ML19296C613

Revised Tech Specs 3/4 2-1,2-8a,2-9,2-10,2-12,3-42,B 3/4 2-1 & B 3/4 2-3 Re Fuel Cycle 4

ML19296C613
Person / Time
Site:	Brunswick
Issue date:	02/20/1980
From:	CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML19296C611	List: ML19296C610 ML19296C613 ML19296C617
References
NUDOCS 8002260710
Download: ML19296C613 (8)

Text

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g 3/4.2 POWER DISTRIBUTION LIMITS 3 /4. 2.1 AVERAGE PLANAR LINEAR HEAT GENERATION FATE LIMITING CCNDITION POR OPERATION 3. 2.1 All AVERAGE PLANAR L INEAR HEAT GENERATION RATES (APLH3R 's) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE snail not exceed the limits shown in Figures 3.2.1 -1, 3. 2.1 -2, 3.2.1 -3, 3.2.1 -4, 3.2.1-5, 3.2.1-6, 3.2.1-7, or 3.2.1-8.

l APPLIC A31LITY -

CONDITION 1, when THERMAL POWER. > 2% of RATED THERMAL P0niii.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, 3.2.1-7, or 3.2.1-8, initiate corrective action l

within 15 minutes and continue corrective action so that APLHGR is within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% 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 REOUIREMENTS 4.2.1 All APLHGR's shall be verified to be equal to.or less than the applicable limit determined from Figure 3. 2.1 -1,' 3'.2.1 -2, 3.2.1 -3, 3.2.1 -4, 3.2.1-5, 3.2.1-6, 3.2.1-7, or 3.2.1-8.

l a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Whenever THERMAL POWER has been increased by at least IEi of RATED THERMAL POWER and s teady state operating conditions have been established, and c.

]nitially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reacter is operating with a LIMITIN3 CONTROL ROD PATTERN for APLH3R i

260 BRUNSWICK - UNIT 2 3/42-1 O

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H MAXIMUM AVERAGE PIANAR LINEAR HEAT GENERATION RATE BRUNSWICK - UNIT 2 3/4 2-8a

POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The flow biased APRM scram trip setpoint (5) and rod block trip set-point (SRB) shall be established according to ne following relationships:

S 1 (0.66W + 54%) T Sgg 1 (0.66W + 42%) T are in percent of RATED THERFAL POWER, S and S,f recirculation flow in percent of rated ficw, where:

W = Loo T = Lowest value of the ratio of design TPF divided by the MTPF obtained for any class of fuel in the core (T < 1.0), and Design TPF for:

P8x8R fuel = 2.48 8x8R fuel = 2.48 7x7 fuel = 2.60 8x8 fuel = 2.45

%PPLICAEILITY:

CONDITION 1, when THERMAL FOWER > 2E of RATED THERMAL F0WER.

ACTION:

With 5 or S-exceedinc the allowable value, initiate corrective action within 15 m%tes and continue corrective ac' tion so that S and 5 are within the required limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWEbto less than 25% of RATED THERMAL %WER 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 4.2.2 The MTPF for each' class of fuel shall be deternined, the value of T calculated, and the flow biased APRM trip setpoint adjusted, as required:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state cperating concitions have been established, and c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL RO: FACERN for MTPF.

BRUNSWICK - UNIT 2 3/4 2-9

'IPOWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL PCWER RAT!0 LIMITING CONDITION FOR OPEFATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core ficw, shall be equal to or g'reater than MCPR times the X, shown in Figure 3.2.3-1, for l.

MCPR for 7x7 fuel = 1.21*

2.

MCPR for 8x8 fuel = 1.28 3.

MCPR for 8x8R fuel = 1.21 4.

MCPR for P8x8R fuel = 1.21 APPLICASILITY:

CONDITION 1, wher. THERMAL POWER > 25% RATED THERMAL POWER.

ACTION:

With MCPR less than the applicable limit determined from Figure 3.2.3-1, initiate corrective acticn within 15 minutes and continue corrective action so that MCPR is equal to or greater than the applicable limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% 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 4.2.3 MCPR shall be determined to be equal to or greater than the applicable limit determined from Figure 3.2.3-1:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MCPR.

• ?or 7x7 fuei assemolles, tne factor is based on the 112% flow curve of Figure 3.2.3-1 rather than the se point of 102.5';.

BRUNSW!CK - U"IT 2 3/a 2-10

POWER DISTRIBUTION LIMITS 3/a.2.a LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 All LINEAR HEAT GENERATION RATES (LHGR 's) shall not exceed:

a.

For 7 X 7 fuel assemblies, as a functicn of core height for any fuel rod in an assembly, the maximum allowable LHGR shown in Figure 3.2.4-1.

b.

For 8x8, 8x8R, and P8x8R fuel assemblies, 13.4 kw/ft.

APPLICABILITY :

CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL

~

POWER.

ACTION:

With the LHGR of any fuel rod exceeding the above limits, initiate corrective action within 15 minutes and continue corrective action so that the LHGR is within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THED, MAL POWER to less than 25% 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 REOUIREMENTS 4.2.4 LHGR's shall be determined to be equal to er less than the appli-cable above limit:

a.

At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.

When THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and c.

Initially and at least once per 12 heurs unen tne reacter is operating on a LIMITING CONTROL RCD PATTERN for LHGR.

