Text

Amend 138 to License DPR-77,modifying Tech Specs to Permit Use of Vantage 5 Hybrid Fuel in Upcoming Operating Cycle 5
	ML20043A872
Person / Time
Site:	Sequoyah
Issue date:	05/08/1990
From:	Black S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20043A873	List: ML20043A872; ML20043A874;
References
	NUDOCS 9005230223
	Download: ML20043A872 (24)
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8 UNITE D STATES NUCLEAR REGULATORY COMMISSION l

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l TENNESSEE VALLEY AUTHORITY DOCKET NO. 50-327-SEQUOYAH NUCLEAR PLANT.--UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.138 License No. DPR-77 1.

The Nuclear Regulatory Commission (the Commission) has found.that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated January 12, 1990 and the supplemental letter dated April 13, 1990 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act).. and the Commission's rules and regulations set forth in 10 CFR Chapter I; i

B.

The facility will operate in confonnity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (1)~ that the activities authorized by L this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Connission's regulations; D.

The issuance of this amendment will not be inimical to the connon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10' CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment

. and paragraph 2.C.(2) of Facility Operating License No. DPR.77 is hereby

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amended to read as follows:

l (2) Technical ~ Specifications

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l' The Technical Specifications contained in Appendices A and B, as j

revised through Amendment No.138, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

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3.

This license amendment is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION t

e Iit or (TVAProjectsDivision for Projects Office of Nuclear Reactor Regulation j

Attachment:

Changes to the Technical Specifications Date of Issuance:

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ATTACHMENT TO LICENSE AMENDMENT NO.13o FACILITY OPERATING LICENSE NO. OPR-77 DOCKET NO. 50-327 Revise the Appendix A Technical Specifications by removing the pages identified below and interting the enclosed pages.

The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. Overleaf pages* are provided to maintain document completeness.

t REMOVE INSERT V

V XI XI*

XII XII B 2-1 B 2-1 B 2-2 B 2-2*

B 2-3 B 2-3 B.2-4 B 2-4*

B 2-5 8 2-5 B 2-6 B 2-6*

3/4 1-19 3/4 1-19 3/4 2-20 3/4 2-10 3/4 2-11 3/4 2-11 3/4 2-12 3/4 2-12 3/4 2-13 3/4 2-13 3/4 2-14 3/4 2-14 3/4 2-15 3/4 2-15 3/4 2-16 3/4 2-16 3/4 2-17 3/4 2-18 3/4 2-19 B 3/4 2-1 B 3/4 2 - 1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-4

'B 3/4 2-4 B 3/4 2-5 B 3/4 4-1 B 3/4'4-1

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INDEX

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LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS 1

SECTION PAGE 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 Axial Flux Difference.....................................

3/4 2-1 3/4.2.2 Heat Flux Hot Channel Factor..............................

3/4 2-5 3/4.2.3 Nuclear Enthalpy Hot Channel Factor.......................

3/4 2-10 3/4.2.4 Quadrant Power Tilt Ratio.................................-

3/4 2-15 3/4.2.5 DNB Paramsters............................................

3/4 2-18 3/4.3 INSTRUMENTATION 3/4,3.1 REACTOR TRIP SYSTEM INSTRUMENTAT!.0N.......................

3/4 3-1 l

3/4.3.2 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION..........................................

3/4 3-14 3/4.3.3 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation......................

3/4 3-39 i

l Movable Incore Detectors..................................

3/4 3-43 Seismic Instrumentation...................................

3/4 3-44 i

Meteorological Instrumentation............................

3/4 3-47 i

Remote Shutdown Instrumentation...........................

3/4 3-50 Chlorine Detection Systems (Deleted) 3/4 3-54 Accident Monitoring Instrumentation.......................

3/0.,-55 1

Fire Detection Instrumentstion.............................

3/4 3-58 Radioactive Liquid Effluent Monitoring Instrumentation....

3/4 3-69 I

Radioactive Gaseous Effluent Monitoring Instrumentation....

3/4 3-74 l

h SEQUOYAN - UNIT 1 V

Amendment No. 62,138

.c RNDEX

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I LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS i

SECTION PAGE 3/4.11 RADIOACTIVE EFFLUENTS l

3/4.11.1 LIQUID EFFLUENTS Concentration............................................

3/4 11-1 i

Dose.....................................................

