ML18064A733

Revised Proposed Ts,Consisting of Changes to Basis in First Paragraph of Page 3-2 & Rev to List of Approved Methodology Repts

ML18064A733
Person / Time
Site:	Palisades
Issue date:	04/27/1995
From:	CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
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ML18064A732	List: ML18064A731 ML18064A733
References
NUDOCS 9505050211
Download: ML18064A733 (23)

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.l ATTACHMENT I Consumers Power Company Palisades Plant Docket 50-255 CORE OPERATING LIMITS REPORT TECHNICAL SPECIFICATION CHANGE REQUEST Revised Proposed Pages Proposed pages 3-lc, 3-2, & 3-3 replace pages 3-lc, 3-ld, 3-2, 3-3, & 3-3a Proposed pages 3-86 & 3-87 replace pages 3-86, 3-87, 3-88, & 3-89 April 27, 1995 18 Pages

J

~!SADES PLANT TECHNICAL SPECIFI~IONS TABLE OF CONTENTS SECTION DESCRIPTION PAGE NO 1.0 DEFINITIONS 1-1 1.1 OPERATING DEFINITIONS 1-1 1.2 MISCELLANEOUS DEFINITIONS 1-5 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 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 B2.1 Basis - Reactor Core Safety Limit B 2-1 B2.2 Basis - Primary Coolant System Safety Limit B 2-2 B2.3 Basis - Limiting Safety System Settings B 2-3 3.0 LIMITING CONDITIONS FOR OPERATION 3-1 3.0 APPLICABILITY 3-1 3.1 PRIMARY COOLANT SYSTEM 3-lb 3.1.1 Operable Components 3-lb 3.1.2 Heatup and Cooldown Rates 3-4 Figure 3-1 Pressure - Temperature Limits for Heatup 3-9 Figure 3-2 Pressure - Temperature Limits for Cooldown 3-10 3.1.3 Minimum Conditions for Criticality 3-12 3.1.4 Maximum Primary Coolant Radioactivity 3-17 3.1.5 Primary Coolant System Leakage Limits 3-20 3.1.6 Maximum PCS Oxygen and Halogen Concentration 3-23 3.1. 7 Primary and Secondary Safety Valves 3-25 3.1.8 Over Pressure Protection Systems 3-25a Figure 3-4 LTOP Limit Curve 3-25c 3.1. 9 Shutdown Cooling 3-25h 3.2 CHEMICAL AND VOLUME CONTROL SYSTEM 3-26 3.3 EMERGENCY CORE COOLING SYSTEM 3-29 3.4 CONTAINMENT COOLING 3-34 3.5 STEAM AND FEEDWATER SYSTEMS 3-38 3.6 CONTAINMENT SYSTEM 3-40 Table 3.6.l Containment Penetrations and Valves 3-40b 3.7 ELECTRICAL SYSTEMS 3-41 3.8 REFUELING OPERATIONS 3-46 3.9 Deleted 3-49 Amendment No.

I

• "\\.

PALltJtES PLANT TECHNICAL SPECIFICATI~

TABLE OF CONTENTS SECTION DESCRIPTION PAGE NO 3-1 3-50 3-50 3-51 3-51 3-52 3-52 3-53 3-53 3.0 LIMITING CONDITIONS FOR OPERATION (continued) 3.10 3.10.1 3.10.2 3.10.3 3.10.4 3.10.5 3.10.6 3.10.7 CONTROL ROD AND POWER DISTRIBUTION LIMITS Shutdown Margin Requirements Deleted Part-Length Control Rods Misaligned or Inoperable Rod Regulating Group Insertion Limits Shutdown Rod Limits Low Power Physics Testing 3.11 POWER DISTRIBUTION INSTRUMENTATION 3-56 3.11.1 Incore Detectors 3-56 3.11.2 Excore Power Distribution Monitoring System 3-57 Figure 3.11-1 Axial Variation Bounbing Condition 3-59 3.12 MODERATOR TEMPERATURE COEFFICIENT OF REACTIVITY 3.13 Deleted 3.14 CONTROL ROOM VENTILATION 3.15 REACTOR PRIMARY SHIELD COOLING SYSTEM 3.16 ESF SYSTEM INITIATION INSTRUMENTATION SETTINGS Table 3.16.1 ESF System Initiation Instrument Setting Limits B3.16 Basis - ESF System Instrumentation Settings 3.17 INSTRUMENTATION AND CONTROL SYSTEMS 3.17.1 Reactor Protective System Instruments Table 3.17.1 Instrument Requirements for RPS 3.17.2 Engineered Safety Features Instruments Table 3.17.2 Instrument Requirements for ESF Systems 3.17.3 Isolation Functions Instruments Table 3.17.3 Instrument Requirements Isolation Functions 3.17.4 Accident Monitoring Instruments Table 3.17.4 Instrument Requirements for Accident Monitoring 3.17.5 Alternate Shutdown System Instruments Table 3.17.5 Instruments for the Alternate Shutdown System 3.17.6 Other Safety Feature Instruments Table 3.17.6 Instruments for Other Safety Features B3.17 Basis - Instrumentation Systems 3.18 3.19 3.20 3.21 3.22 3.23 3.23.1 3.23.2 3.23.3 Deleted IODINE REMOVAL SYSTEM SHOCK SUPPRESSORS (Snubbers)

