ML17214A412

Proposed Tech Spec Changes Re Peaking Factors

ML17214A412
Person / Time
Site:	Saint Lucie
Issue date:	08/25/1983
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML17214A413	List: L-83-460, Application to Amend License DPR-67,changing Tech Specs to Modify Peaking Factors Consistent W/Current Methodology & Unit 2 Tech Specs & Deleting Load Follow Power Peaking Penalties.Safety Evaluation Encl ML17214A412
References
NUDOCS 8308300655
Download: ML17214A412 (14)

Text

0 DEFINITIONS SECTION PAGE 1.0 D. FI hIT10,'IS Defin d Teens..........

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Re portabl e Occurrence......,............

Cont.ainment Vessel Integr)ty...........

Channel Ca 1 i bra tion....

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'Functional Test Core Alteration.

Shutdown i'Jargin.,

Identi ied Leakage'..

P Unidentified Leakage..............

Pressure Boundary Leakage........

Contr oil ed Leakage Azimuthal Power Tilt.....;........

Dose 'Equivalent I-131..

E - hverage 'Disintegration.Energy

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LUCIE - UNIT 1 ST

'":,l

":. PDR ADocg,., 05000335, p~ ""."',."PDR L

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DEFINITIONS REACTOR TRIP SYSTEM

RESPONSE

TIYiE 1.26 The REACTOR TRIP SYSTEM

RESPONSE

TINE shall be the time interval from when the monitored parameter exceeds its trip setpoint at the 'channel sensor until electrical power is interrupted to the CEA drive mechanism.

ENGINEERED SAFETY FEATURE

RESPONSE

TINE 1.27 The ENGINEERED SAFETY FEATURE

RESPONSE

TI)'iE shall be that time

,interval from when the monitored parameter exceeds its ESF actuation setpoint at the channel sensor until the ESF equipment is capable of performing its safety function (i.e., the valves travel to their required positions, pump discharge pressures.

reach their required values, etc.).

Times shall include diesel generator starting and sequence loading delays where applicable.

PHYSICS TESTS 1.28 PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation and

1) described in Chapter 14.0 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Comm'ssion.

UHRODDED INTEGRATED RADIAL PEAKING FACTOR -

F 1.29 The UHRODDED INTEGRATED RADIAL PEAKING FACTOR is the ratio of'he

,peak pin power to the average pin power in an unrodded

core, excluding tilt.

ST.

LUCIE - UNIT 1

1-6

POWER DISTRI BUTION LIMITS SURVEILLANCE REQUIR=MENTS Continued C.

Verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2, where 100 percent of,maximum allowable power represents the maximum THERMAL POWER allowed by the following expression:

Mx N where:

l.

M is the maximum allowable THERMAL POWFR level for the existing Reactor Coolant Pump combination.

2.

N is the maximum allowable fraction of RATED THERMAL POWER as determined by the F

curve of Figure 3.2-3.

'y 4.2.1.4 Incore Detector Monitorin System - The incore detector moni-toring system may.'be used for monitoring the core power distribution by verifying that the incore detector Local Power Density alarms:

a.

b.

Are adjusted to satisfy the requirements of the core power distribution map which shall be updated at least once per 31 days of accumulated operation in MODE l.

Have their alarm setpoint adjusted to less than or equal to the limits shown on Figure 3.2-1 when the following factors are appropriately included in the setting of these alarms:

I.

A measurement-calculational uncertainty factor of 1. 07, --

An engineering uncertainty factor of 1.03, 3-A linear heat rate uncertainty factor of 1.01 due to axial fuel densification and thermal expansion, and A THERMAL POWER measurement uncertainty factor of 1.02.

A g If thecore system becomes inoperable, reduce power to M x N within 4 hour/and monitor linear heat rate in accordance with Specification 4.2.1.

ST.

LUCIE UNIT 1 '/4 2-2

THIS PAGE INTENTIONALLY BLANK 3/4 2-5

POWER DISTRIBUTION LIMITS 3/4.2.2 TOTAL PLANAR RADIAL PEAKING FACTORS - F LIMITING CONDITION FOR OPERATION 3.g.2 The calculated value of Fx shall be limited to

< l.gO' xy APPLICABILITY:

MODE 1".