3/4 11-5

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Liquid Waste Treatment...................................

3/4 11-6 Liquid Holdup Tanks......................................

3/4 11 7 t

3/4 11.2 GASEOUS EFFLUENTS Dose Rate................................................

3/4 11-8 l

.- i Dose-Noble Gases.........................................

3/4 11-12 Dose-Radioiodines, Particulate, and i

Radionuclides Other than Noble Gases...................

3/4 11-13 i

l Gaseous Radwaste Treatment...............................

3/4 11-14 r

Explosive Gas Mixture....................................

3/4 11-15 l

Gas Decay Tanks..........................................

3/4 11-16 i

3/4 11.3 SOLIO RADIOACTIVE WASTE..................................

3/4 11-17 f

3/4 11.4 TOTAL 00SE...............................................

3/4 11-19 3/4.12 RADIOLOGICAL ENVIRONMENTAL MONITORING 3/4.12.1 MONITORING PR0 GRAM......................................

3/4 12-1 3/4.12.2 LAND USE CENSUS.........................................

3/4 12-10 3/4.12.3 INTERLABORATORY COMPARISON..............................

_3/4 12-11 1

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.XI 1

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BASES 5

SECTION PAGE i

3/4.0- APPLICABILITY............................................

B 3/4 0,

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 BORATION CONTR0L...........................................

B 3/4 1-1 3/4.1.2 B0 RATION SYSTEMS...........................................

B'3/4 1-2 3/4.1.3 MOVABLE CONTROL ASSEMBLIES.............

B 3/4 1-3

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3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE......................................

B 3/4 2-1 3/4.2.2 and 3/4.2.3 HEAT FL'JX AND NUCLEAR ENTHALPY HOT CHANNEL FACT 0RS........................................

B 3/4 2-1 3/4.2.4 QUADRANT POWER TILT RATI0..................................

B 3/4 2-4 1

3/4.2.5 DNB PARAMETERS..............................................

B 3/4 2-4 3/4.3' INSTRUMENTATION I

3/4.3.1 and 3/4.3.2 PROTECTIVE AND ENGINEERED SAFETY FEATURES INSTRUMENTATION...................................

B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION.................................

B 3/4 3-2 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION..............

B' 3/4 4-1 t

3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES.......................

B 3/4 4-1 3/4.4.4 PRESSURIZER................................................

B 3/4 4-2 3/4.4.5 STEAM GENERATORS...........................................

B 3/4 4-2 l

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SEQUOYAH - UNIT.1 XII Amendment No. 138

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i 2.1 $AFETY LIMIT $

BASES i

l 2.1.1 REACTOR CORE i

The restrictions of this safety limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is orevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive _ cladding temperatures because of the onset of departure from nucleate boiling (DN8) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been l

related to DN8 through the WRB-1 correlation and the W-3 correlation for conditions outside the range of WRB-1 correlation.

The DNB correlations have been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions.

The local DNS heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNS at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is as follows:

there must be at'least a 95 percent a

probability that the minimum DNBR of the limiting rod during Condition I and i

11 events is greater than or equal to the DNBR limit of the DNB correlation being used (ths WRB-1 or W-3 correlation in'this application).

The correlation DNBR limit is established based on the entire applicable experimental data set such that there is a 95 percent procability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the.

DNBR limit.

i The curves of Figure 2.1-1 show the loci of. points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DN8R limit, or the average enthalpy at the vessel exit is equal to the enthalpy of. saturated liquid.

The curves are based on an enthalpy hot channel f actor.1 F g,_of 1.55 and a reference cosine with a peak of 1.55 for axial power shape.

An allowance is included hr an increase in Fh at reduced power based on the expression:

N F

= 1.55 (1+ 0.3 (1-P))

where P is the fraction of RATED THERMAL POWER r

5 SEQUOYAH - UNIT 1 8 2-1 Amendment No. 19, 114, 138 i

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SAFETY LIMITS BASES These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f) (Delta I) function of the Overtemperature Delta T trip.

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overttmperature Delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel and pressurizer are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure.

The Reactor Coolant System piping, valves and fittings, are designed to ANSI B 31.1 1967 Edition, which permits a maximum transient pressure of 120% (2985 psig) of component design-pressure.

The. Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the values at which the Reactor Trips are set for each functional unit.

