ML20083M495

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Proposed Tech Specs Re Limiting Conditions for Operation on Reactor Coolant Flow for Cycle 9 Reload Application
ML20083M495
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Site: Fort Calhoun Omaha Public Power District icon.png
Issue date: 04/16/1984
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OMAHA PUBLIC POWER DISTRICT
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NUDOCS 8404180165
Download: ML20083M495 (9)
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l 2.0 LIMITING COUDITI0 tis FOR OPERATIOil 2.10 Reactor Core (Continued) _

2.10.h Power Distribution Limits _(Continued)

(5) DUBR tjarcin Juring Power Operation Above 15% of Rated Power (a) The following DNB related parameters shall be maintained within the limits shown: .

(i) Cold Leg Temperature 1 Sh50F *

(ii) Pressurizer Pressure 3,2075 psia" (iii) Reactor Coolant Flow 3,197,000 gpm**

(iv) Axial Shape Index, Yy i Figure 2-7 (b) With any of the above parameters exceeding the 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 power to less than 15% of rated power within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Basis i

Linear Heat Rate The limitation on linear heat rate ensures that in the event of a LOCA, the peak temperature of the fuel cladding will not exceed 22000F.

Either of the two core power distribution monitoring systems, the Excore Detector Monitoring System, or the Incore Detector Monitoring System, provide adequate monitoring of the core power distribution and are capable of verify'.nc that the linear heat rate does not exceed its limits. The Excore Detector Monitoring System performs this function 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 2-6 as adjusted by Specification 2.10.4.(1).(c) for the allowed linear heat rate of Figure 2-5, RC Pump configuration, and F xyT of Figure 2-9.

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 Specification 2.10.2.(6) and long term insertion limits of frecification 2.10.2.(7) are satisfied, (2) the flux penking augmentation 4 factors are as shown in Figure 2-8, (3) the azimuthal power tilt restrictions of Specification 2.10.h.(h) are satisfied, and (h) the total plannr radial <

peaking factor does not exceed the limits of Specification 2.10.k.(3).

4

  • Limit not applicable during either a thermal power ramp in excess of 5% of ,

rated thermal power per minute or a thermal power step of greater than 10%

of rated thermal power. ,

    • This number is an actual limit and corresponds to an indicated flow rate of 202,500 gpm. All other values in this listing are indicated values and in-clude an allowance for measurement uncertainty (e.g., 5450F, indicated, al- i lows for an actual Te of 547 F).

M41END165 844>416 PDR ADOCK 05000205 P PDR FORT CALHOUN 2-57c Amendment No. 32, W3, $7,70

TABLE B-1 i

Explanation for Cycle 9 Technical Specification Changes t I Change Tech. Spec. Number Changes Reasons 13 Figure 2-9 Replace Figure 2-9 The FxyT and FRT s with enclosed Figure limits have been 2-9 changed to reflect

- higher radial peak-ing factors in con-junction with the statistical conbin-
ation of uncertain-I ties program

~

14 2.10.4(1)(2) Change 1.07 to 1.062 Changed to reflect Page 2-50 CECOR accuracy in

CENPD-measuring 153-P , Rev.1-Fn (P-A, INCA /CECOR Power Peaking Uncertainty) 15 2.10.4(2) Change limited to The FRT changes have Page 2-57a < 1.62 to limited been made to reflect '

to _< 1.73 and with proposed changes in FR T > 1.62 to with

  • Tech. Spec.1.1 FRT > 1.73 16 2.10.4(3) Change limited to The FxyT changes have l Page 2-57a < 1.67 to limited been made to reflect  !'

to < 1.78 and with

- proposed changes in FR T > 1.67 to with Tech. Spec.1.1 FR T > 1.78 ,-

17 2.10.4(5)(a)iii** Change ** to pro- Provide operators with Page 2-57c vide allowed mea- the indicated measured sured value of RCS limit on RCS flow rate

, flow rate not in- for consistency with

cluding uncertain- rest of parameters in ties table.

