ML19354E695

Reload Evaluation Rept for Fort Calhoun Station Unit 1, Cycle 13.

ML19354E695
Person / Time
Site:	Fort Calhoun
Issue date:	01/26/1990
From:	OMAHA PUBLIC POWER DISTRICT
To:
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ML19354E688	List: ML19354E687 ML19354E689 ML19354E691 ML19354E695
References
NUDOCS 9002010151
Download: ML19354E695 (64)

Text

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t Fort Calhoun Station Unit No.1-Cycle 13 i Reload Evaluation >

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I 9002010151 900126 ADOCK 05000285 PDR P PDC .

Page 1 of 64 j

FORT CALHOUN STATION UNIT NO.1 CYCLE 13 '!

LICENSE APPLICATION ,

TABLE OF CONTENTS ,

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1.0 INTRODUCTION

AND

SUMMARY

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2.0 OPERATING HISTORY OF CYCLE 12 5 L

3.0 GENERAL DESCRIPTION 6 4.0 FUEL SYSTEM DESIGN 14 5.0 NUCLEAR DESIGN 15 i

5,1 PHYSICAL CHARACTERISTICS ,15 5.1.1 Fuel Management 15 i 5.1.2 Power Distribution 16 5.1.3 Safety Related Data 16 5.1.3.1 Ejected CEA Data 16 5.1.3.2 Dropped CEA Data 17 5.2 ANALYTICAL INPUT TO INCORE MEASUREMENTS 17 I 5.3 NUCLEAR DESIGN METHODOLOGY 17 5.4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS 17 6.0 THERMAL-HYDRAULIC DESIGN 26 I 6.1 DNBR ANALYSIS 26-6.2 FUEL ROD BOWING 26 l

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i FORT CALHOUN STATION UNIT NO 1 i' CYCLE 13 LICENSE APPLICATION  :

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TABLE OF CONTFM (Continued) ,

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EAIM 7.0 TRANSIENT ANALYSIS 28' 7.1 33 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 1) 7.1.1 Excess Load Event 33 7.1.2 RCS Depressurization Event 35 i 7.2 36 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) t 7.2.1 CEA Withdrawal Event 36' 7.2.2 Loss of Coolant Flow Event 43 i 7.2.3 Full Length CEA Drop Event 45  ;

7.2.4 Boron Dilution Event 52 .j 7.3 POSTULATED ACCIDENTS 55 J 7.3.1 CEA Ejection 55 7.3.2 Steam Line Break Accident 55  !

7.3.3 Seized Rotor Event 58 j 8.0 ECCS PERFORMANCE ANALYSIS 59 9.0 STARTUP TESTING 60

10.0 REFERENCES

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1.0 INTRODUCTION

AND

SUMMARY

f This report provides an evaluation of the design and performance for the operation of Fort l Calhoun Station Unit No.1 during its thirteenth fuel cycle at a full rated power of 1500 MWt. .

Planned operating conditions remain the same as those for Cycle 12, unless otherwise -

I noted in the proposed Technical Specification changes, j i

i The core will consist of 93 presently operating batches L M and N assemblies and 40 '

fresh Batch P assemblies.

The Cycle 13 analysis is based on a Cycle 12 termination point between 10,000 MWD /  !

MTU and 12,000 MWD /MTU. In performing analyses of design basis events, limiting safety  :

system settings ard limiting conditions for operation, limiting values of _ key parameters  ;

were chosen to assure that expected Cycle 13 conditions would be enveloped, provided the Cycle 12 termination point falls within the above range. The analysis presented herein >

will accommodate a Cycle 13 length of up to 14.250 MWD /MTU with a coastdown of an additional 1,000 MWD /MTU.

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The evaluation of the reload core characteristics have been conducted with respect to the Fort Calhoun Station Unit No.1 Cycle 12 safety analysis described in the 1989 update of the USAR, hereafter referred to as the " reference cycle" in this report unless noted other-wise.

Specific core differences have been accounted for in the present analysis. In all cases, it i has been concluded that either the reference cycle analyses envelope the new conditions or the revised analyses presented herein continue to show acceptable results. Where dictated by variations from the previous cycle, proposed modifications to the plant Technl-cal Specifications have been provided.

l The Cycle 13 core has been designed to minimize the neutron flux to critical reactor pressure vessel welds to reduce the rate of RTm shift of these welds, This will maximize the time to reach the screening criteria of the proposed amendment to.10 CFR 50.61 regulations as documented in the Federal Register, Volume 54, No. 246, pages 52,946 through 52,950. Tuesday, December 26,1989, which would change the procedure for calculating the amount of radiation embrittlement that a reactor vessel receives and make it consistent with the procedure given in Regulatory Guide 1.99, Revision 2.

1 The reload analysis presented in this report was performed utilizing the methodology l documented in Omaha Public Power District's reload analysis methodology reports (Ref-

{ erences 1, 2, and 3). These methods were also used to perform the Cycle 12 reload analysis, 4

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l 2.0 OPERATING HISTORY OF CYCLE 12 I

Fort Calhoun Station is presently operating in its twelfth fuel cycle utilizing Batches J K, L. l M rnde N fuel assemblies. Fort Calhoun Cycle 12 operation began with criticality on Janu- l ar/ 29,1989, and reached full power on February 23,1989. The reactor has operated up .

j to the present time with the core reactivity, power distributions, and peaking factors having j closely followed the calculated predictions.

i it is estimated that Cycle 12 will be terminated on or about February 16,1989. The Cycle i 12 termination point can vary between 10,000 MWD /MTU and 12,000 MWD /MTU and still' be within the assumptions of the Cycle 13 analyses. As of December 31,1989, the Cycle ,

12 burnup had reached 9,447 MWD /MTU. l

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3.0 GENERAL DESCRIPTION The Cycle 13 core will consist of the number and type of assemblies and fuel batches shown in Table 3-1. Eight J assemblies, 8 K assemblies,21 L assemblies and 3 M assem-  :

blies will be discharged this outage. They will be replaced by 8 fresh unshimmed Batch P  !

assemblies (3.95 w/o enrichment) and 32 fresh shimmed Batch P assemblies (3.58 w/o  !

average enrichment. 0.027 gm Bio / inch), .

4 Figure 3-1 shows the fuel management pattern to be employed in Cycle 13, Two primary changes to the core are incorporated for Cycle 13. First, the average initial enrichment of the 40 fresh Batch P assemblies is 3.66 w/o U-235, a reduction of 0.04 w/o from Cycle 12.

Second, the fuel assembly zone loading technique is used to lower the radial power peak.

Ing facters within the 32 fresh shimmed Batch P assemblies.since shimmed fuel assom-blies at Fort Calhoun Station have in the past contained the maximum radial peaking factors. Each Batch P shimmed assembly contains 80 fuel rods with an initial enrichment of 3.95 w/o of U-235 and 88 fuel rods with an initial enrichment of 3.25 w/o of U-235. The locations of the zoned fuel rods and poison rods within the lattice of the Batch P shimmed assemblies are shown in Figure 3-2.

Due to the Fort Calhoun fuel assembly design, the fuel rods surrounding the five large water holes produce the highest power peaking factors within an assembly, The fuel rod zone loading technique lowers the initial enrichment of U-235 in those fuel rods while maintaining an assembly avera design core average exposure.ge initial enrichment sufficient to achieve the Cy The fuel rod and poison rod locations in Batches M and N shimmed assemblies are shown i

in Figure 3-3. Figure 3-4 shows the fuel rod locations in all unshimmed assemblies.

Figure 3-5 shows the beginning of Cycle 13 assembly burnup distribution for a Cycle 12 termination burnup of 11,000 MWD /MTU. The fuel average discharge exposure at the end  !

of Cycle 12 ls projected to be 40.090 MWD /MTU, The initial enrichment of the fuel assem-blies is also shown in Figure 3-5. Figure 3-6 shows the projected end of Cycle 13 assem-bly burnup distribution. The end of Cycle 13 core average exposure is approximately 29,205 MWD /MTU.

