ML18011A134

Proposed Tech Spec Re Cycle 6 Fuel Transition

ML18011A134
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Site:	Harris
Issue date:	08/27/1993
From:	CAROLINA POWER & LIGHT CO.
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NUDOCS 9309070281
Download: ML18011A134 (30)

Text

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ENCLOSURE 5

SHEARON HARRIS NUCLEAR POWER PLANT NRC DOCKET NO. 50-400/LICENSE NO. NPF-63 REQUEST FOR LICENSE AMENDMENT CYCLE 6 FUEL TRANSITION - ADDITIONALCHANGES TECHNICAL SPECIFICATION PAGES 9'309070281 930827 PDR ADOCK 05000400 P

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TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 1:

(Continued)

K3 pl 5go. 9

< 588.8 F (Nominal T at RATED THERHAL POWER);

0.001072I ps ig >

Pressurizer

pressure, psig, 2235 psig (Nominal RCS operating pressure);

Laplace transform operator, s

i and fl (dl) is a function of the indicated difference between. top and bottom detectors of the power"range neutron ion chambers>

with gains to be selected based on measured instrument response during plant startup tests such that:

/2.o (1)

For qt - qb between -21.6X and

~, fl (AI) = 0, where qt and qb are percent RATED THERHAL POWER in the top and bottom halves of the core respectively, and qt + qb is total THERHAL POWER in percent of RATED THERHAL POWER; o

NOTE 2:

(2)

For each percent be automatically (3)

For each percent be automatically The channel's maximum than 2.1X AT span.

that the magnitude of qt " qb exceeds

-21.6Z, the hT Trip Setpoint shall reduced by 2.36X of its value at RATED THERMAL POWERI and 2oO that the magnitude of qt qb exceeds

+ 6., the AT Trip Setpoint shall reduced by 1.57X of its value at RATED THERMAL POWER.

Trip Setpoint, shall not exceed its computed Trip Setpoint by more

TABLE 2.2-1 (Continued)

TABLE NOTATIONS NOTE 3!

(Continued)

K6 T

gl 0.002/'F for T > T" and K6 0 for T < T",

As defined in Note 1, Indicated T

at RATED THERHAL POWER (Calibration temperature for AT

~

instrumentation,

< 588. 'F),

iso.'3 As defined in Note 1, an NOTE 4 f2(AI) 0 for all AI~

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.3X hT span.

NOTE 5:

The sensor error for temperature is 1.9 and 1.1 for pressure.

= NOTE 6:

The sensor error for steam flow is 0.9, for feed flow is 1.5, and for steam pressure is 0.75.

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2.1 SAFETY LiMITS BASES 2.1.1 REACTOR CORE 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 prevented 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 (DNB) 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 related to DNB.

This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

The Local DNB heat flux ratio (DNBR) defined as the ratio of the heat flux that would cause DNB at a particular core location to the local x i icativ of the t

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4~c 5le~nS R<k a~A esxgn asks is as ol ows:

there must e a e st a 95 percent probability that the minimum DNBR of the limiting rod during Condition I and

'II events is greater than or equal to the DNBR limit of the DNB correlation exng used he WRB"1 or WRB-2 correlatio c

i.n).

The l

correlation DNBR limit is established base on e entire app aca Le experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit 7

c Qr vJesS gh s

In meet g this design

asis, ncertai ties in plant perat ng par eters nuclea and t rmaL p

ameter

, and f el fabr cation param ters a

e cons'dered stati ically such t t ther is at east a

5 pere nt pr abili y with 95 p

cent c nfiden leveL that t minim DNBR or th Limit'ng rod is gre er tha or eq 1 to t e DNBR imit.

he unc rtain ies in the ab ve pl t

pa meters are us to de ermine he pla DNBR ncert inty.

This D

BR u certain y, corn ned wi h the rrelat' DNBR limit estab ishes desi BR val e whic must b

met i plant fety a alyse usin values of in t

paramet rs wit ut unc rtaint't In additio

, mar in has been intai d in the de ign by ecting safety analysi DNBR 1'mits n perf rming afety analy es.

