature is ~ I70°F and ~ 2I0°F AND S/G secondary temperature is less than 20°F higher than PCS cold leg temperature. Shutdown cooling is isolated from the PCS AND PCS cold leg temperature is ~ I20°F and < I70°F AND S/G secondary temperature is less than I00°F higher than PCS cold leg temperature.

i. The PCS shall not be heated or maintained above 325"F unless a minimum of 375 kW of pressurizer heater capacity is available from both buses ID and IE.

Should heater capacity from either bus ID and IE fall below 375 kW, either restore the inoperable heaters to provide at least 375 kW of heater capacity from both buses ID and IE within 72 hours or be in hot shutdown within the next I2 hours. Basis When primary coolant boron concentration is being changed, the process must be uniform throughout the primary coolant system volume to prevent stratification of primary coolant at lower boron concentration which could result in a reactivity insertion. Sufficient mixing of the primary coolant is assured if one shutdown cooling or one primary cool ant pump is in operation. '11 The shutdown cooling pump will circulate the primary system volume in less than 60 minutes when operated at rated capacity. By imposing a minimum shutdown cooling pump flow rate ef 281Q §~~, sufficient time is provided for the operator to terminate the boron dilution under asymmetric flow conditions. 151 The pressurizer volume is relatively inactive, therefore will tend to have a boron concentration higher than rest of the primary coolant system during a dilution operation. Administrative procedures will provide for use of pressurizer sprays to maintain a nominal spread between the boron concentration in the pressurizer and the primary s ste during the addition of boron.'~ 3-Id Amendment No %7, ~~' JJl, JJ~,~ A1:iril 26, 1990-

3.1 PRIMARY COOLANT SYSTEM (contd) Basis (Cont'd) The FSAR safety analysis was performed assuming four primary coolant pumps were operating for accidents that occur during reactor operation. Therefore, reactor startup above hot shutdown is not permitted unless all four primary coolant pumps are operating. Operation with three primary coolant pumps is permitted for a limited time to allow the restart of a stopped pump or for reactor internals vibration monitoring and testing. Requiring the plant to be in hot shutdown with the reactor tripped from the C-06 panel, opening the 42-01 and 42-02 circuit breakers, assures an inadvertent rod bank withdrawal will not be initiated by the control room operator. Both steam generators are required to be operable whenever the temperature of the primary coolant is greater than the design temperature of the shutdown cooling system to assure a redundant heat removal system for the reactor. The transient analyses were performed assuming a vessel flow at hot zero power {532°F) of 140.7 x 106 lb/hr minus 6% to account for flow f measurement uncertainty and core flow bypass. A DNB analysis was performed in a parametric fashion to determine the core inlet temperature as a function of pressure and flow for which the minimum DNBR is equal to the DNB correlation safety limit. This analysis ~ includes the following uncertainties and allowances: 2% of rated power for power 3-2 Amendment No.~~' ~J, JJ~, J~J, J~~'~ February 20, 1991-

3.1 PRIMARY COOLANT SYSTEM (Cont'd} Basis (Cont'd) (E,&_ \\) 1'.. 1"1 o,._( ,4-N 1/ llf..xl {Y1 o u/f.. '> -~o C6Lf: measurement; +/-0.06 for ASI measurement; +/-50 psi for pressurizer pressure; +/-7°f )for inlet temperature; and 3% measurement and 3% bypass~ for core flow<:~ In addition, transient biases were included in the deri vati eR ef the fell ewi r:ig eq1:1ati er:i fer l imiti r:ig reactor inlet,.,, If."'" " te~atl:1re: 'bl£"T'lt, CMI ",..,.IC!.( ~,:; A-l.t.ot..l!Q.~l..tz. te~c .. :~"- INLf:."i 7l1&Mr~e:' \\I S~ ~t~tlf;..1:. Ir.I "THll!. COl-f. < 542.99 +.0580(P-2060) + O.OOOOl(P-2060)**2 + 1.125(W-138) - 5(W-138)**2 With measure mary coolant system flow rates > 150 r+-lbm/hr, limiting the ma

  • a 11 owed in 1 et temperature to the T1nret LCO at ~

150 M lbm/hr increases the margin to DNB for higher PCS flow rates 141

  • The Axial Shape Index alarm channel is being used to monitor the ASI to ensure that the assumed axial power profiles used in the development of the inlet temperature LCO bound measured axial power profiles. The signal representing core power (Q) is the auctioneered higher of the neutron flux power and the Delta-T power. The measured ASI calculated from the excore detector signals and adjusted for shape annealing (Y1) and the core power constitute an ordered pair (Q,Y1).