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TABLE 3-1 FORT CALHOUN UMT NO.1. CYCLE 13 CORE LOADING Irubal BOC EOC Poison Poison Assembly Number of Average Burnup (MWD /MTU) Average Bumup (MWD /MTU) Rods per Loading Desianation Assemblies IEOC12 = 11.000 MWD /MTUI IEOC13 = 14.250 MWD /MTU) Assembly am Bio / inch L 8 33.139 38.322 0 0 M 20 24.882 40.224 0 0 M/ 21 30.821 38.141 8 0.024 2 N 20 11.590 27.465 0 0 N/ 24 14.283 31.006 8 0.020 2

E P 8 0 12.701 0 0 P/ 32 0 17.902 8 0.027

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i AA - Assembly Location BB - Fuel Type 1 2 M/ L 1

3 4 5 6 7 M/ P N P/ N/

8 9 10 11 12 13 W P/ W P/ W P/

14 15 16 17 18 19 P N/ N M N M/

20 21 22 23 24 25 2s N P/ M N/ M P/

M/

27 28 29 30 31 32 33 P/ N/ N M P/ M L

34 35 36 37 38 39 Q W P/ W P/ M W ,

I Cycle 13 Core Loading Pattern Omaha Public Power District Figure Fort Calhoun Station Unit No.1 3-1 Page 8 of 64

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00000000000000  !

OOOOOOOOOOO@OO r OO 000000 00 l 00 000000 GO i 0 0 0.0 9 0 0 0 0 9 0 0 0 0 '

00000000'000000 '

000000 000000-000000 0 0 0 0 0.0  !

00000000000000 -

0000000009990.0 '

OO 000000 90 i 00 000009 @@

09000000000000 .

00000000000000 ,

l 9 - Shim (B C)4 pin (8)

G - Low (3,25%) Enrichment Fuel Pin (88) ,

O - High (3.95 %) Enrichment Fuel Pin (80) >

- Guide Tube Batch P/ Fuel Assembly Omaha Public Power District Figure

'i Fuel Rod and Poison Rod Locations Fort Calhoun Station Unit No.1 3-2 i Page 9 of 64

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. i O0000000000000 l 09000000000.000 i OO 000000 00 '

OO 000000 OO 00009000090000  ;

00000000000000  !

OOOOOO OOOOOO '

000000 000000 00000000000000 )

00009000090000 OO 000000 OO '

OO 000000 OO 09000000000000  !

00000000000000 '

G - Shim (B C)4 pin (8)

O - Fuel Pin (168)

- Guide Tube Batches M/ and N/ Assembly Omaha Public Power District Figure 1 Fuel Rod and Poison Rod Locations Fort Calhoun Station Unit No.-1 3-3  :

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., 00000000000000 00000000000000 OO 000000 00 l OO 000000 OO l

00000000000000 1 00000000000000 1 000000 000000 I 000000 000000 I 00000000000000 i O0000000000000 l OO 000000 OO OO 000000 00 1 00000000000000  ;

00000000000000 l

I O - Fuel Pin (176) l

! - Guide Tube i

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Batches L, M, N and P Omaha Public Power District Figure i Fuel Rod and Guide Tube Locations Fort Calhoun Station Unit No.1 3-4 Page 11 of 64

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- Assembly Location

- Fuel Type .

C.CC - Enrichment DD.DDD - Assembly Av(w/o erage Exposure U-235) (MWD /MTU)

' M/ L 3.80 3.80 31.766 33.024 3 4 5 6 W I P N P/ 7M '

3.80 3.95 3.70 3.58 3.70 30,167 0 10.346 0 15.300 8

M/

3.80

• P/

3.58

' N/

3.70 "P/ N/ P/

3.58 3.70 3.58 30.140 0 12.797 0 14.792 0 14 15 16 17 18 19 3.95 3.70 3.70 3.80 3.70 3.80 0 12.830 12.540 27.028 12.343 30.150 20 21 22 23 24 25 26 3.70 3.58 3.80 3.70 3.80 3.58 M/ 10.356 0 27.025 15.195 20.725 0 ,

3.80 27 28 29 31.796 30 31 32 P/ N/ N M P/ M 3.58 3.70 3.70 3.80 3.58 3.80 33[ 0 14.789 12.364 20.772 O_ 30.284 3.80 34 35 36 33.254 37 38 39 N/ P/ M/ P/ M M/ "

3.70 3.58 3.80 3.58 3.80 15.298 3.80 0 30.261 0 27.439 30.947 Note: EOC 12 Core Average Burnup = 11,000 MWD /MTU -

Cycle 13 BOC Assembly Average Omaha Public Power District Figure Exposure and initial Enrichment .

Fort Calhoun Station Unit No.1 3-5 '

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Assembly Location S8 Fuel Type C.CC - Enrichment (w/o U-235 DD.DDD - Assembly Average Expo)sure (MWD /MTU) '

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M/

3.80 3.80  ;

36.301 30,215 3 4 5 6 7 .

M/ P N P/ N/

2 3.80 3.95 3.70 3.58 3.70 '

i 35.691 12,702 23,775 15,681 29.496 8 9 10 11 12 13 i

3.80 3.58 3.70 3.58 3.70 3.58 35,665 14.027 29.441 19.558 32,673 19,795 .

14 15 16 17 18 19 3,95 3,70 3.70 3,80 3.70 3,80 12,700 29,466 29,768 42,051 29,995 45,248 20 21 22 23 24 25 26 3.70 3.58 3.80 3,70 3.80 3.58

• 23,781 M/ 19,552 42,045 32.297 37,018 19,e78 3.80 27 28 29 30 36.330 31 32 P/ N/ N M P/ M 3.58 3,70 3.70 3.80 3.58 3,80 33[ 15,676 32,665 '30,007 37,061 19,246 44,178 3.80 34 35 36 37 38 429 38 39 N/ P/ M/ P/ M M/

3.70 3.58 3.80 3.58 3.80 3.80 29,493 19,793 45,341 19,687 41,715 43,843 Cycle 13 EOC Assembly Average Omaha Public Power District- Figure Exposure and Initial Enrichment .'

Fort Calhoun Station Unit No.1 3-6 Page 13 of 64

4.0 FUEL SYSTEMS DESIGN 4

The mechanical design for the Batch P fuelis the same as the Batch N fuel supplied by Combustion Engineering (CE) in Cycle 12.

The Batch P fuel is similar in design to the Batch G fuel supplied by Combustion Engineer-ing in Cycle 5 and is mechanically, thermally, and hydraulically compatible with the eight i Batch L Advanced Nuclear Fuel (ANF) supplied fuel assemblies remaining in the core.  !

Reference 4 describes the Batch M fuel characteristics and design. In addition, the Batch P fuel assembly design is essentially identical to the CE Batches M and N fuel which are '

also mechanically, thermally and hydraulically compatible with the ANF fuel which will reside in the Cycle 13 core. References 5 through 7 remain valid for describing the design of the ANF-supplied fuel. The maximum pin burnups of 50,000 MWD /MTU for Batch K fuel '

and 52,000 MWD /MTU for Batch L fuel (Reference 8) also remain valid.

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i 5.0 NUCI FAR DESIGN  ;

j 5.1 PHYSICAL CHARACTERISTICS j 5.1.1 Fuel Manacament ,

The Cycle 13 fuel management uses a low-radial leakage design, with ,

twice and thrice burned assemblies predominantly loaded on the periphery i of the core. This low-radial leakage fuel loading pattern is utilized to minl.  ;

mize the flux to the pressure vessel wclds and achieve the maximum in neutron economy. Use of this type of fuel managemant to achieve reduced _

pressure vessel flux over a standard out-in-in pattern results in higher radial :

peaking factors. The maximum radial peaking factors for Cycle 13 have e

been reduced by lowering the enrichment of the fuel pins adjacent to the  !

assembly water holes as described in Section 3.0..

Also described in Section 3.0 is the Cycle 13 loading pattern which is com-posed of 40 fresh Batch P assemblies of which 32 are shimmed and split i enrichment assembtles designated P/ (with a high pin enrichment of 3.95 w/o and a low pin enrichment of 3.25 w/o) and 8 are unshimmed 3.95 w/o enriched assemblies. Eight thrice bumed L assemblies are being returned to the core, along with 41 twice burned M assemblies and 44 once burned N assemblies to produce a Cycle 13 loading pattern with a cycle energy of 14,250 1,000 MWD /MTU. The Cycle 13 core characteristics have been examined for a Cycle 12 termination between 10,000 MWD /MTU and 12,000 MWD /MTU and limiting values established for the safety analysis.

The Cycle 13 loading pattern is valid for any Cycle 12 endpoint between these values.

Physics characteristics including reactivity coefficients for Cycle 13 are listed in Table 5-1 along with the corresponding values from Cycle 12. It l should be noted that the values of parameters actually employed in safety l

analyses are different from those displayed in Table 5-1 and are typically chosen to conservatively bound predicted values with accommodation for -

approrlate uncertainties and allowances.