The curves of Figure 2.1-1 show the Loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

SHEARON HARRIS - UNIT 1 B 2-1 Amendment No. J,

2. 1 SAFETY LIMITS BASES

2. l. 1 REACTOR CORE (Con t inued )

f s ph s+i&Q These curves are based on an enthalpy hot channel factor, F

, specified in the CORE OPERATINC LIMITS REPORT (COLR) and a r e en e

o in w t a

p k o 5 ~o axial power shape.

An al.lowance is included for an increase in ca culated F

at reduced power based on the expression'.

4H F

=

F

[1 + PF (1"P))

4H 4H 4H Where P is the fraction of RATED THERMAL POWER, F

=

F limit at RATED THERMAL POWER specified in the

COLR, and 4H 4H PF

= Power Factor Multiplier for F H specified in the COLR.

4H 4H 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 fl (4I) function of the Overtemperature trip.

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Over-temperature 4T trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

SHEARON HARRIS - UNIT 1 B 2-la Amendment No.

1

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LIMITING SAFETY SYSTEM SETTINGS BASES Power Ran e, Neutron Flux (Continued)

The Low Setpoint trip may be manually blocked above P-10 (a pover LeveL of approximately LOX oE RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint

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Power Ran e, Neutron Flux, Hi h Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

SpeciEicalLy, this trip complements the Power Range Neutron FLux High and Lov trips to ensure that the criteria are met Eor rod ejection from mid-power.

The Pover Range Negative Rate trip provides protection for controL rod drop accidents.

At high powe'r a single or multiple rod drop accident could cause l.ocal flux peaking which could cause an unconservative local DNBR to exist.

The Power Ran e Ne ative Rate tri vill revent this from occurrin b tri in the reactor.

No credit 'aken Eor ope ion oE the Power nge Negative Rate p

Eor those con rod drop acc's Eor which DNB vill be greater t the desi n

vaLue.

Intermediate and Source Ran e

Neutron FLux The Intermediate and Source

Range, Neutron FLux trips provide core protection during reactor startup to mitigate the consequences of an uncontrolled rod

'l.uster controL assembLy bank withdrawaL Erom a subcritical. condition.

These trips provide redundant protection to the Lov Setpoint trip of the Pover

Range, Neutron fLux channels.

The Source Range channels viLl. initiate a Reactor tr ip at about 10 counts per second unless manuaLLy blocked vhen P"6 becomes active.

The Intermediate Range channel.s vilL initiate a Reactor trip at a current level.

equivaLent to approximateLy 25X of RATED THERMAL POWER unLess manually blocked when P-10 becomes active.

Overtem erature 4T The Overtemperature 4T trip provides core protection to prevent DNB for alL combinations of pressure, power, coolant temperature, and axiaL po~er distribu-tion, provided that the transient is slov with respect to piping transit delays from the core to the temperature detectors (about 4 seconds);

and pressure is vithin the range between the Pressurizer High and Lov Pressure trips'he Set-point is automatically varied vith:

(1) coolant temperature to correct Eor temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping deLays from the core to the loop temperature detectors

, (2) pressurizer

pressure, and (3) axial pover distribution.

With normal axiaL pover distribution, this Reactor trip limit is alvays belov the core Safety Limit as shown in Figure 2.1"1.

IE axial peaks are greater than

design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automaticall.y reduced according to the notations in Table 2.2-1.

SHEARON HARRIS - UNIT 1 B 2-4 Amendment No.

LIMITING SAFETY SYSTEM SETTINGS BASKS The Overpower 4T trip provides assurance of'uel integrity (e.g.,

no fuel pellet melting and less than 1X cladding strain) under aLL possibLe overpower conditions, limits the required range for Overtemperature 4T trip, and provides a backup to the High Neutron Flux trip.

The Setpoint is automatically varied with:

( 1) coolant temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the alLowable heat generation rate (kW/ft) is not exceeded.

Overpo r 4T trap pr ides prote on to mate e the conseque of varxo size st m breaks as ported in P-9226, "

actor Core Res se to E

ssive Se dary Steam leases."

Pressurizer Pressure In each of the pressurizer pressure

channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against low pressure which could Lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint "trip is'automatically blocked by the loss o'f P"7 (a power Level of approximately LOX of RATED THERMAL POWER or turbine impulse chamber pressure at approximately LOZ of full power equivalent);

and on increasing power, automatically reinstated by P"7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer High Water LeveL trip is provided to prevent water relief through the pressurizer safety valves.