The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or less than the drif t allowance assumed for each trip in the safety analyses.-

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SEQUOYAH - UNIT 1 B 2-2 Revised 08/18/87 Bases Change I

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. SAFETY LIMITS l

BASES Manual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

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Power Range, Neutron Flux i

The Power Range, Neutron Flux channel high setpoint provides reactor core l

protection against reactivity excursions which are too rapid to be protected v

by temperature and pressure protective circuitry.

The low set point provides i

redundant protection in the power range'for a power excursion beginning from low power.

The trip associated with the low setpoint may be manually bypassed when P-10 is active (two of.the four power range channels indicate a power level of above approximately 10 percent of RATED THERMAL POWER) and is auto-matica11y reinstated when P-10 becomes inactive (three of the four channels

. indicate a power level below approximately 9 percent of RATED THERMAL POWER).

Power' Range, Neutron Flux, High Rates l

The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level.

Specifically, this trip complements the Power Range Neutron Flux High and low trips to ensure that the criteria are met for rod ejection from partial power.

The Power Range Negative Rate trip provides protection to ensure that the minimum DNBR is maintained above the safety analysis DNBR limit for control rod drop accidents.

At high power a single or multiple rod drop accident could cause local flux peaking which, when in conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist.

The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor for all single or multiple dropped rods.

Intermediate and Source Range, Nuclear Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup.

These trips provide redundant protec-tion to the low setpoint trip of the Power Range, Neutron Flux channels.' The source Range Channels will initiate a reactor trip at about 10 counts per second unless manually blocked when P-6 becomes active.

The Intermediate j

l SEQUOYAH - UNIT 1 B 2-3 Amendment No.138 Revised 08/18/87 l

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3AFETY LIMITS m

BASES Range Channels will initiate a reactor trip at approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature Delta T The Overtemperature Delta T trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the High and Low Pressure reactor trips.

This setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature and dynamic compensation for pipin-delays from the core to the loop temperature detec-tors.

With normal axial power distribution, this reactor trip limit is always 2

2--

below the core safety limit as shown in Figure 2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip is automatically reduced according to the notations in Table 2.2-1.

Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature Delta T setpoint.

Tnree loop operation above the 4 loop P-8 setpoint is permissible after resetting the K1, K2 and K3 inputs to the Overtemperature Delta T channels and -*.ising the P-8 setpoint to its 3 loop value.

In this mode of operation, the /-8 inter-

-b lock and trip functions as a High Neutron Flux trip at the reduced power level.

_g Overpower Delta T The Overpower Delta T reactor trip provides assurance of fuel integrity, e.g., no melting, under all possible overpower conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to the High Neutron Flux trip.

The setpoint includes corrections for changes in density and heat capacity of water with temperature, and dynamic compensation for piping delays from the core to the loop temperature detectors.

No credit

'I was taken for operation of this trip in the accident SEQUOYAH - UNIT 1 B 2-4 Amendment No. 136 15

SAFETY LIMITS l

BASES 1

analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System..

Pressurizer Pressure i

The Pressur(Zer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted.

The High Pressure I

trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig).

The Low Pressure trip provides protection by tripping the reactor i

in the event of a loss of reactor coolant pressure.

l Pressurizer Water level L

The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves.

No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Loss of Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drop below 89% of nominal full loop flow.

Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 89% of nominal fu1111oop flow.

This latter trip will prevent the minimum value of the DNBR'from going below the safety analysis DNBR limit during normal operational transients and anticipated transients when 3 loops are in operation and the Overtemperature Delta T trip set point is adjusted to the value specified for all loops in operation.

With the Overtemperature Delta T trip set point adjusted to the value specified for 3 loop operation, the P-8 trip at 76% RATED THERMAL POWER will prevent the minimum value of the DNBR from going below the safety analysis DNBR limit during l

normal operational transients and anticipated transients with 3 loops in operation, f

SEQUOYAH - UNIT 1 B 2-5 Amendment No.138

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. SAFETY LIMITS BASES Steam Generator Water Level The Steam Generator Water Level Low-Low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity.

The specified setpoint provides allowance that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedwater system.

Steam /Feedwater Flow Mismatch and Low Steam Generator Water Level The Steam /Feedwater Flow Mismatch in coincidence with a Steam Generator Low Water Level trip is not used in the transient and accident analyses but is g

included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System.

This trip is redundant to the Steam Generator Water Level Low-Low trip.