The BOC, HZP Main Steam Line Break accident is the most limiting acci-dent of those used in the determination of regulred shutdown margin for compliance with Technical Specifications. Table 5-2 presents a summary of CEA shutdown worths and reactivity allowances for the Cycle 13 BOC, HZP MSLB accident.- The Cycle 13 values, calculated for minimum scram worth, exceed the minimum value required by Technical Specifications and thus provide an adequate shutdown margin.

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5.0 NUCl FAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.2 Power Distribution Figures 5-1 through 5-3 illustrate the all rods out (ARO) planar radial power distributions at BOC13, MOC13, and EOC13, respectively, and are an aver.

age of the low and high burnup timepoints of the Cycle 12 shutdown win-dow, These radial power densities are assembly averages representative of the entire core length. The high bumup end of the Cycle 12 shutdown win.

dow tends to increase the power peaking in the high power assemblies in the Cycle 13 fuel loading pattern. The radial power distributions, with Bank 4 fully inserted at beginning and end of Cycle 13 are shown in Figures 5-4 '

and 5-5, respectively.

The radial power distributions described in this section are calculated data without uncertainties or other allowances. However, the single rod power peaking values do include the increased peaking that is characteristic of fuel rods adjoining the water holes in the fuel assembly lattice. For both DNB and kw/ft safety and setpoint analyses in either rodded or unrodded configu-rations, the power peaking values actually used are higher than those ex.

pected to occur at any time during Cycle 13. These conservative values, which are used in Section 7.0 of this document, establish the allowable limits for power peaking to be observed during operation; Figures 3-5 and 3-6 show the integrated assembly burnup values at 0 and 14,250 MWD /MTU, based on an EOC12 core average burnup of 11,000 MWD /MTU.

The range of allowable axial peaking is defined by the limiting conditions for operation and their axial shape index (ASI). Within these ASilimits, the neo-essary DNBR and kw/ft margins are maintained for a wide range of possible axial shapes. The maximum three-dimensional or total peaking factor (Fe) anticipated in Cycle 13 during normal base load, all rods out operation at full power is 1.82, not including uncertainty allowances.

5.1.3 Safety Related Data 5.1.3.1 Ejected CEA Data The maximum reactivity worth and planar power peaking factors  ;

associated with an ejected CEA event are shown in Table 5-3 for <

both the beginning and end of Cycle 13. These values encompass the worst conditions anticipated during Cycle 13 for an expected Cycle 12 termination point between 10,000 MWD /MTU and 12,000 MWD /MTU. The values shown for Cycle 13 are calculated in ac-cordance with Reference 3. In addition, Table 5-3 lists these values used from Cycle 12 for comparison.

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5.0 NUCI PAR DESIGN (Continued) 5.1 PHYSICAL CHARACTERISTICS (Continued) 5.1.3 Salitty Related Data (Continued) .

5.1.3.2 Dropped CEA Data The Cycle 13 safety related data for the dropped CEA analysis were calculated identically with the methods used in Cycle 12.

5.2 ANALYTICAL INPUT TO INCORE MEASUREMENTS Incore detector measurement constants to be used in evaluating the reload cycle power distributions will be calculated in the same manner as for Cycle 12, 5.3 NUCLEAR DESIGN METHODOLOGY Analyses have been performed in the manner and with the methodologies docu-mented in References 1 and 2.

5,4 UNCERTAINTIES IN MEASURED POWER DISTRIBUTIONS The power distribution measurement uncertainties which are applied to Cycle 13 are the same as those presented in Reference 2.

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TABLE 5-1  !

FORT CALHOUN UNIT NO.1. CYCLE 13 .

NOMINAL PHYSICS CHARACTERISTICS  !

1)0g3 Cvele 12 Cvele 13 4 Critical Boron Concentration Hot Full Power ARO, I Equilibrium Xenon, BOC ppm 1081 1187 1 l

Inverse Boron Worth

Hot Full Power, BOC ppm /%Ap 113 112 -

Hot Full Power, EOC ppm /%Ap 90 84 ReactMty Coefficients with All CEAs Withdrawn Moderator Temperature l Coefficient (MTC)

Beginning of Cycle HZP 10" Ap/'F +0.25 +0.51 '

End of Cycle, HFP 10* Ap/'F -2.49 -2.47 Donoler Coefficient (FTC)

Hot Zero Power BOC 10ap/'F -1.97 -2.06 i

Hot Full Power. BOC 10 Ap/'F -1.47 -1.66 Hot Full Power, EOC 10ap/'F -1.57 -1.85 Total Delayed Neutron Fraction. G..

BOC 0.00607 0.00614 EOC 0.00521 0.00519 l

Neutron Generation Time. l' l

BOC 10*sec 22.2 21.6 1

EOC 10*sec 28.0 28.8 This value exceeds the Technical Specification limit of +0.50 x 10" Ap/'F, however, CEA Regu- '

lating Group 4 insertion will be utillzed to maintain the MTC within the Technical Specification limit, Page 18 of 64 i

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TABLE 5-2 FORT CALHOUN UNIT NO.1. CYCLE 13 LIMITING VALUES OF REACTIVITY WORTHS AND ALLOWANCES FOR HOT ZERO POWER MAIN STEAM LINE BREAK. %Ap

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Cvels 12 Cvele 13 f

1. Worth of all CEAs Inserted 8.70 9.23 f

2. Stuck CEA Allowance 1.42 1.83 l

3. Worth of all CEAs Less Worth  !

of Most Reactive CEA Stuck Out 7.28 7.40  !

.1 4.- Power Dependent insertion  !

Limit CEA Worth 1.41 1.23

5. Calculated Scram Worth 5.87 6.17

6. Physics Uncertainty plus Blas 0.59 0.80*

7. Net Available Scram Worth 5.28 5.37 i

8. Technical Specification Shutdown Margin 4.00 4.00

9. Margin in Excess of Technical Specification Shutdown Margin 1.28 1.37

'13% of calculated scram worth.

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TABLE 5-3 FORT CALHOUN UNIT NO.1. CYCLE 13 CEA EJECTION DATA

BOC 12 Value EOC 12 Value BOC 13 Value EOC 13 Value

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Maximum Radial Power Peakina Factor i Full Power with Bank 4 inserted: worst CEA ejected 2.38 2.15 2.41 2.68 q

Zero Power with Banks 4+3 inserted; . _

worst CEA ejected 4.85 4.82 3.43 3.99 '

Maximum Ejected CEA Woph (%Ao)

• Full Power with Bank 4 inserted worst CEA ejected 0.39 0.29 0.22 0.30 Zero Power with Banks 4+3 Inserted worst CEA ejected 0.56 0.62 0.28 0.48

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AA - Assembly Location B.BBBB - Assembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.2674 0.2954

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3 4 5 6 7 0.3494 0.8754 0.9164 1.0099 0.9260 i 8 9 10 11 12 13

• 0.3494 0.9238 1.1798- 1.3646. 1.2773 1.3589 14 15 16 17. 18 19 0.8751 1.1791 1.2760 1.1148 1.3278 1.1052 20 21 22 23 24 25 --

26 O.2672 27 28 29 30 31' 32 --

1.0094 1.2767 1~.3268 1.2157- 1.3619 0.9828 33 1~. 596 34 35 36 37 38 39' O.9258 1.3587 1.1035 1.4072 1.0122- 0.9025 ,

Maximum 1-Pin Peak at 60% Core Height i 1

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. l Cycle 13 Assembly RPD Omaha Public Power District- Figure  !

! O MWD /T, HFP, Equilibrium Xenon Fort Calhoun Station Unit No.'1 5 <

Page 21 of 64 1

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AA - Assembly Location . ,

B.BBBB - Ascembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 2 0.3199 0.3590 3 4 5 6 7-0.3895 0.8920 0.9451 1.1053 1,0058 8 9 10 11 12 13 0.3897 0.9868 1.1742 1.3761 1.2603- 1.3952 1.557 14 15 16 17' 18 19 0.8919 1.1735 1.2066 1.0500 1.2335 1.0602 20 21 22 23 24 25

.0499 1.1950 1.1385 1.3799 26 1.557 0.3198 27 28 29 30 31 32

1. 50 1.2601 1.2329 1.1383 1.3501 0.9754 33 0.3580 34 35 36 37 38 39 1.0057 1.3951 1.0589 1.3806 1.0023 0.9083.