An alarm signal is activated before the ordered pair exceed the boundaries specified in H -Fi g1:1r~ 3.:"(). TMIZ:- CO\\.-\\Z. f The requirement that the steam generator temperature be ~ the PCS temperature when forced circulation is initiated in the PCS ensures that an energy addition caused by heat transferred from the secondary system to the PCS will not occur. This requirement applies only to the initiation of forced circulation (the start of the first primary coolant pump) when the PCS cold leg temperature is < 430°F. However, analysis (Reference 6) shows that under limited conditions when the Shutdown Cooling System is isolated from the PCS, forced circulation may be initiated when the steam generator temperature is higher than the PCS cold leg temperature. References (1) Updated FSAR, Section 14.3.2. (2) Updated FSAR, Section 4.3.7. (3) Deleted 3 (4) EMF-92-178, Revision/, Section 15.0.7.1 (5) ANF-90-078 (6) Consumers Power Company Engineering Analysis EA-A-NL-89-14-1 3-3 Amendment No. 3-1-, l, H-7, HS, -l-3-l-, 34' B-7-' -l-43-, ~' '1 ~

3.1 PRIMARY COOLANT SYSTEM (continued) ASI Limit for Tinlet function L Q) ~ 0 0.9

a.

\\J 4'

o. e

~ ~ It-o 0.7 c Q.6 0 ~ u o.s <<] L u.. 0.41 Unacceptable Operations Break Points: -0.550, 0.250 -0.300, o. 700 _- -0.080, 1.00 -0.0fJC>}d.O 5*.. +O. 400}- l 65; --* +o.41Q~:;~.2so>

  • Acceptable Operations I

I / F1 G \\j <<.I~. tA D_ u,~I), \\U CO-I.-~ FIGURE 3-0 ?ft- ~If_ ~ ~ 3~ f.L. l" N\\ 1 II il-~ b \\, ,/

I __

  • 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Contd) 3.10.4 Misaligned or Inoperable Control Rod or Part-Length Rod
a.

A control rod or a part-length rod is considered misaligned if it is out of position from the remainder of the bank by more than 8 inches.

b.

A control rod is considered inoperable if it cannot be moved by its operator or if it cannot be tripped. A part-length rod is considered inoperable if it is not fully withdrawn from the core and cannot be moved by its operator. If more than one control rod or part-length rod becomes misaligned or inoperable, the reactor shall be placed in the hot shutdown condition within 12 hours.

c.

If a control rod or a part-length rod is misaligned, hot channel factors must promptly be shown to be within design limits or reactor power shall be reduced to 75% or less of rated power within two hours. In addition, shutdown margin and individual rod worth limits must be met. Individual rod worth calculations will consider the effects of xenon redistribution and reduced fuel burnup in the region of the misaligned control rod or part-length rod. 3.10.5 rvf Oltlt.. '{.,.. I M IT$ Regulating Group Insertion Limits Cl'.)L-~ f(ef'l.P.ot... A-c.:11~ ((Jr.LL 'i3 "'r I,_ !"' S'~t-.\\T,._,.Ja.t.. or- ?~iti4 IJ.-. II j WI i i.f {LG T"l6"-f

a.
b.

To implement the limits on shutdown margin, individual rod orth and hot channel factors, the limits on control rod re lating group insertion shall be established as shown on Figu 3-6. The 4-pump operation limits of Figure 3-6 o not app for decreasing power level rapidly when s a decrease needed to avoid or minimize a situat* harmful to the plant ersonnel or equipment. Once sue a power decrease is ac 'eved, the limits of Figure 6 will be returned to by bo ting the control rods ove the insertion limit within two ho Limits more r trictive than Figure 3-6 may be implemente uring fuel cle life based on physics calculations and ysics ata obtained during plant start-up and subsequent ope t' n. New limits shall be submitted to the NRC within ays. The sequence of withdr al of the r ulating groups shall be 1, 2, 3, 4.

d.