ACTiDH:

Mith Fxy a.

> l.70 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

Reduce THERMAL POMER to bring the combination of THERMAL,POWER a'nd F

to within the limits of Figure 3.2-3 and withdraw the full length CEAs to or beyond the Long Term.Steady State Insertion Limits of Specification

3. 1.3.6; or b.

Be in HOT STANDBY.

SURVEILLANCE REQUIREMENTS 4.2.2. 1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F

shall be calculated by the expression F

=

F (1+T ) when T.

T xy xy q

. F is calculated with a non-full core power distribution analysis code and xy xy shall be calculated as F

=

F when calculations are performed with a full core power distribution analysis code.

F shall be determined to be within xy its limit at the following intervals:

a.

Prior to operation above 7N of RATED THERMAL POMER after each fuel

loading, b.

At least once per 31 days of accumulated operation in MODE 1, and c.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the AZIMUTHAL POWER TILT (T ) is

> 0.03.

See Special Test Exception 3.10.2.

'I ST.

LUCIE - UNIT I

3/4 2-5

POMER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) 4.2.2.3 F

shall be determined each time a calculation of F is required by using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing reactor coolant pump combination.

This determination shall be limited to core planes between l5" and 85~ of full core height and shall exclude regions influenced by grid effects.

4 ~ 2.2.4 T

shall be determined each time a calculation of F is made using q

xy a non full core power distribution analysis code.

The value of T used in this case to determine F

shall be the measured value of T xy

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q'T.

LUCIE - UNIT 1

3/4 2-7

POWER DISTRIBUTION LIMITS TOTAL INTEGRATED RADIAL PEAKING FACTOR -

Fr LIYIITING CONDITION FOR, OPERATION T

3.2.3 The calculated value of F, shall be limited to ( 1.70.

'PPLICABILITY:

MODE 1*.

ACTIDN:

With Fr > 1.70, within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> either:

'a 0 b.

Be in at least HOT STANDBY, or Reduce THERMAL POMER to bring the combination of THERMAL POWER and F

to within the limits of Figure 3.2-3 and withdraw the full-length r

CEAs to or beyond the Long Term Steady State Insertion Limits of Specification 3.1.3.6.

The THERMAL POMER limit determined from Figure 3.2-3 shall then be used to establish a revised upper THERt1AL POWER level limit on'Figure 3.2-4 (truncate Figure 3.2-4 at the allowable fractioh of RATED THERMAL POMER determined by Figure 3.2-3) and subsequent operation shall be maintained within the reduced acceptable operation region of Figure 3.2-4.

SURYEILLANCE REOUIREHENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F

shall be calculated by the expression F

=

F (1+T )

T T=

",when Fr is calculated with a non-.full core power distribution

'analysis code and shall be calculated as F

= F

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when r

r calculations are performed with a full core power distribution analysis code.

F shall be determined to be within its limit at the'ollowing intervals.

a.

Prior to operation above 7(C of RATED THERMAL POWER after each fuel loading.

b.

At least once per 31 days of accumulated operation in MODE 1, and c.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> if the AZIMUTHAL POWER TILT (T ) is

> 0.03.

See Special Test Exception 3.10.2.

ST.

LUCIE - UNIT, I'/4 2-9

POWER DISTRIBUTION LIMITS SURVEILLANCE REOUIREMENTS (Continued) r

4. 2. 3. 3 F

shall be determined each time a calculation of F

is required by r

using the incore detectors to obtain a power distribution map with all full length CEAs at or above the Long Term Steady State Insertion Limit for the existing reactor coolant pump combi'nation.,

4.2.3.4 T

shall be determined each. time a calculation of F is made using r

a non-full core power distribution analysis code.

The value of T used to T

~

q determine Fr in this case shall be the measured value of T ST.

LUCIE - UNIT I

3/4 2-10

3/4 2

POki. R DISTRIBUTION LIYiITS BASES 3/4.2.1 LINEAR HEAT RATE The limitation on linear heat rate ensures that in the ev nt of a LOCA, the peak temperature of the fuel cladding will not exceed 2200'F.

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System and the Incore Detector Monitoring

System, provide adequate monitoring of the core power distribution and are capable of verifying that the linear heat rate does not exceed its limits.