Maximum 1-Pin Peak at 33% Core Height Cycle 13 Assembly RPD Omaha Public Power District Figure 7,000 MWD /T, HFP, Eq. Xenon Fort Calhoun Station Unit No.1 5-2 i Page 22 of 64 I

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l AA - Assembly Location ,

B.BBBB - Assembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly 1 1

1 2 0.3815 0.4304 J l

3 4 5 6' 7 0.4384 0.9124 0.9720- 1.1942 1.0617 8 9 10' 11 12 13 0.4386 1.0576 1.1497 1.3640 1.2237 1.3951 1.563 14 15 16 17 18- 19 0.9124 1.1493- 1.1426 0.9990. 1.1558 1.0222 j 20 21 22 23 24 25 26

'7'8 ' 84 ' ' 'S '8" ' 484 l

j 0.3814 27 28 29 i' 30 31 32-1.1940 1.2237 1.1554 1.0810 1.3346 0.9733 4- 33 4 O4291 34 35 36 37 38 39

1.0616 1.3951

' 1.0211 1.3489 0.9973 0.9159 ,

1.563 j 1 Maximum 1-Pin Peak at 23% Core Height Cycle 13 Assembly RPD Omaha Public Power District. Figure 14,250 MWDfr, HFP, Eq. Xenon Fort Calhoun Station Unit No.-1 5-3 Page 23 of 64 ,

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i AA - Assembly Location -

B.BBBB - Assembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly ,

l 1 2 1'

0.2872 0.3242 l

3 4 5 6 7 0.2504 0.7936 0.9351 1.0875- 1.0131

! 8 9?

10 11 12 13 .

0.2504 M45964; -1.0481 1.3918 1.3696 1.4774-  !

gm 1.692 14 15 16 17 18. 19 i 0.7933 '1.0474 1.2330' 1.1535 --1.4170' 1.1896  !

20 1 21 22 23 24 25 l 26 0.2870  !

27 28 29 30 31 32  ;

1.0870 1.3689 1.4158 1.2740 1.3549 0.9126 i 1

33  !

34 35 36 37 38 39k l 1.0129 1.4772 1.1878 1.4674 0.9395- [0.A516!

1.692 $>, a l

Maximum 1-Pin Peak at 23% Core Height

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- Bank 4 Locations Cycle 13 Ascembly RPD Bank 4 In . Omaha Public Power District Figure O MWD /T, HFP, Equilibrium Xenon  !

Fort Calhoun Station Unit No.1 5-4 Page 24 of 64

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AA - Assembly Location B.BBBB - Assembly Relative Power Density C.CCC - Maximum 1-Pin Peak Assembly ,

i 1 2 0.4316 0.4972 3 4 5 6 7 ,

0.2946 0.8133 1.0107 1.3253 1.1946 I 8 9 '; 10 11 12 13  !

0.2948 "':0,4684i: 0.9832 1.3934- 1.3290 1.5395 1.752-14 15 16 17 18 19 0.8133 0.9828 1.0802- 1.0351- 1.2437 1.1128 20 21 22 23 24 25 j 26  !

0.4314 27 28 29 30 31 32 1.325, 1.3290- 1.2433 1.1359- 1.3180 33 0.8914.

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0.4957 34 35 36 37 38 39A 1.1945 1.5395 1.1117 1.4073 0.9128 h 0)4.1h7:I 1.752 y9y 1 Maximum 1-Pin Peak at 17% Core Height

y - Bank 4 Locations Cycle 13 Assembly RPD Bank 4 in Omaha Public Power District Figure 15,250 MWD /T, HFP, Eq. Xenon Fort Calhoun Station Unit No.1 5-5 Page 25 of 64

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i 6.0 THERMAL-HYDRAULIC DESIGN 1 i

6.1 DNBR ANALYSIS i l

Steady state DNBR analyses of Cycle 13 at the rated power of 1500 MWt have been performed using the TORC computer code described in Reference.1, the-  ;

CE-1 critical heat flux correlation described in Reference 2, and the CETOP-D computer code described in Reference 3 This combination was used in the Cycle 8 through 12 Fort Calhoun reload analyses (References 4 through 8) and the reload methodology can be found in Reference 9.

Table 6-1 contains a list of pertinent thormal-hydraulic parameters used in both safety analyses and for generating reactor protective system setpoint information.

The calculational factors (engineering heat flux factor, engineering factor on hot channel heat input, rod pitch and clad diameter factor) listed in Table 6-1 have -

been combined statistically with other uncertainty factors at the 95/95 confidence / t probability level (Reference 10) to define the design limit on CE-1 minimum DNBR,  ;

6.2 FUEL ROD BOWING The fuel rod bow penalty accounts for the adverse impact on MDNBR of random variations in spacing between fuel rods, The penalty at 45,000 MWD /MTU burnup is 0.5% in MDNBR. This penalty was applied to the MDNBR design-Ilmit of .1.18 (References 6 and 11) in the statistical combination of uncertaintles (Reference.

10),

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.; q TABLE 6-1 FORT CALHOUN UNIT NO.1. CYCLE 13 THERMAL HYDRAULIC PARAMETERS AT FULL POWER' .

1]Dit Cvele 13" Total Heat Output (Core Only) MWt 1500 10' BTU /hr 5119 Fraction of Heat Generated in Fuel Rod 0.975 '

Primary System Pressure Nominal psia 2100' Minimum In Steady State psia 2075-Maximum in Steady State psia 2150 '

i inlet Temperature 'F~ 543 Total Reactor Coolant Flow ppm 202,500 (Steady State)  :'

10 lbm/hr 76.49 (Through the Core) 10' Ibm /hr 73.08!  !

Hydraulic Diameter (Nominal Channel) ft .044 Average Mass Velocity 10* lbm/hr-ft' 2.244 Core Average Heat. Flux '

(Accounts for Heat Generated BTU /hr-ft' 182,757 in Fuel Rod)

Total Heat Transfer Surface Area tt' 28,013 "

Average Core Enthalpy Rise BTU /lbm 70.5 Average Linear Heat Rate kw/ft 6.17 "

a Engineering Heat Flux Factor 1.03" * ' .

Engineering Factor on Hot Channel Heat input 1.03 * '

I Rod Pitch and Bow 1.065 '

• Fuel Densification Factor (Axial) 1.002

• Design inlet temperature and nominal primary system pressure were used to calculate these parameters.

' Based on Cycle 13 specific value of 616 shims. 'l

"" These factors were combined statistically (Reference 8) with other uncertainty .l factors at 95/95 confidence / probability level to define a design limit on CE-1 l minimum DNBR. '

Page 27 of 64 i

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7.0 TRANSIENT ANALYSIS This section presents the results of the Omaha Public Power District Fort Calhoun Station Unit 1, Cycle 13 Non-LOCA safety analyses at 1500 MWt. l The Design Bases Events (DBEs) considered in the safety analysis are listed in Table 7-1, .

These events were categorized in the following groups:

1. Anticipated Operational Occurrences (AOOs) for which the intervention of the Re-

actor Protection System (RPS) is necessary to prevent exceeding acceptable lim-its.

2.. AOOs for which the initial steady state thermal margin, maintained by Limiting Con-ditions for Operation (LCO), are necessary to prevent exceeding acceptable lim-its.

3. Postulated Accidents.

Core parameters input to the safety analyses for evaluating approaches to DNB and cen-terline temperature to melt fuel design limits are presented in-Table 7-2. l As indicated in Table 7-1, no reanalysis was performed for the DBEs for which key tran -  ;

slent input parameters are within the bounds (i.e... conservative with respect to) of the l reference cycle values (Fort Calhoun Updated Safety Analysis Report including Cycle 12' analyses, Reference 1). For these DBEs the results and conclusions quoted in the refer - ,

ence cycle analy, sis remain valid for Cycle 13.- i For those analyses indicated as reviewed, calculations were performed in accordance- i with Reference 6 until a 10 CFR 50.59 determination could be made that Cycle 13 results .j would be bounded by Cycle 12 or the USAR reference cycle. 1 Events were evaluated for up to a total of 6% steam generator tube plugging in Cycle 11 where conservative. Fort Calhoun Station currently has 1.08% steam generator tubes i plugged; thus, no additional analysis is required. '

For the events reanalyzed. Table 7-3 shows the' reason for the reanalysis, the acceptance criterion to be used in judging the results and a summary of the results obtained. Detailed i presentations of the results of the reanalyses are provided in Sections 7.1 through 7.3. i A

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TABLE 7-1 FORT CALHOUN UNIT NO.1. CYCLE 13 .