If the eactor is subcritical, the rod position at hich cri 'cality could be achieved if the control rods wer

  • hdrawn in normal sequence shall not be lower than the insertion limit for zero power shown on Figure 3-6.

3-60 Amendment No "3i- ~November lT 1977~

  • Corrected-- Sep:£-cmbe-r 9, H}93

INSERT #3 Page 3-60

a.

The regulating groups sha 11 be 1 i mi ted to the wi thdrawa 1 sequence, overlap, and insertion limits specified in the COLR.

b.

With any regulating group inserted beyond its limit,

1.

Restore all regulating groups to within insertion limit within 2 hours.

a.

The regulating groups shall be 1 imited to the withdrawal

  • sequence, overlap, and insertion limits specified in the COLR.
b.

With any regulating group inserted beyond its limit,

1.

Restore all regulating groups to within insertion limit within 2 hours.

, 3.10 CONTROL ROD AND POWER DISTRIBUTION LIMITS (Contd) 3.10.6 Shutdown Rod Limits

a.

All shutdown rods shall be withdrawn before any regulating rods are withdrawn.

b.

The shutdown rods shall not be withdrawn until normal water level is established in the pressurizer.

c.

The shutdown rods shall not be inserted below their exercise limit until all regulating rods are inserted. 3.10.7 Low Power Physics Testing T;.?-T fbx~te,,,, ot'1 Nb Lo J 6 !J=-fC. ((.~ote~~* Sections 3.10.1.a, 3.10.1.b, 3.10.3, 3.10.4.b, 3.10.5 and 3.10.6 may be deviated from during low power physics testing and CROM exercises if necessary to perform a test but only for the time necessary to perform the test. Basis Sufficient control rods shall be withdrawn at all times to assure that the reactivity decrease from a reactor trip provides adequate shutdown margin. The available worth of withdrawn rods must include the reactivity defect of power and the failure of the withdrawn rod of highest worth to insert. The requirement for a shutdown margin of 2.0% in reactivity with 4-pump operation, and of 3.75% in reactivity with less than 4-pump operation, is consistent with the assumptions used in the analysis of accident conditions (including steam line break) as reported in Reference 1 and additional analysis. Requiring the ~ boron concentration to be at cold shutdown boron concentration at ~ less than hot shutdown assures adequate shutdown margin exists to ensure a return to power does not occur if an unanticipated cooldown accident occurs. This requirement applies to normal operating situations and not during emergency conditions where it is necessary to perform operations to mitigate the consequences of an accident. By imposing a minimum shutdown cooling pump flow rate of 2810 gpm, sufficient time is provided for the operator to terminate a boron dilution under asymmetric conditions. For operation with no primary coolant pumps operating and a recirculating flow rate less than 2810 gpm the increased shutdown margin and controls on charging pump operability or alternately the surveillance of the char~ing pumps will ensure that the acceptance criter,ia, for an inadvertent boron dilution event will not be violated. 11 The change in insertion limit with reactor I power s~ewR eR F"i§1:tl"e 3 6 insures that the shutdown i 3-61 Amendment No. ~J, %~, %7, ~~' JJ~, t37- ~9n1ary 20, 1991 -

~ ! [ i I (. i I I': : ( I

L POWER DISTRIBUTION INSTRUMENTATION 3.11.2 EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION The excore monitoring system shall be operable with:

a.
b.
c.

The target Axial Offset {AO) and the Excore Monitoring Allowable Power Level (APL) determined within the grevious 31 days using the incore detectors, and the measured A not deviated from the..,.1 target AO by more than O. 05 in the previous 24 hours. 11-uf.- A' i.- SM!i-Lt.. 'lilt-T.:>11=-i'lt"-""'""ei:i. A--s.. sPee.1F1e:c '"' -n-1E:. CoLe. The AO measured by the excore detectors calibrated with the AO measured by the incore detectors. The quadrant tilt measured by the excore detectors calibrated with the quadrant tilt measured by the incore detectors. APPLICABILITY: (1) Items a., b. and c. above are applicable when the excore detectors are used for monitoring LHR. (2) Item c. above is applicable when the excore detectors are used for monitoring quadrant tilt. (3) Item b., above is applicable for each channel of the TM/LP trip and the Axial Shape Index (ASI) alarm. ACTION 1: With the excore monitoring system inoperable, do not use the system for monitoring LHR. ACTION 2: If the measured quadrant tilt has not been calibrated with the incores, do not use the system for monitoring quadrant tilt. ACTION 3: When the measured AO uncertainty is greater than specified in Specification 4.18.2, the TM/LP trip function and the ASI alarm setpoints shall be conservatively adjusted within twelve {12) hours or that channel shall be declared inoperable. The operability requirements for TM/LP and ASI are given in Table 3.17.1 and 3.17.4, respectively. Basis The excore power distribution monitoring system consists of Power Range Detector Channels 5 through 8. The operability of the excore monitoring SYn}em ensures that the assumptions employed in the PDC-II analysis for determining AO limits that ensure operation within allowable LHR limits are valid. ~ Amendment No ~~o~~~, 1~~,'l~ MO\\Jf:.- /1-Ll Tie.')(\\ n PA-cG!E:... 3-b 1, ort s-C.8 f?1J-~ ~ °3-b~A... /E;I... l/\\AINh-Tft.'1":>,