On decreasing power the Pressurizer High Water Level. trip is automatically blocked by the loss of P"7 (a power Level of approximately LOX of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10Z of full power equivalent);

and on increasing power,, automatically reinstated by P-7

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Reactor CooLant Flow The Reactor Coolant Low Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the Loss of one or more reactor cooLant pumps.

On increasing power above P-7 (a power Level of approximately 10Z of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10X of full power equivaLent),

an automatic Reactor trip will occur if the fLow in more than one Loop drops beLow 90.5Z of nominaL full loop flow.

Above P-8 SHEARON HARRIS - UNIT 1

8 2-5 Amendment No.

POWER DISTRIBUTION LIMITS 3/4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB"related parameters shall be maintained within the following limits:

584. l a.

Indicated Reactor Coolant System T

~ 'F after addition for instrument uncertainty, and b.

Indicated Pressuriaer Pressure

> 2185 psig* after subtraction for instrument uncertainty.

APPLICABILITY:

MODE lo ACTION:

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

SURVEILLANCE RE UIREMENTS 4.2.5 Each of the parameters shown in Specification 3.2.5 shall be verified to be within its limit at least once per 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />'.

""This limit is not applicable during either a Thermal Power Ramp in excess of

+5Z Rated Thermal Power per minute or a Thermal Power step change in excess of +10X Rated Thermal Power.

SHEARON HARRIS UNIT 1 3/4 2"14 Amendment No.

Vt,e p:2..Z zooo~

05 gg 60tph Qgpycg~ff~y Irzg TABLE 3.9-1 v 8'0

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<R Ingrain g +~ /egg

+ an ar eqnal do 0.9g an s ecrCiej ADMINISTRATIVE CONTROLS

> '+ P C~ LQ TO PREVENT DILUTION DURING REFUELING CP6L VALVE NOa (Ebasco Valve No.)

1CS-149 (CS-D121SN)

DESCRIPTION Reactor Makeup Water to CVCS Makeup Control System RE UIRED POSITION Lock closed; may be opened to permit makeup to Refueling Water Storage Tank provided valves 1CS-156 and 1CS-155 are maintained closed with their main control board control switches in "shut" position, and manual valves 1CS-274, 1CS-265 and 1CS-287 are locked closed.

1CS-510 (CS-D631SN)

Boric Acid Batch Tank Outlet Locked closed; may be opened provided the boron concentration of the boric acid batch tank > ~9~~ and valve 1CS-503 in etneed.

1CS-503 (CS-D251)

Reactor Makeup Water to Boric Acid Batch Tank Lock closed, may be opened provided valve 1CS-510 is closed.

1CS-93 (CS-D51SN) 1CS-320 (CS-D641SN)

Resin Sluice to CVCS Demineralizers r

Boron Recycle Evaporator Feed Pump to Charging/SI Pumps Lock closed.

Lock closed.

1CS-570 (CS-D575SH) 1CS"670 (CS-D599SN) 1CS-649 (CS-D198SH)

CVCS Letdown to Boron Thermal Regeneration System Reactor Makeup Water to Boron Thermal Regeneration System Resin Sluice to BTRS Demineralizers Closed with main control board control switch in "shut" position, and BTRS function selector switch maintained in "off" position',

no lock required.

Lock closed.

Lock closed.

1CS-98 (CS-D740SN)

Boron Thermal Regeneration System Bypass Opened with main control board control switch maintained in "open" position; no lock required.

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REACTIVITY CONTROL SYSTEMS BASES MODERATOR TEHPERATURE COEFFICIENT (Continued)

The most negativ HTC, value equ'lent to the most pos ive moderator densi y

coet ficient (M

),

was obtaine by incremental ly corr cting the MDC used the FSAR analyse to nominal op ating conditions.

Th e corrections invol (1) a conv sion of the M

used in the FSAR saf y analyses to its e

ivalent

HTC, bas on the rate change of moderator ensity with temperat re at RATED ERHAL POWER c

dir ions, and (2) sub acting from this va e the larg t di fferences in HTC observed betwe EOL, all rods with

awn, RATED TH HAL POWER con itions, and those mo adverse conditions moderator mperature an

pressure, rod insert n, axial power skew g,

and xenon concentratio that can occur in n mal operation and le to a significantly more negat' EOL MTC at RATED ERHAL POWER.