DESIGN BASIS EVENTS CONSIDERED IN THE NON-LOCA SAFETY ANALYSIS 1 7.1 Anticipated Operational Occurrences for which intervention of the RPS is necessary  !

to prevent exceeding acceptable limits:

,7.1.1 Excess Load Reanalyzed 7.1.2 Reactor Coolant System Depressurization Not Roanalyzed5 7.1.3 Loss of Load Not Reanalyzed 5 ,

7.1.4 Loss of Feedwater Flow Not Reanalyzed 5 .

7.1.5- Excess Heat Removal due to Feedwater Malfunction Not Reanalyzed 5 <

7.1.6 Startup of an inactive Reactor Coolant Pump Not Reanalyzedl <

7.2 Anticipated Operational Occurrences for which sufficient initial steady state thermal margin, maintained by the LCOs, is necessary to prevent exceeding the acceptable -

limits: '

7.2.1 Sequential CEA Group Withdrawal Reanalyzed 2 7.2.2 Loos of Coolant Flow Revieweds.s l

7.2.3 CEA Drop Reanalyzed- 3 7.2.4 Boron Dilution Reanalyzed 7.2.5 Transients Resulting from the Malfunction of One Steam Generator Not Reanalyzedd 7.3 Postulated Accidents 7.3.1 CEA Ejection Reviewed 5 i 7.3.2 Steam Line Break Reviewed 5 -  ;

l 7.3.3 Seized Rotor Reviewed 3 5 l 7.3.4 Steam Generator Tube Rupture Not Reanalyzede '

NOTE: Events evaluated or reanalyzed for the effect of increased steam generator tube i plugging to 6%/SG where conservative.

1 Technical Specifications preclude this event during operation.

2 Requires High Power and Variable High Power Trip.

3 Requires Low Flow Trip.

f dRequires trip on high differential steam generator pressure.

5 Event bounded by reference cycle analysis. A negative determination utilizing the 10 CFR 50.59 criteria was made for this event, eAwaiting approval of proposed methodology change by the NRC.

Page 29 of 64

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TABLE 7 .

FORT CALHOUN UNIT NO.1. CYCLE 13 CORE PARAMETERS INPUT TO SAFETY ANALYSES FOR DNB AND CTM (CENTERLINE TO MLLT) DESIGN LIMITS Physics Parameters .UDits Cvele 12 Values Cvele 13 Values Radial Peaking Factors For DNB Margin Analyses (F4)

Unrodded Region 1.80' 1.70' r Bank 4 Inserted 1.90* 1.73* i For Planar Radial Component (FIv) of 3-D Peak t (CTM Limit Analyses)

Unrodded Region 1.85' 1.75' Bank 4 Inserted 1.94* 1.77*

Maximum Augmentation Factor 1.000 1.000 Moderator Temperature Coefficient 10 tp/*F -2.7 to +0.5 -2.7 to +0.5 Shutdown Margin (Value Assumed in Limiting EOC Zero Power SLB) Mp -4.0 - -4.0 -

'The DNBR analyses utilized the methods discussed in Section 6.1 of this report.

The procedures used in the Statistical Combination of Uncertainties (SCU) as they pertain to DNB and CTM limits are detailed in References 2-5.

Page 30 of 64

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TABLE 7-2 (Continued)

Safety Parameters .U.nfig Cvele 12 Values Cvels 13 Values Power Level MWt 1500* 1500* -~

Maximum Steady State Temperature 'F 545' 543' Minimum Steady State -

Pressurizer Pressure psia 2075* 2075' Maximum Augmentation Factor 1.000 1.000 Reactor Coolant Flow gpm 202,500* '202.500*

Negative Axial Shape LCO Extreme Assumed at Full Power (Ex-Cores) Ip -0.18 -0.18 Maximum CEA Insertion.  % insertion at Full Power of Bank 4 25 25 Maximum initial Linear Heat Rate for Transient Other than LOCA KW/ft 15.22 14.4-Steady State Linear Heat Rate for Fuel CTM l Assumed in the Safety Analysis KW/ft 22.0- 22.0 CEA Drop Time to 100%

Including Holding Coll Delay sec 3.1 3.1 1

Minimum DNBR (CE-1) 1.18* 1.18'

'The effects of uneartainties on these parameters were accounted for statistically-in the DNBR and CTM calculations. The DNBR analysis utilized the methods -

s discussed in Section 6.1 of this report. The procedures used in the Statistical ~ j Combination of Uncertainties (SCU) as they pertain to DNB and CTM limits are detailed in References 2-5.

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TABLE 7-3 FORT CALHOUN LNT NO.'1 DESIGN BASIS EVENTS REANALYZED FOR CYCLE 13 Acceptance Summary Reason for Reanalysis Categon of Results Event ,

increased reactivity insertion Muumtan DNBR greater MDNBR =1.72 Sequential CEA Group PLHGR < 22kW/ft rate and rod shadowing . than 1.18 using the CB-1 Withdrawal factor. correlation. Transient PLHGR < 22 kW/ft.

Minimum DNBR greater MDNBR = 1.36 CEA Drop increase in distortion than 1.18 using CE-1 PLHGR < 22 kW/ft factors nonconservative correlation and Transient.

2 with lower rod worth PLHGR < 22 kW/ft O

N$ Transient PLHGR < 22 kW/ft MDNBR = 1.22 o Excess Load Reduced minimum scram and Minimum DNBR greater PLHGR < 22 kW/ft worth.

g than 1.18 Criterion for' minimum time ' Actual time to lose Boron Dilution increase in Critical Boron to lose prescribed ' shutdown margin >

Concentration.

shutdown margin.' Minimum time.

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3 7.0 TRANSIENT ANALYSIS (Continued)

'L 7.1 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 1)  !

7.1.1 Excess Load Event

.a The Excess Load event was reanalyzed for Cycle 13 to determine the pres-sure bias term for the TM/LP trip setpoint. .

The Excess Load event is one of the Design Basis Events analyzed to deter - (

mine the maximum pressure bias term input to the TM/LP trip. The method- >

ology used for Cycle 13 is described in References 6 and 7. The pressure bias term accounts for margin degradation attributable to measurement and ,

trip system processing delay times. Changes in core power, inlet tempera-ture and RCS pressure during the transient are monitored by the TM/LP trip directly. Consequently, with TM/LP trip setpoints and the bias term deterp .

mined in this analysis, adequate protection will be provided for the Excess  :

Load event to prevent the acceptable DNBR design limit from being exceedJ ed. Table.7.1.1-1 provides a sequence of events for the Excess Load  :

analysis. (

l The analysis of this event shows that a pressure bias term of 90 psia is -

l required compared to the 65 psia value utilized in Cycle 12. This is greater L than that input from the RCS Depressurization event, the other. event for

! which a pressure bias term is calculated. The pressure bias term from the l TM/LP Par equation will be~ increased to 90' psia and incorporated into:

Technical Specification Figure 1-3. The increase in the pressure bias term-more accurately accounts for the transient power decalibration value.,

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TABLE 7.1.1-1 .

FORT CALHOUN UNIT NO.1. CYCLE 13 SEQUENCE OF EVENTS FOR EXCESS LOAD Time (sec) Event Setooint or Valug - -

t 0.1 Steam Dump and Bypass valves open ----

23.21 Reactor trip on High pcwer' 112% of 1500 MWt 23.71 CEAs begin to drop into the core ---

1 23.97 Maximum core heat flux 116.83 %

23.97 Minimum DNBR is reached 1.219 CE-1 correlation

• The reactor trip is a simultaneous Ex-core and A-T Power trip 1

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7.0 TRANSIENT ANALYSIS (Continued).. i 7.1 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 1) (Continued)-

7.1.2 RCS Deoressurization Event The RCS Depressurization event was not reanalyzed for. Cycle _-13 as a bounding pressure bias term was calculated for the Excess Load event.

Since a negative determination utilizing the 10 CFR 50.59 criteria was made -

, for this event for Cycle 13, the conclusions of the Excess Load event calcu-lation of the Peies term remain bounding for Cycle 13.-

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7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES LCATEGORY 2) l l

7.2.1 CEA Withdrawal Event .

. The CEA Withdrawal (CEAW) event was reanalyzed for Cycle 13'to deter- a mine the inillal margins that must be maintained by the Limiting Conditions i

for Operations (LCOs) such that the DNBR and fuel _ centerline to melt l (CTM) design limits will not be exceeded in conjunction with the RPS (Varl.

able High Power, High Pressurizer Pressure, or Axlal Power Distribution Trips).