POWER DISTRIBUTION INSTRUMENTATION 3.11.2 EXCORE POWER DISTRIBUTION MONITORING SYSTEM LIMITING CONDITION FOR OPERATION l'A6\\).it- "6'A? I~ (lf:'l(I Tu j?,A&~ ~-b8' Mo\\JI!. Afl-1!:. &. \\,) p. 1\\0,J (1-N T> ~icl Tc coi..f( Basis (Contd) Surveillance requirements ensure that the instruments are calibrated to agree with the incore measurements and that the target AO is based on the current operating conditions. Updating the Excore Monitoring APL ensures that the core LHR limits are protected within the +/-0.05 band on AO. The APL considers LOCA based LHR limits, and factors are included to account for changes in radial power shape and LHR limits over the calibration interval. is determined from the following: LHR(Zhs

] x Rated Power121 HR(Z)M~ x V(Z) x 1.02 M~

Where: (1) LHR(Z)Ts is the lim "ng LHR vs Section 3.23.1), (2) (3) V(Z) is the functi 1.02 of upburn, The ntity in brackets is the minimum value for the ire core any elevation (excluding the top and bottom 103 of core consiaering limits for peak rods. If the quantity in brae greater than one, the APL shall be the rated power level. References (1) XN-NF-80-47 -{2) ~MF 91-177 7~ (oft;_ -:s-to~ 6 fL-f Ml NA-Tt..!) Amendment No. ~~' ~~t 11~, -+4 -MafcH 2~, 1992. ""Co 1 *, ected * ~RO>Ct f'~ 3 66d1--

~ 1.12 ~ ~* ?-.. 1.10 (0.1.11) -i ~ I Gr 6' 6"\\ 1.08 n_ () D N - () Cf" ~ 1.06 I' r c /Yi *H ~ - I.Oil ~ 2-. ~ ~ 1.02 1.00 0.2 0.3 O.lt 0.6 0.6 0.7 0.8 FRACTION Of ACTIVE FUEL HEIGHT