CRANE OPERATIONS AND MOVEMENT HEAVY LOADS Deleted POWER DISTRIBUTION LIMITS Linear Heat Rate Radial Peaking Factors Quadrant Power Tilt - Tq ii 3-60 3-60 3-61 3-62 3-63 3-63 B 3.16-1 3-64 3-64 3-65 3-66 3-67 3-68 3-69 3-70 3-71 3-72 3-73 3-74 3-77 B 3.17-1 3-79 3-79 3-80 3-81 3-84 3-84 3-84 3-86 3-87 Amendment No.

.1 SECTION I

e LISADES PLANT TECHNICAL SPECIFICATIONS TABLE OF CONTENTS DESCRIPTION PAGE NO 6.0 ADMINISTRATIVE CONTROLS (Continued) 6.6 6.7 6.8 6.9 6.9.1 6.9.1.a 6.9.1.b 6.9.1.c 6.9.1.d 6.9.1.e 6.9.1.f 6.9.2 6.9.3 6.9.4 6.10 6.11

. 6.12 6.13 6.14 6.15 6.16 6.17 6.18 6.19 6.20 6.21 6.22 Deleted SAFETY LIMIT VIOLATION PROCEDURES AND PROGRAMS REPORTING REQUIREMENTS Routine Reports Start-up Report Annual Report Monthly Operating Report Radioactive Effluent Release Report Radiological Environmental Operating Report Core Operating Limits Report Reportable Events Nonroutine Reports Special Reports RECORD RETENTION RADIATION PROTECTION PROGRAM HIGH RADIATION AREA Deleted Deleted SYSTEMS INTEGRITY IODINE MONITORING POST ACCIDENT SAMPLING OFFSITE DOSE CALCULATION MANUAL PROCESS CONTROL PROGRAM Deleted SEALED SOURCE CONTAMINATION SECONDARY WATER CHEMISTRY v

6-10 6-10 6-11 6-14 6-14 6-14 6-14 6-15 6-15 6-15 6-15 6-17 6-17 6-26 6-26 6-28 6-28 6-33 6-33 6-33 6-33 6-34 6-35 6-35 6-36 6-37 6-38 Amendment No.

1.2 MISCELLANEO~DEFINITIONS 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.*

MEMBER(S) OF THE PUBLIC MEMBER(S) OF THE PUBLIC shall include all persons who are not occupationally associated with the plant. This category does not include employees of the utility, its contractors, or its vendors.

Also excluded from this category are persons who enter the site to service equipment or to make deliveries.

OFFSITE DOSE CALCULATION MANUAL (ODCM)

The OFFSITE DOSE CALCULATION MANUAL shall contain the current methodology and parameters used in the calculation of offsite doses due to radioactive gaseous and liquid effluents, in the calculation of gaseous and liquid effluent monitoring alarm and trip setpoints, and in the conduct of the Radiological Environmental Monitoring Program.

The ODCM shall also contain the (1) Radioactive Effluent Controls and Radiological Environmental Monitoring Programs required by Specification 6.8.4 and (2) descriptions of the information to be included in the Radiological Environmental Operating Report and the Radioactive Effluent Release Report required by Specification 6.9.3.

PROCESS CONTROL PROGRAM The PROCESS CONTROL PROGRAM shall contain the current formula, sampling, analyses, tests, and determinations to be made to ensure that the processing and packaging of solid radioactive wastes based on demonstrated processing of actual or simulated wet solid wastes will be accomplished in such a way as to assure compliance with 10 CFR 20, 10 CFR 71, Federal and State regulations, and other requirements governing the disposal of the radioactive waste.