The steam /Feedwater Flow Mismatch portion of this trip is activated when the steam flow exceeds the feedwater flow by greater than or equal to 1.5 x 108 lbs/ hour.

The Steam Generator Low Water level portion of the trip is activated when the water level drops below 24 percent, as indicated by the narrow range instrument.

These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but will initiate a reactor trip before the steam generators are dry.

Therefore, the required capacity and starting time requirements of the auxiliary feedwater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or under-frequency to more than one reactor coolant pump.

The specified set points assure a reactor trip signal is generated before the low flow trip set point is reached.

Time delays are incorporated in the underfrequency and undervoltage trips to prevent spurious reactor trips from momentary electrical power transients.

For undervoltage, the delay is set so that the time required for a signal to reach the reactor trip breakers following the simultaneous trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds.

For underfrequency, the delay is set so that the time required for a signal to reach the reactor trip breakers after the underfrequency trip set point is reached shall not exceed L.6 seconds.

SEQUOYAH - UNIT 1 B 2-6 Revised 03/18/87

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. REACTIVITY CONTROL SYSTEMS I

f ROD OROP TIME I

LIMITING CONDITION FOR OPERATION 3.1. 3. 4 The individual full length (shutdown and control) rod drop time from the fully withdrawn position # shall be less than or equal to 2.7 seconds from l

beginning of decay of stationary gripper coil voltage to dashpot entry with; 5-T,yg greater than or equal to 541'F, and a.

b.

All reactor coolant pumps operating.

APPLICABILITY:

MODES 1 and 2 ACTION:

a.

With the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b.

With the rod drop times within limits but determined with 3 reactor coolant pumps operating, operation may proceed provided THERMAL POWER is restricted to less than or equal to 71% of RATED THERMAL POWER SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full length rods shall~be demonstrated throup,h measurement prior to reactor criticality:

a.

For all rods following each removal of'the reactor vessel head' b.

For specifically affected individual rods following any main-tenance on or modification to-the control rod drive system which could affect the drop time of those specific rods, and c.

At least once per 18 months.

Fully withdrawn shall be the condition where shutdown and control banks are at a position within the interval of > 222 and 5,231 steps withdrawn, inclusive.

SEQUOYAH - UNIT 1 3/4 1-19 Amenoment No. 108, 138

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PO';!ER DISTRIBUTION LIMITS 3/4.'2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR i

LIMITING CONDITION FOR OPERATION i

i 3.2.3 The Nuclear Enthalpy Hot Channel Factor, F H, shall be limited by the following relationship:

Where:

l a.

F g i 1.55 [1.0 + 0.3 (1.0-P)]

THERMAL POWER b*

P

=

RATED THERMAL POWER '

i A_PPMCABILITY:

MODE 1 i

P ACTION:

With Fh exceeding its limit:

a.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to 1 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , l

b.

Demonstrate thru in-core mapping that F is within its limit g

within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER i

to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and c.

Identify and correct the cause of the out of limit condition prior

,to increasing THERMAL POWER above the reduced limit required by a.

or b above; subsequent POWER OPERATION may proceed provided that Ffg is demonstrated through in-core mapping to be within its limit

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at a nominal 50% of RATED THERMAL POWER prior to' exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within.24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% of greater RATED THERMAL POWER.

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'SEQUOYAH - UNIT 1 3/4 2-10 Amendment No.19,138-

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'.; POMER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F shall be determined to be within its limit by using the movable H

incore detectors to obtain a power distribution map:

Prior to operation above 75% of RATED THERMAL POWER after each fuel a.

loading, and b.

At least once per 31 Effective Full Power Days, c.

The measured F shall'be increased by 4% for measurement g

uncertainty.

SEQUOYAH - UNIT 1 3/4 2-11 Amendment'No. 138

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P0 DER DISTRIBUTION LIMITS 3/4.2.4 QUADRANT POWER TILT RATIO LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed 1.02.

APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER

ACTION:

With the QUADRANT POWER TILT RATIO determined to exceed 1.02 a.

but less than or equal to 1.09:

1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until:

a)

Either the QUADRANT TILT RATIO is reduced to within its limit, or b)

THERMAL POWER is reduced to less than 50% of RATED THERMAL

POWER, 2.

Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

a)

Either reduce the QUADRANT POWER TILT RATIO to within its limit, or b)

Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for each 1% of indicated QUADRANT POWER TILT RATIO in excess of 1.0 and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , 3.

Verify that the QUADRANT POWER TILT. RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL.

POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

"See Special Test Exception 3.10.2.

SEQUOYAH - UNIT 1 3/4 2-12 Amendment No. 138

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TOWER DISTRIBUTION LIMITS ACTION::l(Continue'd)

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.b.

- With the QUADRANT POWER TILT RATIO determined to exceed 1.09-due to 1

- misalignment of_either a shutdown or contro1~ rod:

i 1.

Calculate the QUADRANT. POWER TILT RATIO at least once per hour f

until:

a).

Either the QUADRANT POWER. TILT RATIO-ir' reduced to within its limit, or -

j i-b)

THERMAL POWER is'. reduced to less than 50%'of RATED THERMAL POWER..

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2..

. Reduce THERMAL POWER at least 3% from RATED-THERMAL POWER for-each 1% of ; indicated QUADRANT POWER TILT RATIO in excess of-1.0,-within 30 minutes, 3.

Verify that the QUADRANT POWER TILT RATIO is within its-limit-within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to-less than 50% of RATED' THERMAL' POWER within the next'2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High trip Setpoints-to less than or equal to 55% of-RATED THERMAL POWER within the

-next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.

Identify and correct the cause of the,out of limit' condition prior'to increasing THERMAL POWER;' subsequent POWER OPERATION above 50% of RATED THERMAL POWER may. proceed provided that-the QUADRANT POWER TILT RATIO is verified within;itsL11mit at?.least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable'at 95%

or greater RATED THERMAL POWER.

With-the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to t

causes other than the misalignment of either a shutdown'or control' rod:

1.

Calculate the QUADRANT POWER TILT. RATIO at least once per hour until:

a)

Either the QUADRANT POWER TILT-RADIO is reduced to within l

its limit, or J

b)

THERMAL POWER is reduced to less than 50% of' RATED: THERMAL j

POWER.

lSEQUOYAH - UNIT 1 3/4 2-13 Amendment No. 138 i,~'

4

.L..

- f

..f.s p.f l

\\

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POWER DISTRIBUTION' LIMITS-l l-ACTION:

(Continued)~

-- 2.

Reduce' THERMAL POWER lto less than 50% of RATEDETHERMAL: POWER-l l

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High

-Trip Setpoints to.less than or equal-to 55% of RATED THERMAL-

' POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3.

Identify and correct.the cause of..the.out of limit condition

prior to increasing THERMAL. POWER; subsequent POWER'0PERATION above 50% of RATED THERMAL POWER may proceed provided.that the QUADRANT POWER TILT. RATIO!is verified within its limit atlleast i

once per hour for.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or;until verifled at 95% or greater RATED THERMAL POWER c d.

WiththeindicatedQUADRANTPOWERTILTRATIOnotconfirmehas:

required by. Surveillance' Requirement 4.2.4.2.Lreduce THERMAL POWER-to less.than 75 percent: RATED; THERMAL POWER within.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

L With the QUADRANT POWER TILT RATIO not monitore'd as required by-

}

e.

Surveillan:e Requirement 4.2.4.1, reduce; THERMAL: POWER to less than 50 percent. of RATED ; THERMAL POWER-within-the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

f.

The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE'JUIREMENTS 4.2.4.1.The QUADRANT POWER TILT RATIO shall be determined to be'within the' limit above. 50% of RATED' THERMAL POWER by:-

a.

Calculating the ratio ~ at _least once per,7 days.'when _the alarm is OPERABLE.

b.

' Calculating the ratio at least'onceiper -12 hoursi during steady state operation when the' alarm is.. inoperable.

j 4.2.4.2' The QUADRANT POWER TILT RATIO shall be determined'to be within.the limit when above 75' percent of RATED THERMAL' POWER with one _ Power Range ~

m Channel inoperable by using the movable incore detectors;to confirm that the normalized symmetric power distribution,~obtained from-the 4l pairs of symmetric thimble locations or from performance of a full core map, is. consistent with the indicated QUADRANT POWER TILT. RATIO at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , a

I

?

SEQUOYAH - UNIT 1 3/4-.2-14 Amendment'No. 135, 138.