BRUNSWICX - UNIT 2 3/4 2-12

TABLE 3.3.4-2 C0flTROL R0D lilTilDRAllAl BLOCK IllSTRUMEllIATI0ff SEIP0 lilts E'

,lRIP fullCT10tl Afl0 IllSTilVMEllT iluMBEll TRIP SETP0liff All.OllABLE VALUE O

l.

APRM (CSI-april-Cll.A.II,C,D E.F) 5 a.

Upscale (Flow Blased)

< (0.66 W e 42%)

T*

< (0.66 11

• 42%) T

• 13.

Inoperative flA MIPF flA IIIPF c.

Downscale

> 3/125 of full scale

> 3/125 of full scale d.

Upscale (Fixed) 712% of RATED lilEftt1Al. P0llER

< 12% of RATED lilERilAL P0llER 2.

ROD fl10CK N0filTOR (CSI-RBM-Cll.A,B) a.

Lipscale

< (0.66W t 40%)

T*

< (0.66 11 + 40%)

T*

l l>.

Inoperative llA MlPF flA

~HTliff c.

Downscale

> 3/125 of full scale

> 3/125 of full scale w

3.

S_OllitCE RAtlGE fl0tilTORS (C51-SRM-K600A,B,C,0) o a.

Detector not full in flA flA b

13.

Lipscale 5

5

,_ 1 x 10 s

1 10 cps c.

Inoperative flA flA d.

Downscale

> 3 cps

> 3 cps 4.

IllTEIU1EDI ATE RA!!GE M0ttlTORS (C51-IRil-K601 A,0,C,0,E,F,G,ll) a.

Detector not full in flA flA l>.

Upscale c.

Inoperative

_< 100/125 of full scale

< 108/125 of full scale ilA flA d.

Downscale

> 3/125 of full scale

> 3/125 of full scale i

T=2.60 for 7x7 fuel T=2.45 for 8x8 fuel T=2.48 for 8x8R fuel T=2.48 for P8x8R fuel

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200'F limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated effe:ts of fuel pellet densification.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATI0fl RATE This specification assures that the peak cladding tempeature following the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50, Appendix K.

The peak cladding temperature (PCT) following a postulated less-of-coolant accident is primarily a function of the average heat genera-tion rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within a assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap The Technical Speci-conductance and rod-to-rod local peaking f actor.

f,ication APHGR is this LHGR of the highest powered rod divided by its 1ccal peaking factor.

The limiting value for APLHGR is shown in Figures 3. 2.1 -1, 3. 2.1-2, 3.2.1 -3, 3. 2.1 -4, 3. 2.1-5, 3. 2.1-6, 3. 2.1-7, and 3.2.1-8.

The calculational procedure used to establish the APLHGR shown on

3. 2.1 -1, ' 3.2.1-2, 3. 2.1 -3, 3. 2.1-4, 3. 2.1 - 5, 3. 2.1-6, 3. 2.1-7, and Figures is based on a loss-of-coolant accident analysis. The analysis 3.2.1-8 was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50.

A complete discussion of each code employed in the analysis is presented in Differences in this analysis compared to previous analyses Reference 1.

(1) The analysis assumes a fuel assembly performed with Reference 1 are:

planar power consistent with 102" of the MAPLHGR shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, 3.2.1-7, and 3.2.1-8; (2) Fission product decay is computed assuming an energy release rate of 200 MEV/ Fission; (3) Pool boiling is assumed af ter nucleate boiling is lost during the flow stagnation period; (4) The effects of core spray entrainmen and counter-current flow limitation as described in Reference 2, are included in the reflooding calculations.

A lisc of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.

BRUNSWICK - UNIT 2 B 3/4 2-1

POWER DISTRIBUTION LIMITS t

BASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity safety limits of Specifhation 2.1 were based on a TOTAL PEAKING FACTOR of 2.60 for 7x7 fuel, 2.45 for 8x8 fuel, and 2.48 for 8x8R and P8x8R fuel.

The scram setting and rod block functions of the APRM instruments must be adjusted to ensure that the MCPR coes not become less than 1.0 in the degraded situation. The sc"am settings and rod block settings are adjusted in accordance with the femula in this specification when the combination of THERMAL POWER and peak flux indicates a TOTAL PEAKING FACTOR greater than 2.60 for 7 x 7 fuel, 2.45 for 8x8 fuel, and 2.48 for 8x8R and P8x8R fuel.

The method used to l

determine the design TPF shall be consistent with the method used to determine the MTPF.

3/4.213 F.INIMUM CRITICAL POWER RATIO The required operating limit MCPR's at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Lg.jt MCPR of 1.07, and an analysis of abnorral operational transients.

For any abnormal operating tran.

sient analysis evaluation with the initial condition of the reacter being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrunent trip setting as given in Speci fication 2.2.1.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the cost limiting transients have been analyzed to detemine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertien, and coolant temperature decrease.

The limiting transient which detemines the required steady state MCPR limit is the turbine trip with failure of the turbine bypass.

This transient yields the largest a MCPR.

When added to the Safety Limit NCPR of 1.07 the required minimum operating limit n:PR of Specificati:n 3,2,3 is obtained.

Prict to the analysis of abnormal operaticnal t an-sients an initial fuel bundle MCPR was determined. This parameter is based on the bundle flow calculated by a GE multi-channel steady gte flow distribution model as described in Section 4.4 of NED0-20360 and on core parameters shown in Reference 3, response to Ite s 2 and 9.

BRUNSWICK - UNIT 2 B 3/4 2 -3