SITE BOUNDARY The SITE BOUNDARY shall be that line beyond which the land is neither owned nor otherwise controlled by the licensee.

UNRESTRICTED AREA An UNRESTRICTED AREA shall be any area at or beyond the SITE BOUNDARY access to which is not controlled by the licensee for purposes of protection of individuals from exposure to radiation and radioactive materials or, any area within the SITE BOUNDARY used for residential quarters or for industrial, commercial, institutional, or recreational purposes.

Amendment No. SS., 64, &2-,

1-5

3.1 3.1.1 PRIMARY COO~T SYSTEM Applicability Applies to the operable status of the primary coolant system.

Objective To specify certain conditions of the primary coolant system which must be met to assure safe reactor operation.

Specifications Operable Components

a.

b.

c.

d.

e.

At least one primary coolant pump or one shutdown cooling pump with a flow rate greater than or equal to 2810 gpm shall be in operation whenever a change is being made in the boron concentration of the primary coolant and the plant is operating in cold shutdown or above, except during an emergency loss of coolant flow situation.

Under these circumstances, the boron concentration may be increased with no primary coolant pumps or shutdown cooling pumps running.

Four primary coolant pumps shall be in operation whenever the reactor is operated above hot shutdown, with the following exception:

Before removing a pump from service, thermal power shall be reduced as specified in Table 2.3.1 and appropriate corrective action implemented.

With one pump out of service, return the pump to service within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> (return to four-pump operation) or be in hot shutdown (or below) with the reactor tripped (from the C-06 panel, opening the 42-01 and 42-02 circuit breakers) within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Start-up (above hot shutdown) with less than four pumps is not permitted and power operation with less than three pumps is not permitted.

The measured four primar~ coolant pumps operating reactor vessel flow shall be 140.7 x 10 lb/hr or greater, when corrected to 532°F.

Both steam generators shall be capable of performing their heat transfer function whenever the average temperature of the primary coolant is above 300°F.

The AXIAL SHAPE INDEX (ASI) shall be maintained within the limits specified in the COLR.

(1)

When the ASI exceeds the limits specified in the COLR, within 15 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.'

3-lb Amendment No. 3-l, &£, -l-l-8, -1+9, -l-34, -l-3-7, i:-6-+/--,

3.I 3; I. I PRIMARY COOLANT-STEM e

Operable Components (Continued}

f.

Nominal primary system operation pressure shall not exceed 2IOO psi a.

g.

The indicated reactor inlet temperature (T~} shall not exceed the value given by the following equation at s~eady state power operation:

Tc~ 542.99 +.0580(P-2060} +.OOOOI(P-2060} 2 + I.I25(W-I38} -.0205(W-I38} 2

h.

i.

j.

Where:

reactor inlet temperature in °F nominal operatin~ pressure in psia total recirculating mass flow in I06 lb/h corrected to the operating temperature conditions.

If the measured primary coolant system flow rate is greater than I50 M lbm/hr, the maximum inlet temperature shall be less than or equal to the Tc LCO at I50 M lbm/hr.

(I}

When reactor inlet temperature exceeds the limit, restore reactor inlet temperature to within limits within 30 minutes.

Forced circulation fstarting the first primary coolant pump} shall not be initiated un ess one of the following conditions is met:

(I}

(2}

(3}

(4}

PCS cold leg temperature (Tc} is > 430°F.

S/G secondary temperature is ~ Tc.

S/G secondary temperature is < I00°F above Tc, and shutdown cooling is isolated from the PCS, and PCS heatup/cooldown rate is ~ I0°F/hour.

S/G secondary temperature is < I00°F above Tc, and shutdown cooling is isolated from the PCS, and pressurizer level is ~ 57%.

When the PCS cold leg temperature is < 300°F, primary coolant pumps P-50A and* P-508 shall not be operated simultaneously.

The PCS shall not be heated or maintained above 300°F unless a mfoimum of 375 kW of pressurizer heater capacity is available from both buses ID and IE.

Should heater capacity from either bus ID or 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT SHUTDOWN within the next I2 hours.

3-Ic Amendment No. 3+,.§.!, 8£, H-1, H-8, l-34, 7-, l-, 63-,

3.1 PRIMARY CO~NT 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. 111 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 of 2810 gp~, sufficient time is provided for the operator to terminate the boron dilution under asymmetric flow conditions. 151 The pressurizer volume is relatively inactive, therefore it will tend to have a boron concentration higher than the 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.'21 The 57% pressurizer level, in section 3.1.lh(4), is not an analytical result, but simply a decision point between having and not having a bubble.