The Excore Detector Monitoring System performs this unction by continuously monitoring the AXIAL SHAPE INDEX with the OPERABLE quadrant symmetric excore neutron flux detectors and verifying that the AXIAL SHAPE INDEX is maintained within the allowable limits of Figure 3.2-2.

In conjunction with the use of the excore monitoring system and in establishing the AXIAL SHAPE INDEX limits, the following assumptions are made:

1) the CEA insertion limits of Specifications

3. 1. 3. 5 and 3. 1. 3. 6 are satisfied, ") the AZIMUTHAL POWER TILT restrictions of Specification 3.2.4 are satisfied, and 3) the TOTAL PLANAR RADIAL PEAKING FACTOR does not exceed the limits of Speci>ication 3.2.2.

The Incor e Detector Monitoring System continuously provides a

direct measure of the peaking factors ant the alarms which have been established for the individual incore detector segments ensure that the peak linear heat rates will be maintained within the allowable limits of Figure 3.2-1.

The setpoints for these alarms include allowances, set in the conservative directions,A~:

l) a measurement-calculational uncertainty factor.

of 1.07, Z) an engineering uncertainty factor of 1.03,

3) an allowance of 1.01 for axial fuel densification and thermal expansion, and 9) a THERMAL POHER measurement uncertainty factor of 1.02.

3/4.2.2, 3/4.2.3 and 3/4.2.4 TOTAL PLANAR AND INTEGRATED RADIAL PEAKING FACTORS -

F AND F AND AZIMUTHAL PO'lJER TILT - T T

T x

r The limitations on F and T are'rovided to ensure that the assump-tions used in the analysiPfor establishing the Linear Heat Rate and Local Power Densi ty - Hi gh LCOs and LSSS setpoints r emain val i d duri ng operation at theTvarious allowable CEA group insertion limits.

The.

limitations on F

and T

are provided to ensure that the assumptions r

q ST.

LUCIE -, UNIT 1

B 3/4 2-1

POWER DISTRIBUTION LIMITS BASES used in the analysis of establishing the DNB Margin LCO, the Thermal Margin/Low Pressure LSSS setpoints remain valid during operation at the various al lowable CEA group insertion limits.

If F, Fr or T

exceed their basic limitations, operation may continue T

T under the additsonal restrictions imposed by the ACTION statements since these additional restrictions provide adequate provisions to assure that the assumptions used in establishing the Linear Heat-Rate, Thermal Margin/Low Pressure and Local Power Density - High LCDs and LSSS setpoi nts remain valid.

An AZIMUTHAI POWER TILT ) O. 10 is not expected and if it should occur, subsequent operation would be restricted to only those operations required to identify the cause of this unexpected tilt.

The requirement that the measured value of (1

+ Tq) be multiplied by the calculated values of F

and Fx to determine Fx is applicable only wnen Fr T

and F

are calculated with a Ion-full core poQr distribution analysis.

With r

a full core power distribution analysis code the azimuthal tilt is explicitly accounted for as part of the radial power distribution used to calculate Fx and tr.

The Surveillance Requirements for verifying that Fx

, Fr and T

are T

T within their limits provide assurance that the actual values of F, Fr and Tq does not exceed the assumed values.

Verifying F and Fr after earth fuel loading prior to exceeding 75$ of RATED THERMAL RWER provides additional assurance that the core was properly loaded.

3/4.2. 5 DNB PARAMETERS The limits o'n the DNB related parameters assure tht each of the parameters are maintained within the normal steady state envelope of operation asumed in the transient and accident analyses.

The limits are consistent with the safety analyses assumptions and have been analytically demonstrated adequate to maintain a minimum DNBR of 1.23 throughout each analyzed trans'ient.

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored witnin their limits following load changes and other expected transient operation.

The 18 month periodic measurement of the RCS total flow rate is adequate to detect flow degradation and ensure correlation of the flow indication channels with measured flow such that the indicated percent flow will provide sufficient verification of flow rate on a

12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> basis.

ST.

LUCIE - UNIT 8 3/4 2-2 Amendment No.

SAFETY EVALUATION I.

Removal of References to Load Follow Node Since canmercial operation of St. Lucie Unit 1 in 1976, Florida Power 8

Light Co.

has operated Unit 1 in a base loaded mode.