These orrections transform the HDC lue used in the FS safety analyses int the Negative MTC Lim'he 30 ppm surveillance l it MTC value represe s

a conservative HTC alue at a core condition of ppm equilibrium bor concentration, and 's ob ained by making co ections for burnup a

soluble boron to th Negative C Limit.

The Survei llance Requirements for measurement of the HTC at the beginning and near the end of the fuel cycle are adequate to confirm that the HTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 HINIHUH TEHPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551'F.

This limitation is required to ensure:

( 1) the moderator temperature coefficient is within its analyzed temperature

range, (2) the trip instrumentation is within its normal operating

range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum RTNDT temperature.

3/4.1.2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is available during each mode of facility operation.

The components required to perform this function include:

( 1) borated water sources, (2) charging/safety injection

pumps, (3) separate flow paths, (4) boric acid transfer

pumps, and (5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 350'F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.

The boration capa-bility of either flow path is sufficient to provide the required SHUTDOWN MARGIN as defined by Specification 3/4.1.1.2 after xenon decay and cooldown to 200'F.

The maximum expected boration capability requirement occurs at BOL SHEARON HARRIS UNIT 1

B 3/4 1-2 Amendment No.

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'POWER DISTRIBUTION LIMITS BASES UADRANT POWER TILT RATIO (Continued)

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles'he preferred sets of four sym-metric thimbles is a unique set of eight detector locations.

These locations are C-B, E-5> E-lli H-3, H-13, L-5, L-ll, N"8 ~ If other locations must be

used, a special report to HRC should be submitted within 30 days in accordance with 10CFR50.4.

3/4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady"state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the ini-tial FSAR assumptions and have b'een analytically demonstrated adequate to maintain a minimum DNBR that is equal to or greater than the design DNBR value throughout each analyzed transient.

The indicated T

value and the indic ted 'pressurizer pressure value are compared to analytical limits of

'F and 2185 psig, respectively, after an allo~ance for measurement uncertainty is included.

The 12-hour periodic surveillance of these parameters through instrument read" out is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

SHEARON HARRIS - UNIT 1 B 3/4 2-6 Amendment No. J3~

DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internaL pressure of 45.0 psig and a peak air temperature of 380'F.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The coze shall contain 157 fuel assemblies with each fueL assembly normally containing 264 fuel rods clad with Zircaloy"4 except that limited substitution of fuel rods by filler rods consisting of Zircaloy-4, stainless steeL, or by vacancies may be made in fuel assemblies if justified by a cycle-specific evaluation.

Should more than a total of 30 fuel rods or more than 10 fuel rods in any one assembly be replaced per refueling a Special Report describing the number of rods replaced will be submitted to the Comaission, pursuant to Specification 6.9.2, within 30 days after cycle startup.

Each fuel rod shall have a nominal active fuel length of 144 inches.

The initial core loading shall have a maximum enrichment of 3.5 weight percent U"235.

Reload fuel shall be similar in physical design to the initial core Loading and shall have a maximum enrichment of 5.0 weight percent U-235.

Fuel with enrichments greater than 4.20 weight percent U-235 shall contain sufficient integral burnable absorbers such that the requirement of Specification 5.6.1.a.2 is met ~

CONTROL ROD ASSEMBLIES 5.3.2 The core shall contain 52 shutdown and control rod assemblies.

The shutdown and zod assembLies shall contain a nominal 142 inches of absorber materiaL.

The nominal values of absorber material shall be 80X silver, 15Z indium, and 5Z cadmium, or 95Z hafnium with the remainder zirconium.

All control rods shall be clad with stainLess steel tubing.

5 '

REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:

a.

1n accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normaL degradation puzsuant to the applicable Surveillance Requirements, b.

For a pressure of 248$ psig, and c.

For a temperature of 650'F, except for the pressurizer which is 680'F VOLUHE 5.4.2 The total water and steam volume of the R

c or Coolant System is 9410

+ 100 cubic feet at a nominal T of

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~so.S SHEARON HARRIS - UNIT 1 5"6 Amendment No.