The methodology contained in Reference 6 was employed in analyzing the CEAW event. This event is classified as one for which the acceptable DNBR and CTM limits are not violated by virtue of maintenance of sufficient -

initial steady state thermal margin provided by the DNBR and Linear Heat Rate (LHR) related LCOs.

For the HFP CEAW DNBR analysis, an MTC Identical to that utilized in Refer- 'I ence 8 and the gap thermal conductivity consistent with the assumption of Reference 6 were used in conjunction with a variable reactivity insertion rate.

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The HFP case for Cycle 13 is considered to meet the 10 CFR 50.59 criteria since the results show that the required overpower margin is less than the available overpower margin required by the Technical Specifications for the 1 DNB and PLHGR LCOs. Since a negative _10 CFR 50.59 determination was made for Cycle 13, the conclusions for Cyclo 12 remain valid and applica-ble to Cycle 13. ,

The zero power case was analyzed to demonsirate that acceptable DNBR and centerline melt limits are.not exceeded. For the zero power case,'a reactor trip, Initiated by the Variable High Power Trip at 29.1% ( 19.1% plus 10% uncertainty of rated thermal power) was assumed in the analysis. .

The 10 CFR 50.59 criteria is satisfied for the HZP-event if the minimum DNBR is greater than that reported in the reference cycle.

The zero power case initiated at the limiting conditions of operation results +

in a a minimum CE-1 DNBR of 5.24 which is less than the Cycle 12 'value of 6.99, but still far in excess of the minimum -1.18 DNBR limit. The analysis shows that the fuel to centerline melt temperatures are well below those corresponding to the acceptable fuel to centerline melt limit. , The sequence of events for the zero power case is presented in Table 7.2.1-1. Figures 7.2.1-1 to 7.2.1-4 present the transient behavior of the core power, core -

average heat flux, RCS coolant temperatures and the RCS pressure for the .

zero power case.

Page '36-of 64 -

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7.0 TRANSIENT ANALYSIS (Continued) 7 '2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) (Continued) 7.2.1 CEA Withdrawal Event (Continued) , }

It may be concluded that the CEA Withdrawal event, when initiated from the -

Technical Specification LCOs (in conjunction with the Variable High Power = s:

Trip, if required), will not lead to a DNBR or fuel temperature which violates <

the DNBR and CTM design limits. -

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. l TABLE 7.2.1-1 FORT CALHOUN UNIT NO.1. CYCLE 13-KEY PARAMETERS ASSUMED IN THE HZP CEA WITHDRAWAL ANALYSIS -

!~ Parameter 1,10lig Cvele 12 Ovele 13

• I i Initial Core Power Level MWt 1 1'

. Core Inlet Coolant '

Temperature 'F 532 532'

Pressurizer Pressure psia 2053 2075' '

1 Moderator Temperature Coefficient x 10" Ap/'F +0. 5 " +0. 5 '

Doppler Coefficient Multiplier . 0.85 0.85 i

CEA Worth at Trip %Ap 5.28- 5.367-Reactivity Insertion Rate Range x 10" Ap/sec 0 to 1.0 0 to 2.3 CEA Group Withdrawal Rate in/ min 46 46 Holding Coll Delay Time sec 0.5 0.5

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• The DNBR calculations used the methods discussed in Section 6.1 of this document and 4 detailed in References 2 through 5. The effects of uncertainties on these parameters were accounted for statistically in the DNBR and CTM calculations.

"DNBR analysis assumes a MTC' consistent with Reference 8.

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Page 38 of 64 -

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Core Power vs. Time Fort Calhoun Station Unit No.1 7.2.1 -1 Page 39 of 64

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Page 42 of 64-

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.i 7,0 TRANSIEhTT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 21 (Continued) ,

7.2.2 Loss of Coolant Flow Event:

The Loss of Coolant flow event was reviewed for Cycle 13 and it was deter - ,

mined that the event was bounded by the Cycle 12 analysis. The input parameters are listed for Cycles 12 and 13 for comparison In Table 7.2.2-1.

Thus, it was concluded that the Cycle 12 analysis Is bounding for Cycle 13 - I operation.

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h TABLE 7.2.2-1 FORT CALHOUN UNIT NO.1, CYCLE 13 KEY PARAMETERS ASSUMED IN THE LOSS OF COOLANT FLOW ANALYSIS Parameter . Units Cvele 12 Cvele 13 -

Initial Core Power Level MWt 1500* 1500*

Initial Core inlet Coolant Temperature 'F 545* 543*

Initial RCS Flow Rate gpm 208,280* 202,500*

Pressurizer Pressure psia 2075* 2075" Moderator Temperature Coefficient x 10" Ap/'F +0.5 +0.5 Doppler Temperature  !

Multiplier 0.85 0.85 l

j CEA Worth at Trip (ARO) %Ap -6.50 -6.94 .

LFT Analysis Setpoint  % of initial flow 93 93 1 LFT Response Time sec 0.65 -0.65

, CEA Holding Coll Delay sec 0.5 0.5  ;

4 CEA Time to 100% insertion sec 3.1 - 3.1 (including Holding Coil Delay Total Unrodded Radial Peaking Factor (Fl) 1.80 1.70

• The uncertainties on these parameters were combined statistically rather than deterministically. '

The values listed represent the bounds included in the statistical combination.

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7.0' TRANSIENT ANALYSIS (Continued)-

7.2 ANTICIPATED OPERATIONAL' OCCURRENCES (CATEGORY 2) (Continued) 7.2.3 Full Lenath CEA Dron Event The Full Length CEA Drop event was reanalyzed for Cycle 13 to determine the initial margins that must be maintained by the Limiting Conditions for_  ;

Operations (LCOs) such that the DNBR and fuel CTM design limits will not .

be exceeded.  !

This event was analy.ted parametr!cally in initial axial shape and rod config- i uration using the methods described in Reference 6. Table 7.2.3-1. lists the >

key input parameters used for Cycle 13 and compares them to the refer-  !

ence cycle (Cycle 11) values while Table.7.2.3-2 contains a sequence of events for the CEA Drop analysis. Figures 7.2.3_-1 to 7.2.3-4 present the transient behavior of the core power, core average heat flux, RCS coolant _  ;

temperatures and the RCS pressure for the limiting full power case.

The transient was conservatively analyzed at full power with an ASI of

-0.182, which is outside of the LCO limit of -0.06. This results in a minimum .

CE-1 DNBR of 1.36. A maximum allowable initiallinear heat generation rate of 15.18 KW/ft could exist as an initial condition without exceeding the ac.

4 ceptable fuel CTM limit of 22 KW/ft during this transient. This amount of margin is assured by setting the LHR related LCOs based on the more limiting allowable LOCA linear heat rate.

it can be concluded that the CEA Drop event, when initiated from the Tech-nical Specification LCOs will not exceed the DNBR CTM design limits.

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1 TABLE 7.2.3-1 l

FORT CALHOUN UNIT NO.1.- CYCLE 13 KEY PARAMETERS ASSUMED IN THE HFP CEA DROP ANALYSIS Parameter 1)DllS Cvels 11 Cvele 13 -

initial Core Power Level MWt '1500* 1500*

Core inlet Coolant

-Temperature 'F 543' 543' Pressurizer Pressure psla 2075'- 2075' i Core Mass Flow Rate gpm 202,500* 202,500* i l Moderator Temperature '

Coefficient x 10" Ap/'F. -2.7 " -2. 7 "

Doppler Coefficient Multiplier 1.15 1.40' CEA Insertion at Maximum Allowed Power  % Insertion of Bank 4 25 25 1

Dropped CEA Worth Unrodded, %Ap -0.2337 -0.2584 PDIL, %Ap

-0.2295 -0.2571

Maximum Allowed Power Shape index at Negative Extreme of LCO Band -0.18 -0.18 .

Radial Peaking Distortion  !

Factor Unrodded Region 1.1566' -1.2269.

Bank 4 inserted 1.1598 1.2266 '

'The DNBR calculations used the methods discussed in Section 6.1 of this document and' detailed in References 2 through 5. The effects of uncertainties on these parameters were accounted for statistically in the DNBR and CTM calculations.