18 Table 2-2 Add steam generator See Change 6 Page 2-67 differential pressure 19 Table 3-1 Add steam generator See Change 6 Page 3-5 differential pressure as Item 11 20 Table 3-1 Change reactor pro- Maintain consistent Page 3-5 tective system log- numbering scheme ic units from Item 11 to Item 12

TABLE 7-2 FORT CALHOUN UNIT 1, CYCLE 9 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MELT) DESIGN LIMITS Reference Cycle (Cycle 8)

Physics Parameters Units Values Cycle 9 Values Radial Peaking Factors For DNB Margin Analyses (FRT)

Unrodded Region ^

1.65 1.75+

  • Bank 4 Inserted 1.69 1.79+,*

For Planar Radial Component (FxyT) of 3-0 Peak (CTM Limit Analyses)

Unrodded Region 1.72 1.78+

  • Bank 4 Inserted 1.81 1.93+
  • Maximum Augmentation Factor 1.057 1.057 Moderator Temperature Coefficient 10-4ap/'F -2.5 to +0.5 -2.7 to +0.5 Shutdown Margin (Value Assumed in Limiting EOC Zero Power SLG) %Ap -4.0 ,,

-4.0 Tilt Allowance  % 3.0 3.0

  • For the Loss of Coolant flow and CEA Drop Events, the ef fects of uncertainties on these parameters were accounted for statistically in the DNBR and CTM calcu-lations. The DNBR analysis utilized the methods discussed in Section 6.1 of this report. The procedures used in the Statistical Combination of Uncertain-ties (SCU) as they pertain to DNB and CTM limits are detailed in References 2a, 2b, 2c. ,

+The values assumed are conservative with respect to the Technical Specifica-tion limits.

TABLE 7-2 (Continued)

Safety Parameters Units Cycle 8 Values Cycle 9 Values Power Level MWt 1530 1530*

Maximum Steady State Temperature 'F 547 547*

Minimum Steady State Pressurizer Pressure psia 2053 2053*

Reactor Coolant Flow gpm 197,000 202,500*

Negative Axial Shape -

LC0 Extreme Assumed at Full Power (Ex-Cores) Ip -0.20 -0.18 Maximum CEA Insertion  % Insertion l at Full Power of Bank 4 25 25 Maximum Initial Linear Heat Rate for Transient l Other than LOCA KW/ft 15.22 15.22 Steady State Linear Heat Rate for Fuel CTM Assumed in the Safety l Analysis KW/ft 21.0 21.0 CEA Drop Time to 100%

i including Holding Coll Delay sec 3.1 3.1 MinimumDNBR(CE-1) 1.19 1.22*

l *For the Loss of Coolant Flow and CEA Drop Events, the effects of uncertainties on these parameters were accounted for statistically in the DNBR and CTM calcu-i lations. The DNBR analysis utilized the methods discussed in Section 6.1 of The procedures used in the Statistical Combination of Uncertain-this report.

ties (SCU) as they pertain to DNB and CTM limits are detailed in References 2a, 2b, 2c.

l l

e'

  • TABLE 7.2.1-1 FORT CALHOUN CYCLE 9 f

KEY PARAMETERS ASSUMED IN THE CEA WITilDRAWAL ANALYSIS Parameter Units HZP HFP Initial Core Power Level MWt 1 102% of 1500*

Core Inlet Coolant  :

Temperature 'F 532* 547*  !

i Pressurizer Pressure psia 2053* 2053*

Moderator Temperature Coefficient - x10-4ap/*F +0.5 +0 . 5 *

  • Doppler Coefficient Multiplier 0.85 0.85 CEA Worth at Trip 10-2ap -5.25 -6.66 j Reactivity insertion  ;

Rate Range x10-4ap/sec 0 to 1.0 0 to 1.0 '

) CEA Group Withdrawal l Rate in/ min 46 46 ,

j Holding Coil Delay Time sec 0.5 0.5 1

  • The ef fects of uncertainties on these parameters were accounted for deter-ministically and the DNBR calculations used the methods discussed in Sec-tion 6.1 of this document and detailed in References 2a,2b, and 2c.