Due to the significant economic advantages associated with maximizing Unit 1

electrical output, operation in the base loaded mode will continue for the foreseeable futur e.

For this reason, references to load follow oper ation are not applicable to St. Lucie Unit 1 and should be deleted.

Additionally, the specific penalty factors to be applied to calculated nuclear peaking factors (Technical Specifications 3.2-1, 3.2-2, 3.2-3) in the load follow mode were intended to be interim values until NRC approval of the CECOR power distribution analysis topical report CENPD-153.

CECOR will be available for plant use during cycle 6.

Discussions with Combustion Engineering indicate that following the approval of CENPD-153 these penalty factors are covered by the results of the topical report and therefore are not required.

II.

Removal of the Azimuthal Tilt (T ) Penalty Factor It is customary to perform all nuclear peaking factor calculations for St. Lucie Units 1 and 2 with a full core power distribution analysis code.

As such, any tilt component in the radial power distribution is explicitly factored into the calculated peaking factors.

It is therefore unnecessary to multiply the values derived fra~

a full core power distribution analysis by the tiltfactor.

The tilt penalty factor will continue to be included on radial peaking factor calculations performed with a non-full core power 8stribution analysis code.

The original basis for the inclusion of the tilt multiplier on radial

'eaking factor values goes back to cycle 1 at St. Lucie Unit 1

when the only power distribution analysis code avail ab'le was INCA which made use of octant core symmetry.

Because of the radial smoothing effect present when using folded geonetry (INCA), it was necessary to include the tilt multi plier separately.

During later cycles more advanced i ncore analysis codes have been developed which utilize full core geometry.

Peaking factors and linear heat rates (Fq) for both St.

Luci e units are calculated with full core codes.

The continued applicability of the tilt multiplier to peaking factors calculations is needed ony for non-full core analysis code results.

DETE ATION 'OF NO SI GNIF ICANT HAZA The proposed amendment would change Technical Specifications 3.2. 1, 3.2.2, and 3.2.3 to renove references to the load follow mode of operation and limit the applicability of the azimuthal tilt (Tq) multiplier to peaking factor calculations perfonaed with non-full core power di str ibution analysis codes.

The acceptability of these

changes, in that they involve no significant hazard considerations as defined by 10 CFR.50.92(c) is discussed below:

l.

10 CFR 50.92(c)(i)

The modifications proposed will not involve a

significant increase in the probability or consequences of an accident previously evaluated because neither of the changes proposed require a

change in analysis input or assumptions for any St. Lucie Unit 1

transient.

Therefore, acceptable results will continue to be shown for all previously analyzed transients.

2.

3.

10 CFR 50.92(c)(2):

The proposed changes do not create the possibility of a

new or di fferent kind of accident frcm any accident previously evaluated because they do not modify the conf ig'uration of the plant or the manner in which it is operated,

.Since no changes to the plant or its operation are made to the proposed

change, there i's no increase in the possibility of creating an accident of a

new or di fferent type over what currently exists without the proposed change.

10 CFR 50.92(c)(3):

The proposed changes do not involve any reduction in the margin of safety because neither of these cnanges involve any cnanges in al lowable modes of plant operation or al lowable envelopes for plant operational parameters.

Additional ly, none of the changes proposed either represents or requires change in input to plant safety analysis.

Hased on the discussion presented above and the enclosed safety evaluation, Florida Power 8 Light Company has concluded that none of the proposed changes to St. Lucie 1 Technical Specifications would represent a significant hazard as di scussed in 10 CFR 50. 92(c).

STATE OF FLORIDA

)

)

ss.

COUNTY OF PALM BEACH )

Robert E. Uhri being first duly sworn, deposes and says:

That he is Vice President Licensee herein; of Florida Power & Light Company, the That he has executed the foregoing document; that the statements made in this document are true and correct to the best of his knowledge, information, and belief, and that he is authorized to execute the document on behalf of said Licensee.

Robert E. Uhrig Subscribed and sworn to before me this

"'~ -day of NOTARY 'PUBLIC, in and for the County of'Palm Beach, State of Florida.

Notory Public, Stato of Flonda at Largo My Conrrni" >ion Expiraa Octobor 3D,.t%9 My COmmiSSIOn eXpll eS:

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