  • Palisades T echnlcal Specifications

~~~~~~~~~~~~~-L~~~~~~~~~--~~~-:---~~* AXIAL VARIATION BOUNDING CONDITION

3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY Applicability Applies to the moderator temperature coefficient of*reactivity for the core. Objective To specify a limit for the positive moderator coefficient. Specifications The moderator temperatu~e coefficient (MTC) shall be less positive than +0.5 x 10-A.p/°F at :.:;:; 2% of rated power. Bases The limitations on moderator temperature coefficient (MTC) are provi ded11}o ensure that the assumptions used in the safety analysis remain valid. Reference 3 (1) EMF-92-178, Revision./, Section 15.0.5 7p..6r?:- R' ~"l v /'!'\\rs e n:...

===~:;'> 1-b ~ 3-67 Amendment No. -l-l-8, -l-3-7-1 -143, 56, 1"5"§- S-eptembe1 ~l, 1993-inext page is 3-69}

3.23 POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE CLHRl LIMITING CONDITION FOR OPERATION The LHR in the peak power fuel rod at the peak power elevation Z shall Aet exeee~ the valije iA Tahle 3.23 1 times FA(Z) [the fijAetieA FA(Z) I is she"A iA Fi~ijl"e 3.23 1]. ~It. t"l'\\~1a.1~~1i..112.-"':;) wl'rJ..1*" If.I~ A.tM1-r-s. s f'o<e c. I F, '£:,. T::> IN..,,..,,z... COL. R. APPLICABILITY: Power operation above 50% of rated power. ACTION 1: When using the incore alarm system to monitor LHR, and with four or more coincident incore alarms, initiate within 15 minutes corrective action to reduce the LHR to within the limits and restore the incore readings to less than the alarm setpoints within 1 hour or failing this, be at less than 50% of rated power within the following 2 hours. ACTION 2: When using the excore monitoring system to monitor LHR and with the AO deviating from the target AO by more than 0.05, discontinue using the excore monitoring system for monitoring LHR. If the incore alarm system is inoperable, within 2 hours be at 85% (or less} of rated thermal power and follow the procedure in ACTION 3 below. 3-103 Amendment No. ~~' "tt&- -Nevember 1§, 1988,

POWER DISTRIBUTION LIMITS 3.23.l LINEAR HEAT RATE (LHR) LIMITING CONDITION FOR OPERATION N~\\t it:.

4<
.Tlo..t 3 Th '(A-' pt_,

~ -10;:, ACTION 3: If the incore alarm system is inoperable and the excore monitoring system is not being used to monitor LHR, operation at less than or equal to 85% of rated power may continue provided that incore readings are recorded manually. Readings shall be taken on a minimum of 10 individual detectors per quadrant (to include a total number of 160 detectors in a IO-hour period) within 4 hours and at least every 2 hours thereafter. If readings indicate a local power level equal to or greater than the alarm setpoints, the action specified in ACTION I ~ above shall be taken. Basis The limitation of LHR ensures that, in the event of a LOCA, the peak temperature of the cladding wi 11 not exceed 2200° F. 111 Either of the two core power distribution monitoring systems (the incore alarm system or the excore monitoring system) provides adequate monitoring of the core power distribution and is capable of verifying that the LHR does not exceed its limits. The incore alarm system performs this function by continuously monitoring the local power at many points throughout the core and comparing the measurements to predetermined setpoints above which the limit on LHR could be exceeded. The excore monitoring system performs this function by providing comparison of the measured core AO with predetermined AO limits based on incore measurements. An Excore Monitoring Allowable Power Level (APL}, which may be less than rated power, is applied when using the excore monitoring system to ensure that the AO limits adequately restrict the LHR to less than the limiting values. 1~ If the incore alarm system and the excore monitoring system are both inoperable, power will be reduced to provide margin between the actual peak LHR and the LHR limits and the incore readings will be manually collected at the terminal blocks in the control room utilizing a suitable signal detector. If this is not feasible with the manpower available, the reactor power will be reduced to a point below which it is improbable that the LHR limits could be exceeded. 3-104 Amendment %~, ~i, 11~, -tztrf-- April 3, 1992*