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0 NOTE 1:

(Continued)

TABLE 2.2-1 Continued TABLE NOTATIONS M

I CO T'

580.8'F (Nominal T, at RATED THERMAL POWER);

K>

=

0. 001072/ps ig; P

=

Pressurizer

pressure, psig; P'

2235 psig (Nominal RCS operating pressure);

S

=

Laplace transform operator, s';

and f> (h,I) is a function of the indicated difference between top and bottom detectors of the power-range neutron ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

I ~

O (1)

For q, -

q> between

-21.6% and +12.0%,

f> (6,1)

= 0, where q, and q> are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + q> is total THERMAL t

POWER in percent of RATED THERMAL POWER; (2)

For each percent that the magnitude of q, -

q> exceeds

-21.6%,

the hT Trip Setpoint shall be automatically reduced by 2.36% of its value at RATED THERMAL POWER; and (3)

For each percent that the magnitude of q, -

q> exceeds

+ 12.0%, the hT Trip Setpoint shall be automatically reduced by 1.57% of its value at RATED THERMAL POWER.

NOTE 2:

The channel's maximum Trip Setpoint shall not exceed its -computed Trip Setpoint by more than

2. 1% hT span.

CD TABLE 2.2-1 Continued TABLE NOTATIONS C

NOTE 3:

(Continued)

K6 0.002/'F for T > T" and K6 = 0 for T x T",

As defined in Note 1, Indicated Tgyg at RATED THERMAL POWER (Calibration temperature for bT instrumentat~ion, x 580.8'F),

As defined in Note 1, and PJ I

CD fz(GI)

=

0 for all 6,1.

NOTE 4:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2.3% hT span.

NOTE 5:

The sensor error for temperature is 1.9 and

1. 1 for pressure.

NOTE 6:

The sensor error for steam flow is 0.9, for feed flow is 1.5, and for steam pressure is 0.75.

1 l

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'.1 SAFETY LIMITS BASES

2. 1. 1 REACTOR CORE 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 prevented 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 (DNB) 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 related to DNB. This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and nonuniform heat flux distributions.

The local DNB heat flux ratio (DNBR) defined as the ratio of the heat flux that would cause DNB 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 probability that the minimum DNBR of the limiting rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (in this application, the HTP correlation for Siemens Fuel and the WRB-1 or WRB-2 correlation for Westinghouse Fuel).

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

The. curves of Figure 2. 1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature below which the calculated DNBR is no less than the design DNBR value or the average enthalpy at the vessel exit is less than the enthalpy of saturated liquid.

SHEARON HARRIS - UNIT 1

B 2-1 Amendment No.

l 2:1'AFETY LIMITS BASES

2. 1. 1 REACTOR CORE (Continued)

These curves are based on an enthalpy hot channel factor, Fz<, specified in the CORE OPERATING LIMITS REPORT (COLR) and a limiting axial power shape.

An allowance is included for an increase in calculated F>> at reduced power based on the expression:

F~~

=

Fgg

[I + PF~~ (1-P) ]

ATP Where P is the fraction of RATED THERMAL POWER, F><

F<< limit at RATED THERMAL POWER specified in the

COLR, and PF>>

= Power Factor Hultiplier for F>< specified in the COLR.

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>(EI) function of the Overtemperature trip.

When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature hT trips will reduce the Setpoints to provide protection consistent with core Safety Limits.

SHEARON HARRIS - UNIT 1

B 2-la

, Amendment No.

l

+

I

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'i LIHITING SAFETY SYSTE SETTINGS BASES Power Ran e

Neutron Flux Continued The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Ran e

Neutron Flux Hi h Rates The Power Range Positive Rate trip provides protection against rapid flux increases which are characteristic of a rupture of a control rod drive housing.

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

The Power Range Negative Rate trip provides protection for control rod drop accidents.

At high power a single or multiple rod drop accident could cause local flux peaking which could cause an unconservative local DNBR to exist.

The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.

Intermediate and Source Ran e

Neutron Flux The Intermediate and Source

Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncontrolled rod cluster control assembly bank withdrawal from a subcritical condition.

These trips provide redundant protection 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 Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

Over tern erature hT The Overtemperature hT 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 Pressurizer High and L'ow Pressure trips.

The Setpoint is automatically varied with: (1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors, (2) pressurizer

pressure, and (3) axial power distribution.

With normal axial power distribution, this Reactor trip limit is always 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.

SHEARON HARRIS - UNIT 1

B 2-4 Amendment No.