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.' POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS 1

LIMITING CONDITION FOR OPERATION l

3.2.5 The following DNB related parameters shall be' maintained within the limits shown on Table 3.2-1:

Reactor' Coolant System (RCS) T,yg.

a.-

l b.

Pressurizer Pressure c.

RCS Total Flow Rate APPLICABILITY:

MODE 1 ACTION:

With any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 'or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

i SURVEILLANCE REQUIREMENTS l

i l-4.2.5.1 Each of the parameters of Table 3.2-1 shall be verified to be within l

I their limits at least once per 12 ' hours.

4.2.5.2 The RCS total flow rate shall be determined by measurement at least once per 18 months.

4.2.5.3 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

l SEQUOYAH - UNIT 1 3/4 ?-15 Amendment No. 138

N TABLE 3.~2-1

. 'e DNB PARAMETERS i

LIMITS L-4 Loops In1 t

PARAMETER Operation ReactorCoolant;SystemT,yg

< 583*F-l 2220' psia

li Pressurizer Pressure 378400 gpm#

Reactor Coolant System Total Flow i

i P

Limit notLapplicable during:either a THERMAL POWER' ramp in excess of. 5%' RATED:

THERMAL; POWER-per minutefor a THERMAL POWER. step _in excess of-10% RATED.

THERMAL POWER, physics test..or performance of-surveillance requirement 4.1.l'3.b.

Includes--a 3.5% flow measurement-uncertain'.y.-

q i

L f

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N

. j

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-SEQUOYAH - UNIT 1 3/4 2-16 Amendment.No. 41.138 i

$ ' O A g

O

'3/4.2-POWER DISTRIBUTION' LIMITS-BASES The specifications of this section provide assurance'of fuel integrity.

' during Condition I (Normal Operation) and II (Incidents of Moderate Frequency)'

events by:

(a). maintaining the calculated DNBR in the core.at or above design during normal operation and in short term transients,. and (b)' limiting the fission gas' release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. =In-addition, limiting the peak linear

. power density during Condition I events provides assurance that the-initial-conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of-2200*Ftis not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

F (Z)

Heat Flux Hot Channel Factor, is defined as the maximum local-9 heat flux on the surface of a fuel rod at core elevation,Z divided by the average-fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.'

N

-F Nuclear Enthalpy Rise Hot Channel Factorz is defined as the ratio ~ of the iNegral of linear power along,the rod with the highest integrated power to the avera0e rod power.

3/4.2.1 AXIALFLUXDIFFERENCE(AF0J The limits on AXIAL FLUX DIFFERENCE' assure that the F '(Z) upper bound

-d 9

envelope of 2.32 times the normalized axial peaking factor 1s not-exceeded during either normal operation or in the event of ~ xenon redistribution follow-ing power changes.

Provisions for' monitoring the AFD on an: automatic basis'are. derived from-the plant process computer through the AFD Monitor Alarm. The computer deter-i mines the one minute average of each of the OPERABLE excore detector outputs.

and provides an alarm message immediately if the AFD for at least 2:of 4 or.2 of 3 OPERABLE excore channels are outside the allowed'al-Power operating space

' j and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER.

j 4

3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS j

The limits on the heat flux hot channel factor and the nuclear enthalpy rise hot channel factor' ensure that 1) the design : limits on peak local power density and minimum DNBR are not exceeded and'2)'in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS.

i acceptance criteria limit, i

SEQUOYAH - UNIT 1 B 3/4 2-1 Amendment No.19,138 3

Revised: 05/01/90 1

+

.a.

l

., ~.-

' POWER DISTRIBUTION LIMITS JASES Each of.these hot channel factors'is measurable but will normally only be.

determined periodical.ly as. specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance.is sufficient to insure that the' limits are maintained provided:

a.

Control rods in a single group move together with no individual rod.

insertion differing by more than + 13 steps.from the group demand position.

' b.'

Control rod groups are sequenced with overlapping groups as described' in Specification 3.1.3.6.

c.

The control rod insertion limits of Specifications 3.1.3.5 and' 3.1.3.6 are maintained.

d.

The axial power distribution, expressed in terms of AXIAL FLUX

'0!FFERENCE, is maintained within the limits..

N The F as a function of THERMAL POWER allows changes'in the radial g

power shape for all permissible rod insertion limits.

F g_will be maintained within its limits provided conditions a thru d above, are maintained.

When an F measurement is taken,'both experimental error.and manufacturing 0

~

tolerance must be allowed for.