MoJ~ -r.E.XI POWER DISTRIBUTION LIMITS 3.23.1 LINEAR HEAT RATE {LHRl LIMITING CONDITION FOR OPERATION Basis (Contd) TO ?t46't.. The time interval of 2 hours and the minimum of 10 detectors per quadrant are sufficient to maintain adequate surveillance of the core power distribution to detect significant changes until the monitoring systems are returned to service. 3 -ID'6/ To ensure that the design margin of safety is maintained, the determination of both the incore alarm setpoints and the APL takes into account the local LHGR measurement uncertainty factors'~ given in Table 3.23-3, an engineering uncertainty factor of 1.03, a thermal power measurement uncertainty factor of 1.02. References (1) EMF-91-77 (2) (Deleted) (3) (Deleted) (4) XN-NF-80-47 (5) FSAR Section 3.3.2.5 -. *(6) FSAR Seeti en 7. 6. 2. 4- ~ fl\\-Glt 3-105"' ifr.1Mt1<ffl.""r!.."'1> Amendment No. ~~, JJ~, Jpl, J%'/r-, 144-April 3, 1992 .{next ii age i s J 10 7 )..

<;/J~ .-}INV18 lJJl AllVNOI+/-N3+/-NI

.-:.1. TABLE 3.23-1 LINEAR HEAT RATE LIMIT Peak Rod TABLE 3.23-2 RADIAL PEAKING FACTOR LIMITS, Peaking Factor Reload L & M Reload O Assembly F~

1. 76 Peak Rod FT r 2.04 LHR/Peaking Factor Parameter Measurement Uncerta i nty'01 (a)

(b) LHR 0.0795 0.0401 0.0695 0.0455 0.0526 Meas ement uncertainty forcreload cores using all de ctors. asurement uncertainty for reload cores using and once-burned incore detectors. Measurement uncertainty when quadrant power tilt, as determine using i ncore measurements and an i ncore analysis computer program161, 2.8% but is less than or equal to 5%. ~ Amendment No. 68, -l-l-8,.J-4.d., -144, &e, ~ Septemberc 21, 1993 4

~ ~ ~ \\.i-J 1 D:- °"1 \\\\\\ *. r-... ~ ~

  • ~

l1 I rt g: -~

t1

~ F\\) ~ ~ c:::: \\" \\) -/ c ~* () r f\\J. ~ 0 z 0.... f'4 u cC ~ ~ p::

i:

..:I rz1 ..:I ill ~ 0 ....l ..:I < UNAC PTABLE OP ATION 10----~~~~~~~---~~~~~~--..- I 0.8 0.8 ACCEPTABLE OPERATION . 2. 1.0,.93. 0.2 0.4 0.8 0.8 FRACTION OF ACTIVE FUEL HEIGHT FIGURE 3.23-1 ALLOWABLE LHR AS A FUNCT.ON Of PEAK POWER LOCATION .~

886I 'gy ~aqmaAOfit' BII 'Z0 *oN 111ampuawv

711-<9~ 7,-(/() tf/_/MIA/~re,..b 6 .:zAmeAamel'lt Ne. ~~, i 1 a N.wt.emeer 1§, 1928

POWER DISTRIBUTION LIMITS 3.23.2 RADIAL PEAKING FACTORS LIMITING CONDITION FOR OPERATION The radial peaking factors FA, and FT shall be less thaA el" e~ual te the value I iA Ta~le 3.23 2 times the feflewiAg ~uaAtity. The ~uantity is [1.0 + 0.3 (1 - P)] fel" P ). § aAE:i the ~Yafltity is 1. Hi fel" P < * §. P is the eel"e thel"mal 13ewer ifl fractieR ef rated pe~\\19"*2 fv1P,.lf>.('TA-1,.J'2..."':> wt'rl-fl/\\/ '"Tl-I~ 1-fMl"'TS S~~t:J,C/ie:1::> '"' /lo/~ CC>L. rl... APPLICABILITY: Power operation above 25% of rated power. ACTION:

1.

For P < 50% of rated with any radial peaking factor exceeding its limit, be in at least hot shutdown within 6 hours.

2.

For P ~ 50% of rated with any radial peaking factor exceeding its limit, reduce thermal power within 6 hours to less than the lowest value of: [1 F - 3.33( r - 1) ] x Rated Power F L Where Fr is the measured value of either F~, or F~ and FL is the corresponding limit fl"em.:+ael e 3. 23 2. CoL 2. Sp~ I t:=!I f:. 't:> I /\\I Tl-(/?:- Basis The limitations on F~, and F~ are provided to ensure that assumptions used in the analysis for establishing DNB margin, LHR and the thermal margin/low- \\ pressure and variable high-power trip set points remain valid during operation. Data from the incore detectors are used for determining the measured radial peaking factors. The periodic surveillance requirements for determining the measured radial peaking factors provide assurance that they remain within prescribed limits. Determining the measured radial peaking factors after each ~ fuel loading prior to exceeding 50% of rated power provides additional assurance that the core is properly loaded. To ensure that the design margin of safety is maintained, the determination of radial peaking factors takes into account the appropriate measurement ( uncertainty factors 111 g.iveR iA Table 3.2~ Srlf:<:. 1~11f;..z:> IN Tl-I'!.- Cct..R.. References (1) FSAR Section 3.3.2.5 3-111 Amendment No ea, 18, J..6-7, l4a-, !44, ""1"50' JYfle 16, 1993

ADMINISTRATIVE CONTROLS 6.9.1 Routine Reports (continued)

c. Monthly Operatinfi Report - Routine reports of operating statistics and s utdown experience shall be submitted on a monthly basis to the NRC to arrive no later than the fifteenth of each month following the calendar month covered by the report.
d.