It was chosen to agree with the maximum programmed level during power operation.

The limitation, in section 3.1.li, on operating P-50A and P-50B together with T0 below 300°F allows the Pressure Temperature limits of Figures 3-1 and 3-2 to be higher than they would be without this limit.

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 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 fl ow14'.

In addition, transient biases were included in the determination of the allowable reactor inlet temperature.

3-2 Amendment No. 6J, 8§., -H-7, -H-8, 3-l-, 3-7, tea-,

3.1 PRIMARY COOLANT SYSTEM Basis (continued)

The limits of validity of the Tinlet equation are:

1800 s pressure s 2200 psia 100.0 x 106 s Vessel Flow s 150 x 106 lb/h ASI as shown in Figure 3.0 With measured primary coo 1 ant system fl ow rates > 150 M 1 bm/hr, limiting the maximum allowed inlet temperature to the T1"l1i LCO. at 150 M lbm/hr increases the margin to DNB for higher PCS flow rates.

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 s 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 (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+, &!-, H-7, H-8, -l-3-1-, !34, +/-3-7, !43, 56, &9,

3.10 3.10.4 3.10.5 CONTROL RO~ND POWER DISTRIBUTION LIMITS 4lt 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 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Amendment No. 1-, 3-!-, 6-2-,

3-52

. 3.10 3.10.6 3.10.7 CONTROL ROD AN~OWER DISTRIBUTION LIMITS 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. 111 The change in insertion limit with reactor power insures that the shutdown requirements for 4-pump operation is met at all power levels. The 2.5-second drop time specified for the CONTROL RODS is the drop time used in the transient analysis. 111 Amendment No. 3+, M,.§..7, 68, HS, i:-3-7, &2-,

3-53

Amendment No 3-l-, -l-18, ~'

3-55

I

\\

3.12 MODERATOR ~PERATURE 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.

Speci fi cat i ans 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 111 remain valid.

Reference (1)

EMF-92-178, Revision 3, Section 15.0.5 3.13 Deleted Amendment No. -l-l+, -l-l-8, -l-3-7, 43-, ~' &9, 62-,

3-60

3.22 3.23 3.23.1 Deleted

• POWER DISTRIBUTION LIMITS LINEAR HEAT RATE (LHRl 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 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or failing this, be at less than 50% RATED POWER within the following 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> be at 85% (or less) RATED 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% 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 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and at least every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> thereafter. If readings indicate a local power level equal to or greater than the alarm setpoints, the action specified in ACTION 1 above sha 11 be taken.

Amendment No. 31-, W, ~' -H-8, -144, -l62-,

3-84

l POWER DISTRIBUTION LIMIT~

3.23.1 LINEAR NEAT RATE CLHRl LIMITING CONDITION FOR OPERATION The limitation of LHR ensures that, in the event of a LOCA, the peak temperature of the cladding will 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.'21 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 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 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'~ 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 Amendment No. eB, ~, H-8, ~' ~' -144, ~'

3-85

)*

3.23 3.23.2 POWER DIST~UTION LIMITS RADIAL PEAKING FACTORS Basis LIMITING CONDITION FOR OPERATION The radial peaking factors FA, 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

2.

For P ~ 50% of rated with any radial peaking factor exceeding its limit, reduce thermal power within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to less than the lowest value of:

F

[1 - 3.33( r - 1) ] x RATED POWER

-F-L A

T Where F is the measured value of either F, or Fr, and FL is the corre$p6nding limit specified in the COLR.r The limitations on FA, and FT are provided to ensure that assumptions used in the analysis for est~blishing 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 specified in the COLR.

References (1)

FSAR Section l.3.2.5 Amendment No. 68, -l-1-8, 3-7,

-146~ !44, -I-Se, &2-,

3-86

\\. -

. 3.23 3.23.3 POWER DISTRIBU~N LIMITS

• QUADRANT POWER TILT - T q LIMITING CONDITION FOR OPERATION The quadrant power tilt (Tq) shall not exceed 5%.

APPLICABILITY:

Power operation above 25% of RATED POWER.

ACTION:

1.

With Tq > 5% but ~ 10%.

a.

Correct Tq within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after exceeding the limit, or

b.

Determine within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> thereafter, that the radial peaking factors are within the limits of Section 3.23.2, or

c.