"DNBR analysis assumes a MTC consistent with Reference 8. -

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Page 46 of 64

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TABLE 7.2.3-2 i FORT CALHOUN UNIT NO.1 CYCLE 13 SEQUENCE OF EVENTS FOR FULL LENGTH CEA DROP.

l Time (sec) Event Setnoint or Vnlue -

t 0.0 CEA Begins to Drop into Core- --- '

1.0 CEA Reaches Fully inserted Position 100% insertion ,

1.24 Cors Power Level Reaches a Minimum 65.7% of 1500 MWt '

and Begins to Return to Power Due to Reactivity Feedbacks ,

168.52 Core inlet Temperature Reaches a 538.49'F Minimum Value 199.5 RCS Pressure Reaches a Minimum 2010.4 psia  !

Value 200.0 Core Power Returns to its Maximum 92.5% of 1500 MWt Value 200.0 Minimum DNBR is Reached-1.36 (CE-1 Correlation) 1

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Page 47 of 64

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l 7.0 TRANSIENT ANALYSIS (Continued) 7.2 ANTICIPATED OPERATIONAL OCCURRENCES (CATEGORY 2) (Continued) ,

i l 7.2.4 Boron Dilution Event The Boron Dilution event was reanalyzed for Cycle 13 to verify that sufficient time is available for an operator to identify the cause and to terminate an .[

approach to criticality for all subcritical modes of operation.

l Table 7.2.4-1 compares the values of tho' key transient parameters as-sumed in each mode of operation for Cycle 13 and the reference cycle, #

Cycle 11. Table 7.2.4-2 provides the results of the time to lose shutdown ,

margin calculations for Cycle 13 and Cycle 11. The Cycle 13 analysis utilized a mass basis in the calculations, rather than a volumetric basis to ensure that all operating temperature ranges for all modes of operation were bounded. -

P As noted in this table, the critical boron concentration for Cycle 13 is greater l

than the corresponding Cycle 12 values for all modes. Therefore, the time to lose critical shutdown margin will decrease from Cycle 11 results due to the inverse relationship between response time and critical boron concen-tration. A Technical Specification amendment was proposed in Reference 13 to increase the refueling boron concentration to 1900 ppm from the exist.

Ing 1800 ppm value to ensure that the minimum time to critical shutdown margin is maintained. -

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TABLE 7.2.4-1 FORT CALHOUN UNIT NO.1. CYCLE 13 KEY PARAMETERS ASSUMED IN THE BORON DILUTION ANALYSIS Parameters Cvele 12 Values Cvels 13 Vahiae ,

Critical Boron Concentration. oom (ARO. No Xenon)

Mode Hot Standby 1580 1622 Hot Shutdown 1580 1622 Cold Shutdown - Normal RCS Volume 1480 1457 Cold Shutdown - Minimum RCS Volume' 1290 1279 Refueling 1400 1454 Inverse Boron Worth. Dom /%Ao Mode Hot Standby -90 --90 Hot Shutdown -55 -55 Cold Shutdown - Normal RCS Volume -55 -55 Cold Shutdown - Minimum RCS Volume -55 -55 Refueling -55 -55 Minimum Shutdown Maroin Assumed. %An t

l MQdB I

Hot Standby -4.0 -4.0 Hot Shutdown -4.0 -4.0 Cold Shutdown - Normal RCS Volume -3.0 -3.0 Cold Shutdown - Minimum RCS Volume * -3.0 -3.0 Refueling (ppm)" 1800 1900 I

• Shutdown Groups A and B out, all Regulating Groups inserted except most reactive rod stuck out.

" includes a 5.0%dp shutdown margin.

4 Page 53 of 64

TABLE 7.2.4-2 FORT CALHOUN UNIT NO.1. CYCLE 13 RESULTS FROM BORON DILUTION EVENT Time to Lose Prescribed Criterion For Minimum Time to -

Mode Shutdown Marain (Min) Lose Prescribed Shutdown Marain (Mini Cvele 11 Cvele 13 Hot Standby 91.6 53.2 15 Hot Shutdown 44,7 33.7 15 Cold Shutdown 36.2 37.5 15 Normal Volume Cold Shutdown 15.2 15.1 15 Minimum Volume Refueling 31.8 33.3 30 i

Page 54 of 64

7.0 TRANSIENT ANALYSIS (Continued) 7.3 POSTULATED ACCIDENTS 7.3,1 CEA Election

~

The CEA Ejection event was reviewed for Cycle 13. A summary containing the results of this analysis was submitted in Reference 11 for Cycle 11 and has been validated for use in Cycle 13, 7.3.2 Steam Line Break Accident This accident was reviewed for Cycle 13 using the methodology discussed in References 6 and 12. The Steam Line Break (SLB) accident was pre.

vlously analyzed in the Fort Calhoun FSAR and satisfactory results were reported therein. The SLB accidents at both HZP and HFP were examined in the re'erence cycle (Cycle 8) safety evaluation with acceptable results-

'obtained. Both the FSAR and reference cycle evaluations are reported in the 1989 update of the Fort Calhoun Station Unit No.1 USAR.

The Cycle 13 Full Power Steam Line Break accident was reviewed for a more negative MTC of -2.7 x 10" Ap/'F than the -2.5 x 10" Ap/'F value that was used in the Cycle 8 analysis. However, the cooldown curve for Cycle 13 is bounded by Cycle 8 (as shown in Figure 7.3.2-1). This figure -

shows that the reactivity insertion for the Cycle 13 core with an MTC of -2.7 x 10" Ap/'F due to a SLB accident at full power is substantially less than the value used in the Cycle 8 analys!s. (This smaller reactivity insertion is

.;

due to the use of,the DlT cross-sections which are valid for a range of moderator temperatures from room temperature to 600'K while the analy-ses prior to Cycle 9 were performed with cooldown curves derived by con-servatively extrapolating CEPAK cross-section - values to low temperatures.) The fuel temperature coefficient (FTC) used in the Cycle 8 analysis is conservative with respect to the fuel temperature coefficient cal-culated for the Cycle 12. core including uncertainties. The Cycle 13 mini-mum available shutdown worth at HFP is 6.83%Ap compared to a Cycle 8 value of 6.68%Ap. This implies a margin increase of 0.15%Ap. The Cycle 13 moderator cooldown reactivity between 574'F and 350'F at HFP is 4.68%Ap compared to 5.37 %Ap in Cycle 8. This implies a margin in-crease of 0.69 %Ap. The Cycle 13 doppler coefficient is more negative than the Cycle 8 doppler including uncertainties. However, this loss in mar.

gin is offset by the gain in margin from the scram worth and moderator cooldown reactivity. The net gain assures that the overall reactivity inser-tion for a Cycle 13 SLB is less than that of the reference cycle analysis.

Therefore, the return to power is less than that of the reference cycle and Cycle 1 FSAR analyses.

Page 55 of 64

7.0 TRANSIENT ANALYSIS (Continued) 7.3 POSTULATED ACCIDENTS (Continued) 7.3.2 Steam Line Break Accident (Continued) -

A similar evaluation was performed for the Hot Zero Power SLB accident.

Again, the Cycle 13 cooldown for an MTC of.-2.7 x 10%p/'F shows a substantially smaller reactivity insertion than was used in the Cycle 8 analy-sis (as seen in Figure 3.2.2-1). Since the minimum available shutdown margin for Cycle 13 remains unchanged from the reference cycle value (4.0%Ap), the overall reactivity insertion for the Cycle 13 SLB accident will be less severe than that reported for the reference cycle and the FSAR (Cycle 1) cases.

Based on the evaluation presented above, it is concluded that the conse-quences of a SLB accident initiated at either zero or full power are less severe than the reference cycle and FSAR (Cycle 1) cases.

Since a negative determination utilizing the 10 CFR 50.59 criteria was made for the Cycle 13 SLB accident, no reanalysis was performed. Thus, it was concluded that the reference cycle analysis is bounding for Cycle 13 opera-tion.

l Page 56 of 64 J

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Steam Line Break Accident Omaha Public Power District Figure :

ReactMty vs. Moderator Temperature Fort Calhoun Station Unit No.1 7.3.2-1 Page 57 of 64 ,

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l l 7.0 TRANSIENT ANALYSIS (Continued)

.i 7.3 POSTULATED ACCIDENTS (Continued) 7.3.3 Salzad Rotor Event .

The Seized Rotor event was reviewed for Cycle 13 to demonstrate that only  !'

a small fraction of fuel pins are predicted to fall during this event. C is bounded by the reference cycle (Cycle 9) analysis because ai of 1.85 was assumed in the Cycle 9 analysis and the Cycle 13 Technical Specifica-tion of 1,70 remains conservative with respect to the Fl value used in the Cycle 9 analysis.