Radioactive Effluent Release Report

e.

The Radioactive Effluent Release Report shall be submitted in accordance with 10 CFR 50.36a. The report shall include a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be (1) consistent with the objectives outlined in the ODCM and PROCESS CONTROL PROGRAM and (2) in conformance with 10 CFR 50.36a and Section IV.B.l of Appendix I to 10 CFR 50. Radiological Environmental Operating Report I N S ~ l(_T -=tt: t..f oL il StcTto,.._/) The Radiological Environmental Operating Report covering the operation of the unit during the previous calendar year shall be submitted before May 1 of each year. The report shall include summaries, interpretations, and analysis of trends of the results of the Radiological Environmental Monitoring Program for the reporting period. The material provided shall be consistent with the objectives outlined in (lJ the ODCM and (2) Sections IV.B.2, IV.B.3, and IV.C of Appendix I to MOll'L T~x:T TO ft%~ 6.9.2 6.9.3 10 CFR 50. Reportable Events The Commission shall be notified of Reportable Events and a report submitted pursuant to the requirements of 10 CFR 50.73. Nonroutine Reports A report shall be submitted in the event that (a) the Radiological Environmental Monitoring Programs are not substantially conducted as described in the ODCM or (b) an unusual or important event occurs from plant operation that causes a significant environmental impact or affects a potential environmental impact. Reports shall be submitted within 30 days. (Next page is 6-=-WJ 6-15 I"' s~ ~ T1 ON 0 F Co L-r( Amendment 1(~&01 l'C'.'~Nt~"'~ ~Dt:>S I f i4~~ S 6-/' f,_I'(

INSERT #4 Page 6-15

f.

Core Operating Limits Report (COLR) Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following: 3.1.1 3.10.5 3.11.2 3.23.l 3.23.2 Tinle Function and ASI Limits. Reguiating Group Insertion Limits Excore Monitoring Allowable Power Level Linear Heat Rate (LHR) Limits Radial Peaking Factor Limits The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the latest approved revision of the following documents:

1.

XN-75-27(A) and Supplements, "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," Exxon Nuclear Company.

2.

ANF-84-73(P) (A), Appendix Band Supplements, "Advanced Nuclear Fuels Corporation Methodology for Pressurized Water Reactors: Analysis of Chapter 15 Events," Advanced Nuclear Fuels Corporation.

3.

XN-NF-82-2l(P)(A}, "Application of Exxon Nuclear Company PWR Thermal Margin Methodology to Mixed Core Configurations," Exxon Nuclear Company.

4.

ANF-84-093(P)(A) and Supplement, "Steamline Break Methodology for PWRs," Advanced Nuclear Fuels Corporation.

5.

XN-75-32(P)(A), Supplements, "Computational Procedure for Evaluating Fuel Rod Bowing," Exxon Nuclear Company.

6.

EXEM PWR Large Break LOCA Model as defined by: a) b) c) XN-NF-82-20(A), and Supplements, "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," Exxon Nuclear Company. XN-NF-82-07(P)(A), "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model," Exxon Nuclear Company. XN-NF-81-58(A) and Supplements, "RODEX2 Fuel Rod Therma 1-Mechani cal Response Evaluation Model," Exxon Nuclear Company.

INSERT #4 (continued) d) XN-NF-85-lG(A), Volume 1 and Supplements, and Volume 2 and Supplement, "PWR 17x17 Fuel Cooling Tests Program," Exxon Nuclear Company. e) XN-NF-85-44(A), and Supplement, "Scaling of FCTF Based Reflood Heat Transfer Correlation for other Bundle Designs," Exxon Nuclear Company.

7.

XN-NF-78-44(A), "A Generic Analysis of the Control Rod Ejection Transient for Pressurized Water Reactors," Exxon Nuclear Company.

8.

ANF-1224(P)(A) and Supplement, "Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Advanced Nuclear Fuels Corporation.

9.

ANF-89-192(P), "Justification of the ANFP DNB Correlation for High Therma 1 Performance Fue 1 in the Pali sades reactor," Advanced Nuclear Fuels Corporation.

10.

ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation.

11.

EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic 1 imits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as shutdown margin, transient analysis limits, and accident analysis limits) of the safety analysis are met. The COLR, including any mid cycle revisions or supplements, shall be provided, upon issuance for each reload cycle, to the NRC.}}

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