J

    • DNBR analysis assumes MTC consistent with Reference 5.

i h

a 1

' TABLE 7.2.2-1 FORT CALHOUN CYCLE 9 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW ANALYSIS Parameter Units Cycle 9 Initial Core Power Level MWt 1530*

l Initial Core Inlet Coolant Temperature *F = 547*

Initial RCS Flow Rate gpm 202,500*

l Pressurizer Pressure psia 2053*

Moderator Temperature Coef ficient 10-4ap/*F +.5 Doppler Coefficient Multiplier -- 0.85 I

LFT Analysis Setpoint  % of initial flow 0.93 LFT Response Time sec 0.65 4-Pump RCS Flow Coastdown Figure 7.2.2-1 CEA Holding Coil Delcy sec 0.5 CEA Time to 100% Insertion sec 3.1 (Including Holding Coil Delay) .

CEA Worth at Trip (all rods out) 10-2ap -6.87 Total Unrodded Radial Peaking 1.75 Factor (FRT) -

  • The uncertainties on these parameters were cabined statistically rather than determinist 1cally. The values listed represent the bounds included in the statistical combination.

s d

EI TABLE 7.2.3-1 FORT CALHOUN CYCLE 9 KEY PARAMETERS ASSUMED IN THE FULL LENGTH CEA DROP ANALYSIS Parameter 'Inits Cycle 8 Cycle 9 Initial Core Power Level MWt 102% of 1500 102% of 1500*

Core Inlet Temperature *F  : 547 547*

Pressurizer Pressure psia 2053 2053*

Core Mass Flow Rate gpm 197,000 202,500*

Moderator Temperature Coefficient x10-4ap/* F -2.7 -2.7 Doppler Coefficient Multiplier --

1.15 1.15 CEA Insertion at Maximum Allowed  % Insertion of 25 25 Power Bank 4 Dropped CEA Worth %Ap unrodded -0.28 -0.2261 PDIL -0.28 -0.2238 Maximum Allowed Power Axial Shape Index at Negative Extreme of LCO Band -0.20, -0.18 Radial Peaking Distortion Factor '

s Integrated Radial Peaking Urirodded Region 1.1579 1.1585 Bank 4 1.1696 1.1557 Inserted Region.,

Planar Radial Peaking Unrodded Region 1.25~ 1.213.

Bank 4 ,1.24 1.205 Inserted Region

  • The uncertainties .on these parameters were combined statistically rather than deterministically. 'The~ values listed represent the bounds included in the statistical combination.
  • 3 .'. , _

s c. ,

N s -l' 1

\

> .s, l

3, , l

)

w : &.A s & s

,. 4 %

. _ > -.O -

TABLE 7.2.5-1 FORT CALHOUN CYCLE 9 KEY PARAMETERS ASSUMED IN THE ANALYSIS OF LOSS OF LOAD TO ONE STEAM GENERATOR Parameter Units Cycle 4 Initial Core Power MWt 102% of 1500*

Initial Core Inlet Temperature *F 547*

Initial Pressurizer Pressure psia 2053*

Moderator Temperature Coefficient Ap/*F -2.7 x 10-4 Doppler Coefficient Multiplier 1.15

  • The effects of uncertainties on these parameters were accounted for deter-ministically and the DNBR calculations used the methods discussed in Sec-2 tion 6.1 of this document and detailed in References 2a, 2b, and 2c.

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DISCUSSION, JUSTIFICATION, AND SIGNIFICANT HAZARDS CONSIDERATION FOR CYCLE 9 RELOAD This application serves to revise the Cycle 9 reload amendment ap-plication. This revision deletes the proposed amendment of using an indicated value rather than an allowed value for the DNBR limit-ing conditions for operation on reactor coolant system flow rate.

indicated reactor j' This revision incorporates the corresponding thus delineating the  !

coolant system flow rate into a footnote, Since this revidion serves to clarify >

measurement uncertainty.

the Technical Specifications and reports the limits in a method acceptable to the staff, the significant hazards consideration dis-cussion is unchanged and remains applicable for this revision.

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E ATTACHMENT B

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