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LIMITING'SAFETY SYS SETTINGS BA'SES Over ower bT The Overpower hT trip provides assurance of fuel integrity (e.g.,

no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for Overtemperature 5T trip, and provides a backup to the High Neutron Flux trip.

The Setpoint is automatically varied with: (I) coolant temperature to correct for temperature induced changes in density and heat capacity of water, and (2) rate of change of temperature for dynamic compensation for piping delays from the core to the loop temperature detectors, to ensure that the allowable heat generation rate (kW/ft) is not exceeded.

Pressurizer Pressure In each of the pressurizer pressure

channels, there are two independent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is permitted.

The Low Setpoint trip protects against low pressure which could lead to DNB by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by the loss of P-7 (a power level of approximately 10% of RATED THERMAL POWER or turbine impulse chamber pressure at approximately 10% of full power equivalent);

and on increasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against system overpressure.

Pressurizer Water Level The Pressurizer High Water Level trip is provided to prevent water relief through the pressurizer safety valves.

On decreasing power the Pressurizer High Water Level trip is automatically blocked by the loss of P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent);

and on increasing power, automatically reinstated by P-.7.

Reactor Coolant Flow The Reactor Coolant Low Flow trips provide core protection to prevent DNB by mitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10% of full power equivalent),

an automatic Reactor trip will occur if the flow in more than one loop drops below 90.5% of nominal full loop flow.

Above P-8 SHEARON HARRIS - UNIT I B 2-5 Amendment No.

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'OWER DISTRIBUTION L TS 3 4.2.5 DNB PARAMETERS LIMITING CONDITION FOR OPERATION 3.2.5 The following DNB-related parameters shall be maintained within the following limits:

a.

Indicated Reactor Coolant System Tgy M 586 1

F after addition for instrument uncertainty, and b.

Indicated Pressurizer Pressure a

2185 psig after subtraction for instrument uncertainty.

APPLICABILITY:

MODE 1.

ACTION:

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

SURVEILLANCE REQUIREMENTS 4.2.5 Each of the parameters shown in Specification 3.2.5 shall be verified to be within its limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

This limit is,not applicable during either a Thermal Power Ramp in excess of +5% Rated Thermal Power per minute or a Thermal Power step change in excess of F10% Rated Thermal Power.

SHEARON HARRIS - UNIT 1

3/4 2-14 Amendment No.

TABLE 3.9-1 ADMINISTRATIVE CONTROLS TO PREVENT DILUTION DURING REFUELING CP&L VALVE NO.

Ebasco Valve No.

1CS-149 (CS-D121SN)

DESCRI PT ION Reactor Makeup Water to CVCS Makeup Control System RE UIRED POSITION Lock closed; may be opened to permit makeup to Refueling Water Storage Tank provided valves 1CS-156 and 1CS-155 are maintained closed, with their main control board control switches in "shut" position, and manual valves 1CS-274, 1CS-265 and ICS-287 are locked closed.

1CS-510 (CS-D631SN)

Bori c Acid Batch Tank Outl et Locked closed; may be opened provided the boron concentration of the boric acid batch tank a the greater of 2000 ppm or the boron concentration required to maintain off less than or equal to 0.95, as specified in the COLR and valve 1CS-503 is closed.

.1CS-503 (CS-D251) 1CS-93 (CS-D51SN) 1CS-320 (CS-D641SN)

Reactor Makeup Water to Boric Acid Batch Tank Lock closed, may be opened provided valve 1CS-510 is closed.

Resin Sluice to CVCS Demineralizers Lock closed.

Boron, Recycle Evaporator Feed Pump to Lock closed.

Charging/SI Pumps

TABLE 3.9-1 (Continued)

ADMINISTRATIVE CONTROLS TO PREVENT DILUTION DURING REFUELING CPKL VALVE NO.

Ebasco Valve No.

DESCRIPTION RE UIRED POSITION 1CS-570 (CS-D575SN) 1CS-670 (CS-0599SN) 1CS-649 (CS-D198SN) 1CS-98 (CS-D740SN)

CVCS Letdown to Boron Thermal Regeneration System Reactor Hakeup Water to Boron Thermal Regeneration System Resin Sluice to BTRS Demineralizers Closed with main control board control switch in "shut" position, and BTRS function selector switch maintained in "off" position; no lock required.