Therefore, the total number of pins predicted to fall will continue to' be less <

than 1% of all of the fuel pins in the core. Based on this result, the resultant site boundary dose would be well within the limits of 10 CFR 100. '

Since a negative determination utilizing the 10 CFR 50.59 criteria was made for the Cycle 13 Seized Rotor event, no reanalysis was performed.

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Page 58 of 64 -

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8.0 ECCS PERFORMANCE ANALYSIS i Both Cycle 11 Large and Small Break Loss of Coolant accident analyses were performed l

using the methodology discussed in Reference 1. A summary containing the results of the i analyses was submitted in Reference 2. The Cycle 11 revised ECCS analysis was verified  !

to be valid for use in Cycle 13 given the bounding input assumptions. The peak linear heat ~  !

generation rate of 15.2 KW/ft was conservatively reduced to 14.4 KW/ft to ensure the CE i fuel mechanical design requirements were valid for the entire Cycle 13 operation. The l ECCS analysis was verified for Cycle 13 wth the reduced peak linear heat generation rate  ;

and determined to be bounded.  !

Thus, the ECCS analysis was reviewed for Cycle 13 and found to be bounded by the .  !

Cycle 11 analysis. Since a negative determination using the 10 CFR 50.59 criteria was -

made for Cycle 13. the conclusions for Cycle 11 remain valid and applicable to Cycle 13.

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Page 59 of 64 *

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9.0 STARTUP TESTING The startup testing program proposed for Cycle 13 is identical to that used in Cycle 12. It is also the same as the program outilned in the Cycle 6 Reload Application, with two exceptions. First, a CEA exchange technique (Reference 1) for zero power rod worth -

measurements will be performed in accordance with Reference 2. replacing the boration/

dilution method. Also, low power CECOR flux maps and pseudo-ejection rod measure-monts will be substituted for the full core symmetry checks.

The CEA exchange technique is a method for measuring rod worths which is both faster and produces less waste than the typical boration/dllution method. The startup testing method used in Cycles 11 and 12 employed the CEA exchange technique exclusively, Results from the CEA exchange technique were within the acceptance and review critoria for low power physics parameters. The combination of the pseudo-ejection technique at zero power and low power CECOR maps provides for a less time consuming but equally valid technique for detecting azimuthal power tilts during reload core physics testing. The pseudo-ejection rod measurement involves the dilution of a bank into the core, borating a CEA out, and then exchanging (rod swap) the CEA apsinst other symmetric CEA's within the bank to measure rod worths. The acceptance and review criteria for these tests are:

Insi Acceptance Criteria Review Criteria CEA Group Worths 15% of predicted- 15% of predicted -

Pseudo-ejection None ..The greater of: 2.5e rod worth deviation from group measurement average or 15% deviation from group average.

Low Power CECOR Technical Specifica- Azimuthal tilt less than maps tion limits of Fl. 20%.

Fiv, and T.

OPPD has reviewed these tests and has concluded that no unreviewed safety question exists for implementation of these procedures.

4 Page 60 of 64 m-

10.0 REFERENCES

References (Chaoters 1-5)

1. " Omaha Public Power District Reload Core Analysis Methodology Overview", -

OPPD-NA-8301-P, Revision 03, April 1988.

2. " Omaha Public Power District Reload Core Analysis Methodology - Neutronics Design Methods and Verification". OPPD-NA-8302-P, Revision 02 April 1988.

3. " Omaha Public Power District Reload Core Analysis Methodology - Transient and Accident Methods and Verification", OPPD-NA-8303-P, Revision 02, April 1988.

4. " Omaha Batch M Reload Fuel Design Report", CEN-347(O)-P Revision 0, No-vember 1986.

5. " Generic Mechanical Design Report for Exxon Nuclear Fort Calhoun 14 x 14 Reload Fuel Assembly", XN-NF-79-70-P September 1979.

6. " Qualification of Exxon Nuclear Fuel for Extended Burnup", SN-NF-82-08(P)(A) & -.

Supplements 2, 4 & 5, Revision 1. October 1986.

7. " Extended Burnup Report for Fort Calhoun Reloads XN-4 and XN-5 (Batches K and L)', ANF-87-133(P), October,1987.

8. " Amendment No.117 to Operating License DPR-40, Cycle 12 License Applea-tion", Docket No. 50-285, December 14, 1988.

l Page 61 of 64

n- -

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10.0 REFERENCES

(Continued)

References (Chanter 6)

1. ^

" TORC Code, A Computer Code for Determining the Thermal Margin of a Reactor Core" CENPD-161-P, July 1975.

2.

" Critical Heat Flux Correlation For CE Fuel Assemblies with Standard Spacer Grids, Part 1. Uniform Axlal Power Distribution", CENPD-152-PA April 1975.

3.

"CETOP-D Code Structure and Modeling Methods for Calvert Cliffs Units 1 and 2",

CEN-191-(B)-P December 1981 4.

Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amend.

ment No. 70 to Facility Operating License No. DPR-40 for the Omaha Public Power district, Fort Calhoun Station, Unit No.1 Docket No. 50-285, March 15,1983.

5.

Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amend-ment No. 77 to Facility Operating License No. DPR-40 for the Omaha Public Power District, Fort Calhoun Station, Unit No.1 Docket No. 50-285, April 25,1984.

6.

Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Arnere ment No. 92 to Facility Operating License No. DPR-40 for the Omaha Public Power District, Fort Calhoun Station, Unit No.1. Docket No. 50-285, November 29,1985.

7.

Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amend-ment No.109 to Facility Operating Licenss No. DPR-40 for Omaha Public Power District, Fort Calhoun Station, Unit No.1. Docket No. 50-285, May 4,1987.

8.

Safety Evaluation by the Office of Nuclear Reactor Reguls. tion Supporting Amend-ment No.117 to Facility Operating License No. DPR-40 for Omaha Public Power District Fort Calhoun Station, Unit No.1. Docket No. 50-285, December 14,1988.

9. j

" Omaha Public Power District Reload Core Analysis Methodology-Overview", '

OPPD-NA-8301-P Revision 03, April 1988.

10. " Statistical Combination of Uncertainties, Part - 2 " Supplement 1-P, CEN-257(O)-P, August 1985. '

11.

Safety Evaluation Report on CENPD-207-P-A, "CE Critical Heat Flux: Part 2 Non-Uniform Axial Power Distribution", letter, Cecil Thomas (NRC) to'A. E. Scherer (Combustion Engineering), November 2,1984.

l 4

Page 62 of 64

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o

10.0 REFERENCES

(Continued)

References (Chanter 7) 1.

" Amendment No.117 to Operating License DPR-40. Cycle 12 License Applica.

tion". Docket No. 50-285, December 14,1988. -

2. " Statistical Combination of Uncertainties Methodology, Part 1: Axial Power Distri.

bution and Thermal Margin / Low Pressure LSSS for Fort Calhoun", CEN-257(0)-P, November 1983." Supplement 1-P, CEN-257(O)-P, August 1985.

3. " Statistical Combination of Uncertainties Methodology, Part 2: Combination of System Parameter Uncertainties in Thermal Margin Analysis for Fort Calhoun Unit 1", CEN-257(0)-P, November 1983.

4. " Statistical Combination of Uncertainties Methodology, Part 3: Departure from Nu-cleate Bolling and Linear Heat Rate Limiting Conditions for Operation for Fort Cal.

houn", CEN-257(0)-P, November 1983.

5. " Statistical Combination of' Uncertainties, Part 2, " Supplement 1-P, CEN-257(O)-P, August 1985.

6. " Omaha Public Power District RMoed Core Analysis Methodology - Transient and Accident Methods and Ven5 cation". OPPD-NA-8303-P, Revision 02, April 1988.

7. "CE Setpoint Methodology", CENPD-199-P, Rev.1-P, March 1982.

8. "CEA Withdrawal Methodology", CEN-121(B)-P, November 1979.

9.

"CESEC. Digital Simulation of a Combustion Engineering Nuclear Steam Supply System", Enciosure 1-P to LD-82-001, January 6,1982.

10.

" Response to Questions on CESEC", Loulslana Power and Light Company, Water.

ford Unit 3. Docket 50-382, CEN-234(C)-P, December.1982.

11.

Letter LIC-86-675, R. L. Andrews (OPPD) to A. C. Thadani (NRC), dated January i 16,1987.

12.

" Omaha Public Power District Reload Core Analysis Methodology - Neutronics Design Methods and Verification". OPPD-NA-8302-P, Revision 02, April 1988.'

13 Letter LIC-89-1172, K. J. Morris (OPPD) to Document Control Desk (NRC), ex-pected date of January 23,1990.

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