The 5% is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3% is the appropriate allowance for manufacturing tolerance.

When an F is measured, experimental error must be allowed for and 4%'is.

g l

the appropriate allowance for a' full core map.taken with the incore detection The specified limit for Fh also contains an 8% allowance for system.

uncertainties which mean that normal operation will result in-F H'1 1.55/1.08.

The 8%. allowance is based on the following considerations.

j a.

abnormal perturbatios in the radial power' shape, such as from rod misalignment, effect F more directly than F.

H q

l b.

although rod movement has a direct influence upon limiting F to q

within its limit, such control is not readily available to. limit-

}

F g, and j

c.

errors in prediction for control power shape detected during startup i

physics test can be compensated for in F by restricting axial flux q

l distribution.

This compensation for F is less readily available.

H i

l SEQUOYAH - UNIT 1 B 3/4 2-2 Amendment No.19,' 138 i

s..

.,

o

-, POWER DISTRIBUTION LIMITS e

BASES Margin has been retaineo Fuel rod bowing reduces the value of DNB ratio.between the DN f

correlation limit (1.17) to comp.etely offset the rod bow penalty, l

l l

The applicable value or rod bow penalty is referenced in the FSAR.

Margin in excess of the rod bow penalty is available for plant design flexibility.

(z)ismeasuredperiodicallyandincreasedbya The hot channel factor F cycleandheightdependentpoherfactor,W(z),to.provideassurancethatthe limit on the hot channel factor, F (z), is met.

l of normal operation transients and was determined from expected power con 0

The W(z) maneuvers over the full range of burnup conditions in the co per Specification 6.9.1.14.

'3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt _ ratio limit assures that the radial power dist bution satisfies-the design values used in the power capability analysis.

Radial power distribution measurements are made during startup-testing periodically during power operation.

The two hour time allowance for operation with a tilt condition greater-i than 1.02 but less than 1.09 is provided to allow' identification and cor-In th rection of a dropped or misaligned rod.

is reinstated by reducing correct the tilt, the margin for uncertainty on Fnthe power b ilt in excess of 1.0.

3/4.2.5 DNB PARAMETERS The limits on the ONB related parameters assure that each of the para-tion meters are maintained within the normal steady state envelope of operaThe limit

~

assumed in the transient and accident analyses.

td with the initial FSAR assumptions and have been ana analysis ONBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrum hi readout is sufficient to ensure that the parameters are. restored within t e r limits following load changes and other expected transient operation.

' Amendment No.19,138 B 3/4 2-4 I

SEQUOYAH - UNIT 2

'l I

4

+.

3.

3/4.4 REACTOR COOLANT SYSTEM BASES i

,3 /4. 4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all reactor coolant loops in opera-tion, and maintain DNBR above-the-safety analysis DNBR limit during all normal

[

In_ MODES 1 and 2 with one reactor coolant operations and anticipated transients.

loop not in operation _this specification requires that the plant be in at least HOT STAN0BY within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

-In MODE 3, two reactor coolant loops provide sufficient-heat removal capability for removing core decay heat even in the event of a bank withdrawal' accident; however, a single reactor coolant loop provides sufficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by-r Single failure considerations-l 4

opening the Reactor Trip System breakers.

require that two loops be OPERABLE at all times.

q l

In'M00E 4, a single reactor coolant loop or residual _ heat removal (RHR) i loop provides sufficient heat removal capability for removing decay heat; but' single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification.

requires two RHR loops to be OPERABLE.

In MODE 5, single failure. considerations require that two RHR loops be

~

OPERABLE.

The operation of one Reactor. Coolant Pump lor one RHR' pump provides adequate-i flow to ensure mixing, prevent stratification and produce gradual' reactivity changes during boron concentration reductions in the Reactor; Coolant System.

The reactivity-change rate associated with boron reduction will, therefore, be q

within the capability of operator recognition and control.

3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES-The pressurizer code safety valves operate to prevent the RCS froin being Each safety valve.is designed pressurized above its Safety Limit of 2735 psig.lbs per hour of saturated steam to relieve 420,000 The relief capacity of a single safety valve is adequate to relieve any over-In the event that no pressure condition which could occur during shutdown.

i B 3/4 4-1 Amendment No. 12, 84, 138 SEQUOYAH - UNIT 1

.

ML20043A872

Text

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