Lock closed.

Lock closed.

Boron Thermal Regeneration System Bypass Opened with main control board control switch maintained in "open" position; no lock required.

a

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I REACTIVITY CONTROL S

STEMS BASES MODERATOR TEMPERATURE COEFFICIENT Continued The Surveillance Requirements for measurement of the HTC at the beginning and near the end of the fuel cycle are adequate to confirm that the HTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3 4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 551'F.

This limitation is required to ensure:

(1) the moderator temperature coefficient is within its analyzed temperature

range, (2) the trip instrumentation is within its normal operating

range, (3) the pressurizer is capable of being in an OPERABLE status with a steam bubble, and (4) the reactor vessel is above its minimum Ropy temperature.

3 4. 1. 2 BORATION SYSTEMS The Boron Injection System ensures that negative reactivity control is

- available during each mode of facility operation.

The components required to perform this function include:

(1) borated water sources, (2) charging/safety injection pumps, (3) separate flow paths, (4) boric acid transfer

pumps, and (5) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above 350'F, a minimum of two boron injection flow paths are required to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable.

The boration capability of either flow path is sufficient to provide the required SHUTDOWN MARGIN as defined by Specification 3/4. 1. 1.2 after xenon decay and cooldown to 200'F.

The maximum expected boration capability requirement occurs at BOL SHEARON HARRIS - UNIT 1 B 3/4 1-2 Amendment No.

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'OWER DISTRIBUTION L11 TS BASES UADRANT POWER TILT RATIO Continued For purposes of monitoring QUADRANT 'POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of four symmetric thimbles.

The preferred sets of four symmetric thimbles is a unique set of eight detector locations.

These locations are C-8, E-5, E-ll, H-3, H-13, L-5, L-11, N-8. If other locations must be used, a special report to NRC should be submitted within 30 days in accordance with 10CFR50.4.

3 4.2.5 DNB PARAMETERS The limits on the DNB-related parameters assure that each of the parameters are maintained within the normal steady-state envelope of operation assumed in the transient and accident analyses.

The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR that is equal to or greater than the design DNBR value throughout each analyzed transient.

The indicated T, value and the indicated pressurizer pressure value are compared to analytical limits of 586. 1'F and 2185 psig, respectively, after an allowance for measurement uncertainty is included.

The 12-hour periodic surveillance of these parameters through instrument read-out is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

SHEARON HARRIS - UNIT 1

B 3/4 2-6 Amendment No.

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g"=SIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The containment building is designed and shall be maintained for a maximum internal pressure of 45.0 psig and a peak air temperature of 380'F.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3. 1 The core shall contain 157 fuel assemblies with each fuel assembly normally containing 264 fuel rods clad with Zircaloy-4 except that limited substitution of fuel rods by filler rods consisting of Zircaloy-4, stainless

steel, or by vacancies may be made in fuel assemblies if justified by a

cycle-specific evaluation.

Should more than a total of 30 fuel rods or more than 10 fuel rods in any one assembly be replaced per refueling a Special Report describing the number of rods replaced will be submitted to the Commission, pursuant to Specification 6.9.2, within 30 days after cycle startup.

Each fuel rod shall have a nominal active fuel length, of 144 inches.

The initial core loading shall have a maximum enrichment of 3.5 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 5.0 weight percent U-235.

Fuel with enrichments greater than 4.20 weight percent U-235 shall contain sufficient integral burnable absorbers such that the requirement of Specification 5.6. l.a.2 is met.

CONTROL ROD ASSEMBLIES E

5.3.2 The core shall contain 52 shutdown and control rod assemblies.

The shutdown and rod assemblies shall contain a nominal 142 inches of absorber material.

The nominal values of absorber material shall be 80% silver, 15% indium, and 5% cadmium, or 95% hafnium with the remainder zirconium.

All control rods shall be clad with stainless steel tubing.

5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4. 1 The Reactor Coolant System is designed and shall be maintained:

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b.

C.

In accordance with the Code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, For a pressure of 2485 psig, and For a temperature of 650'F, except for the pressurizer which is 6800F VOLUME 5.4.2 The total water and steam volume of the Reactor Coolant System is 9410 z

100 cubic feet at a nominal T,pg of 580.8'F.

SHEARON HARRIS - UNIT 1

5-6 Amendment No.