Reduce power, at the normal shutdown rate, to less than 85%

RATED POWER and determine that the radial peaking factors are within the limits of Section 3.23.2.

At reduced power, determine at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that the ra.dial peaking factors are within the limits of Section 3.23.2.

2.

With Tq > 10%:

a.

Correct Tq within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after exceeding the limit, or

b.

Reduce power to less than 50% RATED POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and determine that the radial peaking factors are within the limits of Section 3.23.2. At reduced power, determine at least once every 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> that the radial peaking factors are within the limits of Section 3.23.2.

3.

With Tq > 15%, be in at least hot standby within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Limitations on quadrant power tilt are provided to ensure that design safety margins are maintained.

Quadrant power tilt is determined from excore detector readings which are calibrated using incore detector measurements. 111 Quadrant power tilt calibration factors are determined using incore measurements and an incore analysis computer program.'~

References (1)

FSAR, Section 7.4.2.2 (2)

FSAR, Section 7.6.2.4 3-87 Amendment No. 68, 118, 144, 154, 162 October 26, 1994

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.1 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 {l) 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 3.23.1 3.23.2 ASI Limits.

Regulating Group Insertion Limits Linear Heat Rate (LHR) Limits Radial Peaking Factor Limits 6-15 Amendment No. -!-6, 6, ~' 8£, -!98, -l-54,

ADMINISTRATIVE CONTROLS 6.9.1 Routine Reports

f.

COLR (continued)

The analytical methods used to determine the core operating limits shall be those approved by the NRC, specifically those described in the latest approved revision of the following documents:

1.

XN-75-27(A) "Exxon Nuclear Neutronics Design Methods for Pressurized Water Reactors," and Supplements l(A), 2(A), 3(P)(A),

4(P)(A), and 5(P)(A); Exxon Nuclear Company.

(LCOs 3.1.1, 3.10.1, 3.10.5, 3.23.1, & 3.23.2)

2.

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

and Appendix B(P)(A) and Supplements l(P)(A), 2(P)(A); Advanced Nuclear Fuels Corporation.

(LCOs 3.1.1, 3.10.5, 3.23.1, & 3.23.2)

3.

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

(LCOs 3.1.1, 3.23.1, & 3.23.2)

4.

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

(LCOs 3.10.1, 3.10.5, 3.23.1, & 3.23.2)

5.

XN-75-32(P)(A), "Computational Procedure for Evaluating Fuel Rod Bowing," and Supplements l(P)(A), 2(P){A), 3(P){A}, and 4{P){A);

Exxon Nuclear Company.

(LCOs 3.1.1, 3.10.5, 3.23.1, & 3.23.2)

6.

EXEM PWR Large Break LOCA Model as defined by:

(LCOs 3.10.5, 3.23.l, & 3.23.2) a)

XN-NF-82-20(A), "Exxon Nuclear Company Evaluation Model EXEM/PWR ECCS Model Updates," and Supplements l{P){A},

2(P){A), 3(P){A), and 4(P)(A); Exxon Nuclear Company.

b)

XN-NF-82-07{P){A), "Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model, 11 Exxon Nuclear Company.

c)

XN-NF-81-58(A) "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model, 11 and Supplements l{P)(A), 2(P)(A), 3{P){A),

and 4{P)(A); Exxon Nuclear Company.

d)

XN-NF-85-16{A),

11 PWR 17xl7 Fuel Cooling Tests Program," Volume 1 and Supplements l{P){A), 2{P){A}, and 3{P){A}, and Volume 2 and Supplement l{P){A); Exxon Nuclear Company.

e)

XN-NF-85-105{A), "Scaling of FCTF Based Reflood Heat Transfer Correlation for other Bundle Designs," and Supplement l(P){A);

Exxon Nuclear Company.

6-16 Amendment No.

~

ADMINISTRATIVE CONTROLS 6.9.1 Routine Reports 6.9.2

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.

{LCOs 3.10.5, 3.23.1, & 3.23.2)

8.

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

(LCOs 3.1.1, 3.23.1, & 3.23.2)

9.

ANF-89-151(P)(A), "ANF-RELAP Methodology for Pressurized Water Reactors: Analysis of Non-LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation. (LCOs 3.1.1, 3.10.5, 3.23.1, & 3.23.2)

10.

EMF-92-153(P)(A), "HTP: Departure from Nucleate Boiling Correlation for High Thermal Performance Fuel," Siemens Power Corporation.

(LCOs 3.1.1, 3.23.1, & 3.23.2)

11.

Letter from K. W. Berry (CPC) to USNRC, "Technical Specification Change Request - Incore Analysis Program" dated October 24, 1989 and Letter from G. B. Slade to USNRC "Technical Specification Ch~nge Request - Incore Analysis Program -

R~vision l" dated August 24, 1990.

Approved in USNRC Safety Evaluation for Amendment 144 to the Palisades Facility Operating License, dated April 3, 1992.

(LCOs 3.1.1, 3.23.1, & 3.23.2)

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.

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 Palisades Plant Docket 50-255 CORE OPERATING LIMITS REPORT TECHNICAL SPECIFICATION CHANGE REQUEST Answers to Reviewer's Questions April 27, 1995 2 Pages

Answers To Quest~s and Comments Raised By NRR R~ewer (Tai Huang]

Question 1:

"Why is the Tinlet Function cycle specific?"

Answer 1:

The inlet temperature function provides administrative protection against the most limiting transients which might occur during power operation.

The dropped control assembly is currently the most limiting transient which relies on the Tinlet LCO and is therefore the basis for the LCO.

Each eye le, the Tinlet LCO must be verified to see if it provides the proper protection against DNB.

If it is determined that the T~. LCO does not provide the proper protection for the limiting event, it must be altered in such a manner that does provide the necessary protection.

The Tinlet LCO used at the Pali sades reactor is set using a purely deterministic methodology.

When the Tinlet LCO is deemed inadequate for a particular cycle, the constants must be changed.

Values for Tinlet' generated from the Tinlet Function, are compared with the values from XCOBRA-IIIC.

If the values generated using the Tinlet Function are not conservative compared to the XCOBRA-IIIC values, the constants in the Tinlet Function are changed to ensure conservatism.

The dependence on XCOBRA-IllC, and inputs that affect XCOBRA-IIIC, cause the Tinlet Function to be cycle dependent.

Since XCOBRA-IIIC is used to develop the Tinlet Function, anything that changes XCOBRA-IIIC results could also change the T~. Function.

For example, if peaking factor or ASI limits change, the limiting DNB event could be affected.

This could cause the T~~ Function to change also. This dependence on core power distribution limits causes the T~. Function to be cycle specific.

Question 2:

"Has Reference 9 been approved?"

Answer 2:

No, reference 9 (ANF-89-192(P)) is the report that explains how generic methodology is applied to Palisades.

The generic methodology (ANF-1224(P)(A) Rev. 0, Departure From Nucleate Boiling Correlation For High Thermal Performance Fuel, May 15, 1989) has been approved.

This plant specific detail is not appropriate to be included in the generic document.

Since the former reference 9 is not an NRC approved document, it has been deleted from the list of references.

Question 3:

Answer 3:

"References within the COLR need to be directly associated with a particular parameter."

The list of references has been revised to indicate which TS limits are associated with each reference; see the revised proposed pages in.

1

t.*

Answers To Questions and Comments Raised By NRR Reviewer (continued)

Question 4:

Answer 4:

"Why is Technical Specification Table 3.23-3 (COLR Table 2.1),

Peaking Factor Measurement Uncertainties, cycle specific?"

The uncertainties associated with the PIDAL methodology, listed in Technical Specifications Table 3.23-3, were generated based on cycle 5, 6, and 7 data. This provided a large pool of data for the calculation and set the limits on peaking factor -uncertainties for future cycles.

Because these uncertainties are heavily dependent upon the core loading pattern (i.e. low vs. high leakage), the theoretical power distribution, and the incore conversion constants, each cycle design will generate a new set of uncertainties based on its core design.

It is important to note that the methodology used by our incore monitoring code, PIDAL, has been approved by the NRC and has not been altered since its approval.

Since our conversion to low leakage core designs in cycle 9 and to quarter core rotational patterns in cycle 10, Palisades has reanalyzed the PIDAL uncertainties to ensure that the approved uncertainties were still appropriate.

The resulting calculated uncertainties for cycles 9 and 10 have been below those listed in Table 3.23-3 due largely to the improvement in the theoretical model.

Because of our continually changing core designs, to lower our vessel fluence, and our continuing improvement in the theoretical models for both power distribution and incore constants, it is our belief that these uncertainty factors should be included in the COLR and updated as these improvements are made.

Approval of the uncertainty determination methodology is in the SER of Consumers Power Docket No. 50-250, Amendment No. 144, License-No. DPR-20, dated April 3' 1992.

2